Course Book Safety Officer Radiation Refresher class.pdf · 1 Course Book Radiation Safety Officer...

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Course Book

Radiation Safety Officer

Radiation Safety Officer Refresher

Nevada Technical Associates

P. Andrew Karam, Ph.D., CHP (585) 247‐4510 (voice and fax)

[email protected]

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Table of Contents

1 A BRIEF HISTORY OF RADIATION PROTECTION ........................................................ 4 2 RADIATION PROTECTION BASICS................................................................................. 10 3 NATURALLY OCCURRING RADIATION AND RADIOATIVITY................................... 21 4 THE BIOLOGICAL EFFECTS OF IONIZING RADIATION........................................... 26 5 RADIATION DETECTORS AND DOSIMETERS ............................................................. 40 6 RESPONDING TO RADIOLOGICAL INCIDENTS AND EMERGENCIES...................... 46 7 10 CFR PART 19 ..................................................................................................................... 57 8 10 CFR PART 20 ..................................................................................................................... 66 9 10 CFR PART 40 ................................................................................................................... 154 10 EXAMPLES OF COMMON PROCEDURES.................................................................. 216 11 RUNNING AN EFFECTIVE HEALTH PHYSICS PROGRAM ................................... 245 COURSE SLIDES ..................................................................................................................... 272 APPENDIX A: GLOSSARY OF TERMINOLOGY ............................................................. 386 APPENDIX B: HOW NUCLEAR REACTORS WORK ......................................................... 390 APPENDIX C: THE NATURAL NUCLEAR REACTOR IN OKLO ..................................... 417 APPENDIX D: REGULATORY GUIDE 8.36 - RADIATION DOSE TO THE EMBRYO/FETUS ...................................................................................................................... 425

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A BRIEF HISTORY OF RADIATION PROTECTION The study of radiation began with the discovery of x‐rays in 1895. The initial experimenters did not realize the potential adverse effects of radiation due to a lack of experience with it. They noted that x‐rays could penetrate matter and expose photographic plates, even while being invisible and assumed that they presented a unique way to diagnose and treat diseases from within the body. In the absence of any immediate physical effects (with the exception of reddening of the skin) it was assumed that there would be no adverse biological effects from this treatment. It took only a few years to discover that x‐rays could, indeed, cause bodily harm in large enough doses prompting radiation users to begin to use shielding. The field, however, was still in its infancy and researchers simply had not had the time to amass sufficient information to really come to any understanding about the effects of exposure to x‐rays. In the absence of any negative information, doctors and the public went under the assumption that exposure to x‐rays was more beneficial than harmful. In a classic experiment in that same year, a physicist decided to determine for himself whether or not x‐rays could cause burns. In this experiment, Elihu Thompson exposed his little finger to x‐rays for a short time each day until the skin started to redden and his finger became painful. By so doing, he determined that low doses of x‐rays can cause cumulative harm that may not appear for some time. Discovery of even longer‐term effects, delayed cancers, was years in the future, leading to the general perception that, so long as there was no short‐term effect, there was no risk. (Incidentally, around this time, a common method of calibrating x‐ray machines was to place oneʹs hand in the path of the beam until the skin began to redden.) The next several years saw increasing use of x‐rays and radiation for a variety of ʺtreatmentsʺ. X‐rays were used for treatment of tuberculosis, chronic aches and pains, criminal behavior, and to remove womenʹs unwanted facial hair. This last use alone kept many plastic surgeons in business for a number of years. Even legitimate uses of x‐rays, such as diagnostic imaging, led to great overexposure from poorly‐shielded machines, inefficient film emulsions, and a lack of concern for dose reduction. Adding to the exposure problem was the lack of personnel dosimetry, the lack of a standard unit for the measurement of radiation dose, the lack of any quantitative measurements relating radiation dose with biological damage, and the lack of exposure limits. Without any method of measuring the exposure received and without even a standard unit of measurement, no radiation protection was even possible at this time. This

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overexposure was more of a concern for the technician who worked with the equipment daily than for the patient, but both were (by today’s standards) overexposed routinely. By 1904 radium had been discovered and was hailed, as x‐rays had been, as a new medical miracle. Radium‐bearing medicines were a common treatment for a variety of problems that would expand over the next 30 years to include over 160 ailments, including, ironically, lethargy, impotence, and baldness1. Some radium‐laced products that were sold during this time frame were toothpastes, hair tonics, skin creams, and water treatment units that infused radium or thorium into drinking water as a preventative medicine. Another popular use for radium was in the painting of luminous watch dials. Studies done in the mid 1920ʹs revealed many cases of death from radium poisoning in both the radium watch dial painters, mostly young girls, and among the regular users of radium‐bearing nostrums. These investigations generally had little effect on the official perception of the dangers of radium until the death of a famous industrialist, Eben Byers, in 1934 from radium poisoning. Byers had been drinking a ʺmedicineʺ named Radiothor in an effort to improve his virility, consuming an average of 8 uCi daily for several years. This resulted in an untimely and horrible death and led to the banning of radium in any consumer products in 1936. The estimated body burden of the radium watch dial painters ranged up to 36 uCi, as compared with todayʹs legal body burden limit of 0.1 uCi. This provided the first definite link between radioactive materials and cancer. In the mid and late 1920ʹs came the adoption of a standard unit of measurement (the Roentgen), radiation detection equipment, and the first recommended dose limits of 200 mr/day. This dose limit was based on receiving one tenth of the dose that was needed to produce skin erythema (reddening of the skin). This dose limit recommendation was repeated in 1931 by the National Bureau of Standards. In 1932, recognizing the varying tolerances of different parts of the body to radiation, it was suggested that the daily dose limit be 100 mr to the whole body or 500 mr to the hands. In 1949, this was reduced even further by the National Council on Radiation Protection (NCRP) to 300 mr/wk and, in 1956, the current limits of 5 REM/yr to the whole body for occupationally‐exposed individuals and 0.5 REM/yr for everyone else was adopted. The field of Health Physics was officially born in 1942 when a physicist for the Manhattan Project, Ernest Wollen, was assigned the task of determining the effects of radiation on Project workers, recommending proper techniques to control their exposure, and to develop reliable methods of measuring this exposure. The result was, in the space of just a few years, amassing enough information to found a new profession and to lay the groundwork for years of studies.

1 The irony is that excessive radiation exposure can cause hair loss, temporary sterility, and lethargy.

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Following the atomic bombing of Japan, Manhattan Project scientists were dispatched to Hiroshima and Nagasaki to study the effects of radiation on humans over both short‐term and long‐term time frames. Other health physicists covered the atomic and hydrogen bomb tests in Nevada and in the Pacific, studying the effects of the explosions, the immediate radiation, and the fallout on plant and animal life, including their effects on humans. From these studies, as well as from accidental releases of fission products and from more laboratory work, our present limits for exposure to radiation and radioactive isotopes were derived. Most of todayʹs exposure and concentration limits are a direct result of the work done by the Manhattan Project health physicists. There is still on‐going research with respect to the recommended radiation limits. The Chernobyl accident yielded a large amount of data on the effects of radiation doses on humans, and reassessments of the radiation release from the atomic bombs is still going on, as well. There has been more research done on the effects of ionizing radiation than on any toxic substance and there is more known about radiationʹs effects on humans than about any other harmful phenomenon. It is also interesting to note that, despite radiationʹs being undetectable to our senses, it is more easily detected than virtually any other substance known. Radiation detectors can reliably detect, depending on the isotope, as little as 10‐23 grams of radioactive material. The standards that we follow and the radiation limits to which we adhere are based upon nearly a century of trial and error. Most of the practices of the past seem ridiculous, antiquated, naive, or just plain stupid with our current knowledge of the effects of radiation and radioactivity. However, without this stupidity and naivete we might still be blissfully unaware of the potential hazards of radiation. Eben Byers and the radium watch dial painters died prematurely and horribly but their deaths led to studies that emphasized the importance of minimizing body burdens of radium and, by extension, of all radioisotopes. The atomic bomb survivors suffered greatly but this suffering led to a much more thorough understanding of the short and long term effects of radiation on the body and, shortly thereafter, the adoption of new and lower standards for radiation exposure. Despite any claims that have been made over the decades, radiation is not our friend. However, neither is it an enemy. It is more analogous to a table saw or a milling machine ‐ a useful tool that, if not treated with the proper respect, will injure or kill those who use it without the respect that it is due.

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A Brief Chronology of Radiation Practices And Standards 1784 W. Morgan produces (but does not recognize) x‐rays in an experiment witnessed

by Benjamin Franklin 1895 Roentgen discovers x‐rays (Nov) 1896 Becquerel discovers radioactivity (Feb) ‐first documented damage to eyes from x‐rays (Mar) ‐first use of protective shielding (July) ‐first documented skin burns from x‐rays ‐public exhibition at Central Park in NY lets people use a fluoroscope to view

internal body parts 1897 x‐ray ʺbathsʺ used in treatment of tuberculosis, criminal behavior, and other

ʺailmentsʺ ‐x‐ray therapy used for removal of unwanted facial hair in women ‐ results in

scarring and disfigurement 1898 first use of protective gloves for x‐ray technicians 1899 first malpractice award given for misuse of x‐ray therapy 1901 first alleged lethality from overexposure to x‐rays ‐first documented skin burn from radioactive material 1904 first death attributed to cumulative overexposure to x‐rays ‐introduction of radium and radium‐based medicines for personal use 1907 first film badge used for personal monitoring 1913 first use of radioactive tracers in research (lead compounds used in solubility

research) 1924 investigation of radium watch dial painters ‐ early case of radium poisoning and

radium‐induced cancers painters had body burdens of up to 36 uCi 1925 first recommended daily x‐ray dose limit of 200 mr/day

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1928 adoption of the Roentgen as the unit of radiation exposure 1930ʹs use of water irradiators to infuse radium and thoriuminto water for health ‐

resulted in up to 0.026 uCi/yr of body burden (26,000 pCi/yr) 1931 first national dose limit of 200 mr/day 1932 industrialist Eben Byers died after consuming 8 uCi/day of “Radiothor, a New

and Effective Remedy” (per the AMA) for several years 1934 FDA outlaws the use of radium in food, drink, and drugs ‐first use of radioactive tracers in human research (P‐32) ‐foot fluoroscopy becomes popular, remains so through the 1950ʹs until linked

with cancers ‐ gave doses in the tens of REM, caused some skin burns 1936 allowed dose limit lowered to 100 mr/day ‐US FDA noted increased sales of radium‐containing health and beauty products 1940ʹs a variety of radioactive tablets and nostrums sold as cures for up to 160 various

ailments, including impotence ‐radium and thorium found in hair tonics, skin creams and other cosmetics, and

chocolate bars ‐radium fertilizer used in U.S. ‐radium and thorium toothpastes introduced by Germans 1941 introduction of body burden limits for radium ‐recommended dose limit lowered to 20 mr/day 1942 the field of Health Physics born with the appointment of Ernest Wollan (a

Manhattan Project physicist) to study the biological effects of radiation 1942‐1950 Manhattan Project scientists develop first good survey instruments, study

effects of radiation and radioactive elements on humans, gather dose effect and body burden information, and more

‐recommended treatment for accidental entry of plutonium into a cut or scratch was immediate high amputation of the affected limb

1945‐1946 studies of Japanese atomic bomb survivors led to first statistically valid

exposure limits

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1950ʹs above‐ground nuclear bomb testing and fallout scares heighten public opposition to any uses of radioactive isotopes

1955 first commercial nuclear reactor built 1956 first nuclear submarine built 1960ʹs increased push for the Atoms for Peace and Operation Plowshare programs to

lower public opposition to nuclear power and use of radiation in research 1960ʹs a boom in ordering of civilian nuclear power plants 1979 Three Mile Island accident and growing public anti‐ nuclear sentiment shut down

domestic nuclear power industry 1986 Chernobyl accident

‐beginnings of public concern with respect to radon levels in dwellings Today people still drink radioactive water from hot springs and visit mines to breathe

radon for their health ‐thorium, americium, uranium, and other radioactive elements still used, or

appear as contaminants, in numerous consumer products ‐use of radioactive isotopes in nuclear medicine and for research continues to be valuable

‐ growing body of knowledge points to possible threshold or hormesis (beneficial) effects of exposure to low levels of radiation ‐ research continues

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RADIATION PROTECTION BASICS

Review of Atomic Structure and the Basic Types of Radiation Atoms are the fundamental units of matter. They bond together to form chemical compounds. The sizes of atoms range from one tenth of an angstrom to nearly two angstroms (108 A = 1 cm). Atomic nuclei are composed of protons and neutrons. The number of protons present determines an atom’s chemical properties. Protons all have a positive charge, so they tend to repel each other electrically. Because of this, neutrons are needed to help hold an atom together – neutrons carry the strong nuclear force that overcomes the electrostatic repulsion of the protons. The strong nuclear force has a very short range so, as an atom increases in size, more neutrons are needed to stabilize the nucleus. When neutrons are added to or subtracted from an atomic nucleus, the energy level of the nucleus is altered and the atom may become unstable. So, for example, a carbon atom with 6 protons and 6 neutrons is stable carbon‐12 while a carbon atom with 6 protons and 8 neutrons is unstable carbon‐14. Carbon 12 and ‐14 are called two isotopes of carbon because they have the same number of protons but different numbers of neutrons. The two most important intrinsic properties of these particles are their mass and their electric charge as these both have a bearing on their interactions with matter and their ability to cause damage. The higher the charge and the more mass that a particle contains, the more damage that it can do. Electrons and positrons are the lightest of these particles. They carry a charge of ±12 and have a mass of one two‐thousandth that of the proton or neutron. They can interact with matter either by direct ionization, or by bremsstralung. High energy electrons are referred to as beta radiation. Direct ionization consists of an electron striking an atom and knocking loose one of that atomʹs electrons. This creates an ion pair (a positively charged atom and a negatively charged electron) that can go on to cause more ionizations within the cell. Bremsstralung is German for ʺbraking radiationʺ and is caused by an electron passing near to a heavy atom. The atom and electron interact electrostatically, the atom deflecting the electron

2 Positrons are the anti-matter equivalents of electrons – identical in all ways, but with a positive electrical charge. When positrons meet “normal” electrons, they annihilate each other, converting their mass into energy and emitting twin gamma ray photons, each with an energy of 511 keV

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which gives off radiation (usually in the x‐ray region) as it changes course. A thin lead shield that is placed around a beta source will shield all of the beta radiation but will emit x‐rays due to bremsstralung. Beta radiation is weakly penetrating and usually constitutes a skin dose only, although the lens of the eye is also susceptible. Due to its low mass and charge of ‐1, the beta can (and should) be shielded by plastic. The number of ionizations caused by beta radiation is proportional to the energy (velocity) of the beta radiation and to the mass of the atoms that it is passing through. So, in general, higher‐energy beta particles will cause more ionizations and those that are interact with heavier atoms will cause more bremstralung x‐rays. Another type of particulate radiation is the alpha particle. Alpha particles are helium atoms that have had their electrons removed, giving them a charge of +2. They are also massive with a mass of 4 amu. This means that they are capable of causing more damage than any other form of radiation, and also that they are far less penetrating. A piece of paper is an adequate alpha shield and they generally cannot penetrate the dead layer of skin that we all have. These properties make alpha particles a concern for internal dose because they cause a lot of damage to living cells, but they are not an external risk. A third form of radiation is the gamma ray, a high‐energy photon that is given off by atomic nuclei that have been excited by beta emission, neutron capture, electron capture, or some other means. Most radioactive decays will produce gamma radiation. An atomic nucleus contains protons and neutrons in discrete energy states, much like the electrons surrounding it. During radioactive decay, the decay particles carry off energy. Unless this energy is the exact amount needed for transition to the next lower energy level, the nucleus is still in an excited state. The nucleus will de‐excite by emission of a gamma containing the energy difference between the energy state that the nucleus is in and the next lower energy level. There are a few nuclides, such as tritium (H‐3) that emit a particle containing the exact amount of radiation that is required for this transition to a stable configuration; the rest of the nuclides will emit gamma radiation when they decay. Gammas have no mass and no charge and interact by either direct collision with electrons, knocking them out of their orbits (the photo‐electric effect), production of an electron‐positron pair if it passes near a heavy nucleus (pair production), or by absorption and re‐emission by an atom, usually in a different direction and at a different energy (Compton scattering). Photons interact very weakly with matter and are best shielded by dense materials such as lead. Photons, along with neutrons, are considered to be a whole‐body dose as they will penetrate through the entire body.

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Radioactive Decay The usual reason for an atom to be unstable is either an excess or a deficit of neutrons in the nucleus. The neutrons provide the force which keeps the positively‐charged protons from repelling each other and ripping the atoms apart. An atom with an insufficient number of neutrons is more likely to decay radioactively, as is an atom with an excessive number of neutrons. In addition, when atoms reach a certain size, the distance across the atom becomes great with respect to the range of the force that is trying to hold them together. This makes these elements more likely to be unstable. The two major decay mechanisms for very large atoms are alpha emission (giving off two protons and two neutrons) and spontaneous fission (breaking into two or more parts of similar size). The best way to give an element an excess of neutrons is to bombard it with neutrons from a neutron generator or in the interior of a nuclear reactor. The major decay mode for lighter atoms is either beta emission or electron capture. In beta emission, the unstable atom will emit an electron or a positron (β‐ or β+) from a neutron or a proton, giving the neutron a net positive charge and turning it into a proton or turning a proton into a neutron. Either of these will change the atom from one element to another since the chemical properties of an element are due to the number of protons that the nucleus contains. Another method of releasing energy from atoms is in the form of high‐energy photons, known as gamma radiation. Any change in the nuclear structure of an atom should result in the emission of a gamma as the nucleus reverts to a less‐excited energy state. The energy of the gamma depends upon the excitation state of the nucleus and on whether the nucleus de‐excites by emitting one gamma or several. On occasion, too, the emission of a gamma from the nucleus will knock an orbital electron out, turning the atom into an ion pair. These electrons are known as Auger (pronounced o‐zhay) electrons. Yet another common means for radioactive decay is called electron capture. This occurs when a nucleus ʺcapturesʺ an orbital electron, merging it with a proton to form a neutron. This, too, will change the atomic number and usually results in the emission of a gamma. TYPE MASS

(amu) CHARGE PENETRATING

ABILITY RELATIVE DAMAGE

SHIELDING

alpha 4 +2 very low 20 skin, paper beta ~0.005 ±1 low 1 clothing, plastic gamma 0 0 high 1 lead, water

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Methods of Radiation Exposure Control There are three basic methods for reducing the radiation dose that is received; time, distance, and shielding. Time is largely self‐explanatory. The less time that is spent in the vicinity of a radiation source, the less exposure will be received. Methods of reducing the amount of time that is spent in a radiation field include rapid transit through the areas of highest radiation levels, preplanning of any activities that are to take place in the radiation field, practicing on mockups of a work area prior to the work to be done in order to improve familiarity with the procedure, and thorough training prior to performing any work. Distance is another factor that can be used in reducing radiation dose. The intensity of a radiation source, for the most part, falls off as the square of the distance from that source. Therefore, if you double your distance you will reduce you exposure by a factor of four. Ways to utilize distance include familiarity with the work area, allowing for dose reduction by skirting the higher radiation levels and ʺhot spotsʺ, working as far as possible from the hottest areas, or removing items from the highest radiation areas (if possible) for work in an area with lower radiation levels. Shielding is the final major method used to reduce the received dose. Any material will provide some amount of shielding. The more material that is between you and the source of the radiation, the lower will be the dose that you receive. Denser material provides better shielding than lighter material does for beta and gamma radiation. Neutron is best shielded by hydrogenous material, and alpha radiation can be shielded by virtually anything. Methods of utilizing shielding include hanging temporary shielding around any hot spots (although this may cause more radiation exposure to the installers than is saved by the workers), and making use of installed equipment such as pumps, walls, water‐filled pipes, tanks, and so on to reduce the intensity of the radiation field. ALARA is a term that forms the basis for most radiation safety practices. It stands for keeping your radiation dose As Low As Reasonably Achievable. Utilization of time, distance, and shielding is a major part of ALARA. The other major part is ensuring that each trip into a radiation area is really necessary. ALARA also applies to institutions, as well as people. Institutions are required to maintain their cumulative radiation exposure as low as reasonable achievable. Therefore, it is in the best interests of everyone to do what they can to assist in this goal.

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Physical information for selected radionuclides Nuclide half‐life decay mode

(energy)+ Γ * (R/hr)/Ci

formation method

Uses

3H 12.27 y β‐ (18 KeV) N/A n capture, cosmogenic

research

14C 5930 y β‐ (156 KeV) N/A cosmogenic research 60Co 5.27 y 2γ (1.17,1.33

MeV) 1.37 n capture radiation

therapy 131I 8.0 d β‐(606 KeV)

γ (364 KeV) 0.283 fission

fragment thyroid treatments

40K 1.28x109 y β‐(1.31 MeV) γ (1.46 MeV)

0.0817 primordial NORM

geologic dating

90Sr 28.6 yrs β‐(546 KeV) γ (480 KeV)

0.487 fission fragment

instrument check source

99mTc 6.0 hrs γ (141 KeV) 0.123 99Mo decay (fiss. Frag.)

medical

137Cs 30.17 yrs β‐ (512 KeV) γ (662 KeV)

0.382 fission fragment

soil density gages

222Rn 3.8 days α(5.49 MeV) γ (512 KeV)

2.73x10‐4 238U decay series

patent medicines

226Ra 1600 yrs α(4.78 MeV) γ (186 KeV)

0.0121 238U decay series

luminous products

232Th 1.4x1010 yrs α(4.01 MeV) γ (12 KeV)


gas lantern mantles

235U 7.04x108 yrs α(4.40 MeV) γ (186 KeV)


nuclear reactor fuel,

238U 4.47x109 yrs α(4.20 MeV) γ (13 KeV)


military uses (armor, shells)

238Pu 87.8 yrs α(5.50 MeV) γ (13.6 KeV)

0.0790 n capture pacemakers, RTGs

239Pu 24,100 yrs α(5.16 MeV) γ (13.6 KeV)

0.0301 n capture nuclear weapons

241Am 432.2 yrs α(5.49 MeV) γ (13.9 KeV)

0.314 n capture smoke detectors

208Tl 3.05 min β (1.79 MeV) γ (2.6 MeV)

1.70 232Th decay series

none

+ ‐ beta energies given are maximum decay energy, alpha and gamma energies are for the most probable decay energy * ‐ given gamma constant reflects radiation dose in air as distance of one meter

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Equations and calculations The Law of Radioactive Decay

N N et ot= × −λ

No and Nt are the number of atoms of a radioactive isotope originally and at any given time (t), λ is the isotope’s decay constant and is equal to the natural logarithm of 2 divided by the half‐life of the isotope, and t is the elapsed time between the two measurements. The Law of Radioactivity

A N= λ

This can be combined with the Law of Radioactive Decay to produce the following:

A A et ot= × −λ

Radiation attenuation due to shielding

xosh eII ))(( ρμ−×=

Radiation dose from a point source

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12 rrDD ×=

where D1 and D2 are the radiation dose at distances r1 and r2, respectively. Radiation dose from a line source

DDr

Lr

Lr2

1

1

1 1 1 2= × +⎛⎝⎜

⎞⎠⎟

− −tan tan

where L1 and L2 are the distance from both ends of the line source to a perpendicular line extending to the measuring location.

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Radiation dose from a disk (or plane) source

D Dh r

h2 1

2 2

2= ×+⎛

⎝⎜

⎞⎠⎟ln

where r is the radius of the disk and h is the distance from the center of the disk. For an irregularly‐shaped area an effective radius is calculated by determining the area of the source, dividing by three, and calculating the square root. Radiation dose in air from radioactive material

2rAD Γ

=

where A is the source activity in Bq and D is the dose rate in mSv/hr

Dose conversion factors and risk factors for selected radionuclides Nuclide

Ingestion Dose Conversion Factor* (mrem/μCi)

Inhalation Dose Conversion Factor+ (mrem/μCi)

Risk Factor (μCi‐1)

3H 0.064 (whole body**) 0.064 (whole body) 3.84 x 10‐8 14C 2.09 (whole body) 2.09 (whole body) 1.25 x 10‐6 60Co 26.2 (lower large intestinal wall) 208 (lung) 1.57 x 10‐5 131I 88.5 (thyroid) 54.5 (thyroid) 1.43 x 10‐5 40K 18.6 (stomach wall) 12.3 (lungs) 7.46 x 10‐6 90Sr 115 (bone surface) 1300 (lungs) 9.13 x 10‐4 99mTc 7.41 x 10‐2 (thyroid) 3.02 x 10 ‐2 (lungs) 1.98 x 10‐8 137Cs 49.8 (whole body) 31.7 (whole body) 1.91 x 10‐5 226Ra 831 (bone surface) 8020 (lungs, bone surface) 5.50 x 10‐3 232Th 1370 (bone surface, red marrow) 7.83 x 105 (lungs, red marrow) 6.32 x 10‐4 235U 101 (bone surface, kidneys) 1.23 x 105 (lungs) 5.43 x 10‐5 238U 95.5 (bone surface, kidneys) 1.18 x 105 (lungs) 5.07 x 10‐5 238Pu 1890 (liver, bone surface) 2.28 x 105 (lungs, bone, liver) 8.02 x 10‐4 239Pu 2080 (bone surface, liver) 2.40 x 105 (lungs, bone, liver) 8.85 x 10‐4 241Am 2140 (liver, bone surface) 2.62 x 105 (liver, bone, gonads) 9.12 x 10‐4 * This assumes maximum transfer fraction from gastro‐intestinal tract to blood + This assumes lung residence time of years (most conservative estimate) ** The organ(s) named are the critical organs (the organs receiving the highest dose) The dose conversion factor given is the dose to the whole body from nuclide uptake

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Radiation units and conversion factors A number of units are used in the radiation professions. These are used to measure rates of radioactive decay, the amount of radiation absorbed by an object, and the biological damage caused by exposure to radiation. Since the US is not yet on the metric (SI) system, there are at least two sets of values for each unit, and there are several obsolete units that, officially, are not used but that still appear. This sheet will give a brief description of these units, exact conversion factors from one to the other, and rough conversion factors (for initial estimates). Prefixes: Both US and SI units use multipliers in front of units. For example, km stands for kilometer, where the “kilo” (or “k”) means 1000. A kg is 1000 grams, a cm is 0.01 meters, and so forth. These prefixes and the amount of multiplication or division they represent is: T Tera multiply by 1 trillion (a thousand billion) G Giga multiply by 1 billion M Mega multiply by 1 million k kilo multiply by 1000 c centi divide by 100 m milli divide by 1000 μ micro divide by 1 million n nano divide by 1 billion p pico divide by 1 trillion (a thousand billion) Radioactivity Radioactivity is a measure of the rate at which atoms decay by emitting radiation. It is NOT a measure of weight or mass – three tons of depleted uranium has the same decay rate (the same level of radioactivity) as one gram of radium‐226. Radioactivity is measured in terms of disintegrations per minute. Note: if you are using a radiation meter, you will almost always measure fewer counts per minute than there are disintegrations per minute. This is because virtually all detectors miss some of the radiation given off, so the count rate (what the instrument “sees”) is less than what the material emits. For example, if you are spraying somebody with a hose, only a part of the water coming out of the hose nozzle will hit that person).

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SI: 1 Becquerel (Bq) gives a disintegration rate of 1 disintegration per second (dps) US: 1 Curie (Ci) gives a disintegration rate of 37 billion dps. So 1 Ci = 37 billion Bq

(37 GBq). Roughly speaking, 1 Bq ≅ 30 pCi 1 MBq ≅ 30 μCi 1 GBq ≅ 30 mCi Radiation dose (and dose rate) Radiation dose measures the amount of energy deposited in an object by ionizing radiation. This is important because this energy deposition is what can cause DNA damage that may be harmful. We speak of radiation dose in air, water, human tissue, and many other objects – as long as an object is absorbing energy, it is receiving radiation dose. One of the first units of radiation exposure is the Roentgen (R), which measures how much electric charge is generated in air by ionizing radiation. The Roentgen is now considered an obsolete unit and isn’t used much anymore, although you can still find references to it. 1 R = exposure to that amount of radiation in dry air at standard temperature and pressure that generates an electrical charge of 2.58 * 10‐4 Coulomb/kg SI: 1 Gray (Gy) = deposition of 1 Joule of energy per kilogram of absorber US: 1 rad (r) = deposition of 100 ergs of energy per gram of absorber Conversions: 1 Gy = 100 rad 1 mGy = 0.1 rad

1 rad = 0.01 Gy 1 mr = 10 μGy 1 R = 0.87 rad (in water or tissue)

Dose equivalent: Some kinds of radiation are inherently more damaging than others because of the physical properties of the radiation. For example, alpha particles are heavy and have a relatively high electrical charge and cause more extensive damage to DNA than beta particles do, even with the same amount of energy deposition. So depositing 100 ergs

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per gram of energy from alpha particles is more damaging than depositing 100 ergs per gram of energy from beta particles. Because of this, each type of radiation has what is called a “quality factor” (also called “relative biological effectiveness) that ranges from 1 to 20 or more. Multiplying the absorbed dose (see above) by the quality factor for a given radiation will give the equivalent dose in units of rem or Sieverts (Sv). Or, mathematically, Sv = Gy x QF. Radiation type RBE (or QF) Gamma 1 x‐ray 1 beta 1 alpha 20 neutrons 5‐20 Conversions: 1 Sv = 100 rem 1 rem = 0.01 Sv 1 μSv = 0.1 mrem Dose rate Radiation dose rate is a measure of how quickly radiation is depositing radiation in air, water, our bodies, or any other absorber. Dose rate can be measured in terms of absorbed dose or dose equivalent and in whatever units of time are most convenient. So, for example, some measure dose rate in mrem per hour and others in terms of mGy per year. Since the units of time are the same in SI and US units, it’s only necessary to convert the units of dose mentioned above. A radiation dose rate of 1 mrem per hour (1 mrem/hr) will give a person a radiation dose of 1 mrem in an hour. Since there are a total of 2000 hours in a working year (50 working weeks of 40 hours per week), working every day in a radiation field of 1 mrem/hr will give an annual dose of 2000 mrem (or 2 rem/yr or 20 mSv/yr). There are also about 8766 hours in a calendar year, so living in a background radiation field of about 15 μrem/hr will give you a dose of 131 mrem/yr from background radiation. Conversion factors: 1 mSv/hr = 100 mrem/hr = 0.1 rem/hr 1 nSv/hr = 8.8 μSv/yr = 0.88 mrem/yr (roughly speaking, 1 nSv/hr ≅ 1 mrem/yr) 1 rem/hr = 10 mSv/hr

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Dose conversion factors and risk factors for selected radionuclides Nuclide Ingestion Dose Conversion

Factor* (mrem/μCi) Inhalation Dose Conversion Factor+ (mrem/μCi)

3H 0.064 (whole body**) 0.064 (whole body) 14C 2.09 (whole body) 2.09 (whole body) 60Co 26.2 (lower large intestinal wall) 208 (lung) 131I 88.5 (thyroid) 54.5 (thyroid) 40K 18.6 (stomach wall) 12.3 (lungs) 90Sr 115 (bone surface) 1300 (lungs) 99mTc 7.41 x 10‐2 (thyroid) 3.02 x 10 ‐2 (lungs) 137Cs 49.8 (whole body) 31.7 (whole body) 226Ra 831 (bone surface) 8020 (lungs, bone surface) 232Th 1370 (bone surface, red marrow) 7.83 x 105 (lungs, red marrow) 235U 101 (bone surface, kidneys) 1.23 x 105 (lungs) 238U 95.5 (bone surface, kidneys) 1.18 x 105 (lungs) 238Pu 1890 (liver, bone surface) 2.28 x 105 (lungs, bone, liver) 239Pu 2080 (bone surface, liver) 2.40 x 105 (lungs, bone, liver) 241Am 2140 (liver, bone surface) 2.62 x 105 (liver, bone, gonads) * This assumes maximum transfer fraction from gastro‐intestinal tract to blood + This assumes lung residence time of years (most conservative estimate) ** The organ(s) named are the critical organs (the organs receiving the highest dose) The dose conversion factor given is the dose to the whole body from nuclide uptake

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NATURALLY‐OCCURRING RADIATION We are all exposed to radiation from natural sources on a continuing basis. This natural radiation comes from four primary sources:

1. Radiation from biologically incorporated radionuclides 2. Radioactive materials in geologic materials 3. Cosmic radiation and cosmogenic radionuclides 4. Radon emanating from the ground

On average, we are exposed to nearly 300 mrem/yr from natural sources of radiation, although these values can change greatly from place to place in the US and around the world. Natural radionuclides Naturally‐occurring radionuclides fall into three major categories, those that are primordial, progeny of U and Th, and those that are formed by nuclear reactions in nature. Primordial radionuclides include isotopes of uranium, thorium, and potassium. Uranium and thorium, in turn, give rise to decay series that consist of several shorter‐lived progeny radionuclides. Radionuclides that are formed continually through natural processes include, but are not limited to, 3H, 10Be, 26Al, 14C, and others called cosmogenic radionuclides. Cosmogenic radionuclides form when a high‐energy cosmic ray particle such as a proton or neutron strikes the nucleus of a stable atom and is captured or causes the ejection of neutron(s) and/or proton(s) from the target nuclide. One example of this is the reaction:

14 14N n C p+ → +

where 14N is the target nuclide, n represents a neutron, 14C is the cosmogenic nucleus, and p represents an ejected proton. This is known as an “n‐p reaction” and is abbreviated:

14 14N n p C( , ) Other types of nuclear reactions include (p, n), (n, γ), and (n, α) reactions. Spallation reactions are the primary formation mechanism of 3H, a relatively short‐lived cosmogenic radionuclide (on a geologic time scale). Tritium is formed by the reaction:

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CHnN 12314 +→+

Other cosmogenic nuclides such as 10Be, 26Al, and 36Cl also form via spallation reactions. Radiation from biochemistry Natural potassium contains a small fraction (about 0.01%) of radioactive K‐40, which gives off either a high‐energy beta (1.33 MeV, 89% of the time) or a high‐energy gamma (1.46 MeV, 11% of the time). Potassium is also a vital nutrient, and our cells make use of it for cell signaling and more. This means that we are continually exposed to radiation from internal potassium, which contributes about 35 mrem/yr to our background radiation dose. Radiation dose from internal K‐40 varies according to an organism’s size (in smaller organisms some of the radiation may escape, causing lower radiation dose) and according to potassium concentrations. Our biochemistry also contains hydrogen and carbon. A fraction of these isotopes are also radioactive, created by cosmic ray interactions in the upper atmosphere as described above. We receive another few mrem/yr from cosmogenic radionuclides. This forms the basis of carbon‐14 dating; by measuring the amount of C‐14 in organic material, we can determine the age of the material using the law of radioactive decay; providing we know what the original C‐14 activity was. On average, humans receive about 40 mrem/yr from biologically incorporated radionuclides, and this varies only slightly from person to person. Radiation from geologic materials Potassium is a part of many rocks and minerals and is present in virtually all soils at up to a few percent by weight. Uranium and thorium are also present in most rocks and soils, although in lesser concentrations – usually only a few parts per million. However, U and Th are chemically similar to calcium and similar elements, so they are present in trace quantities in virtually all rocks in the world. As rocks weather and age, they become soil, so the elements present in rock are usually present in the soil derived from that rock. As U and Th decay towards stable Pb‐206 they do so via a series of radioactive isotopes such as radium, radon, and bismuth (to name a few). This means that, in most cases, rocks containing uranium will also contain a number of other radioactive elements, and these can be detected in a careful analysis. Geological and geochemical processes can lead to wide variations in local concentrations of these elements. For example, basaltic rocks may contain only a few

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tenths of a ppm of uranium while uranium ore can contain nearly pure uranium. Similarly, illite clay may contain 7% potassium by weight, while the mineral sylvite (KCl) is over 50% potassium. In general, radioactive elements (K, U, Th) are large ions and, as such, they are concentrated in light‐colored rocks such as granites. Owing to uranium’s insolubility in anoxic conditions, U is also concentrated in rocks that formed in the absence of oxygen, such as coals and dark shales. The net result is that radiation dose from geologic materials can vary widely throughout the world, from about 10 mrem/yr in some areas to over 20 rem/yr in others3. Perhaps the most extreme example of elevated natural radiation from geologic materials was found 2 billion years ago in what is now Oklo, Gabon (in western Africa). Local geochemistry led to very rich uranium mineralization that, in time, caused a self‐sustaining nuclear chain reaction – a natural nuclear reactor. The Oklo reactor operated off and on for at least a million years before being exhausted, and radiation levels in the “core” probably reached tens or hundreds of rem per hour when it was in operation. Appendix C contains more information about the Oklo reactor and its operating characteristics. These latter cases are extreme, however. In general, we are exposed to an average dose of 28 mrem/yr from geologic soures in the US and Canada. Radon One of the decay products of U‐238 is radon‐222, a radioactive noble gas. Radon emanates from the ground where it can be breathed in. Radon also decays to alpha‐emitting progeny nuclides, so inhaling a single radon atom can lead, ultimately to up to 4 alpha decays in the lungs before reaching stability. Since radon comes from the decay of geologic uranium, radon concentrations will vary according to local geology. In general, radon dose will be higher in areas that are underground, poorly‐ventilated, and in areas with high levels of uranium in the rocks. On average, we receive about 200 mrem/yr from radon inhalation. Cosmic radiation The fourth source of natural radiation comes from outer space; cosmic radiation from the sun and the galaxy. The sun produces energy by fusing hydrogen in its core. This

3 Ramsar Iran is the most extreme example of high radiation levels. In Ramsar, radium is brought to the surface from subterranean igneous rocks and is concentrated into freshwater limestone, where it substitutes for calcium. Homes are built from this rock, and some homes have radiation levels of up to 14 mrem/hr on contact with walls. General area radiation levels in parts of Ramsar were measured by me to be over 2 mrem/hr both in homes and in open areas outside.

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reaction is very close to what happens in a hydrogen bomb, but on a hugely larger scale. The amount of energy produced by hydrogen fusion in the sun’s core heats the sun and drives solar activity as it makes its way to the sun’s surface and into space. The surface of the sun is much cooler than the core, but it is still tremendously hot by our standards – about 6000 degrees centigrade. Partly because of this high temperature, gas from the sun is continually driven off into space; we call this the solar wind. The solar wind is simply protons (the nuclei of hydrogen atoms), neutrons, electrons, and alpha particles (the nuclei of helium atoms) – the same materials the sun is made of – that travel through space at velocities of hundreds of kilometers per second, and some of this solar wind reaches the Earth as one component of cosmic radiation (the rest of cosmic radiation comes from outside the solar system and is called galactic cosmic rays, or GCRs). The Earth’s magnetic field helps to shield us from much of the sun’s radiation, funneling the electrons, protons, and alpha particle into radiation belts that surround our planet instead of letting them reach the ground. Our atmosphere also provides fairly substantial shielding, against the cosmic rays that penetrate the magnetic field. In addition to hot gas, the sun’s surface is also penetrated by magnetic fields, as is the Earth’s surface (although the Earth’s magnetic field is not nearly as strong). In places, the solar magnetic field is stronger than in others, and in some of these places, the magnetic field lines of force can become twisted, which is a way of storing energy. Think of a rubber band‐driven airplane; we can hold the airplane steady while twisting a rubber band around and around. As we twist the rubber band, we are storing energy, and when we let go of the propeller, that stored energy causes the propeller to turn quickly enough to let the plane take off. In an analogous manner, the twisted solar magnetic fields store energy and, when that energy is released, it sprays huge amounts of hot gas into space. This is a solar flare. The particles emitted during a solar flare are high‐energy hydrogen and helium atoms that have had all of their electrons removed. But alpha radiation is high‐energy helium nuclei, and hydrogen nuclei are simply protons. Both of these are forms of radiation, as are the electrons (beta radiation) and neutrons found within the gas. In other words, a solar flare is a huge emission of radiation from the sun, and if it’s aimed at the Earth, we will experience higher levels of radiation because of this. Not all of the radiation emitted by a solar flare will reach the earth – some will be dissipated by the interplanetary magnetic field, some will simply miss the Earth entirely, some will be deflected or captured by the Earth’s magnetic field, and some will be absorbed by our atmosphere. Solar flares that can be measured at the Earth’s surface

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are very rare, but it is more common to be able to measure solar flares at the altitudes at which commercial aircraft fly. Even given that, the increase in radiation levels to aircraft crew and passengers is not very large. For example, during normal solar weather, one can expect to receive a radiation dose of about 71 micro‐Sieverts (about 7.1 mrem) flying from the Eastern US to Australia and about 85 micro‐Sieverts flying to Japan (the Japanese flight is further north, where the Earth’s magnetic field is weaker). This level of radiation exposure is about the same as a week’s worth of natural background radiation and it just isn’t enough to cause any harm to passengers or flight crew. According to the National Oceanographic and Atmospheric Administration, radiation dose from solar flares can reach as high as 200 micro Sieverts per hour (20 mrem/hr) for up to a few hours at commercial aircraft altitudes. This would give a radiation dose of up to 40‐60 mrem (400‐600 micro Sieverts) during a 2‐3 hour solar flare. Although this is a higher radiation dose than is normally experienced, such solar flares are expected to occur only a few times during the 11‐year solar cycle and the great majority of passengers simply won’t be in the air when one occurs. Even for those passengers who are exposed to this level of radiation, however, the expected effects still are not significant – this level of radiation exposure is similar to receiving an x‐ray and is far less than what you get in a CT scan or via fluoroscopy. On average, we receive about 27 mrem/yr from cosmic radiation. This increases slightly towards the magnetic poles and with increasing altitude.

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THE BIOLOGICAL EFFECTS OF IONIZING RADIATION Many of us are concerned about the effects of exposure to ionizing radiation. In particular, we worry about developing cancer over the long term, or about radiation burns and radiation sickness over the short term. However, most of us are simply not aware of the risk associated with various levels of exposure. This lack of information can cause undue worry in some cases, or undue lack of concern in others. There are two distinct types of radiation exposure, acute and chronic, and two primary exposure modes, radiation and radioactive contamination. Each exposure type and mode is slightly different and must be treated differently.

Introduction Radiation is ubiquitous; an inescapable part of life on Earth. Background radiation reaches us from outer space, from the rocks and soils we walk on, and from naturally radioactive potassium in our own bodies. Through its entire history, organisms on Earth have been bombarded by radiation, and this will continue for as long as the Earth exists. Today, the average person in the US is exposed to about 360 mrem each year from background radiation – about 1 mrem a day – and this level of radiation exposure seems to have no ill effects. Of the estimated 600 or so mutations that occur in each of our cells each year (about 900 in those cells exposed to UV radiation), only about 5 are due to the effects of background radiation. In short, environmental radiation is a mutagen, but it is not a major source of DNA damage. At higher levels, however, radiation can cause damage, and that is the subject of this section. Continual exposure to low levels of radiation may cause a mutation that can initiate cancer. Brief exposure to high levels of radiation can cause skin burns, radiation sickness, or a number of radiation‐induced syndromes.

Radiation Damage to Cells Radiation can damage cells by directly striking the DNA and causing damage such as single‐ or double‐strand breaks or point mutations. It’s more likely, however, that the radiation will interact with molecules in the cytoplasm, splitting them apart and forming reactive molecules called free radicals. These free radicals, then, go on to cause DNA damage. Free radicals are caused by more than just radiation – our mitochondria

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leak free radicals all the time, metabolizing our food can create free radicals, and even dissolved oxygen in our cells can cause DNA damage. All of this damage is indistinguishable, with the exception of double‐strand DNA breaks – we can’t “look” at a point mutation and tell if it was caused by radiation or mitochondrial free radicals. When radiation passes through a cell the effects can range from non‐existent to profound. There’s a chance, for example, that a gamma ray will pass right through a cell without interacting at all or that the free radicals produced will simply recombine or be scavenged before they can reach the DNA. If radiation (or the free radicals it produces) do interact with the DNA, there are only a few possibilities – either the DNA will be damaged or it won’t. If the DNA is damaged, we have a few further possibilities – the damage may be beneficial, harmful, or neutral (neutral damage is damage that has no effect on the cell – it may be in non‐coding part of the DNA, or to a gene that’s not expressed in that particular cell, for example). Of the harmful damage, it may be either lethal to the cell or sublethal. At this point, the only DNA damage that concerns us is sublethal damage to the DNA in a part of the genome that may be harmful. Although lethal damage is bad for the cell, at least the damage does not get passed on to daughter cells, so a lethally damaged call cannot go on to cause cancer. However, the possibilities do not stop here, because our cells have DNA damage repair mechanisms. Although these mechanisms are very effective, they are not perfect. This means that any bit of DNA damage may be repaired properly, may be repaired improperly, or might not be repaired at all. It is at this point that DNA damage may become a mutation – a mutation is what happens when damage to our DNA becomes “fixed” and is able to be passed on to the next generation of cells. As with DNA damage, mutations may be good, bad, or indifferent (neutral), and the detrimental mutations may be lethal or sublethal. And, as before, it is only the sublethal damage that’s of interest to us, and then, only if it can cause the cell to become cancerous. It has taken three paragraphs to describe the different possibilities of radiation interacting in a cell. Part of this is for the sake of completeness, but it’s also to help drive home an important point – radiation is a weak carcinogen. If we sum up all the possibilities above, there are over 20 different possibilities. Of these, only 1 (sublethal damage that is misrepaired or unrepaired and causes a cell to become carcinogenic) has a chance of causing cancer. Radiation is a carcinogen, but it’s not a very good one – not compared to many of the chemicals we work with. In the next few sections, we will look a little more about the effects of both acute and chronic radiation exposure on the organism, instead of the individual cells.

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Acute exposure Exposing the whole body to very high levels of radiation in a short period of time can be harmful or fatal. Exposing parts of the whole body to very high radiation levels can also cause harm, but is usually not life threatening. Acute radiation injury has been noted in the survivors of the Japanese atomic bombings, among surviving Chernobyl workers, in the wake of nuclear criticality accidents, and among people who have found lost radioactive sources with high levels of activity. Acute radiation injury to limited parts of the body has also been noted, for example, in patients receiving excessive fluoroscopy, mineralogists misusing x‐ray diffraction equipment, industrial employees using linear accelerators, and radiation oncology patients. In an industrial setting, it is most likely that workers will receive radiation burns on their hands from handling radioactive sources or from putting their hands into a radiation beam. Many industries use linear accelerators, and radiation levels inside can be several million rads per hour. X‐ray diffraction devices can give dose rates of thousands or tens of thousands of rads per minute, and large radiation sources (say, for radiography or food/mail irradiation) can give dose rates of thousands of rads per hour or higher, depending on many variables. In other industries, such as the food irradiation industry, workers have ignored or over‐ridden safety devices and received harmful or fatal whole‐body radiation doses by entering irradiation chambers with a radiation source exposed. Sunburn is a mild form of acute exposure to radiation, but it serves as a starting point to acute radiation injury. At a skin dose of a few hundred rem, the patient will exhibit erythema and, at higher doses, blistering and peeling (dry and moist desquamation). Depending on the characteristics of the exposure, one side of the body may be more affected – typically the side facing the radiation source. Very high radiation doses to parts of the body will produce these same symptoms to limited parts of the body. Some victims may exhibit symptoms of both limited and whole‐body radiation exposure. These are typically people who come across abandoned radioactive sources and carry them home. Other effects of acute whole‐body radiation exposure can include depilation (hair loss), nausea, and a variety of radiation syndromes.

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Radiation Syndromes

Prodromal syndrome In some cases, radiation effects may appear within a few hours of radiation exposure and will persist for up to a few days. In general, higher doses result in earlier and more severe symptoms. At lower levels of exposure, symptoms may include fatigue, nausea, and vomiting. At higher (and probably lethal) exposure levels, patients will also experience fever, diarrhea, and hypotension. Patients with prodromal syndrome have likely been exposed to at least 100 rem, but symptoms will appear at any higher level of exposure. Patients exhibiting symptoms within 30 minutes of exposure have likely received a lethal dose of radiation, as have patients experiencing immediate diarrhea.

Hematopoietic syndrome The blood‐forming organs are among the most sensitive to the effects of radiation, so these organs are among the first to show the results of high radiation exposure. Hematopoietic syndrome begins to appear at doses of from 300 to 800 rem, when the precursor cells are sterilized or killed. This leads to a reduction in blood cell counts as older cells die and are not replaced, and it leaves the patient open to infection and other related problems. Following the initial prodromal syndrome, a patient may be relatively free of symptoms for some time, although a great deal is occurring. Patients with lower levels of exposure may recover from their exposure if the bone marrow can regenerate and if the patient receives medical support (typically antibiotic treatment). At higher levels of exposure, the patient will begin to exhibit chills, fatigue, hair loss, petechia, and ulceration of the mouth as well as infection, bleeding, immune system depression, and other symptoms resulting from the loss of blood cells. A dose of about 300 to 400 rem is lethal to 50% of the exposed population without medical support. This is called the LD50 dose. With medical support, the LD50 dose is about 700 to 800 rem. Treatment for patients suffering from hematopoietic syndromes includes replacing blood cells via transfusion, isolation from sources of infection, and antibiotic treatment.

Gastrointestinal syndrome Patients exposed to 1000 rem (10 Gy) or more will experience gastrointestinal syndrome and, most likely, death within 3‐10 days of exposure. Radiation exposures in this range sterilizes dividing crypt cells, leading to loss of cells from the villi. Within a few days, the villi become almost totally flat as the outer surface sloughs off and is not replaced. In one particular case (a man exposed to between 1100 and 2000 rem in 1946) the patient remained in relatively good condition for nearly a week, at which time he began

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suffering bloody diarrhea, circulatory collapse, and severe damage to the epithelial surfaces throughout the intestinal tract. Treatment for patients suffering from gastrointestinal syndrome include antiemetics, sedatives, a bland diet, and fluid replacement. Antibiotic treatment and blood transfusions are sometimes helpful in keeping patients alive through the first few days or weeks.

Cerebrovascular syndrome Exposure to exceptionally high doses of radiation (in excess of 10,000 rem or 100 Gy) will result in damage to the central nervous system, normally among the most radiation‐resistant parts of the body. Cebrovascular syndrome is accompanied by symptoms of all other radiation syndromes, and it usually results in death within several hours to a few days of exposure. Patients exposed to such high levels of radiation will experience almost immediate nausea, vomiting, disorientation, seizures, and other symptoms of neurological distress, followed by coma and death. Although the exact cause of death is not known, it is thought that part of the cause is the buildup of cranial pressure due to leakage of fluid from blood vessels. Treatment for cerbrovascular syndrome is limited to providing pain relief and sedatives to control convulsions and anxiety because the syndrome is invariably.

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Effects of acute radiation exposure

Dose (REM)

Syndrome or effect Comments

~5 Chromosome changes

Increase in dicentric chromosomes and chromosome fragments noted

15‐25 Blood cell changes

Begin to see depression in numbers of red and white blood cells

100 Radiation sickness

Mild at lower doses, severity and rapidity of onset increases rapidly with increasing dose

300‐800 Hematopoeitic syndrome

Changes in blood cell count due to damage to crypt cells, severe radiation sickness, recovery possible with medical support

400 LD50 With medical treatment, LD50 is about 800 rem 1000 GI syndrome,

LD100

Relatively rapid onset for vomiting

10,000 Cerebrovascular syndrome

Rapid incapacitation, death within a few days

Chronic exposure The primary concern with chronic exposure to relatively low levels of radiation is that we will develop cancer. There are two primary competing hypotheses on this matter, and the matter is still far from being settled.

LNT The linear, no‐threshold (LNT) hypothesis suggests that all radiation exposure is potentially harmful (the “no‐threshold” part), and that the risk of getting cancer from radiation is directly proportional to the dose received (the “linear” part). LNT is the most conservative radiation dose‐response model in that it predicts the highest risk from a given amount of radiation exposure. This is one of the reasons that the LNT is the foundation of radiation regulations virtually everywhere in the world – since we really aren’t sure how we respond to low levels of radiation exposure, it makes sense to control dose (and risk) according to the most conservative model. One problem with the LNT is that it can be used to predict cancer risks down to vanishingly small levels of exposure, and so it has been used to calculate expected cancer rates from exposure to radon, “dirty bombs”, and medical x‐rays. For example, say that the risk of getting cancer from a given radiation exposure is 5 additional cancer

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deaths for every 10,000 person‐rem. That means that exposing 10,000 people to 1 rem each should result in an extra 5 cancer deaths among those people. Or, exposing 1 million people to 10 mrem each should also lead to 5 added cancer deaths. It’s easy to see that we can use this model to predict added cancer deaths from any level of radiation exposure, no matter how trivial, if enough people are exposed. By analogy, we can also say that, since a 1000 kg rock will crush someone, throwing a million one‐gram rocks at a million different people will crush someone. This doesn’t make much sense, and both the Health Physics Society and the International Commission for Radiation Protection have advised against this misuse of the LNT model. In fact, we just don’t know what happens at such low levels of exposure, and we can’t make any such predictions for very small levels of exposure. According to the Health Physics Society, in two separate position papers (which can be found on the HPS web page at www.hps.org), we simply can’t calculate a numerical risk estimate from any exposure of less than 10 rem, so even the first calculation runs afoul of HPS recommendations. In a similar vein, the ICRP has suggested that, when looking at the risk from collective dose, if the most highly exposed individual receives a trivial dose, then everyone’s dose should be treated as trivial. In recent years there have been two well‐publicized transgressions of HPS and ICRP recommendations – a paper about pediatric CT exposures in the summer of 2001 suggested a rise in childhood cancers from using adult settings for children, and a study in early 2002 on the effects of a “dirty bomb” detonation in Manhattan came complete with colored contour lines showing the 1% and 0.1% cancer death rates from the radiation exposure. Unfortunately, both of these studies ignored the recommendations of HPS and ICRP, and it’s likely that both studies greatly over‐stated the risks. In fact, about all we can say is that these projections are the highest we are likely to see, but that there may actually be no additional risk. A description of alternate, and no less probable, dose response models is contained in the following section.

Threshold/Hormesis models Virtually all harmful substances exhibit some level below which there are no apparent harmful effects. This is part of the idea behind the No Observable Adverse Effects Level (NOAEL) – below a threshold dose you simply don’t see any effects from exposure to a substance. There are those who feel that radiation probably behaves similarly – that there is a level of exposure below which there are observable effects from radiation exposure. Threshold effects are fairly easy to imagine; examples of hormesis include selenium, aspirin, red wine, and many vitamins.

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In this figure, the area below the diagonal line is the area in which current epidemiological tools are unable to detect the effects of radiation exposure due to the high background incidence of cancer (Gonzalez, 2004). There are also those who think that exposure to low levels of radiation may be beneficial. This is called hormesis and, although it sounds implausible at first blush, there are plenty of examples of hormesis in the world. Two examples are vitamin D and selenium. Both of these substances are vital nutrients, and both are acutely toxic in sufficiently high doses. Low doses of aspirin can help to stave off heart disease (not to mention the beneficial effects on fever, pain, and inflammation), yet high doses of aspirin can be fatal, and people can also die of excessive salt intake or even water intoxication. In short, the idea of hormesis is not outlandish; only the application of hormesis to radiation exposure seems unusual because we are all so steeped in the idea that radiation is uniformly bad.

The number of participants required to show the effects of various radiation exposures.

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The idea behind assuming a threshold in our response to radiation exposure is that, given the variations in Earth’s background radiation field, it makes sense that our cells should be able to adequately repair DNA damage from slightly elevated levels of radiation. And, let’s face it; radiation is not one of the major environmental mutagens (it accounts for about 1%‐5% of background DNA damage). Our biochemistry contains very effective mechanisms for repairing DNA damage, and it is thought that these mechanisms are able to accommodate some level of added damage, such as would result from exposure to low levels of radiation. The thinking behind positing hormesis effects is that, by presenting a continuing challenge to our mutation repair and tumor suppression mechanisms, they are kept at peak operating efficiency. They are better able to contend with the ordinary, garden‐variety damage that is always cropping up in our genome and, as such, our DNA is better protected than if this radiation exposure was removed. The best way to test these hypotheses, of course, is to perform epidemiological studies of exposed populations, and many such studies have been performed with equivocal results. Researchers have looked at radiation workers, residents of natural high‐background areas, radon concentrations versus lung cancer rates, radiologists, and atomic bomb survivors, among others. Some studies show that risks are slightly higher, some show no effects at all, and some show fewer cancers than expected in the study populations. Part of the problem is that the effects are often smaller than the error bars, and this makes it very difficult to pick out what is actually happening. Unfortunately, there is not yet a “gold‐plated” study that everyone can point to and agree that it was properly done, controlled for all confounding factors, and shows a significant result. But the search continues! Given this degree of uncertainty, many health physicists and most governments feel it is best to control radiation exposure under the risks of the highest‐risk model, LNT. The thinking is that, if we maintain risks at a low and acceptable level under LNT, then whichever model is correct, we will be at no more risk than we have agreed we can accept. The only problem with this model is that, if one of the other models better represents reality, we will have spent a lot of time, effort, and money controlling illusory risks, and these resources will have been taken away from more effective risk‐reduction measures. So this question needs to be answered, and we will hopefully be able to do so before too much longer.

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Fetal exposure and reproductive effects of radiation exposure Pregnant women are sometimes exposed to radiation occupationally or through medical procedures. Although even receiving a single x‐ray can cause worry in an expectant mother, the amount of radiation required to cause harm to a developing fetus is actually rather high. In fact, the medical advice is to take no actions for any fetal radiation dose of less than 5 rem and, in some cases, a dose of up to 15 rem to the fetus poses little risk of causing birth defects or other problems. This is not to say that fetal radiation exposure up to this level is acceptable; just that the risks from the radiation exposure are less than the other risks that accompany any pregnancy. I should also note that there seems to be a threshold for radiation‐induced birth defects at about 5 rem fetal dose – there is no evidence that fetal exposures of less than this level can cause birth defects. In general, there are three distinct phases of radiation sensitivity the fetus goes through. During the first two weeks post‐conception, sometimes called the “all or nothing” phase, radiation exposure will either result in a miscarriage or the pregnancy will go forward with no ill effects at all from exposure to radiation. From the second to the 15th week, the organs and tissues are differentiating and growing, and the fetus is much more sensitive to radiation. During this period, if other risk factors are present, the medical advice is to consider terminating the pregnancy for doses greater than 5 rem. In the absence of other risk factors, however, this may not be called for. The woman’s physician will have to weigh all the risk factors present and recommend a course of action to the woman. During this stage of pregnancy, the primary effects of excessive radiation exposure are low birth rate, low organ weight, mental retardation, or microcephaly (small head size). All of these effects can be caused by other teratogens. Once past the 15th week, the fetus is more resistant to the effects of radiation, although damage can still be caused by sufficiently high radiation dose. This knowledge is the result of studying pregnant women receiving diagnostic or therapeutic medical radiation, Japanese women who were pregnant at the time of the atomic bombings in Japan, and pregnant radiation workers over more than a half century. It is also worth noting that there have been no documented fetal health effects from pre‐conception radiation exposure. Exposure to high levels of radiation may cause temporary or permanent sterility, but it does not seem to cause birth defects. Embryos that form from damaged sperm or ova seem to either fail to implant or miscarry quickly.

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Pregnant worker programs In spite of the above, it still makes sense to control fetal radiation exposure to minimize the risk of inadvertently causing harm. To do this, it is necessary to have a formal program in place in your workplace if you have radioactive sources. However, participation in pregnant worker programs must be voluntary, and a woman cannot be forced to participate in such a program. To participate in a pregnant worker program, the woman must declare her pregnancy in writing – in most places this is done by completing and signing an appropriate form letter. Once this step is taken, the woman is limited to a radiation dose of 500 mrem during the entire duration of pregnancy, and she is not to exceed 50 mrem in any single month. In addition to the lower exposure levels, the RSO should discuss the administrative details of the pregnant worker program with the worker, and may need to issue a fetal dosimeter. This badge should be worn on the abdomen and, if the woman wears a lead apron, it should be worn beneath the badge. Women who only work with low‐energy beta‐emitting isotopes or with alpha emitters will probably not be issued a badge because it is difficult or impossible to monitor for them with standard radiation badges. If it is thought that the woman ingested or inhaled radioactive materials, the RSO will have to perform bioassay measurements to calculate internal and fetal dose. Because women do not immediately know that they are pregnant, and many do not immediately declare their pregnancy, it is possible that the fetus was exposed to some radiation before the mother entered the pregnant worker program. This can be done by reviewing her dosimetry records, if they are available, or by calculating a likely dose based on her work history or by comparison to other workers if she was not previously badged. And, of course, all dosimetry records should be filed and kept for 30 years after the worker leaves the company.

Summary Radiation can be harmful. Large doses in a short time period can cause skin burns, radiation sickness, or death, and large radiation doses to the hands or feet can cause complications that will result in amputation. However, following standard radiation safety precautions can minimize the chance that such accidents will occur. More likely is that radiation workers will be exposed to slightly elevated radiation levels for long periods of time and, in such cases, the data are more ambiguous.

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Under the most conservative (i.e. highest calculated risk) model, the LNT hypothesis, the risks of getting cancer from a lifetime of exposure to 100 mrem/yr are about the same as succumbing to any other occupational illness or injury. However, under other possible scenarios, the risks may be lower, and some feel that exposure to low levels of radiation may be somewhat beneficial. Until more definitive data are available, it makes sense to act as though any radiation exposure is potentially harmful (even if the risk is very low) and to act accordingly. This does not mean avoiding all radiation exposure – it simply means managing radiation exposure the same as we manage exposure to all other harmful agents or situations. Finally, it is possible for pregnant women to continue working with radiation, and below a fetal dose of 5 rem, there seems to be no risk to the fetus. Organizations working with radiation must have pregnant worker policies in place that include voluntary participation, more restrictive dose limits, and some method of monitoring dose. Following these steps should suffice to protect the woman and her baby from the teratogenic effects of radiation although, of course, it is always possible that a baby will be born with birth defects or that a pregnancy will end in miscarriage for reasons not due to radiation exposure.

For more information: Radiation workers or radiation safety officers who have any questions about the health or reproductive effects of radiation should contact a qualified radiation professional. If your company is large enough to warrant a full‐time professional health physicist, this is a good place to start. If not, the Health Physics Society web page (www.hps.org) has a wealth of information. In addition, and perhaps more importantly, the HPS page has an “Ask the Expert” feature. Browsing the list of answered questions may reveal that an answer to your question already exists; if not, anyone may post a question to this site and will receive an answer within a few days.

Hazards Related to Inhalation or Ingestion of Radioactive Materials There are special hazards associated with the inhalation or ingestion (uptake) of radioactive materials. First, introducing radioactive materials into the body increases the radiation dose internal organs such as the stomach, lungs, or intestines are exposed to. In addition, once in the body, radioactive materials may enter the bloodstream and be taken to other internal organs, concentrating there. For example, if radium is ingested, it will concentrate in the bones, exposing the bone surface and marrow to long‐term radiation dose. Alpha radiation cannot normally penetrate the skin to deliver radiation dose but,

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once taken internally, is extremely damaging. For these reasons it is important to minimize the amount of radioactive materials ingested or inhaled to the maximum amount possible. Some examples of collective dose PERSON‐REM SOURCE PER YEAR 4000 Naval nuclear power ‐ ~200 reactors and 10,000 personnel 12,000 air travel 100,000 Denver residents (above natural background) 650,000 cooking with natural gas ‐ US population (radon) 17,000,000 medical and dental x‐rays ‐ US population 20,000,000 natural background ‐ US population

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Radiation doses from other activities DOSE (mR/yr) SOURCE 0.3 ‐ 1 watching picture tube‐type TV for 4 hours daily at 12 feet x‐ray emission from screen 4 reading glossy magazines for 1 hr/day uranium and potassium in clay paper coating 5 eating 1 banana per day potassium (K‐40) in banana 8 carrying radium dial pocket watch 12 hrs/day gamma emission from radium paint 10 living in a brick house instead of a wood one potassium, uranium and thorium in clay in the bricks 25 ‐ 4000 wearing enameled jewelry (with/without metal backing) 10 hrs/week uranium compounds in glazing 70 living in Denver instead of at sea level increased background radiation due to elevation and igneous rocks 100 flying ~5000 miles per month reduced atmospheric radiation shielding 100 ‐ 200 radon gas inhalation (national average) 100 ‐ 200 foods and fertilizers naturally‐occurring potassium (K‐40) and uranium in super‐phosphated fertilizers 150 medical technicians (average) combination x‐ray and nuclear medicine 170 flight crews reduced atmospheric radiation shielding 2000‐5000 dose to lungs ‐ smoking 1 pack of cigarettes/daily polonium (Po‐210) and lead (Pb‐210) from U decay series

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1. 1. Radiation strikes cell

2. Does radiation interact within cell?

3. Does radiation (or radiation products) cause DNA damage?

4. Is the damage repaired properly before the cell divides again?

5. Is mis-repaired damage harmful?

No effect on organism

No

Yes

No

No

Yes

6. Is damage lethal to cell?

Yes

No

No

Yes

No

No

Yes

Yes

8. Will organism live long enough to develop cancer (5 – 20 years post-exposure)?

7. Can damage cause cancer?

Yes

Cancer may develop

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Radiation Detectors and Dosimeters Radiation Detection Equipment There are a number of methods of detecting radiation, most of which rely on radiation’s ability to create ion pairs in irradiated materials. Devices such as Geiger counters and ion chambers detect these ionizations directly and measure the electric current generated by radiation bombardment. Other devices, such as scintillation counters, detect photons that are emitted when a substance is irradiated, amplifying these photons in photo‐multiplier tubes to create a signal. The most commonly‐used of these detectors are described in the following section.

G‐M Tubes G‐M tubes are sealed, gas‐filled tubes containing an anode and a cathode with a large voltage applied across them. The interaction of ionizing radiation with this gas causes secondary ionizations to occur, creating a current spike as the positive ions are attracted to the anode and the negative ions to the cathode. This current is amplified and registers as a count on the meter. The ionized gas will quickly recombine, allowing detection of the next ionizing event. There is a small amount of ʺdead timeʺ while this recombination is taking place, usually in the neighborhood of 1‐2 μsec, during which no counting can take place. This imposes an upper limit on the count rate for which the instrument can theoretically be used. The disadvantage of a G‐M tube is that there is, however briefly, a period during which no counts can be registered. This also allows for the tube to become ʺsaturatedʺ and not register accurately if the count rate is too great. The chief advantages of the G‐M tube are its extreme simplicity, leading to great reliability, and its high sensitivity. Geiger‐Muller tubes are primarily used for the detection of gamma and beta radiation and measure in either counts per minute (for contamination detectors) or mr/hr. Geiger counters are not normally used for measuring radiation dose rates, unless the surveyor knows what isotope is present, because each isotope emits a different energy of gamma radiation – Geiger counters detect only counts and not energy deposition, making them over‐respond to low‐energy radiation, and under‐respond to high‐energy radiation.

Air Ionization Chambers In an air ionization chamber, the ion pairs created in the detector are collected and measured before they can recombine. This gives an output current that is proportional to the strength of the radiation field.

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The advantages of the air ionization chamber are that the output current (the meter reading) is largely independent of the operating voltage and that they are, for the most part, easy to use and very portable. The disadvantages of this instrument are that the generated current is very small, leading to a low sensitivity, and it is very sensitive to changes in humidity and barometric pressure. This means that, if an ion chamber is calibrated under different conditions of air pressure, temperature, or humidity than the conditions under which they are used, these differences can cause the instrument to read erroneously. Ion chambers can measure alpha, beta, or gamma radiation but require a special window in order to measure alpha radiation and measure in REM per hour.

Scintillation Counters Another common means of measuring radioactivity is the scintillation counter. The passage of a beta or alpha particle through a scintillating medium will cause the emission of photons from the scintillant. These photons leave the scintillation medium and interact with one or more photo‐multiplier tubes (P‐M tubes) to register as counts. An advantage of scintillation counters is their relative lack of dead time (usually on the order of tens of nano‐seconds), making their major operation limit the speed at which the electronic components can operate. A disadvantage is that the scintillating medium must be kept dark at all times. Even a tiny pinhole can admit enough light to produce apparent high readings, which can be mistaken for contamination or elevated radiation levels. The scintillation counter is primarily used for the detection of particles such as alpha and beta radiation.

Personal Dosimetry There are several types of personnel dosimetry. Three major types, the Thermo Luminescent Dosimeter (TLD), the optically‐stimulated luminescent (OSL), and the film badge, are used for legal records of individual exposure. The others are used for

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informational purposes only, to inform the wearer of the approximate dose that has been received or of the strength of the field that the worker is in.

TLDʹs TLDʹs consist of small crystals, usually calcium fluoride, containing small amounts of impurities. Incident radiation excites atoms in the crystal which are ʺtrappedʺ by these impurities. Upon heating (thermo‐ ), the ʺtrappedʺ excited electrons fall back to the ground state, giving up a photon (‐luminescent) in the process. The number of photons released is proportional to the total radiation dose received. This photon signal is amplified by a photo‐multiplier tube and the output sent to the dosimeter reader to register the dose. TLDʹs have several advantages. They are extremely rugged and are not adversely affected by extremes in temperature. They are accurate over a very large exposure range ‐ from mR to thousands of R. They are relatively quick and easy to read, allowing for speedy monitoring in emergency situations. They can be reused many times each, saving the expense of replacing them continually. And, finally, they are sensitive to beta, gamma, and x‐ray radiation. The disadvantages of TLDʹs over other methods of dosimetry are their higher initial cost and the fact that, once heated, the dose information is erased, unlike film badges, which can be reread if any questions arise at a later date.

Film Badges Film badges make use of the fact that radiation, like light, will interact with silver halide crystals in film emulsion, causing them to darken. Also like light, the amount of the darkening is related to the total exposure. The film is placed in a holder that can also contain two or three shields of varying thicknesses, allowing measurement of skin dose as well as whole body (deep) dose. The film, once developed, is read by a densitometer to determine the overall dose that the wearer received. The advantages of the film badge are its permanence, its cost, and the ability to allow simultaneous recording of exposure to different types or energies of radiation. The disadvantages of film badges are the amount of time that it takes to develop and process the film, the sensitivity of the film to environmental factors such as temperature

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and humidity, and the fading that can occur if the film is not read promptly. Film badges can be used to measure beta, gamma, and x‐ray dose.

Optically stimulated luminescent dosimeters OSL badges are similar to TLDs in their operating principles, except that they are read by scanning with a laser instead of being heated. The laser adds sufficient energy for the electrons to escape from their traps and, as with TLDs, they emit visible photons when they return to the ground state. OSL badges are as sensitive as TLDs and, unlike TLDs, they can be read out multiple times because the laser can be set to scan only a part of the badge. This gives OSL badges the ability to be archived, as with film, for re‐reading at a later time. Unlike film, OSL badges are not susceptible to environmental factors (at least, not to factors that are not fatal to the wearer). In many ways, these badges combine the best features of both film and TLDs with the exception of being more expensive than film. At present, OSL badges are offered by only one company.

Self‐Reading Dosimeter The self‐reading dosimeter is basically a pocket ion chamber. It consists of two quartz fibers that can have a charge applied to them, acting as an electroscope. The end of one of the fibers is left free and is attached to a small hairline indicator that, as the dosimeter charges or discharges, is pulled across the indicating screen. As ionizations occur within the chamber the net charge on the fibers is lowered, causing them to draw together and pulling the hairline across the scale on the screen. The main advantage of these dosimeters is that they are relatively durable, provide a convenient way for personnel to monitor their dose while working, and can be produced to cover a very wide assortment of ranges. They are also, for the most part, energy‐ independent, allowing the counting of most gamma and x‐ray radiation to which the wearer is exposed. The self‐reading dosimeters have several disadvantages, however. They can be easily discharged by dropping them or banging into something, causing them to go off‐scale high and giving erroneous readings. Their orientation while reading them is also important; they may give very different readings depending on whether they are held right‐side up or upside down due to the effects of gravity on what is primarily a mechanical system. They are also not extremely accurate, their accuracy depending upon

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the scale gradations, the initial zeroing accuracy, and the mechanical and electrical properties of the quartz fibers. All in all, the self‐reading dosimeters are not the best way to measure personal dose for a legal record. However, they are not designed to be. Their purpose is to provide personnel working in a radiation area the ability to roughly monitor their absorbed dose and they serve this purpose very well.

Audible Dosimetry (Rad‐Tads, Chirpers) These are small G‐M tubes that give an audible chirp at intervals that reflect the strength of the radiation field. The faster they are chirping, the stronger the field is. They give no indication of total absorbed dose and are only an indication as to the radiation levels that the individual is in at the present time.

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Responding to radiological accidents and emergencies Introduction Accidents happen. Including accidents involving radiation or radioactivity. In the 25+ years I’ve been working with radiation and radioactivity, I’ve lost count of the number of spills, skin contamination, high radiation levels, and other accidents I’ve had to deal with in some capacity or other. And anyone working with radiation or radioactivity will, at some point, be in a position of having their own radiological incident to contend with. And, of course, we now also have to consider the possibility of deliberate mis‐use of radioactive materials as a part of a terrorist attack. This is a good time to dust off the response plans to see if we are ready to properly address any untoward incidents that might occur. In this article, we will address some general procedures that apply to many radiological incidents as well as some general steps to take in some specific cases. These are guidelines only – every incident is unique. But some general guidelines to generic situations can be applied to the specific situations with suitable modifications. So, in this article, I’ll discuss radioactive spills, skin contamination, traffic accidents, the loss of radioactive materials, and the use of radiological dispersal devices (RDDs, also called “dirty” bombs”). General guidelines In most cases, radiological incidents are not life‐threatening and, in fact, pose little actual physical risk. The radioactivity is a nuisance, a regulatory problem, and a complicating factor, but it is rarely potentially harmful. So the first general rule should be “Don’t panic.” Take the opportunity to think through your actions before springing into action to try to save the day. By so doing, you are less likely to take rash actions that could well end up making things worse rather than better. In general, radiological hazards will pose little risk to people involved in a radiological accident. This means that emergency responders and medical personnel are very unlikely to be at risk from victims or patients, no matter how heavily contaminated they are. This means that victims and patients should be cared for as rapidly as required by their injuries – the badly injured or critical should be cared for immediately, and lightly injured patients may be safely decontaminated or wrapped to contain the contamination so that they do not contaminate an ambulance or hospital emergency room.

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Actions should be taken to try to minimize exposure to people near and responding to the incident. Unless actually involved in incident response, everyone should stay at the greatest distance possible, they should minimize the amount of time they’re in a controlled area, they should try to interpose shielding between themselves and the sources of radiation, and they should try to don appropriate PPE (such as shoe covers, gloves, and coveralls when involved in a spill) if possible. Specific incidents There are over 17,000 radioactive materials licensees in the United States and the vast majority of these licensees are relatively small industrial users who possess either radioactive gauges, small radioactive sources, or relatively minor amounts of unsealed radioactive materials. There are also a large number of small medical licensees, colleges, and radiopharmaceutical vendors. Large organizations, such as nuclear power plants, nuclear fuel cycle facilities, or major research universities constitute only a tiny fraction of all licensees and they are likely to have full‐time health physics staff to handle radiological problems. It is the former group, the relatively small licensees, for whom this article is written and they are likely to experience only a relatively small variety of types of incidents – those noted below. Spill of radioactive material It’s easy to cause a spill – knocking over a small vial of radioactive materials can cause one, as can accidentally ejecting the contents of a pipettor or dropping a sample tube or even just having a drop fall from a beaker or bottle. Radioactive spills cause contamination in the area of the spill, they can lead to the contamination of personnel, and they can result in the spread of contamination to office areas or homes. Minor spills can often be cleaned up fairly easily; major spills can cause problems. There are some actions that can be taken immediately in the event of a spill. The acronym we use at our facility (and what we were taught in the nuclear Navy) was SWIM – Stop the spill, Warn others of the spill, Isolate the area, and Minimize exposure to radiation. It is not necessary to follow these steps in this order, but completing these actions will help to reduce the impact of the event. Stopping the spill is not the same as cleaning it up; it is taking actions to keep the spill from getting worse. If a container fell over, right it (wearing protective gloves, hopefully!) cap or cover it, and place it in a pail or deep tray. Next, try to place absorbent materials over the spilled liquid or, if it is a powder, cover it with dampened wipes or rags to keep it from blowing around. You are not trying to clean up the spill at this point, you are simply trying to limit the amount of spilled material and its extent.

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Warning others may be the first thing that happens – most people make some comment when they cause a spill. You will want to warn others not to walk into the spill area, to ask for help with the cleanup, and anyone nearby who might have been contaminated by the spill should stand fast to keep them from spreading contamination. This should also include contacting the radiation safety officer and other radiation safety staff, plus anyone else on your incident call‐up list. Workers should be able to contact the RSO or a competent alternative at any time in the event of a radiological incident, so the RSO’s pager and/or telephone number(s) should be made available to radiation workers or to Security officers as appropriate in the event of after‐hours spills. Isolate the area involved in the spill. There are several reasons to do this; you want to keep people out of the spill so they don’t get contaminated, you will need room to work on cleaning up the spill, and anyone within the spill boundaries should be considered potentially contaminated. You should use rope or tape or some other physical barrier whenever possible, even if you are isolating an entire room. Simply posting a door sign may not work – many people just don’t read door signs – but they will stop before crossing a rope or tape boundary. At my facility, nobody is permitted to cross a spill boundary to enter an area unless they are wearing gloves, shoe covers, and a lab coat; and nobody is allowed to exit a spill area unless they are surveyed out of the area by Radiation Safety staff. We have had incidents in the past in which junior staff have let themselves be intimidated by senior staff, letting them enter or leave spill areas and spreading contamination. Setting this policy and supporting those who enforce it takes the junior staff off the hook and we have had no such problems since its implementation. Once spill boundaries are established, they should be verified by surveying on the “clean” side to confirm that all of the contamination is contained within the boundaries. Minimizing exposure is as much a philosophical point as a procedure. As noted above, spills are not life‐endangering. There is time to consider the best way to address the problem. Think about the situation you are faced with – do you have proper PPE, do you have the materials you need to survey and decontaminated efficiently, are you wearing respiratory protection (if the materials are volatile), do you have dosimetry, do you know where the highest radiation areas are (and how to work around them), and so forth. By taking a moment to consider your situation and planning on how to best address it, you will be helping to reduce your exposure and that of others in the area. Once you have completed these immediate actions, the spill should not worsen, and it is possible to begin survey and clean‐up. In general, it is better to work from the outer spill areas towards the center and, in the case of multi‐level spills (say, a spill on a table

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that drips to the floor) to work from the top towards the bottom. In most cases, spills may be cleaned up with standard commercial cleaners, although spills involving radioactive metals (such as Cs‐137 or Co‐60) may benefit from the use of specialty products. Contamination surveys should be performed with an appropriate detector for the type of radiation emitted by the isotope spilled. This information is summarized in the accompanying table. A direct frisk will reveal the total amount (fixed plus removable) of contamination present in an area while a smear wipe will only show how much removable contamination is there, so there is a value in performing both types of surveys. However, some isotopes (H‐3, in particular) are very difficult to survey for by direct frisk and it’s possible that the only reliable information about contamination levels will be obtained via smear wipe surveys. Skin contamination As with radioactive spills, skin contamination is not life‐endangering although, in rare cases, localized skin burns can result from “hot particles”. This means that workers shouldn’t panic over skin contamination, but also that they should work quickly to remove the contaminants. The immediate actions in case of skin contamination can be remembered as “CCC”: • Contact the RSO to inform him/her about the skin contamination • Count the amount of contamination on the skin with an appropriate detector and

write this number down. This will later be used to help calculate skin dose and/or possible uptake from the contamination

• Clean the contaminated area by going to the nearest sink and washing with mild soap and cool to warm water. While cleaning, a general rule is to not take any actions that are painful or uncomfortable – in most cases, the skin acts as a barrier to keep contamination on the outside of the body, and it is important to not breach this barrier.

While decontaminating, the worker should survey periodically; if the count rate continues to decrease then the decontamination is having an effect and should continue. If, however, the count rate stabilizes or if the skin starts to redden or bleed, decontamination should stop until the RSO or another qualified person arrives to determine what should be done. In some cases, simply wrapping the contaminated area in plastic can help – the contamination is “sweated” out – but this is obviously not a good idea for facial contamination! More drastic decontamination measures should ONLY be taken if there is a need (because of very high contamination levels) AND if advised by a competent radiation safety professional.

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Following decontamination it may be necessary to calculate radiation dose to the skin or to internal organs. These should be done by either a staff health physicist or by a consultant because these calculations can be complex and it is necessary to make sure they are done correctly. There are some software programs that will help with these calculations, but they give the best results in the hands of a radiation safety professional. In the event a person is contaminated by something that will be absorbed through the skin (e.g. tritiated water or many iodine compounds) it may also be necessary to take urine samples or to perform thyroid counts to check for uptake of isotope. This determination can also be made by a health physicist. Traffic accidents involving radioactive materials Every day, radioactive materials are transported in thousands of vehicles. These include soil density gauges, radiopharmaceuticals, small vials of research isotopes, radioactive waste, nuclear reactor fuel, and more. Although rare, these vehicles are sometimes involved in accidents that may or may not release radioactive materials. It is imperative that any vehicular accident involving radioactivity be reported immediately to the company (if appropriate) and to emergency response personnel so that injured people can be cared for and so that the radioactive materials can be recovered and contained. The primary concern for any vehicular accident is the health of the people involved in the accident. Injured personnel must be cared for first, and stabilized if necessary. Contaminated (or potentially contaminated) people should be cared for without regard to their contamination if necessary. However, it may be prudent to inform emergency response and medical personnel of the contamination (and that it poses no risk to them) so that the victim can be wrapped or decontaminated to minimize contamination spread to the ambulance or medical facility. Even this step is not a necessity, but it will help to reduce the chance that a vehicle or medical room will require decontamination prior to use for other patients. After injured personnel are cared for, the radioactive materials must be accounted for, contained, and recovered as appropriate. The physical form of the radioactive materials (e.g. liquid, gas, solid), the manner in which they are contained, and the severity of the accident will determine the amount and spread of contamination. For example, a soil density gauge packed in its case will likely escape unscathed from all but the most severe accidents, while a jug of radioactive liquid may contaminate the inside of the vehicle and the ground it drips onto. This phase of the accident recovery should include donning appropriate protective equipment (say, gloves, protective coveralls, and shoe covers), opening the storage area, and assessing the physical condition of the

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radioactive materials storage container. If there is obvious leakage into the vehicle interior or onto the ground, or if contamination surveys show materials were released, they must be contained and cleaned up as necessary. Loss of radioactive materials

2. Radioactive materials must be accounted for at all times, and the loss of radioactive materials must be taken seriously. At the least, the loss of radioactive materials raises the concern of accidental exposure of the public or release to the environment. At worst, we must also consider the possibility of deliberate misuse. In the past, such misuse included attempted poisoning and deliberate contamination. Today, to these concerns we must also consider that missing radioactive materials may also be used in a terrorist attack. In any event, all cases of missing radioactive materials must be investigated and explained, and every effort should be made to locate and recover the missing materials. And, if enough material is lost, it may need to be reported to regulatory authorities, depending on the amount of radioactivity lost and the potential radiation dose to the public.

3. According to 10 CFR Section 20.2201 (Reports of theft or loss of licensed material):

4. Each licensee shall report by telephone as follows:

(i) Immediately after its occurrence becomes known to the licensee, any lost, stolen, or missing licensed material in an aggregate quantity equal to or greater than 1,000 times the quantity specified in appendix C to part 20 under such circumstances that it appears to the licensee that an exposure could result to persons in unrestricted areas; or

(ii) Within 30 days after the occurrence of any lost, stolen, or missing licensed material becomes known to the licensee, all licensed material in a quantity greater than 10 times the quantity specified in appendix C to part 20 that is still missing at this time.

Written reports. (1) Each licensee required to make a report under paragraph (a) of this section shall, within 30 days after making the telephone report, make a written report setting forth the following information:

(i) A description of the licensed material involved, including kind, quantity, and chemical and physical form; and

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(ii) A description of the circumstances under which the loss or theft occurred; and

(iii) A statement of disposition, or probable disposition, of the licensed material involved; and

(iv) Exposures of individuals to radiation, circumstances under which the exposures occurred, and the possible total effective dose equivalent to persons in unrestricted areas; and

(v) Actions that have been taken, or will be taken, to recover the material; and

(vi) Procedures or measures that have been, or will be, adopted to ensure against a recurrence of the loss or theft of licensed material.

You should have some criteria for determining when radioactive materials are considered to be “lost” and a policy or procedure for trying to locate or account for the lost materials. Although it sounds silly to say so, it’s not always obvious when radioactive materials are actually lost. For example, the necessary files may not be accessible and the person with access may be on vacation, making it hard to locate a particular source. Is the source lost? Not necessarily, if you can prove at some point that it was under control at all times – even if it takes a week or so for the custodian to return from vacation. On the other hand, you may decide a source is lost after only a few hours if you know exactly where it’s supposed to be and find it’s not there. You will need to decide for yourself when radioactive materials are considered lost, and you should be able to justify your decision to your regulators. Your procedure should also include how you plan to determine dose to the public if the materials cannot be located or if it is determined they were discharged to the environment by, say, incineration or discharge into the sanitary sewer system. It’s acceptable to have the dose determination performed by an outside consultant if your organization lacks a full‐time health physicist to perform these calculations, and hiring a consultant is a better idea than attempting them yourself and making a mistake. If the missing radioactive are eventually located, this should be reported to your regulators immediately. Regardless of the outcome of your investigations and attempts to locate the missing materials, you will also need to document all your actions and the results of your investigation, and you should keep a copy of your report and all supporting documentation in your incident files.

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Radiological terrorism Dealing with radiological terrorism could be the subject of a full book in and of itself, and it is simply not possible to go into exhaustive detail in part of a single article. For this reason, only broad generalities will be given here. In the event of radiological terrorism, your facility may be directly involved in the attack, or you may be called upon to provide support to the recovery efforts. In the latter case, your degree of participation will depend on your capabilities and whatever arrangements you may have made with emergency response personnel prior to the attack. In the absence of any pre‐existing emergency response arrangements, you may not be permitted to participate directly in the emergency phase of the attack because the Incident Commander may simply not know what you and your staff can do. This means that, if you want to assist, you should work with local emergency responders in advance to determine your role, rather than simply appearing at the scene, meters in hand, wanting to help. If an attack takes place on or near your facility, you may be directly involved. In this case, your highest priority will probably be protecting your personnel, recovering from any physical damage (fires, blast, etc.), and minimizing the spread of contamination to your facility. In general, a “dirty bomb” may cause extensive property damage, but there may be few, if any health effects from the radiological portion of the attack, although the radioactivity will complicate response efforts. The scene of a radiological attack will have elevated levels of contamination and possibly radiation. It will also be a crime scene, it may pose serious non‐radiological health risks, and emergency response efforts (putting out fires, stabilizing damaged buildings, isolating damaged utility lines, rescuing injured people, etc.) may be taking place. Obviously, rescuing people is the highest priority, along with addressing physical risks such as fires. But you must also remember that radiological concerns are also present, requiring setting up and enforcing radiological boundaries, surveying personnel leaving contamination areas, and performing radiation surveys as necessary to make sure that rescuers are not placing themselves at risk. These guidelines are deliberately general because any terrorist attack or other incident will be unique. The best we can do is to try to keep these general guidelines in mind as we work with our regulators, emergency response personnel, and others involved in the incident to deal with the situation as it unfolds and presents itself.

54

Closing thoughts Most radiological incidents are not life‐endangering. Even terrorist attacks are likely to either inconvenience a lot of people or to endanger a few people due to radiation (although an explosion may, indeed, be deadly). In most cases, you will have the luxury of taking a minute to consider your actions and to think through your response so that you handle the incident appropriately. And remember, people must take the highest priority – you can always decontaminate an area, but we can’t restore lost health or lives. For more information: CRC Handbook of Management of Radiation Protection Programs, Second Edition, Kenneth L. Miller (Editor), CRC Press, Boca Raton FL, 1992 Operational Radiation Safety Program, National Council on Radiation Protection and Measurements Report #127, 1998 Management of Terrorist Events Involving Radioactive Material, National Council on Radiation Protection and Measurements Report #138, 2001 Radiation Protection: A Guide for Scientists, Regulators, and Physicians, Fourth Edition, Jacob Shapiro, Harvard University Press, 2002 Disaster Preparedness for Radiology Professionals, American College of Radiology, 2003 (available on‐line at www.acr.org under “Disaster Planning Information” link) Conducting a contamination survey (to accompany the spill procedure) Performing radiation surveys can take a lot of time, but it is time well‐spent because a hasty survey can lead to spreading contamination around your facility and off‐site. Holding the detector too far away from the surface you’re surveying can give erroneously low readings, as can moving the detector too quickly. The phrase we use at the University of Rochester is “low and slow” – the detector should be help no more than about a half inch (1 cm) from the surface being surveyed and moved no more quickly than 1‐2 inches per second (or 3‐5 cm per second). While surveying, you should keep the audible response turned on because you will want to watch the probe itself – you don’t want to miss a spot, to hold the detector too far from the surface, or to contaminate the detector by bumping it into a contaminated surface. Listen to the count

55

rate and, if you hear an increase, pause for a moment to see if the increase is sustained. If so, look at the meter face to see what the count rate is at that location. When logging results, and in decontamination, you must convert from counts per minute (what the meter reads out in) to disintegration per minute (the amount of contamination present). This is done by dividing the count rate by the meter efficiency (which should be determined when the meter is calibrated). For example, if a geiger counter has 40% detection efficiency for P‐32, a count rate of 80 cpm above background levels corresponds to a disintegration rate of 200 dpm (80 ÷0.4 = 200). Radiological terrorism; general guidelines We have established some general guidelines for responding to on‐site radiological attack in a university or hospital setting. Some or all may be applicable to your facility.

1. Personnel who can neither hear nor see an explosion are probably not at risk. They should stay put if indoors or, if outdoors, go inside to await further information and instructions. People should NOT try to drive away because driving is likely to be more dangerous than staying put.

2. After going indoors, personnel should close open doors and windows, wash hands and face (take a shower if possible), and change your outer clothes if you can.

3. Contaminated injured people should have serious injuries treated without regard to contamination levels – contaminated persons do not endanger emergency response or medical personnel. If injuries are not serious, it may be possible to decontaminate the victims before transporting them, or at least to wrap them in a sheet or blanket to minimize the spread of contamination to vehicles and hospitals. This judgment call must be made on a case‐by‐case basis, depending on the extent of injuries and contamination.

4. You may need to perform surveys to establish radiological boundaries. These boundaries may be for high radiation or high contamination levels. According to regulations, the limit for removable contamination in an unrestricted area is 1000 dpm/100 cm2 and radiation levels in uncontrolled areas cannot exceed 2 mrem in one hour. Radiation surveys are relatively easy to perform, and radiation boundaries can often be established fairly easily. However, contamination boundaries are more difficult to establish because contamination surveys can be difficult and time‐consuming to perform. In some cases, it may be best to simply set contamination control boundaries a few hundred meters downwind and then expand or collapse them as you survey to confirm them. Note: Until you have a

56

good idea of contamination levels, you should dress in contamination control gear (shoe covers, gloves, coveralls, for example) to reduce the risk of personnel contamination.

5. Potentially contaminated people should stay in the controlled area until they can be surveyed and released. If there are only a few people, it may be possible to survey everyone directly and decontaminate them as necessary. However, even a few tens of people who are contaminated can take a great deal of time to survey thoroughly and decontaminate. Depending on your capabilities and those of the emergency responders, you may have little option other than releasing moderately contaminated people with instructions on how to decontaminate themselves and their clothing. However, releasing such people should be a last resort, to be taken only when it is obvious that no other reasonable options exist and with the concurrence of regulatory and emergency response personnel.

6. Eating, drinking, smoking, chewing tobacco, applying cosmetics, and other possible avenues of accidental ingestion or inhalation should be prohibited in any radiologically controlled area, or by any potentially contaminated person.

How to choose a survey meter Type of radiation emitted

Example isotopes

Type of survey

Type of detector to use

Alpha U‐238, Pu‐238, Pu‐239, Ra‐226, Po‐210, Am‐241

Direct frisk, smear wipe

Zinc sulfide (ZnS) or proportional counter

Low‐energy beta H‐3, C‐14, S‐35, Pu‐241 Smear wipe Liquid scintillation counter, proportional counter

Medium to high‐energy beta

P‐32, Sr‐90, I‐131 Direct frisk, smear wipe

Geiger counter, liquid scintillation counter, proportional counter

Low‐energy gamma

I‐125, I‐129, Am‐241 Direct frisk or smear wipe

Thin‐crystal (1”x1mm) sodium iodide

Medium‐ to high‐energy gamma

I‐131, Cs‐137, Co‐60, Ir‐192

Direct frisk or smear wipe

Thick‐crystal (1”x1” or larger) sodium iodide, Geiger counter

57

10 CFR PART 19‐‐NOTICES, INSTRUCTIONS AND REPORTS TO WORKERS: INSPECTION AND INVESTIGATIONS

Part Index

Sec.

19.1 Purpose.

19.2 Scope.

19.3 Definitions.

19.4 Interpretations.

19.5 Communications.

19.8 Information collection requirements: OMB approval.

19.11 Posting of notices to workers.

19.12 Instruction to workers.

19.13 Notifications and reports to individuals.

19.14 Presence of representatives of licensees and workers during inspections.

19.15 Consultation with workers during inspections.

19.16 Requests by workers for inspections.

19.17 Inspections not warranted; informal review.

19.18 Sequestration of witnesses and exclusion of counsel in interviews conducted under subpoena.

19.20 Employee protection.

19.30 Violations.

19.31 Application for exemptions.

19.32 Discrimination prohibited.

19.40 Criminal penalties.

Authority: Secs. 53, 63, 81, 103, 104, 161, 186, 68 Stat. 930, 933, 935, 936, 937, 948, 955, as amended, sec. 234, 83 Stat. 444, as amended, sec. 1701, 106 Stat. 2951, 2952, 2953 (42 U.S.C. 2073, 2093, 2111, 2133, 2134, 2201, 2236, 2282, 2297f); sec. 201, 88 Stat. 1242, as amended (42 U.S.C. 5841); Pub. L. 95‐601, sec. 10, 92 Stat. 2951 (42 U.S.C. 5851); sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note).

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Source: 38 FR 22217, Aug. 17, 1973, unless otherwise noted.

a) § 19.1 Purpose.

The regulations in this part establish requirements for notices, instructions, and reports by licensees to individuals participating in licensed activities and options available to these individuals in connection with Commission inspections of licensees to ascertain compliance with the provisions of the Atomic Energy Act of 1954, as amended, title II of the Energy Reorganization Act of 1974, and regulations, orders, and licenses thereunder regarding radiological working conditions. The regulations in this part also establish the rights and responsibilities of the Commission and individuals during interviews compelled by subpoena as part of agency inspections or investigations pursuant to section 161c of the Atomic Energy Act of 1954, as amended, on any matter within the Commissionʹs jurisdiction.

[55 FR 247, Jan. 4, 1990]

b) § 19.2 Scope.

The regulations in this part apply to all persons who receive, possess, use, or transfer material licensed by the Nuclear Regulatory Commission pursuant to the regulations in parts 30 through 36, 39, 40, 60, 61, 63, 70, or part 72 of this chapter, including persons licensed to operate a production or utilization facility under part 50 of this chapter, persons licensed to possess power reactor spent fuel in an independent spent fuel storage installation (ISFSI) pursuant to part 72 of this chapter, and in accordance with 10 CFR 76.60 to persons required to obtain a certificate of compliance or an approved compliance plan under part 76 of this chapter. The regulations regarding interviews of individuals under subpoena apply to all investigations and inspections within the jurisdiction of the Nuclear Regulatory Commission other than those involving NRC employees or NRC contractors. The regulations in this part do not apply to subpoenas issued pursuant to 10 CFR 2.720.

[66 FR 55789, Nov. 2, 2001]

c) § 19.3 Definitions.

As used in this part:

Act means the Atomic Energy Act of 1954, (68 Stat. 919) including any amendments thereto.

Commission means the United States Nuclear Regulatory Commission.

Exclusion means the removal of counsel representing multiple interests from an interview whenever the NRC official conducting the interview has concrete evidence that the presence of the counsel would obstruct and impede the particular investigation or inspection.

License means a license issued under the regulations in parts 30 through 36, 39, 40, 60, 61, 63, 70, or 72 of this chapter, including licenses to operate a production or utilization facility pursuant to part 50 of this chapter.

Licensee means the holder of such a license.

Restricted area means an area, access to which is limited by the licensee for the purpose of protecting individuals against undue risks from exposure to radiation and radioactive materials. Restricted area does not include areas used as residential quarters, but separate rooms in a residential building may be set apart as a restricted area.

59

Sequestration means the separation or isolation of witnesses and their attorneys from other witnesses and their attorneys during an interview conducted as part of an investigation, inspection, or other inquiry.

Worker means an individual engaged in activities licensed by the Commission and controlled by a licensee, but does not include the licensee.

[38 FR 22217, Aug. 17, 1973, as amended at 40 FR 8783, Mar. 3, 1975; 53 FR 31680, Aug. 19, 1988; 55 FR 247, Jan. 4, 1990; 56 FR 23470, May 21, 1991; 56 FR 65948, Dec. 19, 1991; 57 FR 61785, Dec. 29, 1992; 58 FR 7736, Feb. 9, 1993; 66 FR 55789, Nov. 2, 2001; 69 FR 76600, Dec. 22, 2004]

d) § 19.4 Interpretations.

Except as specifically authorized by the Commission in writing, no interpretation of the meaning of the regulations in this part by any officer or employee of the Commission other than a written interpretation by the General Counsel will be recognized to be binding upon the Commission.

e) § 19.5 Communications.

Except where otherwise specified in this part, all communications and reports concerning the regulations in this part should be addressed to the Regional Administrator of the appropriate U.S. Nuclear Regulatory Commission Regional Office listed in Appendix D of part 20 of this chapter. Communications, reports, and applications may be delivered in person at the Commissionʹs offices at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland.

[67 FR 67098, Nov. 4, 2002]

f) § 19.8 Information collection requirements: OMB approval.

(a) The Nuclear Regulatory Commission has submitted the information collection requirements contained in this part to the Office of Management and Budget (OMB) for approval as required by the Paperwork Reduction Act (44 U.S.C. 3501 et seq.). The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number. OMB has approved the information collection requirements contained in the part under control number 3150‐0044.

(b) The approved information collection requirements contained in this part appear in §§ 19.13 and 19.16.

[62 FR 52185, Oct. 6, 1997]

g) § 19.11 Posting of notices to workers.

(a) Each licensee shall post current copies of the following documents:

(1) The regulations in this part and in part 20 of this chapter;

(2) The license, license conditions, or documents incorporated into a license by reference, and amendments thereto;

(3) The operating procedures applicable to licensed activities;

(4) Any notice of violation involving radiological working conditions, proposed imposition of civil penalty, or order issued pursuant to subpart B of part 2 of this chapter, and any response from the licensee.

60

(b) If posting of a document specified in paragraph (a) (1), (2) or (3) of this section is not practicable, the licensee may post a notice which describes the document and states where it may be examined.

(c)(1) Each licensee and each applicant for a specific license shall prominently post NRC Form 3, ʺNotice to Employees,ʺ dated August 1997. Later versions of NRC Form 3 that supersede the August 1997 version shall replace the previously posted version within 30 days of receiving the revised NRC Form 3 from the Commission.

(2) Additional copies of NRC Form 3 may be obtained by writing to the Regional Administrator of the appropriate U.S. Nuclear Regulatory Commission Regional Office listed in appendix D to part 20 of this chapter, by calling (301) 415‐5877, via e‐mail to [email protected], or by visiting the NRCʹs Web site at http://www.nrc.gov and selecting forms from the index found on the home page.

(d) Documents, notices, or forms posted pursuant to this section shall appear in a sufficient number of places to permit individuals engaged in licensed activities to observe them on the way to or from any particular licensed activity location to which the document applies, shall be conspicuous, and shall be replaced if defaced or altered.

(e) Commission documents posted pursuant to paragraph (a)(4) of this section shall be posted within 2 working days after receipt of the documents from the Commission; the licenseeʹs response, if any, shall be posted within 2 working days after dispatch by the licensee. Such documents shall remain posted for a minimum of 5 working days or until action correcting the violation has been completed, whichever is later.

[38 FR 22217, Aug. 17, 1973, as amended at 40 FR 8783, Mar. 3, 1975; 47 FR 30454, July 14, 1982; 58 FR 52408, Oct. 8, 1993; 60 FR 24551, May 9, 1995; 61 FR 6764, Feb. 22, 1996; 62 FR 48166, Sept. 15, 1997; 68 FR 58801, Oct. 10, 2003]

h) § 19.12 Instruction to workers.

(a) All individuals who in the course of employment are likely to receive in a year an occupational dose in excess of 100 mrem (1 mSv) shall be‐‐

(1) Kept informed of the storage, transfer, or use of radiation and/or radioactive material;

(2) Instructed in the health protection problems associated with exposure to radiation and/or radioactive material, in precautions or procedures to minimize exposure, and in the purposes and functions of protective devices employed;

(3) Instructed in, and required to observe, to the extent within the workers control, the applicable provisions of Commission regulations and licenses for the protection of personnel from exposure to radiation and/or radioactive material;

(4) Instructed of their responsibility to report promptly to the licensee any condition which may lead to or cause a violation of Commission regulations and licenses or unnecessary exposure to radiation and/or radioactive material;

(5) Instructed in the appropriate response to warnings made in the event of any unusual occurrence or malfunction that may involve exposure to radiation and/or radioactive material; and

(6) Advised as to the radiation exposure reports which workers may request pursuant to § 19.13.

(b) In determining those individuals subject to the requirements of paragraph (a) of this section, licensees must take into consideration assigned activities during normal and abnormal situations involving exposure to radiation and/or radioactive material which can reasonably be expected to occur during the life of a licensed facility. The extent of

61

these instructions must be commensurate with potential radiological health protection problems present in the work place.

[60 FR 36043, July 13, 1995]

i) § 19.13 Notifications and reports to individuals.

(a) Radiation exposure data for an individual, and the results of any measurements, analyses, and calculations of radioactive material deposited or retained in the body of an individual, shall be reported to the individual as specified in this section. The information reported shall include data and results obtained pursuant to Commission regulations, orders or license conditions, as shown in records maintained by the licensee pursuant to Commission regulations. Each notification and report shall: be in writing; include appropriate identifying data such as the name of the licensee, the name of the individual, the individualʹs social security number; include the individualʹs exposure information; and contain the following statement:

This report is furnished to you under the provisions of the Nuclear Regulatory Commission regulation 10 CFR part 19. You should preserve this report for further reference.

(b) Each licensee shall advise each worker annually of the workerʹs dose as shown in records maintained by the licensee pursuant to the provisions of § 20.2106 of 10 CFR part 20.

(c)(1) At the request of a worker formerly engaged in licensed activities controlled by the licensee, each licensee shall furnish to the worker a report of the workerʹs exposure to radiation and/or to radioactive material:

(i) As shown in records maintained by the licensee pursuant to § 20.2106 for each year the worker was required to be monitored under the provisions of § 20.1502; and

(ii) For each year the worker was required to be monitored under the monitoring requirements in effect prior to January 1, 1994.

(2) This report must be furnished within 30 days from the time the request is made or within 30 days after the exposure of the individual has been determined by the licensee, whichever is later. This report must cover the period of time that the workerʹs activities involved exposure to radiation from radioactive material licensed by the Commission and must include the dates and locations of licensed activities in which the worker participated during this period.

(d) When a licensee is required pursuant to §§ 20.2202, 20.2203, 20.2204, or 20.2206 of this chapter to report to the Commission any exposure of an individual to radiation or radioactive material the licensee shall also provide the individual a report on his or her exposure data included therein. This report must be transmitted at a time not later than the transmittal to the Commission.

(e) At the request of a worker who is terminating employment with the licensee that involved exposure to radiation or radioactive materials, during the current calendar quarter or the current year, each licensee shall provide at termination to each worker, or to the workerʹs designee, a written report regarding the radiation dose received by that worker from operations of the licensee during the current year or fraction thereof. If the most recent individual monitoring results are not available at that time, a written estimate of the dose must be provided together with a clear indication that this is an estimate.

[38 FR 22217, Aug. 17, 1973, as amended at 40 FR 8783, Mar. 3, 1975; 44 FR 32352, June 6, 1979; 58 FR 67658, Dec. 22, 1993; 59 FR 41642, Aug. 15, 1994]

62

j) § 19.14 Presence of representatives of licensees and workers during inspections.

(a) Each licensee shall afford to the Commission at all reasonable times opportunity to inspect materials, activities, facilities, premises, and records pursuant to the regulations in this chapter.

(b) During an inspection, Commission inspectors may consult privately with workers as specified in § 19.15. The licensee or licenseeʹs representative may accompany Commission inspectors during other phrases of an inspection.

(c) If, at the time of inspection, an individual has been authorized by the workers to represent them during Commission inspections, the licensee shall notify the inspectors of such authorization and shall give the workersʹ representative an opportunity to accompany the inspectors during the inspection of physical working conditions.

(d) Each workersʹ representative shall be routinely engaged in licensed activities under control of the licensee and shall have received instructions as specified in § 19.12.

(e) Different representatives of licensees and workers may accompany the inspectors during different phases of an inspection if there is no resulting interference with the conduct of the inspection. However, only one workersʹ representative at a time may accompany the inspectors.

(f) With the approval of the licensee and the workersʹ representative an individual who is not routinely engaged in licensed activities under control of the license, for example, a consultant to the licensee or to the workersʹ representative, shall be afforded the opportunity to accompany Commission inspectors during the inspection of physical working conditions.

(g) Notwithstanding the other provisions of this section, Commission inspectors are authorized to refuse to permit accompaniment by any individual who deliberately interferes with a fair and orderly inspection. With regard to areas containing information classified by an agency of the U.S. Government in the interest of national security, an individual who accompanies an inspector may have access to such information only if authorized to do so. With regard to any area containing proprietary information, the workersʹ representative for that area shall be an individual previously authorized by the licensee to enter that area.

k) § 19.15 Consultation with workers during inspections.

(a) Commission inspectors may consult privately with workers concerning matters of occupational radiation protection and other matters related to applicable provisions of Commission regulations and licenses to the extent the inspectors deem necessary for the conduct of an effective and thorough inspection.

(b) During the course of an inspection any worker may bring privately to the attention of the inspectors, either orally or in writing, any past or present condition which he has reason to believe may have contributed to or caused any violation of the act, the regulations in this chapter, or license condition, or any unnecessary exposure of an individual to radiation from licensed radioactive material under the licenseeʹs control. Any such notice in writing shall comply with the requirements of § 19.16(a).

(c) The provisions of paragraph (b) of this section shall not be interpreted as authorization to disregard instructions pursuant to § 19.12.

l) § 19.16 Requests by workers for inspections.

(a) Any worker or representative of workers who believes that a violation of the Act, the regulations in this chapter, or license conditions exists or has occurred in license activities with regard to radiological working conditions in

63

which the worker is engaged, may request an inspection by giving notice of the alleged violation to the Administrator of the appropriate Commission Regional Office, or to Commission inspectors. Any such notice shall be in writing, shall set forth the specific grounds for the notice, and shall be signed by the worker or representative of workers. A copy shall be provided the licensee by the Regional Office Administrator, or the inspector no later than at the time of inspection except that, upon the request of the worker giving such notice, his name and the name of individuals referred to therein shall not appear in such copy or on any record published, released or made available by the Commission, except for good cause shown.

(b) If, upon receipt of such notice, the Regional Office Administrator determines that the complaint meets the requirements set forth in paragraph (a) of this section, and that there are reasonable grounds to believe that the alleged violation exists or has occurred, he shall cause an inspection to be made as soon as practicable, to determine if such alleged violation exists or has occurred. Inspections pursuant to this section need not be limited to matters referred to in the complaint.

[38 FR 22217, Aug. 17, 1973, as amended at 40 FR 8783, Mar. 3, 1975; 47 FR 30454, July 14, 1982; 52 FR 31610, Aug. 21, 1987]

m) § 19.17 Inspections not warranted; informal review.

(a) If the Administrator of the appropriate Regional Office determines, with respect to a complaint under § 19.16, that an inspection is not warranted because there are no reasonable grounds to believe that a violation exists or has occurred, he shall notify the complainant in writing of such determination. The complainant may obtain review of this determination by submitting a written statement of position to the Executive Director for Operations, either by mail to the U.S. Nuclear Regulatory Commission, Washington, DC 20555‐0001; by hand delivery to the NRCʹs offices at 11555 Rockville Pike, Rockville, Maryland; or, where practicable, by electronic submission, for example, via Electronic Information Exchange, or CD‐ROM. Electronic submissions must be made in a manner that enables the NRC to receive, read, authenticate, distribute, and archive the submission, and process and retrieve it a single page at a time. Detailed guidance on making electronic submissions can be obtained by visiting the NRCʹs Web site at http://www.nrc.gov/site‐help/eie.html, by calling (301) 415‐6030, by e‐mail to [email protected], or by writing the Office of Information Services, U.S. Nuclear Regulatory Commission, Washington, DC 20555‐0001. The guidance discusses, among other topics, the formats the NRC can accept, the use of electronic signatures, and the treatment of nonpublic information. The Executive Director for Operations will provide the licensee with a copy of such statement by certified mail, excluding, at the request of the complainant, the name of the complainant. The licensee may submit an opposing written statement of position with the Executive Director for Operations who will provide the complainant with a copy of such statement by certified mail. Upon the request of the complainant, the Executive Director for Operations or his designee may hold an informal conference in which the complainant and the licensee may orally present their views. An informal conference may also be held at the request of the licensee, but disclosure of the identity of the complainant will be made only following receipt of written authorization from the complainant. After considering all written and oral views presented, the Executive Director for Operations shall affirm, modify, or reverse the determination of the Administrator of the appropriate Regional Office and furnish the complainant and the licensee a written notification of his decision and the reason therefor.

(b) If the Administrator of the appropriate Regional Office determines that an inspection is not warranted because the requirements of § 19.16(a) have not been met, he shall notify the complainant in writing of such determination. Such determination shall be without prejudice to the filing of a new complaint meeting the requirements of § 19.16(a).

[38 FR 22217, Aug. 17, 1973, as amended at 40 FR 8783, Mar. 3, 1975; 52 FR 31610, Aug. 21, 1987; 67 FR 77652, Dec. 19, 2002; 68 FR 58801, Oct. 10, 2003; 70 FR 69421, Nov. 16, 2005]

64

n) § 19.18 Sequestration of witnesses and exclusion of counsel in interviews conducted under subpoena.

(a) All witnesses compelled by subpoena to submit to agency interviews shall be sequestered unless the official conducting the interviews permits otherwise.

(b) Any witness compelled by subpoena to appear at an interview during an agency inquiry may be accompanied, represented, and advised by counsel of his or her choice. However, when the agency official conducting the inquiry determines, after consultation with the Office of the General Counsel, that the agency has concrete evidence that the presence of an attorney representing multiple interests would obstruct and impede the investigation or inspection, the agency official may prohibit that counsel from being present during the interview.

(c) The interviewing official is to provide a witness whose counsel has been excluded under paragraph (b) of this section and the witnessʹs counsel a written statement of the reasons supporting the decision to exclude. This statement, which must be provided no later than five working days after exclusion, must explain the basis for the counselʹs exclusion. This statement must also advise the witness of the witnessʹ right to appeal the exclusion decision and obtain an automatic stay of the effectiveness of the subpoena by filing a motion to quash the subpoena with the Commission within five days of receipt of this written statement.

(d) Within five days after receipt of the written notification required in paragraph (c) of this section, a witness whose counsel has been excluded may appeal the exclusion decision by filing a motion to quash the subpoena with the Commission. The filing of the motion to quash will stay the effectiveness of the subpoena pending the Commissionʹs decision on the motion.

(e) If a witnessʹ counsel is excluded under paragraph (b) of this section, the interview may, at the witnessʹ request, either proceed without counsel or be delayed for a reasonable period of time to permit the retention of new counsel. The interview may also be rescheduled to a subsequent date established by the NRC, although the interview shall not be rescheduled by the NRC to a date that precedes the expiration of the time provided under § 19.18(d) for appeal of the exclusion of counsel, unless the witness consents to an earlier date.

[55 FR 247, Jan. 4, 1990, as amended at 56 FR 65948, Dec. 19, 1991; 57 FR 61785, Dec. 29, 1992]

o) § 19.20 Employee protection.

Employment discrimination by a licensee (or a holder of a certificate of compliance issued pursuant to part 76) or a contractor or subcontractor of a licensee (or a holder of a certificate of compliance issued pursuant to part 76) against an employee for engaging in protected activities under this part or parts 30, 40, 50, 60, 61, 63, 70, 72, 76, or 150 of this chapter is prohibited.

[66 FR 55789, Nov. 2, 2001]

p) § 19.30 Violations.

(a) The Commission may obtain an injunction or other court order to prevent a violation of the provisions of‐‐

(1) The Atomic Energy Act of 1954, as amended;

(2) Title II of the Energy Reorganization Act of 1974, as amended; or

65

(3) A regulation or order issued pursuant to those Acts.

(b) The Commission may obtain a court order for the payment of a civil penalty imposed under section 234 of the Atomic Energy Act:

(1) For violations of‐‐

(i) Sections 53, 57, 62, 63, 81, 82, 101, 103, 104, 107, or 109 of the Atomic Energy Act of 1954, as amended;

(ii) Section 206 of the Energy Reorganization Act;

(iii) Any rule, regulation, or order issued pursuant to the sections specified in paragraph (b)(1)(i) of this section;

(iv) Any term, condition, or limitation of any license issued under the sections specified in paragraph (b)(1)(i) of this section.

(2) For any violation for which a license may be revoked under section 186 of the Atomic Energy Act of 1954, as amended.

[57 FR 55071, Nov. 24, 1992]

q) § 19.31 Application for exemptions.

The Commission may upon application by any licensee or upon its own initiative, grant such exemptions from the requirements of the regulations in this part as it determines are authorized by law and will not result in undue hazard to life or property.

r) § 19.32 Discrimination prohibited.

No person shall on the ground of sex be excluded from participation in, be denied the benefit of, or be subjected to discrimination under any program or activity licensed by the Nuclear Regulatory Commission. This provision will be enforced through agency provisions and rules similar to those already established, with respect to racial and other discrimination, under Title VI of the Civil Rights Act of 1964. This remedy is not exclusive, however, and will not prejudice or cut off any other legal remedies available to a discriminatee.

[65 FR 54949, Sept. 12, 2000; 68 FR 75389, Dec. 31, 2003]

s) § 19.40 Criminal penalties.

(a) Section 223 of the Atomic Energy Act of 1954, as amended, provides for criminal sanctions for willful violation of, attempted violation of, or conspiracy to violate, any regulation issued under sections 161b, 161i, or 161o of the Act. For purposes of section 223, all the regulations in part 19 are issued under one or more of sections 161b, 161i, or 161o, except for the sections listed in paragraph (b) of this section.

(b) The regulations in part 19 that are not issued under sections 161b, 161i, or 161o for the purposes of section 223 are as follows: §§ 19.1, 19.2, 19.3, 19.4, 19.5, 19.8, 19.16, 19.17, 19.18, 19.30, 19.31, and 19.40.

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10 CFR PART 20‐‐STANDARDS FOR PROTECTION AGAINST RADIATION

Part Index

i) Subpart A‐‐General Provisions

20.1001 Purpose.

20.1002 Scope.

20.1003 Definitions.

20.1004 Units of radiation dose.

20.1005 Units of radioactivity.



20.1008 Implementation.


ii) Subpart B‐‐Radiation Protection Programs

20.1101 Radiation protection programs.

iii) Subpart C‐‐Occupational Dose Limits

20.1201 Occupational dose limits for adults.

20.1202 Compliance with requirements for summation of external and internal doses.

20.1203 Determination of external dose from airborne radioactive material.

20.1204 Determination of internal exposure.

20.1205 [Reserved]

20.1206 Planned special exposures.

20.1207 Occupational dose limits for minors.

20.1208 Dose equivalent to an embryo/fetus.

iv) Subpart D‐‐Radiation Dose Limits for Individual Members of the Public

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20.1301 Dose limits for individual members of the public.

20.1302 Compliance with dose limits for individual members of the public.

v) Subpart E‐‐Radiological Criteria for License Termination

20.1401 General provisions and scope.

20.1402 Radiological criteria for unrestricted use.

20.1403 Criteria for license termination under restricted conditions.

20.1404 Alternate criteria for license termination.

20.1405 Public notification and public participation.

20.1406 Minimization of contamination.

vi) Subpart F‐‐Surveys and Monitoring

20.1501 General.

20.1502 Conditions requiring individual monitoring of external and internal occupational dose.

vii) Subpart G‐‐Control of Exposure From External Sources in Restricted Areas

20.1601 Control of access to high radiation areas.

20.1602 Control of access to very high radiation areas.

viii) Subpart H‐‐Respiratory Protection and Controls to Restrict Internal Exposure in Restricted Areas

20.1701 Use of process or other engineering controls.

20.1702 Use of other controls.

20.1703 Use of individual respiratory protection equipment.

20.1704 Further restrictions on the use of respiratory protection equipment.

20.1705 Application for use of higher assigned protection factors.

ix) Subpart I‐‐Storage and Control of Licensed Material

20.1801 Security of stored material.

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20.1802 Control of material not in storage.

x) Subpart J‐‐Precautionary Procedures

20.1901 Caution signs.

20.1902 Posting requirements.

20.1903 Exceptions to posting requirements.

20.1904 Labeling containers.

20.1905 Exemptions to labeling requirements.

20.1906 Procedures for receiving and opening packages.

xi) Subpart K‐‐Waste Disposal

20.2001 General requirements.

20.2002 Method for obtaining approval of proposed disposal procedures.

20.2003 Disposal by release into sanitary sewerage.

20.2004 Treatment or disposal by incineration.

20.2005 Disposal of specific wastes.

20.2006 Transfer for disposal and manifests.

20.2007 Compliance with environmental and health protection regulations.

xii) Subpart L‐‐Records

20.2101 General provisions.

20.2102 Records of radiation protection programs.

20.2103 Records of surveys.

20.2104 Determination of prior occupational dose.

20.2105 Records of planned special exposures.

20.2106 Records of individual monitoring results.

20.2107 Records of dose to individual members of the public.

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20.2108 Records of waste disposal.

20.2109 [Reserved]

20.2110 Form of records.

xiii) Subpart M‐‐Reports

20.2201 Reports of theft or loss of licensed material.

20.2202 Notification of incidents.

20.2203 Reports of exposures, radiation levels, and concentrations of radioactive material exceeding the constraints or limits.

20.2204 Reports of planned special exposures.

20.2205 Reports to individuals of exceeding dose limits.

20.2206 Reports of individual monitoring.

xiv) Subpart N‐‐Exemptions and Additional Requirements

20.2301 Applications for exemptions.

20.2302 Additional requirements.

xv) Subpart O‐‐Enforcement

20.2401 Violations.


Appendix A to Part 20‐‐Assigned Protection Factors for Respirators

Appendix B to Part 20‐‐Annual Limits on Intake (ALIs) and Derived Air Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations; Concentrations for Release to Sewerage

Appendix C to Part 20‐‐Quantities of Licensed Material Requiring Labeling

Appendix D to Part 20‐‐United States Nuclear Regulatory Commission Regional Offices

Appendix E to Part 20‐‐[Reserved]

Appendix F to Part 20‐‐[Reserved]

Appendix G to Part 20‐‐Requirements for Transfers of Low‐Level Radioactive Waste Intended for Disposal at Licensed Land Disposal Facilities and Manifests

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Authority: Secs. 53, 63, 65, 81, 103, 104, 161, 182, 186, 68 Stat. 930, 933, 935, 936, 937, 948, 953, 955, as amended, sec. 1701, 106 Stat. 2951, 2952, 2953 (42 U.S.C. 2073, 2093, 2095, 2111, 2133, 2134, 2201, 2232, 2236, 2297f), secs. 201, as amended, 202, 206, 88 Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846); sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note).

t) Subpart A‐‐General Provisions

Source: 56 FR 23391, May 21, 1991, unless otherwise noted.

u) § 20.1001 Purpose.

(a) The regulations in this part establish standards for protection against ionizing radiation resulting from activities conducted under licenses issued by the Nuclear Regulatory Commission. These regulations are issued under the Atomic Energy Act of 1954, as amended, and the Energy Reorganization Act of 1974, as amended.

(b) It is the purpose of the regulations in this part to control the receipt, possession, use, transfer, and disposal of licensed material by any licensee in such a manner that the total dose to an individual (including doses resulting from licensed and unlicensed radioactive material and from radiation sources other than background radiation) does not exceed the standards for protection against radiation prescribed in the regulations in this part. However, nothing in this part shall be construed as limiting actions that may be necessary to protect health and safety.

v) § 20.1002 Scope.

The regulations in this part apply to persons licensed by the Commission to receive, possess, use, transfer, or dispose of byproduct, source, or special nuclear material or to operate a production or utilization facility under Parts 30 through 36, 39, 40, 50, 60, 61, 63, 70, or 72 of this chapter, and in accordance with 10 CFR 76.60 to persons required to obtain a certificate of compliance or an approved compliance plan under part 76 of this chapter. The limits in this part do not apply to doses due to background radiation, to exposure of patients to radiation for the purpose of medical diagnosis or therapy, to exposure from individuals administered radioactive material and released under § 35.75, or to exposure from voluntary participation in medical research programs.

[67 FR 20370, Apr. 24, 2002; 67 FR 62872, Oct. 9, 2002, as amended at 67 FR 77652, Dec. 19, 2002]

w) § 20.1003 Definitions.

As used in this part:

Absorbed dose means the energy imparted by ionizing radiation per unit mass of irradiated material. The units of absorbed dose are the rad and the gray (Gy).

Act means the Atomic Energy Act of 1954 (42 U.S.C. 2011 et seq.), as amended.

Activity is the rate of disintegration (transformation) or decay of radioactive material. The units of activity are the curie (Ci) and the becquerel (Bq).

Adult means an individual 18 or more years of age.

Airborne radioactive material means radioactive material dispersed in the air in the form of dusts, fumes, particulates, mists, vapors, or gases.

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Airborne radioactivity area means a room, enclosure, or area in which airborne radioactive materials, composed wholly or partly of licensed material, exist in concentrations‐‐

(1) In excess of the derived air concentrations (DACs) specified in appendix B, to §§ 20.1001‐20.2401, or

(2) To such a degree that an individual present in the area without respiratory protective equipment could exceed, during the hours an individual is present in a week, an intake of 0.6 percent of the annual limit on intake (ALI) or 12 DAC‐hours.

Air‐purifying respirator means a respirator with an air‐purifying filter, cartridge, or canister that removes specific air contaminants by passing ambient air through the air‐purifying element.

ALARA (acronym for ʺas low as is reasonably achievableʺ) means making every reasonable effort to maintain exposures to radiation as far below the dose limits in this part as is practical consistent with the purpose for which the licensed activity is undertaken, taking into account the state of technology, the economics of improvements in relation to state of technology, the economics of improvements in relation to benefits to the public health and safety, and other societal and socioeconomic considerations, and in relation to utilization of nuclear energy and licensed materials in the public interest.

Annual limit on intake (ALI) means the derived limit for the amount of radioactive material taken into the body of an adult worker by inhalation or ingestion in a year. ALI is the smaller value of intake of a given radionuclide in a year by the reference man that would result in a committed effective dose equivalent of 5 rems (0.05 Sv) or a committed dose equivalent of 50 rems (0.5 Sv) to any individual organ or tissue. (ALI values for intake by ingestion and by inhalation of selected radionuclides are given in Table 1, Columns 1 and 2, of appendix B to §§ 20.1001‐20.2401).

Assigned protection factor (APF) means the expected workplace level of respiratory protection that would be provided by a properly functioning respirator or a class of respirators to properly fitted and trained users. Operationally, the inhaled concentration can be estimated by dividing the ambient airborne concentration by the APF.

Atmosphere‐supplying respirator means a respirator that supplies the respirator user with breathing air from a source independent of the ambient atmosphere, and includes supplied‐air respirators (SARs) and self‐contained breathing apparatus (SCBA) units.

Background radiation means radiation from cosmic sources; naturally occurring radioactive material, including radon (except as a decay product of source or special nuclear material); and global fallout as it exists in the environment from the testing of nuclear explosive devices or from past nuclear accidents such as Chernobyl that contribute to background radiation and are not under the control of the licensee. ``Background radiationʹʹ does not include radiation from source, byproduct, or special nuclear materials regulated by the Commission.

Bioassay (radiobioassay) means the determination of kinds, quantities or concentrations, and, in some cases, the locations of radioactive material in the human body, whether by direct measurement (in vivo counting) or by analysis and evaluation of materials excreted or removed from the human body.

Byproduct material means‐‐

(1) Any radioactive material (except special nuclear material) yielded in, or made radioactive by, exposure to the radiation incident to the process of producing or utilizing special nuclear material; and

(2) The tailings or wastes produced by the extraction or concentration of uranium or thorium from ore processed primarily for its source material content, including discrete surface wastes resulting from uranium solution extraction processes. Underground ore bodies depleted by these solution extraction operations do not constitute ʺbyproduct

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materialʺ within this definition.

Class (or lung class or inhalation class) means a classification scheme for inhaled material according to its rate of clearance from the pulmonary region of the lung. Materials are classified as D, W, or Y, which applies to a range of clearance half‐times: for Class D (Days) of less than 10 days, for Class W (Weeks) from 10 to 100 days, and for Class Y (Years) of greater than 100 days.

Collective dose is the sum of the individual doses received in a given period of time by a specified population from exposure to a specified source of radiation.

Commission means the Nuclear Regulatory Commission or its duly authorized representatives.

Committed dose equivalent (HT,50) means the dose equivalent to organs or tissues of reference (T) that will be received from an intake of radioactive material by an individual during the 50‐year period following the intake.

Committed effective dose equivalent (HE,50) is the sum of the products of the weighting factors applicable to each of the body organs or tissues that are irradiated and the committed dose equivalent to these organs or tissues (HE,50 = ΣWTHT.50).

Constraint (dose constraint) means a value above which specified licensee actions are required.

Controlled area means an area, outside of a restricted area but inside the site boundary, access to which can be limited by the licensee for any reason.

Critical Group means the group of individuals reasonably expected to receive the greatest exposure to residual radioactivity for any applicable set of circumstances.

Declared pregnant woman means a woman who has voluntarily informed the licensee, in writing, of her pregnancy and the estimated date of conception. The declaration remains in effect until the declared pregnant woman withdraws the declaration in writing or is no longer pregnant.

Decommission means to remove a facility or site safely from service and reduce residual radioactivity to a level that permits‐‐

(1) Release of the property for unrestricted use and termination of the license; or

(2) Release of the property under restricted conditions and termination of the license.

Deep‐dose equivalent (Hd), which applies to external whole‐body exposure, is the dose equivalent at a tissue depth of 1 cm (1000 mg/cm2).

Demand respirator means an atmosphere‐supplying respirator that admits breathing air to the facepiece only when a negative pressure is created inside the facepiece by inhalation.

Department means the Department of Energy established by the Department of Energy Organization Act (Pub. L. 95‐91, 91 Stat. 565, 42 U.S.C. 7101 et seq.) to the extent that the Department, or its duly authorized representatives, exercises functions formerly vested in the U.S. Atomic Energy Commission, its Chairman, members, officers, and components and transferred to the U.S. Energy Research and Development Administration and to the Administrator thereof pursuant to sections 104 (b), (c), and (d) of the Energy Reorganization Act of 1974 (Pub. L. 93‐438, 88 Stat. 1233 at 1237, 42 U.S.C. 5814) and retransferred to the Secretary of Energy pursuant to section 301(a) of the Department of

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Energy Organization Act (Pub. L. 95‐91, 91 Stat 565 at 577‐578, 42 U.S.C. 7151).

Derived air concentration (DAC) means the concentration of a given radionuclide in air which, if breathed by the reference man for a working year of 2,000 hours under conditions of light work (inhalation rate 1.2 cubic meters of air per hour), results in an intake of one ALI. DAC values are given in Table 1, Column 3, of appendix B to §§ 20.1001‐20.2401.

Derived air concentration‐hour (DAC‐hour) is the product of the concentration of radioactive material in air (expressed as a fraction or multiple of the derived air concentration for each radionuclide) and the time of exposure to that radionuclide, in hours. A licensee may take 2,000 DAC‐hours to represent one ALI, equivalent to a committed effective dose equivalent of 5 rems (0.05 Sv).

Disposable respirator means a respirator for which maintenance is not intended and that is designed to be discarded after excessive breathing resistance, sorbent exhaustion, physical damage, or end‐of‐service‐life renders it unsuitable for use. Examples of this type of respirator are a disposable half‐mask respirator or a disposable escape‐only self‐contained breathing apparatus (SCBA).

Distinguishable from background means that the detectable concentration of a radionuclide is statistically different from the background concentration of that radionuclide in the vicinity of the site or, in the case of structures, in similar materials using adequate measurement technology, survey, and statistical techniques.

Dose or radiation dose is a generic term that means absorbed dose, dose equivalent, effective dose equivalent, committed dose equivalent, committed effective dose equivalent, or total effective dose equivalent, as defined in other paragraphs of this section.

Dose equivalent (HT) means the product of the absorbed dose in tissue, quality factor, and all other necessary modifying factors at the location of interest. The units of dose equivalent are the rem and sievert (Sv).

Dosimetry processor means an individual or organization that processes and evaluates individual monitoring equipment in order to determine the radiation dose delivered to the equipment.

Effective dose equivalent (HE) is the sum of the products of the dose equivalent to the organ or tissue (HT) and the weighting factors (WT) applicable to each of the body organs or tissues that are irradiated (HE = ΣWTHT).

Embryo/fetus means the developing human organism from conception until the time of birth.

Entrance or access point means any location through which an individual could gain access to radiation areas or to radioactive materials. This includes entry or exit portals of sufficient size to permit human entry, irrespective of their intended use.

Exposure means being exposed to ionizing radiation or to radioactive material.

External dose means that portion of the dose equivalent received from radiation sources outside the body.

Extremity means hand, elbow, arm below the elbow, foot, knee, or leg below the knee.

Filtering facepiece (dust mask) means a negative pressure particulate respirator with a filter as an integral part of the facepiece or with the entire facepiece composed of the filtering medium, not equipped with elastomeric sealing surfaces and adjustable straps.

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Fit factor means a quantitative estimate of the fit of a particular respirator to a specific individual, and typically estimates the ratio of the concentration of a substance in ambient air to its concentration inside the respirator when worn.

Fit test means the use of a protocol to qualitatively or quantitatively evaluate the fit of a respirator on an individual.

Generally applicable environmental radiation standards means standards issued by the Environmental Protection Agency (EPA) under the authority of the Atomic Energy Act of 1954, as amended, that impose limits on radiation exposures or levels, or concentrations or quantities of radioactive material, in the general environment outside the boundaries of locations under the control of persons possessing or using radioactive material.

Government agency means any executive department, commission, independent establishment, corporation wholly or partly owned by the United States of America, which is an instrumentality of the United States, or any board, bureau, division, service, office, officer, authority, administration, or other establishment in the executive branch of the Government.

Gray [See § 20.1004].

Helmet means a rigid respiratory inlet covering that also provides head protection against impact and penetration.

High radiation area means an area, accessible to individuals, in which radiation levels from radiation sources external to the body could result in an individual receiving a dose equivalent in excess of 0.1 rem (1 mSv) in 1 hour at 30 centimeters from the radiation source or 30 centimeters from any surface that the radiation penetrates.

Hood means a respiratory inlet covering that completely covers the head and neck and may also cover portions of the shoulders and torso.

Individual means any human being.

Individual monitoring means‐‐

(1) The assessment of dose equivalent by the use of devices designed to be worn by an individual;

(2) The assessment of committed effective dose equivalent by bioassay (see Bioassay) or by determination of the time‐weighted air concentrations to which an individual has been exposed, i.e., DAC‐hours; or

(3) The assessment of dose equivalent by the use of survey data.

Individual monitoring devices (individual monitoring equipment) means devices designed to be worn by a single individual for the assessment of dose equivalent such as film badges, thermoluminescence dosimeters (TLDs), pocket ionization chambers, and personal (ʺlapelʺ) air sampling devices.

Internal dose means that portion of the dose equivalent received from radioactive material taken into the body.

Lens dose equivalent (LDE) applies to the external exposure of the lens of the eye and is taken as the dose equivalent at a tissue depth of 0.3 centimeter (300 mg/cm2).

License means a license issued under the regulations in parts 30 through 36, 39, 40, 50, 60, 61, 63, 70, or 72 of this chapter.

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Licensed material means source material, special nuclear material, or byproduct material received, possessed, used, transferred or disposed of under a general or specific license issued by the Commission.

Licensee means a license issued under the regulations in parts 30 through 36, 39, 40, 50, 60, 61, 63, 70, or 72 of this chapter.

Limits (dose limits) means the permissible upper bounds of radiation doses.

Loose‐fitting facepiece means a respiratory inlet covering that is designed to form a partial seal with the face.

Lost or missing licensed material means licensed material whose location is unknown. It includes material that has been shipped but has not reached its destination and whose location cannot be readily traced in the transportation system.

Member of the public means any individual except when that individual is receiving an occupational dose.

Minor means an individual less than 18 years of age.

Monitoring (radiation monitoring, radiation protection monitoring) means the measurement of radiation levels, concentrations, surface area concentrations or quantities of radioactive material and the use of the results of these measurements to evaluate potential exposures and doses.

Negative pressure respirator (tight fitting) means a respirator in which the air pressure inside the facepiece is negative during inhalation with respect to the ambient air pressure outside the respirator.

Nonstochastic effect means health effects, the severity of which varies with the dose and for which a threshold is believed to exist. Radiation‐induced cataract formation is an example of a nonstochastic effect (also called a deterministic effect).

NRC means the Nuclear Regulatory Commission or its duly authorized representatives.

Occupational dose means the dose received by an individual in the course of employment in which the individual’s assigned duties involve exposure to radiation or to radioactive material from licensed and unlicensed sources of radiation, whether in the possession of the licensee or other person. Occupational dose does not include doses received from background radiation, from any medical administration the individual has received, from exposure to individuals administered radioactive material and released under § 35.75, from voluntary participation in medical research programs, or as a member of the public.

Person means‐‐

(1) Any individual, corporation, partnership, firm, association, trust, estate, public or private institution, group, Government agency other than the Commission or the Department of Energy (except that the Department shall be considered a person within the meaning of the regulations in 10 CFR chapter I to the extent that its facilities and activities are subject to the licensing and related regulatory authority of the Commission under section 202 of the Energy Reorganization Act of 1974 (88 Stat. 1244), the Uranium Mill Tailings Radiation Control Act of 1978 (92 Stat. 3021), the Nuclear Waste Policy Act of 1982 (96 Stat. 2201), and section 3(b)(2) of the Low‐Level Radioactive Waste Policy Amendments Act of 1985 (99 Stat. 1842)), any State or any political subdivision of or any political entity within a State, any foreign government or nation or any political subdivision of any such government or nation, or other entity; and

(2) Any legal successor, representative, agent, or agency of the foregoing.

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Planned special exposure means an infrequent exposure to radiation, separate from and in addition to the annual dose limits.

Positive pressure respirator means a respirator in which the pressure inside the respiratory inlet covering exceeds the ambient air pressure outside the respirator.

Powered air‐purifying respirator (PAPR) means an air‐purifying respirator that uses a blower to force the ambient air through air‐purifying elements to the inlet covering.

Pressure demand respirator means a positive pressure atmosphere‐supplying respirator that admits breathing air to the facepiece when the positive pressure is reduced inside the facepiece by inhalation.

Public dose means the dose received by a member of the public from exposure to radiation or to radioactive material released by a licensee, or to any other source of radiation under the control of a licensee. Public dose does not include occupational dose or doses received from background radiation, from any medical administration the individual has received, from exposure to individuals administered radioactive material and released under § 35.75, or from voluntary participation in medical research programs.

Qualitative fit test (QLFT) means a pass/fail fit test to assess the adequacy of respirator fit that relies on the individualʹs response to the test agent.

Quality Factor (Q) means the modifying factor (listed in tables 1004(b).1 and 1004(b).2 of § 20.1004) that is used to derive dose equivalent from absorbed dose.

Quantitative fit test (QNFT) means an assessment of the adequacy of respirator fit by numerically measuring the amount of leakage into the respirator.

Quarter means a period of time equal to one‐fourth of the year observed by the licensee (approximately 13 consecutive weeks), providing that the beginning of the first quarter in a year coincides with the starting date of the year and that no day is omitted or duplicated in consecutive quarters.

Rad (See § 20.1004).

Radiation (ionizing radiation) means alpha particles, beta particles, gamma rays, x‐rays, neutrons, high‐speed electrons, high‐speed protons, and other particles capable of producing ions. Radiation, as used in this part, does not include non‐ionizing radiation, such as radio‐ or microwaves, or visible, infrared, or ultraviolet light.

Radiation area means an area, accessible to individuals, in which radiation levels could result in an individual receiving a dose equivalent in excess of 0.005 rem (0.05 mSv) in 1 hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates.

Reference man means a hypothetical aggregation of human physical and physiological characteristics arrived at by international consensus. These characteristics may be used by researchers and public health workers to standardize results of experiments and to relate biological insult to a common base.

Rem (See § 20.1004).

Residual radioactivity means radioactivity in structures, materials, soils, groundwater, and other media at a site resulting from activities under the licenseeʹs control. This includes radioactivity from all licensed and unlicensed sources used by the licensee, but excludes background radiation. It also includes radioactive materials remaining at

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the site as a result of routine or accidental releases of radioactive material at the site and previous burials at the site, even if those burials were made in accordance with the provisions of 10 CFR part 20.

Respiratory protective device means an apparatus, such as a respirator, used to reduce the individualʹs intake of airborne radioactive materials.

Restricted area means an area, access to which is limited by the licensee for the purpose of protecting individuals against undue risks from exposure to radiation and radioactive materials. Restricted area does not include areas used as residential quarters, but separate rooms in a residential building may be set apart as a restricted area.

Sanitary sewerage means a system of public sewers for carrying off waste water and refuse, but excluding sewage treatment facilities, septic tanks, and leach fields owned or operated by the licensee.

Self‐contained breathing apparatus (SCBA) means an atmosphere‐supplying respirator for which the breathing air source is designed to be carried by the user.

Shallow‐dose equivalent (Hs), which applies to the external exposure of the skin of the whole body or the skin of an extremity, is taken as the dose equivalent at a tissue depth of 0.007 centimeter (7 mg/cm2).

Sievert (See § 20.1004).

Site boundary means that line beyond which the land or property is not owned, leased, or otherwise controlled by the licensee.

Source material means‐‐

(1) Uranium or thorium or any combination of uranium and thorium in any physical or chemical form; or

(2) Ores that contain, by weight, one‐twentieth of 1 percent (0.05 percent), or more, of uranium, thorium, or any combination of uranium and thorium. Source material does not include special nuclear material.

Special nuclear material means‐‐

(1) Plutonium, uranium‐233, uranium enriched in the isotope 233 or in the isotope 235, and any other material that the Commission, pursuant to the provisions of section 51 of the Act, determines to be special nuclear material, but does not include source material; or

(2) Any material artificially enriched by any of the foregoing but does not include source material.

Stochastic effects means health effects that occur randomly and for which the probability of the effect occurring, rather than its severity, is assumed to be a linear function of dose without threshold. Hereditary effects and cancer incidence are examples of stochastic effects.

Supplied‐air respirator (SAR) or airline respirator means an atmosphere‐supplying respirator for which the source of breathing air is not designed to be carried by the user.

Survey means an evalulation of the radiological conditions and potential hazards incident to the production, use, transfer, release, disposal, or presence of radioactive material or other sources of radiation. When appropriate, such an evaluation includes a physical survey of the location of radioactive material and measurements or calculations of levels of radiation, or concentrations or quantities of radioactive material present.

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Tight‐fitting facepiece means a respiratory inlet covering that forms a complete seal with the face.

Total Effective Dose Equivalent (TEDE) means the sum of the deep‐dose equivalent (for external exposures) and the committed effective dose equivalent (for internal exposures).

Unrestricted area means an area, access to which is neither limited nor controlled by the licensee.

Uranium fuel cycle means the operations of milling of uranium ore, chemical conversion of uranium, isotopic enrichment of uranium, fabrication of uranium fuel, generation of electricity by a light‐water‐cooled nuclear power plant using uranium fuel, and reprocessing of spent uranium fuel to the extent that these activities directly support the production of electrical power for public use. Uranium fuel cycle does not include mining operations, operations at waste disposal sites, transportation of radioactive material in support of these operations, and the reuse of recovered non‐uranium special nuclear and byproduct materials from the cycle.

User seal check (fit check) means an action conducted by the respirator user to determine if the respirator is properly seated to the face. Examples include negative pressure check, positive pressure check, irritant smoke check, or isoamyl acetate check.

Very high radiation area means an area, accessible to individuals, in which radiation levels from radiation sources external to the body could result in an individual receiving an absorbed dose in excess of 500 rads (5 grays) in 1 hour at 1 meter from a radiation source or 1 meter from any surface that the radiation penetrates.

(Note: At very high doses received at high dose rates, units of absorbed dose (e.g., rads and grays) are appropriate, rather than units of dose equivalent (e.g., rems and sieverts)).

Week means 7 consecutive days starting on Sunday.

Weighting factor WT, for an organ or tissue (T) is the proportion of the risk of stochastic effects resulting from irradiation of that organ or tissue to the total risk of stochastic effects when the whole body is irradiated uniformly. For calculating the effective dose equivalent, the values of WT are:

Organ Dose Weighting Factors

Organ or Tissue WT

Gonads 0.25

Breast 0.15

Red bone marrow 0.12

Lung 0.12

Thyroid 0.03

Bone surfaces 0.03

Remainder 10.30

Whole Body 21.001 0.30 results from 0.06 for each of 5 ʺremainderʺ organs (excluding the skin and the lens of the eye) that receive the highest doses. 2 For the purpose of weighting the external whole body dose (for adding it to the internal dose), a single weighting factor, wT=1.0, has been specified. The use of other weighting factors for external exposure will be approved on a

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case‐by‐case basis until such time as specific guidance is issued.

Whole body means, for purposes of external exposure, head, trunk (including male gonads), arms above the elbow, or legs above the knee.

Working level (WL) is any combination of short‐lived radon daughters (for radon‐222: polonium‐218, lead‐214, bismuth‐214, and polonium‐214; and for radon‐220: polonium‐216, lead‐212, bismuth‐212, and polonium‐212) in 1 liter of air that will result in the ultimate emission of 1.3x105 MeV of potential alpha particle energy.

Working level month (WLM) means an exposure to 1 working level for 170 hours (2,000 working hours per year/12 months per year=approximately 170 hours per month).

Year means the period of time beginning in January used to determine compliance with the provisions of this part. The licensee may change the starting date of the year used to determine compliance by the licensee provided that the change is made at the beginning of the year and that no day is omitted or duplicated in consecutive years.

[56 FR 23391, May 21, 1991, as amended at 57 FR 57878, Dec. 8, 1992; 58 FR 7736, Feb. 9, 1993; 60 FR 36043, July 13, 1995; 60 FR 48625, Sept. 20, 1995; 61 FR 65127, Dec. 10, 1996; 62 FR 4133, Jan. 29, 1997; 62 FR 39087, July 21, 1997; 63 FR 39481, July 23, 1998; 64 FR 54556, Oct. 7, 1999; 66 FR 55789, Nov. 2, 2001; 67 FR 16304, Apr. 5, 2002; 67 FR 20370, Apr. 24, 2002; 67 FR 62872, Oct. 9, 2002]

x) § 20.1004 Units of radiation dose.

(a) Definitions. As used in this part, the units of radiation dose are:

Gray (Gy) is the SI unit of absorbed dose. One gray is equal to an absorbed dose of 1 Joule/kilogram (100 rads).

Rad is the special unit of absorbed dose. One rad is equal to an absorbed dose of 100 ergs/gram or 0.01 joule/kilogram (0.01 gray).

Rem is the special unit of any of the quantities expressed as dose equivalent. The dose equivalent in rems is equal to the absorbed dose in rads multiplied by the quality factor (1 rem=0.01 sievert).

Sievert is the SI unit of any of the quantities expressed as dose equivalent. The dose equivalent in sieverts is equal to the absorbed dose in grays multiplied by the quality factor (1 Sv=100 rems).

(b) As used in this part, the quality factors for converting absorbed dose to dose equivalent are shown in table 1004(b).1.

Table 1004(b).1‐Quality Factors and Absorbed Dose Equivalencies

Quality factor Type of radiation

(Q)

Absorbed dose equal to a unit dose equivalenta

X‐, gamma, or beta radiation 1 1

Alpha particles, multiple‐charged particles, fission fragments and heavy particles of unknown charge 20 0.05

Neutrons of unknown energy 10 0.1

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High‐energy protons 10 0.1a Absorbed dose in rad equal to 1 rem or the absorbed dose in gray equal to 1 sievert.

(c) If it is more convenient to measure the neutron fluence rate than to determine the neutron dose equivalent rate in rems per hour or sieverts per hour, as provided in paragraph (b) of this section, 1 rem (0.01 Sv) of neutron radiation of unknown energies may, for purposes of the regulations in this part, be assumed to result from a total fluence of 25 million neutrons per square centimeter incident upon the body. If sufficient information exists to estimate the approximate energy distribution of the neutrons, the licensee may use the fluence rate per unit dose equivalent or the appropriate Q value from table 1004(b).2 to convert a measured tissue dose in rads to dose equivalent in rems.

Table 1004(b).2.‐‐Mean Quality Factors, Q, and Fluence Per Unit Dose Equivalent for Monoenergetic Neutrons

Neutron energy (MeV) Quality factora (Q) Fluence per unit dose equivalentb (neutrons cm‐2 rem ‐1)

2.5 x 10‐8 2 980 x 106

1 x 10‐7 2 980 x 106

1 x 10‐6 2 810 x 106

1 x 10‐5 2 810 x 106

1 x 10‐4 2 840 x 106

1 x 10‐3 2 980 x 106

1 x 10‐2 2.5 1010 x 106

1 x 10‐1 7.5 170 x 106

5 x 10‐1 11 39 x 106

1 11 27 x 106

2.5 9 29 x 106

5 8 23 x 106

7 7 24 x 106

10 6.5 24 x 106

14 7.5 17 x 106

20 8 16 x 106

40 7 14 x 106

60 5.5 16 x 106

1 x 102 4 20 x 106

2 x 102 3.5 19 x 106

3 x 102 3.5 16 x 106

(thermal).....

4 x 102 3.5 14 x 106 a Value of quality factor (Q) at the point where the dose equivalent is maximum in a 30‐cm diameter cylinder tissue‐

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equivalent phantom. b Monoenergetic neutrons incident normally on a 30‐cm diameter cylinder tissue‐equivalent phantom.

y) § 20.1005 Units of radioactivity.

For the purposes of this part, activity is expressed in the special unit of curies (Ci) or in the SI unit of becquerels (Bq), or their multiples, or disintegrations (transformations) per unit of time.

(a) One becquerel=1 disintegration per second (s‐1).

(b) One curie=3.7x1010 disintegrations per second=3.7x1010 becquerels=2.22x1012 disintegrations per minute.

[56 FR 23391, May 21, 1991; 56 FR 61352, Dec. 3, 1991]

z) § 20.1006 Interpretations.

Except as specifically authorized by the Commission in writing, no interpretation of the meaning of the regulations in this part by an officer or employee of the Commission other than a written interpretation by the General Counsel will be recognized to be binding upon the Commission.

aa) § 20.1007 Communications.

Unless otherwise specified, communications or reports concerning the regulations in this part should be addressed to the Executive Director for Operations (EDO), and sent either by mail to the U.S. Nuclear Regulatory Commission, Washington, DC 20555‐0001; by hand delivery to the NRCʹs offices at 11555 Rockville Pike, Rockville, Maryland; or, where practicable, by electronic submission, for example, via Electronic Information Exchange, or CD‐ROM. Electronic submissions must be made in a manner that enables the NRC to receive, read, authenticate, distribute, and archive the submission, and process and retrieve it a single page at a time. Detailed guidance on making electronic submissions can be obtained by visiting the NRCʹs Web site at http://www.nrc.gov/site‐help/eie.html, by calling (301) 415‐6030, by e‐mail to [email protected], or by writing the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555‐0001. The guidance discusses, among other topics, the formats the NRC can accept, the use of electronic signatures, and the treatment of nonpublic information.

[68 FR 58801, Oct. 10, 2003]

bb) § 20.1008 Implementation.

(a) [Reserved]

(b) The applicable section of §§ 20.1001‐20.2402 must be used in lieu of requirements in the standards for protection against radiation in effect prior to January 1, 19941 that are cited in license conditions or technical specifications, except as specified in paragraphs (c), (d), and (e) of this section. If the requirements of this part are more restrictive than the existing license condition, then the licensee shall comply with this part unless exempted by paragraph (d) of this section.

(c) Any existing license condition or technical specification that is more restrictive than a requirement in §§ 20.1001‐20.2402 remains in force until there is a technical specification change, license amendment, or license renewal.

(d) If a license condition or technical specification exempted a licensee from a requirement in the standards for

82

protection against radiation in effect prior to January 1, 1994,1 it continues to exempt a licensee from the corresponding provision of §§ 20.1001‐20.2402.

(e) If a license condition cites provisions in requirements in the standards for protection against radiation in effect prior to January 1, 19941 and there are no corresponding provisions in §§ 20.1001‐20.2402, then the license condition remains in force until there is a technical specification change, license amendment, or license renewal that modifies or removes this condition.

[59 FR 41643, Aug. 15, 1994]

1 See §§ 20.1‐20.602 codified as of January 1, 1993.

cc) § 20.1009 Information collection requirements: OMB approval.

(a) The Nuclear Regulatory Commission has submitted the information collection requirements contained in this part to the Office of Management and Budget (OMB) for approval as required by the Paperwork Reduction Act (44 U.S.C. 3501 et seq.). The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number. OMB has approved the information collection requirements contained in this part under control number 3150‐0014.

(b) The approved information collection requirements contained in this part appear in §§ 20.1003, 20.1101, 20.1202, 20.1203, 20.1204, 20.1206, 20.1208, 20.1301, 20.1302, 20.1403, 20.1404, 20.1406, 20.1501, 20.1601, 20.1703, 20.1901, 20.1904, 20.1905, 20.1906, 20.2002, 20.2004, 20.2005, 20.2006, 20.2102, 20.2103, 20.2104, 20.2105, 20.2106, 20.2107, 20.2108, 20.2110, 20.2201, 20.2202, 20.2203, 20.2204, 20.2205, 20.2206, 20.2301, and appendix G to this part.

(c) This part contains information collection requirements in addition to those approved under the control number specified in paragraph (a) of this section. These information collection requirements and the control numbers under which they are approved are as follows:

(1) In § 20.2104, NRC Form 4 is approved under control number 3150‐0005.

(2) In §§ 20.2106 and 20.2206, NRC Form 5 is approved under control number 3150‐0006.

(3) In § 20.2006 and appendix G to 10 CFR Part 20, NRC Form 540 and 540A is approved under control number 3150‐0164.



[63 FR 50128, Sept. 21, 1998, as amended at 67 FR 67099, Nov. 4, 2002]

dd) Subpart B‐‐Radiation Protection Programs


ee) § 20.1101 Radiation protection programs.

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(a) Each licensee shall develop, document, and implement a radiation protection program commensurate with the scope and extent of licensed activities and sufficient to ensure compliance with the provisions of this part. (See § 20.2102 for recordkeeping requirements relating to these programs.)

(b) The licensee shall use, to the extent practical, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable (ALARA).

(c) The licensee shall periodically (at least annually) review the radiation protection program content and implementation.

(d) To implement the ALARA requirements of § 20.1101 (b), and notwithstanding the requirements in § 20.1301 of this part, a constraint on air emissions of radioactive material to the environment, excluding Radon‐222 and its daughters, shall be established by licensees other than those subject to § 50.34a, such that the individual member of the public likely to receive the highest dose will not be expected to receive a total effective dose equivalent in excess of 10 mrem (0.1 mSv) per year from these emissions. If a licensee subject to this requirement exceeds this dose constraint, the licensee shall report the exceedance as provided in § 20.2203 and promptly take appropriate corrective action to ensure against recurrence.

[56 FR 23396, May 21, 1991, as amended at 61 FR 65127, Dec. 10, 1996; 63 FR 39482, July 23, 1998]

ff) Subpart C‐‐Occupational Dose Limits


gg) § 20.1201 Occupational dose limits for adults.

(a) The licensee shall control the occupational dose to individual adults, except for planned special exposures under § 20.1206, to the following dose limits.

(1) An annual limit, which is the more limiting of‐‐

(i) The total effective dose equivalent being equal to 5 rems (0.05 Sv); or

(ii) The sum of the deep‐dose equivalent and the committed dose equivalent to any individual organ or tissue other than the lens of the eye being equal to 50 rems (0.5 Sv).

(2) The annual limits to the lens of the eye, to the skin of the whole body, and to the skin of the extremities, which are:

(i) A lens dose equivalent of 15 rems (0.15 Sv), and

(ii) A shallow‐dose equivalent of 50 rem (0.5 Sv) to the skin of the whole body or to the skin of any extremity.

(b) Doses received in excess of the annual limits, including doses received during accidents, emergencies, and planned special exposures, must be subtracted from the limits for planned special exposures that the individual may receive during the current year (see § 20.1206(e)(1)) and during the individualʹs lifetime (see § 20.1206(e)(2)).

(c) The assigned deep‐dose equivalent must be for the part of the body receiving the highest exposure. The assigned

84

shallow‐dose equivalent must be the dose averaged over the contiguous 10 square centimeters of skin receiving the highest exposure. The deep‐dose equivalent, lens‐dose equivalent, and shallow‐dose equivalent may be assessed from surveys or other radiation measurements for the purpose of demonstrating compliance with the occupational dose limits, if the individual monitoring device was not in the region of highest potential exposure, or the results of individual monitoring are unavailable.

(d) Derived air concentration (DAC) and annual limit on intake (ALI) values are presented in table 1 of appendix B to part 20 and may be used to determine the individualʹs dose (see § 20.2106) and to demonstrate compliance with the occupational dose limits.

(e) In addition to the annual dose limits, the licensee shall limit the soluble uranium intake by an individual to 10 milligrams in a week in consideration of chemical toxicity (see footnote 3 of appendix B to part 20).

(f) The licensee shall reduce the dose that an individual may be allowed to receive in the current year by the amount of occupational dose received while employed by any other person (see § 20.2104(e)).

[56 FR 23396, May 21, 1991, as amended at 60 FR 20185, Apr. 25, 1995; 63 FR 39482, July 23, 1998; 67 FR 16304, Apr. 5, 2002]

hh) § 20.1202 Compliance with requirements for summation of external and internal doses.

(a) If the licensee is required to monitor under both §§ 20.1502(a) and (b), the licensee shall demonstrate compliance with the dose limits by summing external and internal doses. If the licensee is required to monitor only under § 20.1502(a) or only under § 20.1502(b), then summation is not required to demonstrate compliance with the dose limits. The licensee may demonstrate compliance with the requirements for summation of external and internal doses by meeting one of the conditions specified in paragraph (b) of this section and the conditions in paragraphs (c) and (d) of this section.

(Note: The dose equivalents for the lens of the eye, the skin, and the extremities are not included in the summation, but are subject to separate limits.)

(b) Intake by inhalation. If the only intake of radionuclides is by inhalation, the total effective dose equivalent limit is not exceeded if the sum of the deep‐dose equivalent divided by the total effective dose equivalent limit, and one of the following, does not exceed unity:

(1) The sum of the fractions of the inhalation ALI for each radionuclide, or

(2) The total number of derived air concentration‐hours (DAC‐hours) for all radionuclides divided by 2,000, or

(3) The sum of the calculated committed effective dose equivalents to all significantly irradiated1 organs or tissues (T) calculated from bioassay data using appropriate biological models and expressed as a fraction of the annual limit.

(c) Intake by oral ingestion. If the occupationally exposed individual also receives an intake of radionuclides by oral ingestion greater than 10 percent of the applicable oral ALI, the licensee shall account for this intake and include it in demonstrating compliance with the limits.

(d) Intake through wounds or absorption through skin. The licensee shall evaluate and, to the extent practical, account for intakes through wounds or skin absorption.

Note: The intake through intact skin has been included in the calculation of DAC for hydrogen‐3 and does not need

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to be further evaluated.

[56 FR 23396, May 21, 1991, as amended at 57 FR 57878, Dec. 8, 1992]

1 An organ or tissue is deemed to be significantly irradiated if, for that organ or tissue, the product of the weighting factor, wT, and the committed dose equivalent, HT,50, per unit intake is greater than 10 percent of the maximum weighted value of HT,50, (i.e., wT HT,50) per unit intake for any organ or tissue.

ii) § 20.1203 Determination of external dose from airborne radioactive material.

Licensees shall, when determining the dose from airborne radioactive material, include the contribution to the deep‐dose equivalent, lens dose equivalent, and shallow‐dose equivalent from external exposure to the radioactive cloud (see appendix B to part 20, footnotes 1 and 2).

Note: Airborne radioactivity measurements and DAC values should not be used as the primary means to assess the deep‐dose equivalent when the airborne radioactive material includes radionuclides other than noble gases or if the cloud of airborne radioactive material is not relatively uniform. The determination of the deep‐dose equivalent to an individual should be based upon measurements using instruments or individual monitoring devices.

[56 FR 23396, May 21, 1991, as amended at 60 FR 20185, Apr. 25, 1995; 63 FR 39482, July 23, 1998]

jj) § 20.1204 Determination of internal exposure.

(a) For purposes of assessing dose used to determine compliance with occupational dose equivalent limits, the licensee shall, when required under § 20.1502, take suitable and timely measurements of‐‐

(1) Concentrations of radioactive materials in air in work areas; or

(2) Quantities of radionuclides in the body; or

(3) Quantities of radionuclides excreted from the body; or

(4) Combinations of these measurements.

(b) Unless respiratory protective equipment is used, as provided in § 20.1703, or the assessment of intake is based on bioassays, the licensee shall assume that an individual inhales radioactive material at the airborne concentration in which the individual is present.

(c) When specific information on the physical and biochemical properties of the radionuclides taken into the body or the behavior or the material in an individual is known, the licensee may‐‐

(1) Use that information to calculate the committed effective dose equivalent, and, if used, the licensee shall document that information in the individualʹs record; and

(2) Upon prior approval of the Commission, adjust the DAC or ALI values to reflect the actual physical and chemical characteristics of airborne radioactive material (e.g., aerosol size distribution or density); and

(3) Separately assess the contribution of fractional intakes of Class D, W, or Y compounds of a given radionuclide (see appendix B to part 20) to the committed effective dose equivalent.

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(d) If the licensee chooses to assess intakes of Class Y material using the measurements given in § 20.1204(a)(2) or (3), the licensee may delay the recording and reporting of the assessments for periods up to 7 months, unless otherwise required by §§ 20.2202 or 20.2203, in order to permit the licensee to make additional measurements basic to the assessments.

(e) If the identity and concentration of each radionuclide in a mixture are known, the fraction of the DAC applicable to the mixture for use in calculating DAC‐hours must be either‐‐

(1) The sum of the ratios of the concentration to the appropriate DAC value (e.g., D, W, Y) from appendix B to part 20 for each radionuclide in the mixture; or

(2) The ratio of the total concentration for all radionuclides in the mixture to the most restrictive DAC value for any radionuclide in the mixture.

(f) If the identity of each radionuclide in a mixture is known, but the concentration of one or more of the radionuclides in the mixture is not known, the DAC for the mixture must be the most restrictive DAC of any radionuclide in the mixture.

(g) When a mixture of radionuclides in air exists, licensees may disregard certain radionuclides in the mixture if‐‐

(1) The licensee uses the total activity of the mixture in demonstrating compliance with the dose limits in § 20.1201 and in complying with the monitoring requirements in § 20.1502(b), and

(2) The concentration of any radionuclide disregarded is less than 10 percent of its DAC, and

(3) The sum of these percentages for all of the radionuclides disregarded in the mixture does not exceed 30 percent.

(h)(1) In order to calculate the committed effective dose equivalent, the licensee may assume that the inhalation of one ALI, or an exposure of 2,000 DAC‐hours, results in a committed effective dose equivalent of 5 rems (0.05 Sv) for radionuclides that have their ALIs or DACs based on the committed effective dose equivalent.

(2) When the ALI (and the associated DAC) is determined by the nonstochastic organ dose limit of 50 rems (0.5 Sv), the intake of radionuclides that would result in a committed effective dose equivalent of 5 rems (0.05 Sv) (the stochastic ALI) is listed in parentheses in table 1 of appendix B to part 20. In this case, the licensee may, as a simplifying assumption, use the stochastic ALIs to determine committed effective dose equivalent. However, if the licensee uses the stochastic ALIs, the licensee must also demonstrate that the limit in § 20.1201(a)(1)(ii) is met.

[56 FR 23396, May 21, 1991, as amended at 60 FR 20185, Apr. 25, 1995]

kk) § 20.1205 [Reserved] ll) § 20.1206 Planned special exposures.

A licensee may authorize an adult worker to receive doses in addition to and accounted for separately from the doses received under the limits specified in § 20.1201 provided that each of the following conditions is satisfied‐‐

(a) The licensee authorizes a planned special exposure only in an exceptional situation when alternatives that might avoid the dose estimated to result from the planned special exposure are unavailable or impractical.

(b) The licensee (and employer if the employer is not the licensee) specifically authorizes the planned special exposure, in writing, before the exposure occurs.

87

(c) Before a planned special exposure, the licensee ensures that the individuals involved are‐‐

(1) Informed of the purpose of the planned operation;

(2) Informed of the estimated doses and associated potential risks and specific radiation levels or other conditions that might be involved in performing the task; and

(3) Instructed in the measures to be taken to keep the dose ALARA considering other risks that may be present.

(d) Prior to permitting an individual to participate in a planned special exposure, the licensee ascertains prior doses as required by § 20.2104(b) during the lifetime of the individual for each individual involved.

(e) Subject to § 20.1201(b), the licensee does not authorize a planned special exposure that would cause an individual to receive a dose from all planned special exposures and all doses in excess of the limits to exceed‐‐

(1) The numerical values of any of the dose limits in § 20.1201(a) in any year; and

(2) Five times the annual dose limits in § 20.1201(a) during the individualʹs lifetime.

(f) The licensee maintains records of the conduct of a planned special exposure in accordance with § 20.2105 and submits a written report in accordance with § 20.2204.

(g) The licensee records the best estimate of the dose resulting from the planned special exposure in the individualʹs record and informs the individual, in writing, of the dose within 30 days from the date of the planned special exposure. The dose from planned special exposures is not to be considered in controlling future occupational dose of the individual under § 20.1201(a) but is to be included in evaluations required by § 20.1206 (d) and (e).

[56 FR 23396, May 21, 1991, as amended at 63 FR 39482, July 23, 1998]

mm) § 20.1207 Occupational dose limits for minors.

The annual occupational dose limits for minors are 10 percent of the annual dose limits specified for adult workers in § 20.1201.

nn) § 20.1208 Dose equivalent to an embryo/fetus.

(a) The licensee shall ensure that the dose equivalent to the embryo/fetus during the entire pregnancy, due to the occupational exposure of a declared pregnant woman, does not exceed 0.5 rem (5 mSv). (For recordkeeping requirements, see § 20.2106.)

(b) The licensee shall make efforts to avoid substantial variation above a uniform monthly exposure rate to a declared pregnant woman so as to satisfy the limit in paragraph (a) of this section.

(c) The dose equivalent to the embryo/fetus is the sum of‐‐

(1) The deep‐dose equivalent to the declared pregnant woman; and

(2) The dose equivalent to the embryo/fetus resulting from radionuclides in the embryo/fetus and radionuclides in the declared pregnant woman.

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(d) If the dose equivalent to the embryo/fetus is found to have exceeded 0.5 rem (5 mSv), or is within 0.05 rem (0.5 mSv) of this dose, by the time the woman declares the pregnancy to the licensee, the licensee shall be deemed to be in compliance with paragraph (a) of this section if the additional dose equivalent to the embryo/fetus does not exceed 0.05 rem (0.5 mSv) during the remainder of the pregnancy.


oo) Subpart D‐‐Radiation Dose Limits for Individual Members of the Public


pp) § 20.1301 Dose limits for individual members of the public.

(a) Each licensee shall conduct operations so that —

(1) The total effective dose equivalent to individual members of the public from the licensed operation does not exceed 0.1 rem (1 mSv) in a year, exclusive of the dose contributions from background radiation, from any administration the individual has received, from exposure to individuals administered radioactive material and released under § 35.75, from voluntary participation in medical research programs, and from the licensee’s disposal of radioactive material into sanitary sewerage in accordance with § 20.2003, and

(2) The dose in any unrestricted area from external sources, exclusive of the dose contributions from patients administered radioactive material and released in accordance with § 35.75, does not exceed 0.002 rem (0.02 millisievert) in any one hour.

(b) If the licensee permits members of the public to have access to controlled areas, the limits for members of the public continue to apply to those individuals.

(c) Notwithstanding paragraph (a)(1) of this section, a licensee may permit visitors to an individual who cannot be released, under § 35.75, to receive a radiation dose greater than 0.1 rem (1 mSv) if—

(1) The radiation dose received does not exceed 0.5 rem (5 mSv); and

(2) The authorized user, as defined in 10 CFR Part 35, has determined before the visit that it is appropriate.

(d) A licensee or license applicant may apply for prior NRC authorization to operate up to an annual dose limit for an individual member of the public of 0.5 rem (5 mSv). The licensee or license applicant shall include the following information in this application:

(1) Demonstration of the need for and the expected duration of operations in excess of the limit in paragraph (a) of this section;

(2) The licenseeʹs program to assess and control dose within the 0.5 rem (5 mSv) annual limit; and

(3) The procedures to be followed to maintain the dose as low as is reasonably achievable.

(e) In addition to the requirements of this part, a licensee subject to the provisions of EPAʹs generally applicable environmental radiation standards in 40 CFR part 190 shall comply with those standards.

89

(f) The Commission may impose additional restrictions on radiation levels in unrestricted areas and on the total quantity of radionuclides that a licensee may release in effluents in order to restrict the collective dose.

[56 FR 23398, May 21, 1991, as amended at 60 FR 48625, Sept. 20, 1995; 62 FR 4133, Jan. 29, 1997; 67 FR 20370, Apr. 24, 2002; 67 FR 62872, Oct. 9, 2002]

qq) § 20.1302 Compliance with dose limits for individual members of the public.

(a) The licensee shall make or cause to be made, as appropriate, surveys of radiation levels in unrestricted and controlled areas and radioactive materials in effluents released to unrestricted and controlled areas to demonstrate compliance with the dose limits for individual members of the public in § 20.1301.

(b) A licensee shall show compliance with the annual dose limit in § 20.1301 by‐‐

(1) Demonstrating by measurement or calculation that the total effective dose equivalent to the individual likely to receive the highest dose from the licensed operation does not exceed the annual dose limit; or

(2) Demonstrating that‐‐

(i) The annual average concentrations of radioactive material released in gaseous and liquid effluents at the boundary of the unrestricted area do not exceed the values specified in table 2 of appendix B to part 20; and

(ii) If an individual were continuously present in an unrestricted area, the dose from external sources would not exceed 0.002 rem (0.02 mSv) in an hour and 0.05 rem (0.5 mSv) in a year.

(c) Upon approval from the Commission, the licensee may adjust the effluent concentration values in appendix B to part 20, table 2, for members of the public, to take into account the actual physical and chemical characteristics of the effluents (e.g., aerosol size distribution, solubility, density, radioactive decay equilibrium, chemical form).

[56 FR 23398, May 21, 1991; 56 FR 61352, Dec. 3, 1991, as amended at 57 FR 57878, Dec. 8, 1992; 60 FR 20185, Apr. 25, 1995]

rr) Subpart E‐‐Radiological Criteria for License Termination

Source: 62 FR 39088, July 21, 1987, unless otherwise noted.

ss) § 20.1401 General provisions and scope.

(a) The criteria in this subpart apply to the decommissioning of facilities licensed under Parts 30, 40, 50, 60, 61, 63, 70, and 72 of this chapter, and release of part of a facility or site for unrestricted use in accordance with § 50.83 of this chapter, as well as other facilities subject to the Commissionʹs jurisdiction under the Atomic Energy Act of 1954, as amended, and the Energy Reorganization Act of 1974, as amended. For high‐level and low‐level waste disposal facilities (10 CFR Parts 60, 61, 63), the criteria apply only to ancillary surface facilities that support radioactive waste disposal activities. The criteria do not apply to uranium and thorium recovery facilities already subject to Appendix A to 10 CFR Part 40 or to uranium solution extraction facilities.

(b) The criteria in this subpart do not apply to sites which:

(1) Have been decommissioned prior to the effective date of the rule in accordance with criteria identified in the Site

90

Decommissioning Management Plan (SDMP) Action Plan of April 16, 1992 (57 FR 13389);

(2) Have previously submitted and received Commission approval on a license termination plan (LTP) or decommissioning plan that is compatible with the SDMP Action Plan criteria; or

(3) Submit a sufficient LTP or decommissioning plan before August 20, 1998 and such LTP or decommissioning plan is approved by the Commission before August 20, 1999 and in accordance with the criteria identified in the SDMP Action Plan, except that if an EIS is required in the submittal, there will be a provision for day‐for‐day extension.

(c) After a site has been decommissioned and the license terminated in accordance with the criteria in this subpart, or after part of a facility or site has been released for unrestricted use in accordance with § 50.83 of this chapter and in accordance with the criteria in this subpart, the Commission will require additional cleanup only, if based on new information, it determines that the criteria of this subpart were not met and residual radioactivity remaining at the site could result in significant threat to public health and safety.

(d) When calculating TEDE to the average member of the critical group the licensee shall determine the peak annual TEDE dose expected within the first 1000 years after decommissioning.

[62 FR 39088, July 21, 1997, as amended at 66 FR 55789, Nov. 2, 2001]

tt) § 20.1402 Radiological criteria for unrestricted use.

A site will be considered acceptable for unrestricted use if the residual radioactivity that is distinguishable from background radiation results in a TEDE to an average member of the critical group that does not exceed 25 mrem (0.25 mSv) per year, including that from groundwater sources of drinking water, and the residual radioactivity has been reduced to levels that are as low as reasonably achievable (ALARA). Determination of the levels which are ALARA must take into account consideration of any detriments, such as deaths from transportation accidents, expected to potentially result from decontamination and waste disposal.

uu) § 20.1403 Criteria for license termination under restricted conditions.

A site will be considered acceptable for license termination under restricted conditions if:

(a) The licensee can demonstrate that further reductions in residual radioactivity necessary to comply with the provisions of § 20.1402 would result in net public or environmental harm or were not being made because the residual levels associated with restricted conditions are ALARA. Determination of the levels which are ALARA must take into account consideration of any detriments, such as traffic accidents, expected to potentially result from decontamination and waste disposal;

(b) The licensee has made provisions for legally enforceable institutional controls that provide reasonable assurance that the TEDE from residual radioactivity distinguishable from background to the average member of the critical group will not exceed 25 mrem (0.25 mSv) per year;

(c) The licensee has provided sufficient financial assurance to enable an independent third party, including a governmental custodian of a site, to assume and carry out responsibilities for any necessary control and maintenance of the site. Acceptable financial assurance mechanisms are‐‐

(1) Funds placed into an account segregated from the licenseeʹs assets and outside the licenseeʹs administrative control as described in § 30.35(f)(1) of this chapter;

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(2) Surety method, insurance, or other guarantee method as described in § 30.35(f)(2) of this chapter;

(3) A statement of intent in the case of Federal, State, or local Government licensees, as described in § 30.35(f)(4) of this chapter; or

(4) When a governmental entity is assuming custody and ownership of a site, an arrangement that is deemed acceptable by such governmental entity.

(d) The licensee has submitted a decommissioning plan or License Termination Plan (LTP) to the Commission indicating the licenseeʹs intent to decommission in accordance with §§ 30.36(d), 40.42(d), 50.82 (a) and (b), 70.38(d), or 72.54 of this chapter, and specifying that the licensee intends to decommission by restricting use of the site. The licensee shall document in the LTP or decommissioning plan how the advice of individuals and institutions in the community who may be affected by the decommissioning has been sought and incorporated, as appropriate, following analysis of that advice.

(1) Licensees proposing to decommission by restricting use of the site shall seek advice from such affected parties regarding the following matters concerning the proposed decommissioning‐‐

(i) Whether provisions for institutional controls proposed by the licensee;

(A) Will provide reasonable assurance that the TEDE from residual radioactivity distinguishable from background to the average member of the critical group will not exceed 25 mrem (0.25 mSv) TEDE per year;

(B) Will be enforceable; and

(C) Will not impose undue burdens on the local community or other affected parties.

(ii) Whether the licensee has provided sufficient financial assurance to enable an independent third party, including a governmental custodian of a site, to assume and carry out responsibilities for any necessary control and maintenance of the site;

(2) In seeking advice on the issues identified in § 20.1403(d)(1), the licensee shall provide for:

(i) Participation by representatives of a broad cross section of community interests who may be affected by the decommissioning;

(ii) An opportunity for a comprehensive, collective discussion on the issues by the participants represented; and

(iii) A publicly available summary of the results of all such discussions, including a description of the individual viewpoints of the participants on the issues and the extent of agreement and disagreement among the participants on the issues; and

(e) Residual radioactivity at the site has been reduced so that if the institutional controls were no longer in effect, there is reasonable assurance that the TEDE from residual radioactivity distinguishable from background to the average member of the critical group is as low as reasonably achievable and would not exceed either‐‐

(1) 100 mrem (1 mSv) per year; or

(2) 500 mrem (5 mSv) per year provided the licensee‐‐

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(i) Demonstrates that further reductions in residual radioactivity necessary to comply with the 100 mrem/y (1 mSv/y) value of paragraph (e)(1) of this section are not technically achievable, would be prohibitively expensive, or would result in net public or environmental harm;

(ii) Makes provisions for durable institutional controls;

(iii) Provides sufficient financial assurance to enable a responsible government entity or independent third party, including a governmental custodian of a site, both to carry out periodic rechecks of the site no less frequently than every 5 years to assure that the institutional controls remain in place as necessary to meet the criteria of § 20.1403(b) and to assume and carry out responsibilities for any necessary control and maintenance of those controls. Acceptable financial assurance mechanisms are those in paragraph (c) of this section.

vv) § 20.1404 Alternate criteria for license termination.

(a) The Commission may terminate a license using alternate criteria greater than the dose criterion of §§ 20.1402, 20.1403(b), and 20.1403(d)(1)(i)(A), if the licensee‐‐

(1) Provides assurance that public health and safety would continue to be protected, and that it is unlikely that the dose from all man‐made sources combined, other than medical, would be more than the 1 mSv/y (100 mrem/y) limit of subpart D, by submitting an analysis of possible sources of exposure;

(2) Has employed to the extent practical restrictions on site use according to the provisions of § 20.1403 in minimizing exposures at the site; and

(3) Reduces doses to ALARA levels, taking into consideration any detriments such as traffic accidents expected to potentially result from decontamination and waste disposal.

(4) Has submitted a decommissioning plan or License Termination Plan (LTP) to the Commission indicating the licenseeʹs intent to decommission in accordance with §§ 30.36(d), 40.42(d), 50.82 (a) and (b), 70.38(d), or 72.54 of this chapter, and specifying that the licensee proposes to decommission by use of alternate criteria. The licensee shall document in the decommissioning plan or LTP how the advice of individuals and institutions in the community who may be affected by the decommissioning has been sought and addressed, as appropriate, following analysis of that advice. In seeking such advice, the licensee shall provide for:

(i) Participation by representatives of a broad cross section of community interests who may be affected by the decommissioning;

(ii) An opportunity for a comprehensive, collective discussion on the issues by the participants represented; and

(iii) A publicly available summary of the results of all such discussions, including a description of the individual viewpoints of the participants on the issues and the extent of agreement and disagreement among the participants on the issues.

(b) The use of alternate criteria to terminate a license requires the approval of the Commission after consideration of the NRC staffʹs recommendations that will address any comments provided by the Environmental Protection Agency and any public comments submitted pursuant to § 20.1405.

ww) § 20.1405 Public notification and public participation.

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site pursuant to §§ 20.1403 or 20.1404, or whenever the Commission deems such notice to be in the public interest, the Commission shall:

(a) Notify and solicit comments from:

(1) local and State governments in the vicinity of the site and any Indian Nation or other indigenous people that have treaty or statutory rights that could be affected by the decommissioning; and

(2) the Environmental Protection Agency for cases where the licensee proposes to release a site pursuant to § 20.1404.

(b) Publish a notice in the Federal Register and in a forum, such as local newspapers, letters to State or local organizations, or other appropriate forum, that is readily accessible to individuals in the vicinity of the site, and solicit comments from affected parties.

xx) § 20.1406 Minimization of contamination.

Applicants for licenses, other than renewals, after August 20, 1997, shall describe in the application how facility design and procedures for operation will minimize, to the extent practicable, contamination of the facility and the environment, facilitate eventual decommissioning, and minimize, to the extent practicable, the generation of radioactive waste.

yy) Subpart F‐‐Surveys and Monitoring


zz) § 20.1501 General.

(a) Each licensee shall make or cause to be made, surveys that‐‐

(1) May be necessary for the licensee to comply with the regulations in this part; and

(2) Are reasonable under the circumstances to evaluate‐‐

(i) The magnitude and extent of radiation levels; and

(ii) Concentrations or quantities of radioactive material; and

(iii) The potential radiological hazards.

(b) The licensee shall ensure that instruments and equipment used for quantitative radiation measurements (e.g., dose rate and effluent monitoring) are calibrated periodically for the radiation measured.

(c) All personnel dosimeters (except for direct and indirect reading pocket ionization chambers and those dosimeters used to measure the dose to the extremities) that require processing to determine the radiation dose and that are used by licensees to comply with § 20.1201, with other applicable provisions of this chapter, or with conditions specified in a license must be processed and evaluated by a dosimetry processor‐‐

(1) Holding current personnel dosimetry accreditation from the National Voluntary Laboratory Accreditation Program (NVLAP) of the National Institute of Standards and Technology; and

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(2) Approved in this accreditation process for the type of radiation or radiations included in the NVLAP program that most closely approximates the type of radiation or radiations for which the individual wearing the dosimeter is monitored.


aaa) § 20.1502 Conditions requiring individual monitoring of external and internal occupational dose.

Each licensee shall monitor exposures to radiation and radioactive material at levels sufficient to demonstrate compliance with the occupational dose limits of this part. As a minimum‐‐

(a) Each licensee shall monitor occupational exposure to radiation from licensed and unlicensed radiation sources under the control of the licensee and shall supply and require the use of individual monitoring devices by‐‐

(1) Adults likely to receive, in 1 year from sources external to the body, a dose in excess of 10 percent of the limits in § 20.1201(a),

(2) Minors likely to receive, in 1 year, from radiation sources external to the body, a deep dose equivalent in excess of 0.1 rem (1 mSv), a lens dose equivalent in excess of 0.15 rem (1.5 mSv), or a shallow dose equivalent to the skin or to the extremities in excess of 0.5 rem (5 mSv);

(3) Declared pregnant women likely to receive during the entire pregnancy, from radiation sources external to the body, a deep dose equivalent in excess of 0.1 rem (1 mSv);2 and

(4) Individuals entering a high or very high radiation area.

(b) Each licensee shall monitor (see § 20.1204) the occupational intake of radioactive material by and assess the committed effective dose equivalent to‐‐

(1) Adults likely to receive, in 1 year, an intake in excess of 10 percent of the applicable ALI(s) in table 1, Columns 1 and 2, of appendix B to §§ 20.1001‐20.2402;

(2) Minors likely to receive, in 1 year, a committed effective dose equivalent in excess of 0.1 rem (1 mSv); and

(3) Declared pregnant women likely to receive, during the entire pregnancy, a committed effective dose equivalent in excess of 0.1 rem (1 mSv).


2 All of the occupational doses in § 20.1201 continue to be applicable to the declared pregnant worker as long as the embryo/fetus dose limit is not exceeded.

bbb) Subpart G‐‐Control of Exposure From External Sources in Restricted Areas


ccc) § 20.1601 Control of access to high radiation areas.

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(a) The licensee shall ensure that each entrance or access point to a high radiation area has one or more of the following features‐‐

(1) A control device that, upon entry into the area, causes the level of radiation to be reduced below that level at which an individual might receive a deep‐dose equivalent of 0.1 rem (1 mSv) in 1 hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates;

(2) A control device that energizes a conspicuous visible or audible alarm signal so that the individual entering the high radiation area and the supervisor of the activity are made aware of the entry; or

(3) Entryways that are locked, except during periods when access to the areas is required, with positive control over each individual entry.

(b) In place of the controls required by paragraph (a) of this section for a high radiation area, the licensee may substitute continuous direct or electronic surveillance that is capable of preventing unauthorized entry.

(c) A licensee may apply to the Commission for approval of alternative methods for controlling access to high radiation areas.

(d) The licensee shall establish the controls required by paragraphs (a) and (c) of this section in a way that does not prevent individuals from leaving a high radiation area.

(e) Control is not required for each entrance or access point to a room or other area that is a high radiation area solely because of the presence of radioactive materials prepared for transport and packaged and labeled in accordance with the regulations of the Department of Transportation provided that‐‐

(1) The packages do not remain in the area longer than 3 days; and

(2) The dose rate at 1 meter from the external surface of any package does not exceed 0.01 rem (0.1 mSv) per hour.

(f) Control of entrance or access to rooms or other areas in hospitals is not required solely because of the presence of patients containing radioactive material, provided that there are personnel in attendance who will take the necessary precautions to prevent the exposure of individuals to radiation or radioactive material in excess of the limits established in this part and to operate within the ALARA provisions of the licenseeʹs radiation protection program.

ddd) § 20.1602 Control of access to very high radiation areas.

In addition to the requirements in § 20.1601, the licensee shall institute additional measures to ensure that an individual is not able to gain unauthorized or inadvertent access to areas in which radiation levels could be encountered at 500 rads (5 grays) or more in 1 hour at 1 meter from a radiation source or any surface through which the radiation penetrates.

eee) Subpart H‐‐Respiratory Protection and Controls to Restrict Internal Exposure in Restricted Areas


fff) § 20.1701 Use of process or other engineering controls.

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The licensee shall use, to the extent practical, process or other engineering controls (e.g., containment, decontamination, or ventilation) to control the concentration of radioactive material in air.

[64 FR 54556, Oct. 7, 1999]

ggg) § 20.1702 Use of other controls.

(a) When it is not practical to apply process or other engineering controls to control the concentrations of radioactive material in the air to values below those that define an airborne radioactivity area, the licensee shall, consistent with maintaining the total effective dose equivalent ALARA, increase monitoring and limit intakes by one or more of the following means‐‐

(1) Control of access;

(2) Limitation of exposure times;

(3) Use of respiratory protection equipment; or

(4) Other controls.

(b) If the licensee performs an ALARA analysis to determine whether or not respirators should be used, the licensee may consider safety factors other than radiological factors. The licensee should also consider the impact of respirator use on workersʹ industrial health and safety.

[64 FR 54556, Oct. 7, 1999]

hhh) § 20.1703 Use of individual respiratory protection equipment.

If the licensee assigns or permits the use of respiratory protection equipment to limit the intake of radioactive material,

(a) The licensee shall use only respiratory protection equipment that is tested and certified by the National Institute for Occupational Safety and Health (NIOSH) except as otherwise noted in this part.

(b) If the licensee wishes to use equipment that has not been tested or certified by NIOSH, or for which there is no schedule for testing or certification, the licensee shall submit an application to the NRC for authorized use of this equipment except as provided in this part. The application must include evidence that the material and performance characteristics of the equipment are capable of providing the proposed degree of protection under anticipated conditions of use. This must be demonstrated either by licensee testing or on the basis of reliable test information.

(c) The licensee shall implement and maintain a respiratory protection program that includes:

(1) Air sampling sufficient to identify the potential hazard, permit proper equipment selection, and estimate doses;

(2) Surveys and bioassays, as necessary, to evaluate actual intakes;

(3) Testing of respirators for operability (user seal check for face sealing devices and functional check for others) immediately prior to each use;

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(4) Written procedures regarding‐‐

(i) Monitoring, including air sampling and bioassays;

(ii) Supervision and training of respirator users;

(iii) Fit testing;

(iv) Respirator selection;

(v) Breathing air quality;

(vi) Inventory and control;

(vii) Storage, issuance, maintenance, repair, testing, and quality assurance of respiratory protection equipment;

(viii) Recordkeeping; and

(ix) Limitations on periods of respirator use and relief from respirator use;

(5) Determination by a physician that the individual user is medically fit to use respiratory protection equipment:

(i) Before the initial fitting of a face sealing respirator;

(ii) Before the first field use of non‐face sealing respirators, and

(iii) Either every 12 months thereafter, or periodically at a frequency determined by a physician.

(6) Fit testing, with fit factor > 10 times the APF for negative pressure devices, and a fit factor > 500 for any positive pressure, continuous flow, and pressure‐demand devices, before the first field use of tight fitting, face‐sealing respirators and periodically thereafter at a frequency not to exceed 1 year. Fit testing must be performed with the facepiece operating in the negative pressure mode.

(d) The licensee shall advise each respirator user that the user may leave the area at any time for relief from respirator use in the event of equipment malfunction, physical or psychological distress, procedural or communication failure, significant deterioration of operating conditions, or any other conditions that might require such relief.

(e) The licensee shall also consider limitations appropriate to the type and mode of use. When selecting respiratory devices the licensee shall provide for vision correction, adequate communication, low temperature work environments, and the concurrent use of other safety or radiological protection equipment. The licensee shall use equipment in such a way as not to interfere with the proper operation of the respirator.

(f) Standby rescue persons are required whenever one‐piece atmosphere‐supplying suits, or any combination of supplied air respiratory protection device and personnel protective equipment are used from which an unaided individual would have difficulty extricating himself or herself. The standby persons must be equipped with respiratory protection devices or other apparatus appropriate for the potential hazards. The standby rescue persons shall observe or otherwise maintain continuous communication with the workers (visual, voice, signal line, telephone, radio, or other suitable means), and be immediately available to assist them in case of a failure of the air supply or for any other reason that requires relief from distress. A sufficient number of standby rescue persons must be immediately available to assist all users of this type of equipment and to provide effective emergency rescue if

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needed.

(g) Atmosphere‐supplying respirators must be supplied with respirable air of grade D quality or better as defined by the Compressed Gas Association in publication G‐7.1, ʺCommodity Specification for Air,ʺ 1997 and included in the regulations of the Occupational Safety and Health Administration (29 CFR 1910.134(i)(1)(ii)(A) through (E). Grade D quality air criteria include‐‐

(1) Oxygen content (v/v) of 19.5‐23.5%;

(2) Hydrocarbon (condensed) content of 5 milligrams per cubic meter of air or less;

(3) Carbon monoxide (CO) content of 10 ppm or less;

(4) Carbon dioxide content of 1,000 ppm or less; and

(5) Lack of noticable odor.

(h) The licensee shall ensure that no objects, materials or substances, such as facial hair, or any conditions that interfere with the face‐‐facepiece seal or valve function, and that are under the control of the respirator wearer, are present between the skin of the wearerʹs face and the sealing surface of a tight‐fitting respirator facepiece.

(i) In estimating the dose to individuals from intake of airborne radioactive materials, the concentration of radioactive material in the air that is inhaled when respirators are worn is initially assumed to be the ambient concentration in air without respiratory protection, divided by the assigned protection factor. If the dose is later found to be greater than the estimated dose, the corrected value must be used. If the dose is later found to be less than the estimated dose, the corrected value may be used.

[64 FR 54557, Oct. 7, 1999, as amended at 67 FR 77652, Dec. 19, 2002]

iii) § 20.1704 Further restrictions on the use of respiratory protection equipment.

The Commission may impose restrictions in addition to the provisions of §§ 20.1702, 20.1703, and Appendix A to Part 20, in order to:

(a) Ensure that the respiratory protection program of the licensee is adequate to limit doses to individuals from intakes of airborne radioactive materials consistent with maintaining total effective dose equivalent ALARA; and

(b) Limit the extent to which a licensee may use respiratory protection equipment instead of process or other engineering controls.

[64 FR 54557, Oct. 7, 1999]

jjj) § 20.1705 Application for use of higher assigned protection factors.

The licensee shall obtain authorization from the Commission before using assigned protection factors in excess of those specified in Appendix A to Part 20. The Commission may authorize a licensee to use higher assigned protection factors on receipt of an application that‐‐

(a) Describes the situation for which a need exists for higher protection factors; and

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(b) Demonstrates that the respiratory protection equipment provides these higher protection factors under the proposed conditions of use.

[64 FR 54557, Oct. 7, 1999]

kkk) Subpart I‐‐Storage and Control of Licensed Material


lll) § 20.1801 Security of stored material.

The licensee shall secure from unauthorized removal or access licensed materials that are stored in controlled or unrestricted areas.

mmm) § 20.1802 Control of material not in storage.

The licensee shall control and maintain constant surveillance of licensed material that is in a controlled or unrestricted area and that is not in storage.

nnn) Subpart J‐‐Precautionary Procedures


ooo) § 20.1901 Caution signs.

(a) Standard radiation symbol. Unless otherwise authorized by the Commission, the symbol prescribed by this part shall use the colors magenta, or purple, or black on yellow background. The symbol prescribed by this part is the three‐bladed design:

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(1) Cross‐hatched area is to be magenta, or purple, or black, and

(2) The background is to be yellow.

(b) Exception to color requirements for standard radiation symbol. Notwithstanding the requirements of paragraph (a) of this section, licensees are authorized to label sources, source holders, or device components containing sources of licensed materials that are subjected to high temperatures, with conspicuously etched or stamped radiation caution symbols and without a color requirement.

(c) Additional information on signs and labels. In addition to the contents of signs and labels prescribed in this part, the licensee may provide, on or near the required signs and labels, additional information, as appropriate, to make individuals aware of potential radiation exposures and to minimize the exposures.

ppp) § 20.1902 Posting requirements.

(a) Posting of radiation areas. The licensee shall post each radiation area with a conspicuous sign or signs bearing the radiation symbol and the words ʺCAUTION, RADIATION AREA.ʺ

(b) Posting of high radiation areas. The licensee shall post each high radiation area with a conspicuous sign or signs bearing the radiation symbol and the words ʺCAUTION, HIGH RADIATION AREAʺ or ʺDANGER, HIGH RADIATION AREA.ʺ

(c) Posting of very high radiation areas. The licensee shall post each very high radiation area with a conspicuous sign or signs bearing the radiation symbol and words ʺGRAVE DANGER, VERY HIGH RADIATION AREA.ʺ

(d) Posting of airborne radioactivity areas. The licensee shall post each airborne radioactivity area with a conspicuous sign or signs bearing the radiation symbol and the words ʺCAUTION, AIRBORNE RADIOACTIVITY AREAʺ or

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ʺDANGER, AIRBORNE RADIOACTIVITY AREA.ʺ

(e) Posting of areas or rooms in which licensed material is used or stored. The licensee shall post each area or room in which there is used or stored an amount of licensed material exceeding 10 times the quantity of such material specified in appendix C to part 20 with a conspicuous sign or signs bearing the radiation symbol and the words ʺCAUTION, RADIOACTIVE MATERIAL(S)ʺ or ʺDANGER, RADIOACTIVE MATERIAL(S).ʺ


qqq) § 20.1903 Exceptions to posting requirements.

(a) A licensee is not required to post caution signs in areas or rooms containing radioactive materials for periods of less than 8 hours, if each of the following conditions is met:

(1) The materials are constantly attended during these periods by an individual who takes the precautions necessary to prevent the exposure of individuals to radiation or radioactive materials in excess of the limits established in this part; and

(2) The area or room is subject to the licenseeʹs control.

(b) Rooms or other areas in hospitals that are occupied by patients are not required to be posted with caution signs pursuant to § 20.1902 provided that the patient could be released from licensee control pursuant to § 35.75 of this chapter.

(c) A room or area is not required to be posted with a caution sign because of the presence of a sealed source provided the radiation level at 30 centimeters from the surface of the source container or housing does not exceed 0.005 rem (0.05 mSv) per hour.

(d) Rooms in hospitals or clinics that are used for teletherapy are exempt from the requirement to post caution signs under § 20.1902 if‐‐

(1) Access to the room is controlled pursuant to 10 CFR 35.615; and

(2) Personnel in attendance take necessary precautions to prevent the inadvertent exposure of workers, other patients, and members of the public to radiation in excess of the limits established in this part.

[56 FR 23401, May 21, 1991, as amended at 57 FR 39357, Aug. 31, 1992; 62 FR 4133, Jan. 29, 1997; 63 FR 39482, July 23, 1998]

rrr) § 20.1904 Labeling containers.

(a) The licensee shall ensure that each container of licensed material bears a durable, clearly visible label bearing the radiation symbol and the words ʺCAUTION, RADIOACTIVE MATERIALʺ or ʺDANGER, RADIOACTIVE MATERIAL.ʺ The label must also provide sufficient information (such as the radionuclide(s) present, an estimate of the quantity of radioactivity, the date for which the activity is estimated, radiation levels, kinds of materials, and mass enrichment) to permit individuals handling or using the containers, or working in the vicinity of the containers, to take precautions to avoid or minimize exposures.

(b) Each licensee shall, prior to removal or disposal of empty uncontaminated containers to unrestricted areas, remove or deface the radioactive material label or otherwise clearly indicate that the container no longer contains

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radioactive materials.

sss) § 20.1905 Exemptions to labeling requirements.

A licensee is not required to label‐‐

(a) Containers holding licensed material in quantities less than the quantities listed in appendix C to part 20; or

(b) Containers holding licensed material in concentrations less than those specified in table 3 of appendix B to part 20; or

(c) Containers attended by an individual who takes the precautions necessary to prevent the exposure of individuals in excess of the limits established by this part; or

(d) Containers when they are in transport and packaged and labeled in accordance with the regulations of the Department of Transportation,3 or

(e) Containers that are accessible only to individuals authorized to handle or use them, or to work in the vicinity of the containers, if the contents are identified to these individuals by a readily available written record (examples of containers of this type are containers in locations such as water‐filled canals, storage vaults, or hot cells). The record must be retained as long as the containers are in use for the purpose indicated on the record; or

(f) Installed manufacturing or process equipment, such as reactor components, piping, and tanks.


3 Labeling of packages containing radioactive materials is required by the Department of Transportation (DOT) if the amount and type of radioactive material exceeds the limits for an excepted quantity or article as defined and limited by DOT regulations 49 CFR 173.403 (m) and (w) and 173.421‐424.

ttt) § 20.1906 Procedures for receiving and opening packages.

(a) Each licensee who expects to receive a package containing quantities of radioactive material in excess of a Type A quantity, as defined in § 71.4 and appendix A to part 71 of this chapter, shall make arrangements to receive‐‐

(1) The package when the carrier offers it for delivery; or

(2) Notification of the arrival of the package at the carrierʹs terminal and to take possession of the package expeditiously.

(b) Each licensee shall‐‐

(1) Monitor the external surfaces of a labeled3a package for radioactive contamination unless the package contains only radioactive material in the form of a gas or in special form as defined in 10 CFR 71.4;

(2) Monitor the external surfaces of a labeled3a package for radiation levels unless the package contains quantities of radioactive material that are less than or equal to the Type A quantity, as defined in § 71.4 and appendix A to part 71 of this chapter; and

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(3) Monitor all packages known to contain radioactive material for radioactive contamination and radiation levels if there is evidence of degradation of package integrity, such as packages that are crushed, wet, or damaged.

(c) The licensee shall perform the monitoring required by paragraph (b) of this section as soon as practical after receipt of the package, but not later than 3 hours after the package is received at the licenseeʹs facility if it is received during the licenseeʹs normal working hours, or not later than 3 hours from the beginning of the next working day if it is received after working hours.

(d) The licensee shall immediately notify the final delivery carrier and the NRC Operations Center (301‐816‐5100), by telephone, when‐‐

(1) Removable radioactive surface contamination exceeds the limits of § 71.87(i) of this chapter; or

(2) External radiation levels exceed the limits of § 71.47 of this chapter.

(e) Each licensee shall‐‐

(1) Establish, maintain, and retain written procedures for safely opening packages in which radioactive material is received; and

(2) Ensure that the procedures are followed and that due consideration is given to special instructions for the type of package being opened.

(f) Licensees transferring special form sources in licensee‐owned or licensee‐operated vehicles to and from a work site are exempt from the contamination monitoring requirements of paragraph (b) of this section, but are not exempt from the survey requirement in paragraph (b) of this section for measuring radiation levels that is required to ensure that the source is still properly lodged in its shield.

[56 FR 23401, May 21, 1991, as amended at 57 FR 39357, Aug. 31, 1992; 60 FR 20185, Apr. 25, 1995; 63 FR 39482, July 23, 1998]

3a Labeled with a Radioactive White I, Yellow II, or Yellow III label as specified in U.S. Department of Transportation regulations, 49 CFR 172.403 and 172.436‐440.

uuu) Subpart K‐‐Waste Disposal


vvv) § 20.2001 General requirements.

(a) A licensee shall dispose of licensed material only‐‐

(1) By transfer to an authorized recipient as provided in § 20.2006 or in the regulations in parts 30, 40, 60, 61, 63, 70, and 72 of this chapter;

(2) By decay in storage; or

(3) By release in effluents within the limits in § 20.1301; or

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(4) As authorized under §§ 20.2002, 20.2003, 20.2004, or § 20.2005.

(b) A person must be specifically licensed to receive waste containing licensed material from other persons for:

(1) Treatment prior to disposal; or

(2) Treatment or disposal by incineration; or

(3) Decay in storage; or

(4) Disposal at a land disposal facility licensed under part 61 of this chapter; or

(5) Disposal at a geologic repository under part 60 or part 63 of this chapter.

[56 FR 23403, May 21, 1991, as amended at 66 FR 55789, Nov. 2, 2001]

www) § 20.2002 Method for obtaining approval of proposed disposal procedures.

A licensee or applicant for a license may apply to the Commission for approval of proposed procedures, not otherwise authorized in the regulations in this chapter, to dispose of licensed material generated in the licenseeʹs activities. Each application shall include:

(a) A description of the waste containing licensed material to be disposed of, including the physical and chemical properties important to risk evaluation, and the proposed manner and conditions of waste disposal; and

(b) An analysis and evaluation of pertinent information on the nature of the environment; and

(c) The nature and location of other potentially affected licensed and unlicensed facilities; and

(d) Analyses and procedures to ensure that doses are maintained ALARA and within the dose limits in this part.

xxx) § 20.2003 Disposal by release into sanitary sewerage.

(a) A licensee may discharge licensed material into sanitary sewerage if each of the following conditions is satisfied:

(1) The material is readily soluble (or is readily dispersible biological material) in water; and

(2) The quantity of licensed or other radioactive material that the licensee releases into the sewer in 1 month divided by the average monthly volume of water released into the sewer by the licensee does not exceed the concentration listed in table 3 of appendix B to part 20; and

(3) If more than one radionuclide is released, the following conditions must also be satisfied:

(i) The licensee shall determine the fraction of the limit in table 3 of appendix B to part 20 represented by discharges into sanitary sewerage by dividing the actual monthly average concentration of each radionuclide released by the licensee into the sewer by the concentration of that radionuclide listed in table 3 of appendix B to part 20; and

(ii) The sum of the fractions for each radionuclide required by paragraph (a)(3)(i) of this section does not exceed unity; and

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(4) The total quantity of licensed and other radioactive material that the licensee releases into the sanitary sewerage system in a year does not exceed 5 curies (185 GBq) of hydrogen‐3, 1 curie (37 GBq) of carbon‐14, and 1 curie (37 GBq) of all other radioactive materials combined.

(b) Excreta from individuals undergoing medical diagnosis or therapy with radioactive material are not subject to the limitations contained in paragraph (a) of this section.


yyy) § 20.2004 Treatment or disposal by incineration.

(a) A licensee may treat or dispose of licensed material by incineration only:

(1) As authorized by paragraph (b) of this section; or

(2) If the material is in a form and concentration specified in § 20.2005; or

(3) As specifically approved by the Commission pursuant to § 20.2002.

(b) (1) Waste oils (petroleum derived or synthetic oils used principally as lubricants, coolants, hydraulic or insulating fluids, or metalworking oils) that have been radioactively contaminated in the course of the operation or maintenance of a nuclear power reactor licensed under part 50 of this chapter may be incinerated on the site where generated provided that the total radioactive effluents from the facility, including the effluents from such incineration, conform to the requirements of appendix I to part 50 of this chapter and the effluent release limits contained in applicable license conditions other than effluent limits specifically related to incineration of waste oil. The licensee shall report any changes or additions to the information supplied under §§ 50.34 and 50.34a of this chapter associated with this incineration pursuant to § 50.71 of this chapter, as appropriate. The licensee shall also follow the procedures of § 50.59 of this chapter with respect to such changes to the facility or procedures.

(2) Solid residues produced in the process of incinerating waste oils must be disposed of as provided by § 20.2001.

(3) The provisions of this section authorize onsite waste oil incineration under the terms of this section and supersede any provision in an individual plant license or technical specification that may be inconsistent.

[57 FR 57656, Dec. 7, 1992]

zzz) § 20.2005 Disposal of specific wastes.

(a) A licensee may dispose of the following licensed material as if it were not radioactive:

(1) 0.05 microcurie (1.85 kBq), or less, of hydrogen‐3 or carbon‐14 per gram of medium used for liquid scintillation counting; and

(2) 0.05 microcurie (1.85 kBq), or less, of hydrogen‐3 or carbon‐14 per gram of animal tissue, averaged over the weight of the entire animal.

(b) A licensee may not dispose of tissue under paragraph (a)(2) of this section in a manner that would permit its use either as food for humans or as animal feed.

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(c) The licensee shall maintain records in accordance with § 20.2108.

aaaa) § 20.2006 Transfer for disposal and manifests.

(a) The requirements of this section and appendix G to 10 CFR Part 20 are designed to‐‐

(1) Control transfers of low‐level radioactive waste by any waste generator, waste collector, or waste processor licensee, as defined in this part, who ships low‐level waste either directly, or indirectly through a waste collector or waste processor, to a licensed low‐level waste land disposal facility (as defined in Part 61 of this chapter);

(2) Establish a manifest tracking system; and

(3) Supplement existing requirements concerning transfers and recordkeeping for those wastes.

(b) Any licensee shipping radioactive waste intended for ultimate disposal at a licensed land disposal facility must document the information required on NRCʹs Uniform Low‐Level Radioactive Waste Manifest and transfer this recorded manifest information to the intended consignee in accordance with appendix G to 10 CFR Part 20.

(c) Each shipment manifest must include a certification by the waste generator as specified in section II of appendix G to 10 CFR Part 20.

(d) Each person involved in the transfer for disposal and disposal of waste, including the waste generator, waste collector, waste processor, and disposal facility operator, shall comply with the requirements specified in section III of appendix G to 10 CFR Part 20.

[63 FR 50128, Sept. 21, 1998]

bbbb) § 20.2007 Compliance with environmental and health protection regulations.

Nothing in this subpart relieves the licensee from complying with other applicable Federal, State, and local regulations governing any other toxic or hazardous properties of materials that may be disposed of under this subpart.

cccc) Subpart L‐‐Records


dddd) § 20.2101 General provisions.

(a) Each licensee shall use the units: curie, rad, rem, including multiples and subdivisions, and shall clearly indicate the units of all quantities on records required by this part.

(b) In the records required by this part, the licensee may record quantities in SI units in parentheses following each of the units specified in paragraph (a) of this section. However, all quantities must be recorded as stated in paragraph (a) of this section.

(c) Not withstanding the requirements of paragraph (a) of this section, when recording information on shipment manifests, as required in § 20.2006(b), information must be recorded in the International System of Units (SI) or in SI and units as specified in paragraph (a) of this section.

107

(d) The licensee shall make a clear distinction among the quantities entered on the records required by this part (e.g., total effective dose equivalent, shallow‐dose equivalent, lens dose equivalent, deep‐dose equivalent, committed effective dose equivalent).

[56 FR 23404, May 21, 1991, as amended at 60 FR 15663, Mar. 27, 1995; 63 FR 39483, July 23, 1998]

eeee) § 20.2102 Records of radiation protection programs.

(a) Each licensee shall maintain records of the radiation protection program, including:

(1) The provisions of the program; and

(2) Audits and other reviews of program content and implementation.

(b) The licensee shall retain the records required by paragraph (a)(1) of this section until the Commission terminates each pertinent license requiring the record. The licensee shall retain the records required by paragraph (a)(2) of this section for 3 years after the record is made.

ffff) § 20.2103 Records of surveys.

(a) Each licensee shall maintain records showing the results of surveys and calibrations required by §§ 20.1501 and 20.1906(b). The licensee shall retain these records for 3 years after the record is made.

(b) The licensee shall retain each of the following records until the Commission terminates each pertinent license requiring the record:

(1) Records of the results of surveys to determine the dose from external sources and used, in the absence of or in combination with individual monitoring data, in the assessment of individual dose equivalents. This includes those records of results of surveys to determine the dose from external sources and used, in the absence of or in combination with individual monitoring data, in the assessment of individual dose equivalents required under the standards for protection against radiation in effect prior to January 1, 1994; and

(2) Records of the results of measurements and calculations used to determine individual intakes of radioactive material and used in the assessment of internal dose. This includes those records of the results of measurements and calculations used to determine individual intakes of radioactive material and used in the assessment of internal dose required under the standards for protection against radiation in effect prior to January 1, 1994; and

(3) Records showing the results of air sampling, surveys, and bioassays required pursuant to § 20.1703(c)(1) and (2). This includes those records showing the results of air sampling, surveys, and bioassays required under the standards for protection against radiation in effect prior to January 1, 1994; and

(4) Records of the results of measurements and calculations used to evaluate the release of radioactive effluents to the environment. This includes those records of the results of measurements and calculations used to evaluate the release of radioactive effluents to the environment required under the standards for protection against radiation in effect prior to January 1, 1994.

[56 FR 23404, May 21, 1991, as amended at 60 FR 20185, Apr. 25, 1995; 66 FR 64737, Dec. 14, 2001]

gggg) § 20.2104 Determination of prior occupational dose.

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(a) For each individual who is likely to receive in a year, an occupational dose requiring monitoring pursuant to § 20.1502 the licensee shall‐‐

(1) Determine the occupational radiation dose received during the current year; and

(2) Attempt to obtain the records of cumulative occupational radiation dose.

(b) Prior to permitting an individual to participate in a planned special exposure, the licensee shall determine‐‐

(1) The internal and external doses from all previous planned special exposures; and

(2) All doses in excess of the limits (including doses received during accidents and emergencies) received during the lifetime of the individual.

(c) In complying with the requirements of paragraph (a) of this section, a licensee may‐‐

(1) Accept, as a record of the occupational dose that the individual received during the current year, a written signed statement from the individual, or from the individualʹs most recent employer for work involving radiation exposure, that discloses the nature and the amount of any occupational dose that the individual may have received during the current year;

(2) Accept, as the record of cumulative radiation dose, an up‐to‐date NRC Form 4, or equivalent, signed by the individual and countersigned by an appropriate official of the most recent employer for work involving radiation exposure, or the individualʹs current employer (if the individual is not employed by the licensee); and

(3) Obtain reports of the individualʹs dose equivalent(s) from the most recent employer for work involving radiation exposure, or the individualʹs current employer (if the individual is not employed by the licensee) by telephone, telegram, electronic media, or letter. The licensee shall request a written verification of the dose data if the authenticity of the transmitted report cannot be established.

(d) The licensee shall record the exposure history of each individual, as required by paragraph (a) of this section, on NRC Form 4, or other clear and legible record, including all of the information required by NRC Form 44. The form or record must show each period in which the individual received occupational exposure to radiation or radioactive material and must be signed by the individual who received the exposure. For each period for which the licensee obtains reports, the licensee shall use the dose shown in the report in preparing the NRC Form 4. For any period in which the licensee does not obtain a report, the licensee shall place a notation on the NRC Form 4 indicating the periods of time for which data are not available.

(e) If the licensee is unable to obtain a complete record of an individualʹs current and previously accumulated occupational dose, the licensee shall assume‐‐

(1) In establishing administrative controls under § 20.1201(f) for the current year, that the allowable dose limit for the individual is reduced by 1.25 rems (12.5 mSv) for each quarter for which records were unavailable and the individual was engaged in activities that could have resulted in occupational radiation exposure; and

(2) That the individual is not available for planned special exposures.

(f) The licensee shall retain the records on NRC Form 4 or equivalent until the Commission terminates each pertinent license requiring this record. The licensee shall retain records used in preparing NRC Form 4 for 3 years after the

109

record is made. This includes records required under the standards for protection against radiation in effect prior to January 1, 1994.

[56 FR 23404, May 21, 1991, as amended at 57 FR 57878, Dec. 8, 1992; 60 FR 20186, Apr. 25, 1995; 60 FR 36043, July 13, 1995]

4 Licensees are not required to partition historical dose between external dose equivalent(s) and internal committed dose equivalent(s). Further, occupational exposure histories obtained and recorded on NRC Form 4 before January 1, 1994, might not have included effective dose equivalent, but may be used in the absence of specific information on the intake of radionuclides by the individual.

hhhh) § 20.2105 Records of planned special exposures.

(a) For each use of the provisions of § 20.1206 for planned special exposures, the licensee shall maintain records that describe‐‐

(1) The exceptional circumstances requiring the use of a planned special exposure; and

(2) The name of the management official who authorized the planned special exposure and a copy of the signed authorization; and

(3) What actions were necessary; and

(4) Why the actions were necessary; and

(5) How doses were maintained ALARA; and

(6) What individual and collective doses were expected to result, and the doses actually received in the planned special exposure.

(b) The licensee shall retain the records until the Commission terminates each pertinent license requiring these records.

iiii) § 20.2106 Records of individual monitoring results.

(a) Recordkeeping requirement. Each licensee shall maintain records of doses received by all individuals for whom monitoring was required pursuant to § 20.1502, and records of doses received during planned special exposures, accidents, and emergency conditions. These records5 must include, when applicable‐‐

(1) The deep‐dose equivalent to the whole body, lens dose equivalent, shallow‐dose equivalent to the skin, and shallow‐dose equivalent to the extremities;

(2) The estimated intake of radionuclides (see § 20.1202);

(3) The committed effective dose equivalent assigned to the intake of radionuclides;

(4) The specific information used to assess the committed effective dose equivalent pursuant to § 20.1204(a) and (c), and when required by § 20.1502;

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(5) The total effective dose equivalent when required by § 20.1202; and

(6) The total of the deep‐dose equivalent and the committed dose to the organ receiving the highest total dose.

(b) Recordkeeping frequency. The licensee shall make entries of the records specified in paragraph (a) of this section at least annually.

(c) Recordkeeping format. The licensee shall maintain the records specified in paragraph (a) of this section on NRC Form 5, in accordance with the instructions for NRC Form 5, or in clear and legible records containing all the information required by NRC Form 5.

(d) Privacy protection. The records required under this section should be protected from public disclosure because of their personal privacy nature. These records are protected by most State privacy laws and, when transferred to the NRC, are protected by the Privacy Act of 1974, Public Law 93‐579, 5 U.S.C. 552a, and the Commissionʹs regulations in 10 CFR part 9.

(e) The licensee shall maintain the records of dose to an embryo/fetus with the records of dose to the declared pregnant woman. The declaration of pregnancy shall also be kept on file, but may be maintained separately from the dose records.

(f) The licensee shall retain the required form or record until the Commission terminates each pertinent license requiring this record. This includes records required under the standards for protection against radiation in effect prior to January 1, 1994.


5 Assessments of dose equivalent and records made using units in effect before the licenseeʹs adoption of this part need not be changed.

jjjj) § 20.2107 Records of dose to individual members of the public.

(a) Each licensee shall maintain records sufficient to demonstrate compliance with the dose limit for individual members of the public (see § 20.1301).

(b) The licensee shall retain the records required by paragraph (a) of this section until the Commission terminates each pertinent license requiring the record.

kkkk) § 20.2108 Records of waste disposal.

(a) Each licensee shall maintain records of the disposal of licensed materials made under §§ 20.2002, 20.2003, 20.2004, 20.2005, 10 CFR part 61 and disposal by burial in soil, including burials authorized before January 28, 1981.6

(b) The licensee shall retain the records required by paragraph (a) of this section until the Commission terminates each pertinent license requiring the record. Requirements for disposition of these records, prior to license termination, are located in §§ 30.51, 40.61, 70.51, and 72.80 for activities licensed under these parts.

[56 FR 23404, May 21, 1991, as amended at 60 FR 20186, Apr. 25, 1995; 61 FR 24673, May 16, 1996]

6 A previous § 20.304 permitted burial of small quantities of licensed materials in soil before January 28, 1981, without

111

specific Commission authorization.

llll) § 20.2109 [Reserved] mmmm) § 20.2110 Form of records.

Each record required by this part must be legible throughout the specified retention period. The record may be the original or a reproduced copy or a microform provided that the copy or microform is authenticated by authorized personnel and that the microform is capable of producing a clear copy throughout the required retention period. The record may also be stored in electronic media with the capability for producing legible, accurate, and complete records during the required retention period. Records, such as letters, drawings, and specifications, must include all pertinent information, such as stamps, initials, and signatures. The licensee shall maintain adequate safeguards against tampering with and loss of records.

nnnn) Subpart M‐‐Reports


oooo) § 20.2201 Reports of theft or loss of licensed material.

(a) Telephone reports. (1) Each licensee shall report by telephone as follows:

(i) Immediately after its occurrence becomes known to the licensee, any lost, stolen, or missing licensed material in an aggregate quantity equal to or greater than 1,000 times the quantity specified in appendix C to part 20 under such circumstances that it appears to the licensee that an exposure could result to persons in unrestricted areas; or

(ii) Within 30 days after the occurrence of any lost, stolen, or missing licensed material becomes known to the licensee, all licensed material in a quantity greater than 10 times the quantity specified in appendix C to part 20 that is still missing at this time.

(2) Reports must be made as follows:

(i) Licensees having an installed Emergency Notification System shall make the reports to the NRC Operations Center in accordance with § 50.72 of this chapter, and

(ii) All other licensees shall make reports by telephone to the NRC Operations Center (301)‐816‐5100.

(b) Written reports. (1) Each licensee required to make a report under paragraph (a) of this section shall, within 30 days after making the telephone report, make a written report setting forth the following information:

(i) A description of the licensed material involved, including kind, quantity, and chemical and physical form; and

(ii) A description of the circumstances under which the loss or theft occurred; and

(iii) A statement of disposition, or probable disposition, of the licensed material involved; and

(iv) Exposures of individuals to radiation, circumstances under which the exposures occurred, and the possible total effective dose equivalent to persons in unrestricted areas; and

(v) Actions that have been taken, or will be taken, to recover the material; and

112

(vi) Procedures or measures that have been, or will be, adopted to ensure against a recurrence of the loss or theft of licensed material.

(2) Reports must be made as follows:

(i) For holders of an operating license for a nuclear power plant, the events included in paragraph (b) of this section must be reported in accordance with the procedures described in § 50.73(b), (c), (d), (e), and (g) of this chapter and must include the information required in paragraph (b)(1) of this section, and

(ii) All other licensees shall make reports to the Administrator of the appropriate NRC Regional Office listed in appendix D to part 20.

(c) A duplicate report is not required under paragraph (b) of this section if the licensee is also required to submit a report pursuant to §§ 30.55(c), 40.64(c), 50.72, 50.73, 70.52, 73.27(b), 73.67(e)(3)(vii), 73.67(g)(3)(iii), 73.71, or § 150.19(c) of this chapter.

(d) Subsequent to filing the written report, the licensee shall also report any additional substantive information on the loss or theft within 30 days after the licensee learns of such information.

(e) The licensee shall prepare any report filed with the Commission pursuant to this section so that names of individuals who may have received exposure to radiation are stated in a separate and detachable part of the report.

[56 FR 23406, May 21, 1991, as amended at 58 FR 69220, Dec. 30, 1993; 60 FR 20186, Apr. 25, 1995; 66 FR 64738, Dec. 14, 2001; 67 FR 3585, Jan. 25, 2002]

pppp) § 20.2202 Notification of incidents.

(a) Immediate notification. Notwithstanding any other requirements for notification, each licensee shall immediately report any event involving byproduct, source, or special nuclear material possessed by the licensee that may have caused or threatens to cause any of the following conditions‐‐

(1) An individual to receive‐‐

(i) A total effective dose equivalent of 25 rems (0.25 Sv) or more; or

(ii) A lens dose equivalent of 75 rems (0.75 Sv) or more; or

(iii) A shallow‐dose equivalent to the skin or extremities of 250 rads (2.5 Gy) or more; or

(2) The release of radioactive material, inside or outside of a restricted area, so that, had an individual been present for 24 hours, the individual could have received an intake five times the annual limit on intake (the provisions of this paragraph do not apply to locations where personnel are not normally stationed during routine operations, such as hot‐cells or process enclosures).

(b) Twenty‐four hour notification. Each licensee shall, within 24 hours of discovery of the event, report any event involving loss of control of licensed material possessed by the licensee that may have caused, or threatens to cause, any of the following conditions:

(1) An individual to receive, in a period of 24 hours‐‐

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(i) A total effective dose equivalent exceeding 5 rems (0.05 Sv); or

(ii) A lens dose equivalent exceeding 15 rems (0.15 Sv); or

(iii) A shallow‐dose equivalent to the skin or extremities exceeding 50 rems (0.5 Sv); or

(2) The release of radioactive material, inside or outside of a restricted area, so that, had an individual been present for 24 hours, the individual could have received an intake in excess of one occupational annual limit on intake (the provisions of this paragraph do not apply to locations where personnel are not normally stationed during routine operations, such as hot‐cells or process enclosures).

(c) The licensee shall prepare any report filed with the Commission pursuant to this section so that names of individuals who have received exposure to radiation or radioactive material are stated in a separate and detachable part of the report.

(d) Reports made by licensees in response to the requirements of this section must be made as follows:

(1) Licensees having an installed Emergency Notification System shall make the reports required by paragraphs (a) and (b) of this section to the NRC Operations Center in accordance with 10 CFR 50.72; and

(2) All other licensees shall make the reports required by paragraphs (a) and (b) of this section by telephone to the NRC Operations Center (301) 816‐5100.

(e) The provisions of this section do not include doses that result from planned special exposures, that are within the limits for planned special exposures, and that are reported under § 20.2204.

[56 FR 23406, May 21, 1991, as amended at 56 FR 40766, Aug. 16, 1991; 57 FR 57879, Dec. 8, 1992; 59 FR 14086, Mar. 25, 1994; 63 FR 39483, July 23, 1998]

qqqq) § 20.2203 Reports of exposures, radiation levels, and concentrations of radioactive material exceeding the constraints or limits.

(a) Reportable events. In addition to the notification required by § 20.2202, each licensee shall submit a written report within 30 days after learning of any of the following occurrences:

(1) Any incident for which notification is required by § 20.2202; or

(2) Doses in excess of any of the following:

(i) The occupational dose limits for adults in § 20.1201; or

(ii) The occupational dose limits for a minor in § 20.1207; or

(iii) The limits for an embryo/fetus of a declared pregnant woman in § 20.1208; or

(iv) The limits for an individual member of the public in § 20.1301; or

(v) Any applicable limit in the license; or

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(vi) The ALARA constraints for air emissions established under § 20.1101(d); or

(3) Levels of radiation or concentrations of radioactive material in‐‐

(i) A restricted area in excess of any applicable limit in the license; or

(ii) An unrestricted area in excess of 10 times any applicable limit set forth in this part or in the license (whether or not involving exposure of any individual in excess of the limits in § 20.1301); or

(4) For licensees subject to the provisions of EPAʹs generally applicable environmental radiation standards in 40 CFR part 190, levels of radiation or releases of radioactive material in excess of those standards, or of license conditions related to those standards.

(b) Contents of reports. (1) Each report required by paragraph (a) of this section must describe the extent of exposure of individuals to radiation and radioactive material, including, as appropriate:

(i) Estimates of each individualʹs dose; and

(ii) The levels of radiation and concentrations of radioactive material involved; and

(iii) The cause of the elevated exposures, dose rates, or concentrations; and

(iv) Corrective steps taken or planned to ensure against a recurrence, including the schedule for achieving conformance with applicable limits, ALARA constraints, generally applicable environmental standards, and associated license conditions.

(2) Each report filed pursuant to paragraph (a) of this section must include for each occupationally overexposed1 individual: the name, Social Security account number, and date of birth. The report must be prepared so that this information is stated in a separate and detachable part of the report and must be clearly labeled ʺPrivacy Act Information: Not for Public Disclosure.ʺ

(c) For holders of an operating license for a nuclear power plant, the occurrences included in paragraph (a) of this section must be reported in accordance with the procedures described in § 50.73(b), (c), (d), (e), and (g) of this chapter and must also include the information required by paragraph (b) of this section. Occurrences reported in accordance with § 50.73 of this chapter need not be reported by a duplicate report under paragraph (a) of this section.

(d) All licensees, other than those holding an operating license for a nuclear power plant, who make reports under paragraph (a) of this section shall submit the report in writing either by mail addressed to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555‐0001; by hand delivery to the NRCʹs offices at 11555 Rockville Pike, Rockville, Maryland; or, where practicable, by electronic submission, for example, Electronic Information Exchange, or CD‐ROM. Electronic submissions must be made in a manner that enables the NRC to receive, read, authenticate, distribute, and archive the submission, and process and retrieve it a single page at a time. Detailed guidance on making electronic submissions can be obtained by visiting the NRCʹs Web site at http://www.nrc.gov/site‐help/eie.html, by calling (301) 415‐6030, by e‐mail to [email protected], or by writing the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555‐0001. A copy should be sent to the appropriate NRC Regional Office listed in appendix D to this part.

[56 FR 23406, May 21, 1991, as amended at 60 FR 20186, Apr. 25, 1995; 61 FR 65127, Dec. 10, 1996; 68 FR 14309, Mar. 25, 2003; 68 FR 58802, Oct. 10, 2003]

115

1 With respect to the limit for the embryo‐fetus (§ 20.1208), the identifiers should be those of the declared pregnant woman.

rrrr) § 20.2204 Reports of planned special exposures.

The licensee shall submit a written report to the Administrator of the appropriate NRC Regional Office listed in appendix D to part 20 within 30 days following any planned special exposure conducted in accordance with § 20.1206, informing the Commission that a planned special exposure was conducted and indicating the date the planned special exposure occurred and the information required by § 20.2105.


ssss) § 20.2205 Reports to individuals of exceeding dose limits.

When a licensee is required, pursuant to the provisions of §§ 20.2203, 20.2204, or 20.2206, to report to the Commission any exposure of an identified occupationally exposed individual, or an identified member of the public, to radiation or radioactive material, the licensee shall also provide a copy of the report submitted to the Commission to the individual. This report must be transmitted at a time no later than the transmittal to the Commission.

[60 FR 36043, July 13, 1995]

tttt) § 20.2206 Reports of individual monitoring.

(a) This section applies to each person licensed by the Commission to‐‐

(1) Operate a nuclear reactor designed to produce electrical or heat energy pursuant to § 50.21(b) or § 50.22 of this chapter or a testing facility as defined in § 50.2 of this chapter; or

(2) Possess or use byproduct material for purposes of radiography pursuant to Parts 30 and 34 of this chapter; or

(3) Possess or use at any one time, for purposes of fuel processing, fabricating, or reprocessing, special nuclear material in a quantity exceeding 5,000 grams of contained uranium‐235, uranium‐233, or plutonium, or any combination thereof pursuant to part 70 of this chapter; or

(4) Possess high‐level radioactive waste at a geologic repository operations area pursuant to part 60 or 63 of this chapter; or

(5) Possess spent fuel in an independent spent fuel storage installation (ISFSI) pursuant to part 72 of this chapter; or

(6) Receive radioactive waste from other persons for disposal under part 61 of this chapter; or

(7) Possess or use at any time, for processing or manufacturing for distribution pursuant to parts 30, 32, 33 or 35 of this chapter, byproduct material in quantities exceeding any one of the following quantities:

Radionuclide Quantity of

radionuclide1 in curies

Cesium‐137 1

116

Cobalt‐60 1

Gold‐198 100

Iodine‐131 1

Iridium‐192 10

Krypton‐85 1,000

Promethium‐147 10

Techetium‐99m 1,000

1 The Commission may require as a license condition, or by rule, regulation, or order pursuant to § 20.2302, reports from licensees who are licensed to use radionuclides not on this list, in quantities sufficient to cause comparable radiation levels.

(b) Each licensee in a category listed in paragraph (a) of this section shall submit an annual report of the results of individual monitoring carried out by the licensee for each individual for whom monitoring was required by § 20.1502 during that year. The licensee may include additional data for individuals for whom monitoring was provided but not required. The licensee shall use Form NRC 5 or electronic media containing all the information required by Form NRC 5.

(c) The licensee shall file the report required by § 20.2206(b), covering the preceding year, on or before April 30 of each year. The licensee shall submit the report to the REIRS Project Manager by an appropriate method listed in § 20.1007 or via the REIRS Web site at http://www.reirs.com.

[56 FR 23406, May 21, 1991, as amended at 56 FR 32072, July 15, 1991; 66 FR 5578, Nov. 2, 2001; 68 FR 58802, Oct. 10, 2003]

uuuu) Subpart N‐‐Exemptions and Additional Requirements


vvvv) § 20.2301 Applications for exemptions.

The Commission may, upon application by a licensee or upon its own initiative, grant an exemption from the requirements of the regulations in this part if it determines the exemption is authorized by law and would not result in undue hazard to life or property.

wwww) § 20.2302 Additional requirements.

The Commission may, by rule, regulation, or order, impose requirements on a licensee, in addition to those established in the regulations in this part, as it deems appropriate or necessary to protect health or to minimize danger to life or property.

xxxx) Subpart O‐‐Enforcement 5. § 20.2401 Violations.


117






(i) Sections 53, 57, 62, 63, 81, 82, 101, 103, 104, 107 or 109 of the Atomic Energy Act of 1954, as amended;


(iii) Any rule, regulation, or order issued pursuant to the sections specified in paragraph (b)(1)(i) of this section; and


(2) For any violation for which a license may be revoked under Section 186 of the Atomic Energy Act of 1954, as amended.

[56 FR 23408, May 21, 1991; 56 FR 61352, Dec. 3, 1991, as amended at 57 FR 55071, Nov. 24, 1992]

yyyy) § 20.2402 Criminal penalties.

(a) Section 223 of the Atomic Energy Act of 1954, as amended, provides for criminal sanctions for willful violation of, attempted violation of, or conspiracy to violate, any regulation issued under sections 161b, 161i, or 161o of the Act. For purposes of section 223, all the regulations in §§ 20.1001 through 20.2402 are issued under one or more of sections 161b, 161i, or 161o, except for the sections listed in paragraph (b) this section.

(b) The regulations in §§ 20.1001 through 20.2402 that are not issued under Sections 161b, 161i, or 161o for the purposes of Section 223 are as follows: §§ 20.1001, 20.1002, 20.1003, 20.1004, 20.1005, 20.1006, 20.1007, 20.1008, 20.1009, 20.1405, 20.1704, 20.1903, 20.1905, 20.2002, 20.2007, 20.2301, 20.2302, 20.2401, and 20.2402.

[57 FR 55071, Nov. 24, 1992]

zzzz) Appendix A to Part 20‐‐Assigned Protection Factors for Respiratorsa Operating mode Assigned Protection

Factors

I. Air Purifying Respirators [Particulateb only]c:

Filtering facepiece disposabled Negative Pressure (d)

Facepiece, half e Negative Pressure 10

Facepiece, full Negative Pressure 100

Facepiece, half Powered air‐purifying 50

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respirators

Facepiece, full Powered air‐purifying respirators

1000

Helmet/hood Powered air‐purifying respirators

1000

Facepiece, loose‐fitting Powered air‐purifying respirators

25

II. Atmosphere supplying respirators [particulate, gases and vaporsf]:

1. Air‐line respirator:

Facepiece, half Demand 10

Facepiece, half Continuous Flow 50

Facepiece, half Pressure Demand 50

Facepiece, full Demand 100

Facepiece, full Continuous Flow 1000

Facepiece, full Pressure Demand 1000

Helmet/hood Continuous Flow 1000

Facepiece, loose‐fitting Continuous Flow 25

Suit Continuous Flow (g)

2. Self‐contained breathing Apparatus (SCBA):

Facepiece, full Demand h100

Facepiece, full Pressure Demand i10,000

Facepiece, full Demand, Recirculating h100

Facepiece, full Positive Pressure Recirculating i10,000

III. Combination Respirators:

Any combination of air‐purifying and atmosphere‐supplying respirators

Assigned protection factor for type and mode of operation as listed above.

a These assigned protection factors apply only in a respiratory protection program that meets the requirements of this Part. They are applicable only to airborne radiological hazards and may not be appropriate to circumstances when chemical or other respiratory hazards exist instead of, or in addition to, radioactive hazards. Selection and use of respirators for such circumstances must also comply with Department of Labor regulations.

Radioactive contaminants for which the concentration values in Table 1, Column 3 of Appendix B to Part 20 are based on internal dose due to inhalation may, in addition, present external exposure hazards at higher concentrations. Under these circumstances, limitations on occupancy may have to be governed by external dose limits.

b Air purifying respirators with APF <100 must be equipped with particulate filters that are at least 95 percent efficient. Air purifying respirators with APF = 100 must be equipped with particulate filters that are at least 99

119

percent efficient. Air purifying respirators with APFs >100 must be equipped with particulate filters that are at least 99.97 percent efficient.

c The licensee may apply to the Commission for the use of an APF greater than 1 for sorbent cartridges as protection against airborne radioactive gases and vapors (e.g., radioiodine).

d Licensees may permit individuals to use this type of respirator who have not been medically screened or fit tested on the device provided that no credit be taken for their use in estimating intake or dose. It is also recognized that it is difficult to perform an effective positive or negative pressure pre‐use user seal check on this type of device. All other respiratory protection program requirements listed in § 20.1703 apply. An assigned protection factor has not been assigned for these devices. However, an APF equal to 10 may be used if the licensee can demonstrate a fit factor of at least 100 by use of a validated or evaluated, qualitative or quantitative fit test.

e Under‐chin type only. No distinction is made in this Appendix between elastomeric half‐masks with replaceable cartridges and those designed with the filter medium as an integral part of the facepiece (e.g., disposable or reusable disposable). Both types are acceptable so long as the seal area of the latter contains some substantial type of seal‐enhancing material such as rubber or plastic, the two or more suspension straps are adjustable, the filter medium is at least 95 percent efficient and all other requirements of this Part are met.

f The assigned protection factors for gases and vapors are not applicable to radioactive contaminants that present an absorption or submersion hazard. For tritium oxide vapor, approximately one‐third of the intake occurs by absorption through the skin so that an overall protection factor of 3 is appropriate when atmosphere‐supplying respirators are used to protect against tritium oxide. Exposure to radioactive noble gases is not considered a significant respiratory hazard, and protective actions for these contaminants should be based on external (submersion) dose considerations.

g No NIOSH approval schedule is currently available for atmosphere supplying suits. This equipment may be used in an acceptable respiratory protection program as long as all the other minimum program requirements, with the exception of fit testing, are met (i.e., § 20.1703).

h The licensee should implement institutional controls to assure that these devices are not used in areas immediately dangerous to life or health (IDLH).

i This type of respirator may be used as an emergency device in unknown concentrations for protection against inhalation hazards. External radiation hazards and other limitations to permitted exposure such as skin absorption shall be taken into account in these circumstances. This device may not be used by any individual who experiences perceptible outward leakage of breathing gas while wearing the device.

[64 FR 54558, Oct. 7, 1999; 64 FR 55524, Oct. 13, 1999]

aaaaa) Appendix B to Part 20‐‐Annual Limits on Intake (ALIs) and Derived Air Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations; Concentrations for Release to Sewerage

Introduction

For each radionuclide Table 1 indicates the chemical form which is to be used for selecting the appropriate ALI or DAC value. The ALIs and DACs for inhalation are given for an aerosol with an activity median aerodynamic diameter (AMAD) of 1 μm and for three classes (D,W,Y) of radioactive material, which refer to their retention

120

(approximately days, weeks or years) in the pulmonary region of the lung. This classification applies to a range of clearance half‐times of less than 10 days for D, for W from 10 to 100 days, and for Y greater than 100 days. The class (D, W, or Y) given in the column headed ʺClassʺ applies only to the inhalation ALIs and DACs given in Table 1, columns 2 and 3. Table 2 provides concentration limits for airborne and liquid effluents released to the general environment. Table 3 provides concentration limits for discharges to sanitary sewer systems.

Notation

The values in Tables 1, 2, and 3 are presented in the computer ʺEʺ notation. In this notation a value of 6E‐02 represents a value of 6x10‐2 or 0.06, 6E+2 represents 6x102 or 600, and 6E+0 represents 6x100 or 6.

Table 1 ʺOccupationalʺ

Note that the columns in Table 1, of this appendix captioned ʺOral Ingestion ALI,ʺ ʺInhalation ALI,ʺ and ʺDAC,ʺ are applicable to occupational exposure to radioactive material.

The ALIs in this appendix are the annual intakes of a given radionuclide by ʺReference Manʺ which would result in either (1) a committed effective dose equivalent of 5 rems (stochastic ALI) or (2) a committed dose equivalent of 50 rems to an organ or tissue (non‐stochastic ALI). The stochastic ALIs were derived to result in a risk, due to irradiation of organs and tissues, comparable to the risk associated with deep dose equivalent to the whole body of 5 rems. The derivation includes multiplying the committed dose equivalent to an organ or tissue by a weighting factor, wT. This weighting factor is the proportion of the risk of stochastic effects resulting from irradiation of the organ or tissue, T, to the total risk of stochastic effects when the whole body is irradiated uniformly. The values of wT are listed under the definition of weighting factor in § 20.1003. The non‐stochastic ALIs were derived to avoid non‐stochastic effects, such as prompt damage to tissue or reduction in organ function.

A value of wT=0.06 is applicable to each of the five organs or tissues in the ʺremainderʺ category receiving the highest dose equivalents, and the dose equivalents of all other remaining tissues may be disregarded. The following parts of the GI tract‐‐stomach, small intestine, upper large intestine, and lower large intestine‐‐are to be treated as four separate organs.

Note that the dose equivalents for extremities (hands and forearms, feet and lower legs), skin, and lens of the eye are not considered in computing the committed effective dose equivalent, but are subject to limits that must be met separately.

When an ALI is defined by the stochastic dose limit, this value alone, is given. When an ALI is determined by the non‐stochastic dose limit to an organ, the organ or tissue to which the limit applies is shown, and the ALI for the stochastic limit is shown in parentheses. (Abbreviated organ or tissue designations are used: LLI wall = lower large intestine wall; St. wall = stomach wall; Blad wall = bladder wall; and Bone surf = bone surface.)

The use of the ALIs listed first, the more limiting of the stochastic and non‐stochastic ALIs, will ensure that non‐stochastic effects are avoided and that the risk of stochastic effects is limited to an acceptably low value. If, in a particular situation involving a radionuclide for which the non‐stochastic ALI is limiting, use of that non‐stochastic ALI is considered unduly conservative, the licensee may use the stochastic ALI to determine the committed effective dose equivalent. However, the licensee shall also ensure that the 50‐rem dose equivalent limit for any organ or tissue is not exceeded by the sum of the external deep dose equivalent plus the internal committed dose to that organ (not the effective dose). For the case where there is no external dose contribution, this would be demonstrated if the sum of the fractions of the nonstochastic ALIs (ALIns) that contribute to the committed dose equivalent to the organ receiving the highest dose does not exceed unity (i.e., (intake (in μCi) of each radionuclide/ALIns) < 1.0). If there is an external deep dose equivalent contribution of Hd then this sum must be less than 1 ‐ (Hd/50) instead of being < 1.0.

121

The derived air concentration (DAC) values are derived limits intended to control chronic occupational exposures. The relationship between the DAC and the ALI is given by: DAC=ALI(in μCi)/(2000 hours per working year x 60 minutes/hour x 2 x 104 ml per minute)=[ALI/2.4x109] μCi/ml, where 2x104 ml is the volume of air breathed per minute at work by ʺReference Manʺ under working conditions of ʺlight work.ʺ

The DAC values relate to one of two modes of exposure: either external submersion or the internal committed dose equivalents resulting from inhalation of radioactive materials. Derived air concentrations based upon submersion are for immersion in a semi‐infinite cloud of uniform concentration and apply to each radionuclide separately.

The ALI and DAC values relate to exposure to the single radionuclide named, but also include contributions from the in‐growth of any daughter radionuclide produced in the body by the decay of the parent. However, intakes that include both the parent and daughter radionuclides should be treated by the general method appropriate for mixtures.

The value of ALI and DAC do not apply directly when the individual both ingests and inhales a radionuclide, when the individual is exposed to a mixture of radionuclides by either inhalation or ingestion or both, or when the individual is exposed to both internal and external radiation (see § 20.1202). When an individual is exposed to radioactive materials which fall under several of the translocation classifications (i.e., Class D, Class W, or Class Y) of the same radionuclide, the exposure may be evaluated as if it were a mixture of different radionuclides.

It should be noted that the classification of a compound as Class D, W, or Y is based on the chemical form of the compound and does not take into account the radiological half‐life of different radioisotopes. For this reason, values are given for Class D, W, and Y compounds, even for very short‐lived radionuclides.

Table 2

The columns in Table 2 of this appendix captioned ʺEffluents,ʺ ʺAir,ʺ and ʺWater,ʺ are applicable to the assessment and control of dose to the public, particularly in the implementation of the provisions of § 20.1302. The concentration values given in Columns 1 and 2 of Table 2 are equivalent to the radionuclide concentrations which, if inhaled or ingested continuously over the course of a year, would produce a total effective dose equivalent of 0.05 rem (50 millirem or 0.5 millisieverts).

Consideration of non‐stochastic limits has not been included in deriving the air and water effluent concentration limits because non‐stochastic effects are presumed not to occur at the dose levels established for individual members of the public. For radionuclides, where the non‐stochastic limit was governing in deriving the occupational DAC, the stochastic ALI was used in deriving the corresponding airborne effluent limit in Table 2. For this reason, the DAC and airborne effluent limits are not always proportional as was the case in appendix B to §§ 20.1‐20.601.

The air concentration values listed in Table 2, Column 1, were derived by one of two methods. For those radionuclides for which the stochastic limit is governing, the occupational stochastic inhalation ALI was divided by 2.4 x 109ml, relating the inhalation ALI to the DAC, as explained above, and then divided by a factor of 300. The factor of 300 includes the following components: a factor of 50 to relate the 5‐rem annual occupational dose limit to the 0.1‐rem limit for members of the public, a factor of 3 to adjust for the difference in exposure time and the inhalation rate for a worker and that for members of the public; and a factor of 2 to adjust the occupational values (derived for adults) so that they are applicable to other age groups.

For those radionuclides for which submersion (external dose) is limiting, the occupational DAC in Table 1, Column 3, was divided by 219. The factor of 219 is composed of a factor of 50, as described above, and a factor of 4.38 relating occupational exposure for 2,000 hours per year to full‐time exposure (8,760 hours per year). Note that an additional factor of 2 for age considerations is not warranted in the submersion case.

122

The water concentrations were derived by taking the most restrictive occupational stochastic oral ingestion ALI and dividing by 7.3 x 107. The factor of 7.3 x 107 (ml) includes the following components: the factors of 50 and 2 described above and a factor of 7.3 x 105 (ml) which is the annual water intake of ʺReference Man.ʺ

Note 2 of this appendix provides groupings of radionuclides which are applicable to unknown mixtures of radionuclides. These groupings (including occupational inhalation ALIs and DACs, air and water effluent concentrations and sewerage) require demonstrating that the most limiting radionuclides in successive classes are absent. The limit for the unknown mixture is defined when the presence of one of the listed radionuclides cannot be definitely excluded either from knowledge of the radionuclide composition of the source or from actual measurements.

Table 3 ʺSewer Disposalʺ

The monthly average concentrations for release to sanitary sewers are applicable to the provisions in § 20.2003. The concentration values were derived by taking the most restrictive occupational stochastic oral ingestion ALI and dividing by 7.3 x 106(ml). The factor of 7.3 x 106(ml) is composed of a factor of 7.3 x 105(ml), the annual water intake by ʺReference Man,ʺ and a factor of 10, such that the concentrations, if the sewage released by the licensee were the only source of water ingested by a reference man during a year, would result in a committed effective dose equivalent of 0.5 rem.

[56 FR 23409, May 21, 1991; 56 FR 61352, Dec. 3, 1991, as amended at 57 FR 57879, Dec. 8, 1992. Redesignated at 58 FR 67659, Dec. 22, 1993]

bbbbb) Appendix C to Part 20‐‐Quantities1 of Licensed Material Requiring Labeling Radionuclide Abbreviation Quantity (μCi)

Hydrogen‐3 H‐3 1,000

Beryllium‐7 Be‐7 1,000

Beryllium‐10 Be‐10 1

Carbon‐11 C‐11 1,000

Carbon‐14 C‐14 100

Fluorine‐18 F‐18 1,000

Sodium‐22 Na‐22 10

Sodium‐24 Na‐24 100

Magnesium‐28 Mg‐28 100

Aluminum‐26 Al‐26 10

Silicon‐31 Si‐31 1,000

Silicon‐32 Si‐32 1

Phosphorus‐32 P‐32 10

Phosphorus‐33 P‐33 100

Sulfur‐35 S‐35 100

Chlorine‐36 Cl‐36 10

Chlorine‐38 Cl‐38 1,000

123

Chlorine‐39 Cl‐39 1,000

Argon‐39 Ar‐39 1,000

Argon‐41 Ar‐41 1,000

Potassium‐40 K‐40 100

Potassium‐42 K‐42 1,000




Calcium‐41 Ca‐41 100



Scandium‐43 Sc‐43 1,000

Scandium‐44m Sc‐44m 100

Scandium‐44 Sc‐44 100




Scandium‐49 Sc‐49 1,000

Titanium‐44 Ti‐44 1

Titanium‐45 Ti‐45 1,000

Vanadium‐47 V‐47 1,000

Vanadium‐48 V‐48 100

Vanadium‐49 V‐49 1,000

Chromium‐48 Cr‐48 1,000



Manganese‐51 Mn‐51 1,000

Manganese‐52m Mn‐52m 1,000

Manganese‐52 Mn‐52 100


Manganese‐54 Mn‐54 100


Iron‐52 Fe‐52 100

Iron‐55 Fe‐55 100

Iron‐59 Fe‐59 10

124

Iron‐60 Fe‐60 1

Cobalt‐55 Co‐55 100



Cobalt‐58m Co‐58m 1,000




Cobalt‐61 Co‐61 1,000


Nickel‐56 Ni‐56 100




Nickel‐65 Ni‐65 1,000


Copper‐60 Cu‐60 1,000

Copper‐61 Cu‐61 1,000

Copper‐64 Cu‐64 1,000

Copper‐67 Cu‐67 1,000

Zinc‐62 Zn‐62 100

Zinc‐63 Zn‐63 1,000

Zinc‐65 Zn‐65 10

Zinc‐69m Zn‐69m 100

Zinc‐69 Zn‐69 1,000

Zinc‐71m Zn‐71m 1,000

Zinc‐72 Zn‐72 100

Gallium‐65 Ga‐65 1,000

Gallium‐66 Ga‐66 100




Gallium‐72 Ga‐72 100


Germanium‐66 Ge‐66 1,000

125


Germanium‐68 Ge‐68 10






Arsenic‐69 As‐69 1,000


Arsenic‐71 As‐71 100







Selenium‐70 Se‐70 1,000

Selenium‐73m Se‐73m 1,000

Selenium‐73 Se‐73 100



Selenium‐81m Se‐81m 1,000



Bromine‐74m Br‐74m 1,000

Bromine‐74 Br‐74 1,000


Bromine‐76 Br‐76 100


Bromine‐80m Br‐80m 1,000


Bromine‐82 Br‐82 100



Krypton‐74 Kr‐74 1,000

126





Krypton‐83m Kr‐83m 1,000

Krypton‐85m Kr‐85m 1,000




Rubidium‐79 Rb‐79 1,000

Rubidium‐81m Rb‐81m 1,000


Rubidium‐82m Rb‐82m 1,000

Rubidium‐83 Rb‐83 100






Strontium‐80 Sr‐80 100

Strontium‐81 Sr‐81 1,000


Strontium‐85m Sr‐85m 1,000


Strontium‐87m Sr‐87m 1,000


Strontium‐90 Sr‐90 0.1



Yttrium‐86m Y‐86m 1,000

Yttrium‐86 Y‐86 100





127





Yttrium‐94 Y‐94 1,000

Yttrium‐95 Y‐95 1,000

Zirconium‐86 Zr‐86 100






Niobium‐88 Nb‐88 1,000

Niobium‐89m (66 min) Nb‐89m 1,000

Niobium‐89 (122 min) Nb‐89 1,000


Niobium‐90 Nb‐90 100

Niobium‐93m Nb‐93m 10


Niobium‐95m Nb‐95m 100





Molybdenum‐90 Mo‐90 100

Molybdenum‐93m Mo‐93m 100



Molybdenum‐101 Mo‐101 1,000

Technetium‐93m Tc‐93m 1,000

Technetium‐93 Tc‐93 1,000



Technetium‐96m Tc‐96 1,000

Technetium‐96 Tc‐96 100

128

Technetium‐97m Tc‐97m 100







Ruthenium‐94 Ru‐94 1,000


Ruthenium‐103 Ru‐103 100


Ruthenium‐106 Ru‐106 1

Rhodium‐99m Rh‐99m 1,000

Rhodium‐99 Rh‐99 100




Rhodium‐102m Rh‐102m 10





Rhodium‐107 Rh‐107 1,000

Palladium‐100 Pd‐100 100

Palladium‐101 Pd‐101 1,000




Silver‐102 Ag‐102 1,000

Silver‐103 Ag‐103 1,000

Silver‐104m Ag‐104m 1,000

Silver‐104 Ag‐104 1,000

Silver‐105 Ag‐105 100

Silver‐106m Ag‐106m 100

Silver‐106 Ag‐106 1,000

129



Silver‐111 Ag‐111 100

Silver‐112 Ag‐112 100

Silver‐115 Ag‐115 1,000

Cadmium‐104 Cd‐104 1,000

Cadmium‐107 Cd‐107 1,000

Cadmium‐109 Cd‐109 1

Cadmium‐113m Cd‐113m 0.1


Cadmium‐115m Cd‐115m 10


Cadmium‐117m Cd‐117m 1,000

Cadmium‐117 Cd‐117 1,000

Indium‐109 In‐109 1,000

Indium‐110 (69.1 min.) In‐110 1,000

Indium‐110 (4.9h) In‐110 1,000

Indium‐111 In‐111 100

Indium‐112 In‐112 1,000

Indium‐113m In‐113m 1,000

Indium‐114m In‐114m 10


Indium‐115 In‐115 100



Indium‐117 In‐117 1,000


Tin‐110 Sn‐110 100

Tin‐111 Sn‐111 1,000

Tin‐113 Sn‐113 100

Tin‐117m Sn‐117m 100

Tin‐119m Sn‐119m 100

Tin‐121m Sn‐121m 100

Tin‐121 Sn‐121 1,000

Tin‐123m Sn‐123m 1,000

130

Tin‐123 Sn‐123 10

Tin‐125 Sn‐125 10

Tin‐126 Sn‐126 10

Tin‐127 Sn‐127 1,000

Tin‐128 Sn‐128 1,000

Antimony‐115 Sb‐115 1,000

Antimony‐116m Sb‐116m 1,000



Antimony‐118m Ab‐118m 1,000

Antimony‐119 Ab‐119 1,000

Antimony‐120 (16 min.) Ab‐120 1,000

Antimony‐120 (5.76d) Ab‐120 100

Antimony‐122 Ab‐122 100







Antimony‐128 (10.4 min.) Ab‐128 1,000

Antimony‐128 (9.01h) Ab‐128 100




Tellurium‐116 Te‐116 1,000

Tellurium‐121m Te‐121m 10

Tellurium‐121 Te‐121 100








131







Iodine‐120m I‐120m 1,000

Iodine‐120 I‐120 100

Iodine‐121 I‐121 1,000

Iodine‐123 I‐123 100

Iodine‐124 I‐124 10

Iodine‐125 I‐125 1

Iodine‐126 I‐126 1

Iodine‐128 I‐128 1,000

Iodine‐129 I‐129 1

Iodine‐130 I‐130 10

Iodine‐131 I‐131 1

Iodine‐132m I‐132m 100

Iodine‐132 I‐132 100

Iodine‐133 I‐133 10

Iodine‐134 I‐134 1,000

Iodine‐135 I‐135 100

Xenon‐120 Xe‐120 1,000

Xenon‐121 Xe‐121 1,000

Xenon‐122 Xe‐122 1,000

Xenon‐123 Xe‐123 1,000

Xenon‐125 Xe‐125 1,000

Xenon‐127 Xe‐127 1,000

Xenon‐129m Xe‐129m 1,000

Xenon‐131m Xe‐131m 1,000

Xenon‐133m Xe‐133m 1,000

Xenon‐133 Xe‐133 1,000

Xenon‐135m Xe‐135m 1,000

Xenon‐135 Xe‐135 1,000

Xenon‐138 Xe‐138 1,000

132

Cesium‐125 Cs‐125 1,000

Cesium‐127 Cs‐127 1,000

Cesium‐129 Cs‐129 1,000

Cesium‐130 Cs‐130 1,000

Cesium‐131 Cs‐131 1,000

Cesium‐132 Cs‐132 100

Cesium‐134m Cs‐134m 1,000

Cesium‐134 Cs‐134 10

Cesium‐135m Cs‐135m 1,000

Cesium‐135 Cs‐135 100



Cesium‐138 Cs‐138 1,000

Barium‐126 Ba‐126 1,000

Barium‐128 B‐128 100

Barium‐131m Ba‐131m 1,000

Barium‐131 Ba‐131 100

Barium‐133m Ba‐133m 100

Barium‐133 Ba‐133 100

Barium‐135m Ba‐135m 100

Barium‐139 Ba‐139 1,000

Barium‐140 Ba‐140 100

Barium‐141 Ba‐141 1,000

Barium‐142 Ba‐142 1,000

Lanthanum‐131 La‐131 1,000

Lanthanum‐132 La‐132 100








Cerium‐134 Ce‐134 100

Cerium‐135 Ce‐135 100

133

Cerium‐137m Ce‐137m 100

Cerium‐137 Ce‐137 1,000

Cerium‐139 Ce‐139 100

Cerium‐141 Ce‐141 100

Cerium‐143 Ce‐143 100

Cerium‐144 Ce‐144 1

Praseodymium‐136 Pr‐136 1,000

Praseodymium‐137 Pr‐137 1,000

Praseodymium‐138m Pe‐138m 1,000

Praseodymium‐139 Pe‐139 1,000

Praseodymium‐142m Pe‐142m 1,000

Praseodymium‐142 Pe‐142 100





Neodymium‐136 Nd‐136 1,000

Neodymium‐138 Nd‐138 100

Neodymium‐139m Nd‐139m 1,000



Neodymium‐147 Nd‐147 100



Promethium‐141 Pm‐141 1,000

Promethium‐143 Pm‐143 100





Promethium‐148m Pm‐148m 10



Promethium‐150 Pm‐150 1,000


134

Samarium‐141m Sm‐141m 1,000

Samarium‐141 Sm‐141 1,000


Samarium‐145 Sm‐145 100







Europium‐145 Eu‐145 100





Europium‐150 (12.62h) Eu‐150 100

Europium‐150 (34.2y) Eu‐150 1

Europium‐152m Eu‐152m 100






Europium‐158 Eu‐158 1,000

Gadolinium‐145 Gd‐145 1,000

Gadolinium‐146 Gd‐146 10


Gadolinium‐148 Gd‐148 0.001






Terbium‐147 Tb‐147 1,000

Terbium‐149 Tb‐149 100

135

Terbium‐150 Tb‐150 1,000


Terbium‐153 Tb‐153 1,000


Terbium‐155 Tb‐155 1,000

Terbium‐156m (5.0h) Tb‐156m 1,000

Terbium‐156m (24.4h) Tb‐156m 1,000






Dysprosium‐155 Dy‐155 1,000


Dysprosium‐159 Dy‐159 100


Dysprosium‐166 Dy‐166 100

Holmium‐155 Ho‐155 1,000

Holmium‐157 Ho‐157 1,000

Holmium‐159 Ho‐159 1,000

Holmium‐161 Ho‐161 1,000

Holmium‐162m Ho‐162m 1,000

Holmium‐162 Ho‐162 1,000

Holmium‐164m Hp‐164m 1,000

Holmium‐164 Ho‐164 1,000

Holmium‐166m Ho‐166m 1

Holmium‐166 Ho‐166 100

Holmium‐167 Ho‐167 1,000

Erbium‐161 Er‐161 1,000

Erbium‐165 Er‐165 1,000

Erbium‐169 Er‐169 100

Erbium‐171 Er‐171 100

Erbium‐172 Er‐172 100

Thulium‐162 Tm‐162 1,000

Thulium‐166 Tm‐166 100

136






Thulium‐175 Tm‐175 1,000

Ytterbium‐162 Yb‐162 1,000

Ytterbium‐166 Yb‐166 100






Lutetium‐169 Lu‐169 100





Lutetium‐174m Lu‐174m 10


Lutetium‐176m Lu‐176m 1,000


Lutetium‐177m Lu‐177m 10


Lutetium‐178m Lu‐178m 1,000

Lutetium‐178 Lu‐178 1,000

Lutetium‐179 Lu‐179 1,000

Hafnium‐170 Hf‐170 100


Hafnium‐173 Hf‐173 1,000


Hafnium‐177m Hf‐177m 1,000

Hafnium‐178m Hf‐178m 0.1

Hafnium‐179m Hf‐179m 10


137



Hafnium‐182 Hf‐182 0.1

Hafnium‐183 Hf‐183 1,000


Tantalum‐172 Ta‐172 1,000




Tantalum‐176 Ta‐176 100




Tantalum‐180m Ta‐180m 1,000


Tantalum‐182m Ta‐182m 1,000






Tungsten‐176 W‐176 1,000

Tungsten‐177 W‐177 1,000

Tungsten‐178 W‐178 1,000

Tungsten‐179 W‐179 1,000

Tungsten‐181 W‐181 1,000

Tungsten‐185 W‐185 100



Rhenium‐177 Re‐177 1,000

Rhenium‐178 Re‐178 1,000

Rhenium‐181 Re‐181 1,000

Rhenium‐182 (12.7h) Re‐182 1,000

Rhenium‐182 (64.0h) Re‐182 100

Rhenium‐184m Re‐184m 10

138

Rhenium‐184 Re‐184 100

Rhenium‐186m Re‐186m 10


Rhenium‐187 Re‐187 1,000

Rhenium‐188m Re‐188m 1,000



Osmium‐180 Os‐180 1,000

Osmium‐181 Os‐181 1,000

Osmium‐182 Os‐182 100

Osmium‐185 Os‐185 100

Osmium‐189m Os‐189m 1,000

Osmium‐191m Os‐191m 1,000

Osmium‐191 Os‐191 100

Osmium‐193 Os‐193 100

Osmium‐194 Os‐194 1

Iridium‐182 Ir‐182 1,000

Iridium‐184 Ir‐184 1,000

Iridium‐185 Ir‐185 1,000

Iridium‐186 Ir‐186 100

Iridium‐187 Ir‐187 1,000



Iridium‐190m Ir‐190m 1,000


Iridium‐192 (73.8d) Ir‐192 1

Iridium‐192m (1.4 min.) Ir‐192m 10

Iridium‐194m Ir‐194m 10


Iridium‐195m Ir‐195m 1,000

Iridium‐195 Ir‐95 1,000

Platinum‐186 Pt‐186 1,000

Platinum‐188 Pt‐188 100



139

Platinum‐193m Pt‐193m 100


Platinum‐195m Pt‐195m 100

Platinum‐197m Pt‐197m 1,000




Gold‐193 Au‐193 1,000

Gold‐194 Au‐194 100

Gold‐195 Au‐195 10

Gold‐198m Au‐198m 100

Gold‐198 Au‐198 100

Gold‐199 Au‐199 100

Gold‐200m Au‐200m 100

Gold‐200 Au‐200 1,000

Gold‐201 Au‐201 1,000

Mercury‐193m Hg‐193m 100

Mercury‐193 Hg‐193 1,000

Mercury‐194 Hg‐194 1


Mercury‐195 Hg‐195 1,000


Mercury‐197 Hg‐197 1,000

Mercury‐199m Hg‐199m 1,000

Mercury‐203 Hg‐203 100

Thallium‐194m Tl‐194m 1,000

Thallium‐194 Tl‐194 1,000



Thallium‐198m Tl‐198m 1,000





Thallium‐202 Tl‐202 100

140

Thallium‐204 Tl‐204 100

Lead‐195m Pb‐195m 1,000

Lead‐198 Pb‐198 1,000

Lead‐199 Pb‐199 1,000

Lead‐200 Pb‐200 100

Lead‐201 Pb‐201 1,000

Lead‐202m Pb‐202m 1,000

Lead‐202 Pb‐202 10

Lead‐203 Pb‐2023 1,000

Lead‐205 Pb‐205 100

Lead‐209 Pb‐209 1,000

Lead‐210 Pb‐210 0.01

Lead‐211 Pb‐211 100

Lead‐212 Pb‐212 1

Lead‐214 Pb‐214 100

Bismuth‐200 Bi‐200 1,000

Bismuth‐201 Bi‐201 1,000

Bismuth‐202 Bi‐202 1,000

Bismuth‐203 Bi‐203 100




Bismuth‐210m Bi‐210m 0.1





Polonium‐203 Po‐203 1,000



Polonium‐210 Po‐210 0.1

Astatine‐207 At‐207 100

Astatine‐211 At‐211 10

Radon‐220 Rn‐220 1

Radon‐222 Rn‐222 1

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Francium‐222 Fr‐222 100

Francium‐223 Fr‐223 100

Radium‐223 Ra‐223 0.1

Radium‐224 Ra‐224 0.1

Radium‐225 Ra‐225 0.1

Radium‐226 Ra‐226 0.1

Radium‐227 Ra‐227 1,000

Radium‐228 Ra‐228 0.1

Actinium‐224 Ac‐224 1

Actinium‐225 Ac‐225 0.01



Actinium‐228 Ac‐228 1

Thorium‐226 Th‐226 10

Thorium‐227 Th‐227 0.01

Thorium‐228 Th‐228 0.001

Thorium‐229 Th‐229 0.001

Thorium‐230 Th‐230 0.001




Thorium‐natural 100

Protactinium‐227 Pa‐227 10


Protactinium‐230 Pa‐230 0.01

Protactinium‐231 Pa‐231 0.001




Uranium‐230 U‐230 0.01

Uranium‐231 U‐231 100

Uranium‐232 U‐232 0.001

Uranium‐233 U‐233 0.001

Uranium‐234 U‐234 0.001

Uranium‐235 U‐235 0.001

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Uranium‐236 U‐236 0.001

Uranium‐237 U‐237 100

Uranium‐238 U‐238 100

Uranium‐239 U‐239 1,000

Uranium‐240 U‐240 100

Uranium‐natural 100

Neptunium‐232 Np‐232 100

Neptunium‐233 Np‐233 1,000



Neptunium‐236 (1.15x105y) Np‐236 0.001

Neptunium‐236 (22.5h) Np‐236 1

Neptunium‐237 Np‐237 0.001



Neptunium‐240 Np‐240 1,000

Plutonium‐234 Pu‐234 10

Plutonium‐235 Pu‐235 1,000

Plutonium‐236 Pu‐236 0.001







Plutonium‐243 Pu‐243 1,000



Americium‐237 Am‐237 1,000

Americium‐238 Am‐238 100



Americium‐241 Am‐241 0.001

Americium‐242m Am‐242m 0.001


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Americium‐243 Am‐243 0.001

Americium‐244m Am‐244m 100



Americium‐246m Am‐246 1,000


Curium‐238 Cm‐238 100

Curium‐240 Cm‐240 0.1

Curium‐241 Cm‐241 1

Curium‐242 Cm‐242 0.01

Curium‐243 Cm‐243 0.001

Curium‐244 Cm‐244 0.001

Curium‐245 Cm‐245 0.001

Curium‐246 Cm‐246 0.001

Curium‐247 Cm‐247 0.001

Curium‐248 Cm‐248 0.001

Curium‐249 Cm‐249 1,000

Berkelium‐245 Bk‐245 100


Berkelium‐247 Bk‐247 0.001

Berkelium‐249 Bk‐249 0.1


Californium‐244 Cf‐244 100

Californium‐246 Cf‐246 1

Californium‐248 Cf‐248 0.01







Any alpha emitting radionuclide not listed above or mixtures or alpha emitters of unknown composition

0.001

Einsteinium‐250 Es‐250 100

Einsteinium‐251 Es‐251 100

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Einsteinium‐253 Es‐253 0.1

Einsteinium‐254m Es‐254m 1

Einsteinium‐254 Es‐254 0.01

Fermium‐252 Fm‐252 1




Fermium‐257 Fm‐257 0.01

Mendelevium‐257 Md‐257 10

Mendelevium‐258 Md‐258 0.01

Any radionuclide other than alpha emitter radionuclides not listed above, or mixtures of beta emitters of unknown composition

0.01

1 The quantities listed above were derived by taking 1/10th of the most restrictive ALI listed in table 1, columns 1 and 2, of appendix B to §§ 20.1001‐20.2401 of this part, rounding to the nearest factor of 10, and arbitrarily constraining the values listed between 0.001 and 1,000 μCi. Values of 100 μCi have been assigned for radionuclides having a radioactive half‐life in excess of 109 years (except rhenium, 1000 μCi) to take into account their low specific activity.

NOTE: For purposes of §§ 20.1902(e), 20.1905(a), and 20.2201(a) where there is involved a combination of radionuclides in known amounts, the limit for the combination should be derived as follows: determine, for each radionuclide in the combination, the ratio between the quantity present in the combination and the limit otherwise established for the specific radionuclide when not in combination. The sum of such ratios for all radionuclides in the combination may not exceed ʺ1ʺ (i.e., ʺunityʺ).

[56 FR 23465, May 21, 1991; 56 FR 61352, Dec. 3, 1991. Redesignated and amended at 58 FR 67659, Dec. 22, 1993; 60 FR 20186, Apr. 25, 1995]

ccccc) APPENDIX D TO PART 20‐‐UNITED STATES NUCLEAR REGULATORY COMMISSION REGIONAL OFFICES

Region Address Telephone (24 hour) E‐Mail

NRC Headquarters Operations Center USNRC, Division of Incident Reponse Operations, Washington, DC 20555‐0001

(301) 816‐5100 (301) 951‐0550 (301) 816‐5151 (fax)

[email protected]

Region I: Connecticut, Delaware, District of Columbia, Maine, Maryland, Massachusetts, New Hampshire, New Jersey, New York, Pennsylvania, Rhode Island, and Vermont.

USNRC, Region I, 475 Allendale Road, King of Prussia, PA 19406‐1415.

(610) 337‐5000, (800) 432‐1156 TDD: (301) 415‐5575

[email protected]

Region II: Alabama, Florida, Georgia, Kentucky, Mississippi, North Carolina,

USNRC, Region II, Sam Nunn Atlanta

(404) 562‐4400, (800) 877‐8510

[email protected]

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Puerto Rico, South Carolina, Tennessee, Virginia, Virgin Islands, and West Virginia.

Federal Center, Suite 23T85, 61 Forsyth Street, SW, Atlanta, GA 30303‐8931.

TDD: (301) 415‐5575

Region III: Illinois, Indiana, Iowa, Michigan, Minnesota, Missouri, Ohio, and Wisconsin.

USNRC, Region III, 801 Warrenville Road, Lisle, IL 60532‐4351.

(630) 829‐9500 (800) 522‐3025 TDD: (301) 415‐5575

[email protected]

Region IV: Alaska, Arizona, Arkansas, California, Colorado, Hawaii, Idaho, Kansas, Louisiana, Montana, Nebraska, Nevada, New Mexico, North Dakota, Oklahoma, Oregon, South Dakota, Texas, Utah, Washington, Wyoming, and the U.S. territories and possessions in the Pacific.

USNRC, Region IV, 611 Ryan Plaza Drive, Suite 400, Arlington, TX 76011‐4005.

(817) 860‐8100 (800) 952‐9677 TDD: (301) 415‐5575

[email protected]

[56 FR 23468, May 21, 1991, as amended at 56 FR 41449, Aug. 21, 1991; 58 FR 64111, Dec. 6, 1993; 59 FR 17465, Apr. 13, 1994; 60 FR 24551, May 9, 1995; 62 FR 22880, Apr. 28, 1997; 67 FR 67099, Nov. 4, 2002; 67 FR 77652, Dec. 19, 2002; 68 FR 58802, Oct. 10, 2003]

ddddd) Appendix E to Part 20‐‐[Reserved] eeeee) Appendix F to Part 20‐‐[Reserved] fffff) Appendix G to Part 20‐‐Requirements for Transfers of Low‐Level Radioactive Waste Intended for Disposal at Licensed Land Disposal Facilities and Manifests

I. Manifest

A waste generator, collector, or processor who transports, or offers for transportation, low‐level radioactive waste intended for ultimate disposal at a licensed low‐level radioactive waste land disposal facility must prepare a Manifest (OMB Control Numbers 3150‐0164,‐0165, and‐0166) reflecting information requested on applicable NRC Forms 540 (Uniform Low‐Level Radioactive Waste Manifest (Shipping Paper)) and 541 (Uniform Low‐Level Radioactive Waste Manifest (Container and Waste Description)) and, if necessary, on an applicable NRC Form 542 (Uniform Low‐Level Radioactive Waste Manifest (Manifest Index and Regional Compact Tabulation)). NRC Forms 540 and 540A must be completed and must physically accompany the pertinent low‐level waste shipment. Upon agreement between shipper and consignee, NRC Forms 541 and 541A and 542 and 542A may be completed, transmitted, and stored in electronic media with the capability for producing legible, accurate, and complete records on the respective forms. Licensees are not required by NRC to comply with the manifesting requirements of this part when they ship:

(a) LLW for processing and expect its return (i.e., for storage under their license) prior to disposal at a licensed land disposal facility;

(b) LLW that is being returned to the licensee who is the ʺwaste generatorʺ or ʺgenerator,ʺ as defined in this part; or

(c) Radioactively contaminated material to a ʺwaste processorʺ that becomes the processorʹs ʺresidual waste.ʺ

For guidance in completing these forms, refer to the instructions that accompany the forms. Copies of manifests required by this appendix may be legible carbon copies, photocopies, or computer printouts that reproduce the data in the format of the uniform manifest.

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NRC Forms 540, 540A, 541, 541A, 542 and 542A, and the accompanying instructions, in hard copy, may be obtained by writing or calling the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555‐0001, telephone (301) 415‐5877, or by visiting the NRCʹs Web site at http://www.nrc.gov and selecting forms from the index found on the home page.

This appendix includes information requirements of the Department of Transportation, as codified in 49 CFR part 172. Information on hazardous, medical, or other waste, required to meet Environmental Protection Agency regulations, as codified in 40 CFR parts 259, 261 or elsewhere, is not addressed in this section, and must be provided on the required EPA forms. However, the required EPA forms must accompany the Uniform Low‐Level Radioactive Waste Manifest required by this chapter.

As used in this appendix, the following definitions apply:

Chelating agent has the same meaning as that given in § 61.2 of this chapter.

Chemical description means a description of the principal chemical characteristics of a low‐level radioactive waste.

Computer‐readable medium means that the regulatory agencyʹs computer can transfer the information from the medium into its memory.

Consignee means the designated receiver of the shipment of low‐level radioactive waste.

Decontamination facility means a facility operating under a Commission or Agreement State license whose principal purpose is decontamination of equipment or materials to accomplish recycle, reuse, or other waste management objectives, and, for purposes of this part, is not considered to be a consignee for LLW shipments.

Disposal container means a container principally used to confine low‐level radioactive waste during disposal operations at a land disposal facility (also see ʺhigh integrity containerʺ). Note that for some shipments, the disposal container may be the transport package.

EPA identification number means the number received by a transporter following application to the Administrator of EPA as required by 40 CFR part 263.

Generator means a licensee operating under a Commission or Agreement State license who (1) is a waste generator as defined in this part, or (2) is the licensee to whom waste can be attributed within the context of the Low‐Level Radioactive Waste Policy Amendments Act of 1985 (e.g., waste generated as a result of decontamination or recycle activities).

High integrity container (HIC) means a container commonly designed to meet the structural stability requirements of § 61.56 of this chapter, and to meet Department of Transportation requirements for a Type A package.

Land disposal facility has the same meaning as that given in § 61.2 of this chapter.

NRC Forms 540, 540A, 541, 541A, 542, and 542A are official NRC Forms referenced in this appendix. Licensees need not use originals of these NRC Forms as long as any substitute forms are equivalent to the original documentation in respect to content, clarity, size, and location of information. Upon agreement between the shipper and consignee, NRC Forms 541 (and 541A) and NRC Forms 542 (and 542A) may be completed, transmitted, and stored in electronic media. The electronic media must have the capability for producing legible, accurate, and complete records in the format of the uniform manifest.

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Package means the assembly of components necessary to ensure compliance with the packaging requirements of DOT regulations, together with its radioactive contents, as presented for transport.

Physical description means the items called for on NRC Form 541 to describe a low‐level radioactive waste.

Residual waste means low‐level radioactive waste resulting from processing or decontamination activities that cannot be easily separated into distinct batches attributable to specific waste generators. This waste is attributable to the processor or decontamination facility, as applicable.

Shipper means the licensed entity (i.e., the waste generator, waste collector, or waste processor) who offers low‐level radioactive waste for transportation, typically consigning this type of waste to a licensed waste collector, waste processor, or land disposal facility operator.

Shipping paper means NRC Form 540 and, if required, NRC Form 540A which includes the information required by DOT in 49 CFR part 172.

Source material has the same meaning as that given in § 40.4 of this chapter.

Special nuclear material has the same meaning as that given in § 70.4 of this chapter.

Uniform Low‐Level Radioactive Waste Manifest or uniform manifest means the combination of NRC Forms 540, 541, and, if necessary, 542, and their respective continuation sheets as needed, or equivalent.

Waste collector means an entity, operating under a Commission or Agreement State license, whose principal purpose is to collect and consolidate waste generated by others, and to transfer this waste, without processing or repackaging the collected waste, to another licensed waste collector, licensed waste processor, or licensed land disposal facility.

Waste description means the physical, chemical and radiological description of a low‐level radioactive waste as called for on NRC Form 541.

Waste generator means an entity, operating under a Commission or Agreement State license, who (1) possesses any material or component that contains radioactivity or is radioactively contaminated for which the licensee foresees no further use, and (2) transfers this material or component to a licensed land disposal facility or to a licensed waste collector or processor for handling or treatment prior to disposal. A licensee performing processing or decontamination services may be a ʺwaste generatorʺ if the transfer of low‐level radioactive waste from its facility is defined as ʺresidual waste.ʺ

Waste processor means an entity, operating under a Commission or Agreement State license, whose principal purpose is to process, repackage, or otherwise treat low‐level radioactive material or waste generated by others prior to eventual transfer of waste to a licensed low‐level radioactive waste land disposal facility.

Waste type means a waste within a disposal container having a unique physical description (i.e., a specific waste descriptor code or description; or a waste sorbed on or solidified in a specifically defined media).

Information Requirements

A. General Information

The shipper of the radioactive waste, shall provide the following information on the uniform manifest:

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1. The name, facility address, and telephone number of the licensee shipping the waste;

2. An explicit declaration indicating whether the shipper is acting as a waste generator, collector, processor, or a combination of these identifiers for purposes of the manifested shipment; and

3. The name, address, and telephone number, or the name and EPA identification number for the carrier transporting the waste.

B. Shipment Information

The shipper of the radioactive waste shall provide the following information regarding the waste shipment on the uniform manifest:

1. The date of the waste shipment;

2. The total number of packages/disposal containers;

3. The total disposal volume and disposal weight in the shipment;

4. The total radionuclide activity in the shipment;

5. The activity of each of the radionuclides H‐3, C‐14, Tc‐99, and I‐129 contained in the shipment; and

6. The total masses of U‐233, U‐235, and plutonium in special nuclear material, and the total mass of uranium and thorium in source material.

C. Disposal Container and Waste Information

The shipper of the radioactive waste shall provide the following information on the uniform manifest regarding the waste and each disposal container of waste in the shipment:

1. An alphabetic or numeric identification that uniquely identifies each disposal container in the shipment;

2. A physical description of the disposal container, including the manufacturer and model of any high integrity container;

3. The volume displaced by the disposal container;

4. The gross weight of the disposal container, including the waste;

5. For waste consigned to a disposal facility, the maximum radiation level at the surface of each disposal container;

6. A physical and chemical description of the waste;

7. The total weight percentage of chelating agent for any waste containing more than 0.1% chelating agent by weight, plus the identity of the principal chelating agent;

8. The approximate volume of waste within a container;

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9. The sorbing or solidification media, if any, and the identity of the solidification media vendor and brand name;

10. The identities and activities of individual radionuclides contained in each container, the masses of U‐233, U‐235, and plutonium in special nuclear material, and the masses of uranium and thorium in source material. For discrete waste types (i.e., activated materials, contaminated equipment, mechanical filters, sealed source/devices, and wastes in solidification/stabilization media), the identities and activities of individual radionuclides associated with or contained on these waste types within a disposal container shall be reported;

11. The total radioactivity within each container; and

12. For wastes consigned to a disposal facility, the classification of the waste pursuant to § 61.55 of this chapter. Waste not meeting the structural stability requirements of § 61.56(b) of this chapter must be identified.

D. Uncontainerized Waste Information

The shipper of the radioactive waste shall provide the following information on the uniform manifest regarding a waste shipment delivered without a disposal container:

1. The approximate volume and weight of the waste;

2. A physical and chemical description of the waste;

3. The total weight percentage of chelating agent if the chelating agent exceeds 0.1% by weight, plus the identity of the principal chelating agent;

4. For waste consigned to a disposal facility, the classification of the waste pursuant to § 61.55 of this chapter. Waste not meeting the structural stability requirements of § 61.56(b) of this chapter must be identified;

5. The identities and activities of individual radionuclides contained in the waste, the masses of U‐233, U‐235, and plutonium in special nuclear material, and the masses of uranium and thorium in source material; and

6. For wastes consigned to a disposal facility, the maximum radiation levels at the surface of the waste.

E. Multi‐Generator Disposal Container Information

This section applies to disposal containers enclosing mixtures of waste originating from different generators. (Note: The origin of the LLW resulting from a processorʹs activities may be attributable to one or more ʺgeneratorsʺ (including ʺwaste generatorsʺ) as defined in this part). It also applies to mixtures of wastes shipped in an uncontainerized form, for which portions of the mixture within the shipment originate from different generators.

1. For homogeneous mixtures of waste, such as incinerator ash, provide the waste description applicable to the mixture and the volume of the waste attributed to each generator.

2. For heterogeneous mixtures of waste, such as the combined products from a large compactor, identify each generator contributing waste to the disposal container, and, for discrete waste types (i.e., activated materials, contaminated equipment, mechanical filters, sealed source/devices, and wastes in solidification/stabilization media), the identities and activities of individual radionuclides contained on these waste types within the disposal container. For each generator, provide the following:

(a) The volume of waste within the disposal container;

150

(b) A physical and chemical description of the waste, including the solidification agent, if any;

(c) The total weight percentage of chelating agents for any disposal container containing more than 0.1% chelating agent by weight, plus the identity of the principal chelating agent;

(d) The sorbing or solidification media, if any, and the identity of the solidification media vendor and brand name if the media is claimed to meet stability requirements in 10 CFR 61.56(b); and

(e) Radionuclide identities and activities contained in the waste, the masses of U‐233, U‐235, and plutonium in special nuclear material, and the masses of uranium and thorium in source material if contained in the waste.

II. Certification

An authorized representative of the waste generator, processor, or collector shall certify by signing and dating the shipment manifest that the transported materials are properly classified, described, packaged, marked, and labeled and are in proper condition for transportation according to the applicable regulations of the Department of Transportation and the Commission. A collector in signing the certification is certifying that nothing has been done to the collected waste which would invalidate the waste generatorʹs certification.

III. Control and Tracking

A. Any licensee who transfers radioactive waste to a land disposal facility or a licensed waste collector shall comply with the requirements in paragraphs A.1 through 9 of this section. Any licensee who transfers waste to a licensed waste processor for waste treatment or repackaging shall comply with the requirements of paragraphs A.4 through 9 of this section. A licensee shall:

1. Prepare all wastes so that the waste is classified according to § 61.55 and meets the waste characteristics requirements in § 61.56 of this chapter;

2. Label each disposal container (or transport package if potential radiation hazards preclude labeling of the individual disposal container) of waste to identify whether it is Class A waste, Class B waste, Class C waste, or greater then Class C waste, in accordance with § 61.55 of this chapter;

3. Conduct a quality assurance program to assure compliance with §§ 61.55 and 61.56 of this chapter (the program must include management evaluation of audits);

4. Prepare the NRC Uniform Low‐Level Radioactive Waste Manifest as required by this appendix;

5. Forward a copy or electronically transfer the Uniform Low‐Level Radioactive Waste Manifest to the intended consignee so that either (i) receipt of the manifest precedes the LLW shipment or (ii) the manifest is delivered to the consignee with the waste at the time the waste is transferred to the consignee. Using both (i) and (ii) is also acceptable;

6. Include NRC Form 540 (and NRC Form 540A, if required) with the shipment regardless of the option chosen in paragraph A.5 of this section;

7. Receive acknowledgement of the receipt of the shipment in the form of a signed copy of NRC Form 540;

8. Retain a copy of or electronically store the Uniform Low‐Level Radioactive Waste Manifest and documentation of acknowledgement of receipt as the record of transfer of licensed material as required by 10 CFR Parts 30, 40, and 70

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of this chapter; and

9. For any shipments or any part of a shipment for which acknowledgement of receipt has not been received within the times set forth in this appendix, conduct an investigation in accordance with paragraph E of this appendix.

B. Any waste collector licensee who handles only prepackaged waste shall:

1. Acknowledge receipt of the waste from the shipper within one week of receipt by returning a signed copy of NRC Form 540;

2. Prepare a new manifest to reflect consolidated shipments that meet the requirements of this appendix. The waste collector shall ensure that, for each container of waste in the shipment, the manifest identifies the generator of that container of waste;

3. Forward a copy or electronically transfer the Uniform Low‐Level Radioactive Waste Manifest to the intended consignee so that either: (i) Receipt of the manifest precedes the LLW shipment or (ii) the manifest is delivered to the consignee with the waste at the time the waste is transferred to the consignee. Using both (i) and (ii) is also acceptable;

4. Include NRC Form 540 (and NRC Form 540A, if required) with the shipment regardless of the option chosen in paragraph B.3 of this section;


6. Retain a copy of or electronically store the Uniform Low‐Level Radioactive Waste Manifest and documentation of acknowledgement of receipt as the record of transfer of licensed material as required by 10 CFR parts 30, 40, and 70 of this chapter;

7. For any shipments or any part of a shipment for which acknowledgement of receipt has not been received within the times set forth in this appendix, conduct an investigation in accordance with paragraph E of this appendix; and

8. Notify the shipper and the Administrator of the nearest Commission Regional Office listed in appendix D of this part when any shipment, or part of a shipment, has not arrived within 60 days after receipt of an advance manifest, unless notified by the shipper that the shipment has been cancelled.

C. Any licensed waste processor who treats or repackages waste shall:

1. Acknowledge receipt of the waste from the shipper within one week of receipt by returning a signed copy of NRC Form 540;

2. Prepare a new manifest that meets the requirements of this appendix. Preparation of the new manifest reflects that the processor is responsible for meeting these requirements. For each container of waste in the shipment, the manifest shall identify the waste generators, the preprocessed waste volume, and the other information as required in paragraph I.E. of this appendix;

3. Prepare all wastes so that the waste is classified according to § 61.55 of this chapter and meets the waste characteristics requirements in § 61.56 of this chapter;

4. Label each package of waste to identify whether it is Class A waste, Class B waste, or Class C waste, in accordance with §§ 61.55 and 61.57 of this chapter;

152

5. Conduct a quality assurance program to assure compliance with §§ 61.55 and 61.56 of this chapter (the program shall include management evaluation of audits);

6. Forward a copy or electronically transfer the Uniform Low‐Level Radioactive Waste Manifest to the intended consignee so that either: (i) Receipt of the manifest precedes the LLW shipment or (ii) the manifest is delivered to the consignee with the waste at the time the waste is transferred to the consignee. Using both (i) and (ii) is also acceptable;

7. Include NRC Form 540 (and NRC Form 540A, if required) with the shipment regardless of the option chosen in paragraph C.6 of this section;


9. Retain a copy of or electronically store the Uniform Low‐Level Radioactive Waste Manifest and documentation of acknowledgement of receipt as the record of transfer of licensed material as required by 10 CFR parts 30, 40, and 70 of this chapter;

10. For any shipment or any part of a shipment for which acknowledgement of receipt has not been received within the times set forth in this appendix, conduct an investigation in accordance with paragraph E of this appendix; and


D. The land disposal facility operator shall:

1. Acknowledge receipt of the waste within one week of receipt by returning, as a minimum, a signed copy of NRC Form 540 to the shipper. The shipper to be notified is the licensee who last possessed the waste and transferred the waste to the operator. If any discrepancy exists between materials listed on the Uniform Low‐Level Radioactive Waste Manifest and materials received, copies or electronic transfer of the affected forms must be returned indicating the discrepancy;

2. Maintain copies of all completed manifests and electronically store the information required by 10 CFR 61.80(l) until the Commission terminates the license; and


E. Any shipment or part of a shipment for which acknowledgement is not received within the times set forth in this section must:

1. Be investigated by the shipper if the shipper has not received notification or receipt within 20 days after transfer; and

2. Be traced and reported. The investigation shall include tracing the shipment and filing a report with the nearest Commission Regional Office listed in Appendix D to this part. Each licensee who conducts a trace investigation shall file a written report with the appropriate NRC Regional Office within 2 weeks of completion of the investigation.

[60 FR 15664, Mar. 27, 1995, as amended at 60 FR 25983, May 16, 1995; 68 FR 58802, Oct. 10, 2003]

153

10 CFR PART 40‐‐DOMESTIC LICENSING OF SOURCE MATERIAL

Part Index

General Provisions

Sec.

40.1 Purpose.

40.2 Scope.

40.2a Coverage of inactive tailings sites.

40.3 License requirements.

40.4 Definitions.



40.7 Employee protection.


40.9 Completeness and accuracy of information.

40.10 Deliberate misconduct.

Exemptions

40.11 Persons using source material under certain Department of Energy and Nuclear Regulatory Commission contracts.

40.12 Carriers.

40.13 Unimportant quantities of source material.

40.14 Specific exemptions.

General Licenses

40.20 Types of licenses.

40.21 General license to receive title to source or byproduct material.

40.22 Small quantities of source material.

154

40.23 General license for carriers of transient shipments of natural uranium other than in the form of ore or ore residue.

40.24 [Reserved]

40.25 General license for use of certain industrial products or devices.

40.26 General license for possession and storage of byproduct material as defined in this part.

40.27 General license for custody and long‐term care of residual radioactive material disposal sites.

40.28 General license for custody and long‐term care of uranium or thorium byproduct materials disposal sites.

License Applications

40.31 Application for specific licenses.

40.32 General requirements for issuance of specific licenses.

40.33 Issuance of a license for a uranium enrichment facility.

40.34 Special requirements for issuance of specific licenses.

40.35 Conditions of specific licenses issued pursuant to § 40.34.

40.36 Financial assurance and recordkeeping for decommissioning.

40.38 Ineligibility of certain applicants.

Licenses

40.41 Terms and conditions of licenses.

40.42 Expiration and termination of licenses and decommissioning of sites and separate buildings or outdoor areas.

40.43 Renewal of licenses.

40.44 Amendment of licenses at request of licensee.

40.45 Commission action on applications to renew or amend.

40.46 Inalienability of licenses.

Transfer of Source Material

40.51 Transfer of source or byproduct material.

Records, Reports, and Inspections

155

40.60 Reporting requirements.

40.61 Records.

40.62 Inspections.

40.63 Tests.

40.64 Reports.

40.65 Effluent monitoring reporting requirements.

40.66 Requirements for advance notice of export shipments of natural uranium.

40.67 Requirement for advance notice for importation of natural uranium from countries that are not party to the Convention on the Physical Protection of Nuclear Material.

Modification and Revocation of Licenses

40.71 Modification and revocation of licenses.

Enforcement

40.81 Violations.


Appendix A to Part 40‐‐Criteria Relating to the Operation of Uranium Mills and the Disposition of Tailings or Wastes Produced by the Extraction or Concentration of Source Material From Ores Processed Primarily for Their Source Material Content

Authority: Secs. 62, 63, 64, 65, 81, 161, 182, 183, 186, 68 Stat. 932, 933, 935, 948, 953, 954, 955, as amended, secs. 11e(2), 83, 84, Pub. L. 95‐604, 92 Stat. 3033, as amended, 3039, sec. 234, 83 Stat. 444, as amended (42 U.S.C. 2014(e)(2), 2092, 2093, 2094, 2095, 2111, 2113, 2114, 2201, 2232, 2233, 2236, 2282); sec. 274, Pub. L. 86‐373, 73 Stat. 688 (42 U.S.C. 2021); secs. 201, as amended, 202, 206, 88 Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846); sec. 275, 92 Stat. 3021, as amended by Pub. L. 97‐415, 96 Stat. 2067 (42 U.S.C. 2022); sec. 193, 104 Stat. 2835, as amended by Pub. L. 104‐134, 110 Stat. 1321, 1321‐349 (42 U.S.C. 2243); sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note).

Section 40.7 also issued under Pub. L. 95‐601, sec. 10, 92 Stat. 2951 (42 U.S.C. 5851). Section 40.31(g) also issued under sec. 122, 68 Stat. 939 (42 U.S.C. 2152). Section 40.46 also issued under sec. 184, 68 Stat. 954, as amended (42 U.S.C. 2234). Section 40.71 also issued under sec. 187, 68 Stat. 955 (42 U.S.C. 2237).

Source: 26 FR 284, Jan. 14, 1961, unless otherwise noted.

ggggg) General Provisions hhhhh) § 40.1 Purpose.

(a) The regulations in this part establish procedures and criteria for the issuance of licenses to receive title to, receive, possess, use, transfer, or deliver source and byproduct materials, as defined in this part, and establish and provide

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for the terms and conditions upon which the Commission will issue such licenses. (Additional requirements applicable to natural and depleted uranium at enrichment facilities are set forth in § 70.22 of this chapter.) These regulations also provide for the disposal of byproduct material and for the long‐term care and custody of byproduct material and residual radioactive material. The regulations in this part also establish certain requirements for the physical protection of import, export, and transient shipments of natural uranium. (Additional requirements applicable to the import and export of natural uranium are set forth in part 110 of this chapter.)

(b) The regulations contained in this part are issued under the Atomic Energy Act of 1954, as amended (68 Stat. 919), title II of the Energy Reorganization Act of 1974, as amended (88 Stat. 1242), and titles I and II of the Uranium Mill Tailings Radiation Control Act of 1978, as amended (42 U.S.C. 7901).

[55 FR 45597, Oct. 30, 1990, as amended at 56 FR 55997, Oct. 31, 1991]

iiiii) § 40.2 Scope.

Except as provided in §§ 40.11 to 40.14, inclusive, the regulations in this part apply to all persons in the United States. This part also gives notice to all persons who knowingly provide to any licensee, applicant, contractor, or subcontractor, components, equipment, materials, or other goods or services, that relate to a licenseeʹs or applicantʹs activities subject to this part, that they may be individually subject to NRC enforcement action for violation of § 40.10.

[56 FR 40689, Aug. 15, 1991]

jjjjj) § 40.2a Coverage of inactive tailings sites.

(a) Prior to the completion of the remedial action, the Commission will not require a license pursuant to 10 CFR chapter I for possession of residual radioactive materials as defined in this part that are located at a site where milling operations are no longer active, if the site is covered by the remedial action program of title I of the Uranium Mill Tailings Radiation Control Act of 1978, as amended. The Commission will exert its regulatory role in remedial actions primarily through concurrence and consultation in the execution of the remedial action pursuant to title I of the Uranium Mill Tailings Radiation Control Act of 1978, as amended. After remedial actions are completed, the Commission will license the long‐term care of sites, where residual radioactive materials are disposed, under the requirements set out in § 40.27.

(b) The Commission will regulate byproduct material as defined in this part that is located at a site where milling operations are no longer active, if such site is not covered by the remedial action program of title I of the Uranium Mill Tailings Radiation Control Act of 1978. The criteria in appendix A of this part will be applied to such sites.

[45 FR 65531, Oct. 3, 1980, as amended at 55 FR 45598, Oct. 30, 1990]

kkkkk) § 40.3 License requirements.

A person subject to the regulations in this part may not receive title to, own, receive, possess, use, transfer, provide for long‐term care, deliver or dispose of byproduct material or residual radioactive material as defined in this part or any source material after removal from its place of deposit in nature, unless authorized in a specific or general license issued by the Commission under the regulations in this part.

[55 FR 45598, Oct. 30, 1990]

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lllll) § 40.4 Definitions.

Act means the Atomic Energy Act of 1954 (68 Stat. 919), including any amendments thereto;

Agreement State means any State with which the Atomic Energy Commission or the Nuclear Regulatory Commission has entered into an effective agreement under subsection 274b. of the Atomic Energy Act of 1954, as amended.

Alert means events may occur, are in progress, or have occurred that could lead to a release of radioactive material but that the release is not expected to require a response by offsite response organizations to protect persons offsite.

Byproduct Material means the tailings or wastes produced by the extraction or concentration of uranium or thorium from any ore processed primarily for its source material content, including discrete surface wastes resulting from uranium solution extraction processes. Underground ore bodies depleted by such solution extraction operations do not constitute ʺbyproduct materialʺ within this definition.

With the exception of ʺbyproduct materialʺ as defined in section 11e. of the Act, other terms defined in section 11 of the Act shall have the same meaning when used in the regulations in this part.

Commencement of construction means any clearing of land, excavation, or other substantial action that would adversely affect the natural environment of a site but does not include changes desirable for the temporary use of the land for public recreational uses, necessary borings to determine site characteristics or other preconstruction monitoring to establish background information related to the suitability of a site or to the protection of environmental values.

Commission means the Nuclear Regulatory Commission or its duly authorized representatives.

Corporation means the United States Enrichment Corporation (USEC), or its successor, a Corporation that is authorized by statute to lease the gaseous diffusion enrichment plants in Paducah, Kentucky, and Piketon, Ohio, from the Department of Energy, or any person authorized to operate one or both of the gaseous diffusion plants, or other facilities, pursuant to a plan for the privatization of USEC that is approved by the President.

Decommission means to remove a facility or site safely from service and reduce residual radioactivity to a level that permits‐‐

(1) Release of the property for unrestricted use and termination of the license; or

(2) Release of the property under restricted conditions and termination of the license.

Department and Department of Energy means the Department of Energy established by the Department of Energy Organization Act (Pub. L. 95‐91, 91 Stat. 565, 42 U.S.C. 7101 et seq.) to the extent that the Department, or its duly authorized representatives, exercises functions formerly vested in the U.S. Atomic Energy Commission, its Chairman, members, officers and components and transferred to the U.S. Energy Research and Development Administration and to the Administrator thereof pursuant to sections 104 (b), (c) and (d) of the Energy Reorganization Act of 1974 (Pub. L. 93‐438, 88 Stat. 1233 at 1237, 42 U.S.C. 5814) and retransferred to the Secretary of Energy pursuant to section 301(a) of the Department of Energy Organization Act (Pub. L. 95‐91, 91 Stat. 565 at 577‐578, 42 U.S.C. 7151).

Depleted uranium means the source material uranium in which the isotope uranium‐235 is less than 0.711 weight percent of the total uranium present. Depleted uranium does not include special nuclear material.

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Effective kilogram means (1) for the source material uranium in which the uranium isotope uranium‐235 is greater than 0.005 (0.5 weight percent) of the total uranium present: 10,000 kilograms, and (2) for any other source material: 20,000 kilograms.

Foreign obligations means the commitments entered into by the U.S. Government under Atomic Energy Act (AEA) section 123 agreements for cooperation in the peaceful uses of atomic energy. Imports and exports of material or equipment pursuant to such agreements are subject to these commitments, which in some cases involve an exchange of information on imports, exports, retransfers with foreign governments, peaceful end‐use assurances, and other conditions placed on the transfer of the material or equipment. The U.S. Government informs the licensee of obligations attached to material.

Government agency means any executive department, commission, independent establishment, corporation, wholly or partly owned by the United States of America which is an instrumentality of the United States, or any board, bureau, division, service, office, officer, authority, administration, or other establishment in the executive branch of the Government.

License, except where otherwise specified, means a license issued pursuant to the regulations in this part.

Persons means: (1) Any individual, corporation, partnership, firm, association, trust, estate, public or private institution, group, Government agency other than the Commission or the Department of Energy except that the Department of Energy shall be considered a person within the meaning of the regulations in this part to the extent that its facilities and activities are subject to the licensing and related regulatory authority of the Commission pursuant to section 202 of the Energy Reorganization Act of 1974 (88 Stat. 1244) and the Uranium Mill Tailings Radiation Control Act of 1978 (92 Stat. 3021), any State or any political subdivision of, or any political entity within a State, any foreign government or nation or any subdivision of any such government or nation, or other entity; and (2) any legal successor, representative, agent or agency of the foregoing.

Pharmacist means an individual registered by a state or territory of the United States, the District of Columbia or the Commonwealth of Puerto Rico to compound and dispense drugs, prescriptions and poisons.

Physician means a medical doctor or doctor of osteopathy licensed by a State or Territory of the United States, the District of Columbia, or the Commonwealth of Puerto Rico to prescribe drugs in the practice of medicine.

Principal activities, as used in this part, means activities authorized by the license which are essential to achieving the purpose(s) for which the license was issued or amended. Storage during which no licensed material is accessed for use or disposal and activities incidental to decontamination or decommissioning are not principal activities.

Residual radioactive material means: (1) Waste (which the Secretary of Energy determines to be radioactive) in the form of tailings resulting from the processing of ores for the extraction of uranium and other valuable constituents of the ores; and (2) other waste (which the Secretary of Energy determines to be radioactive) at a processing site which relates to such processing, including any residual stock of unprocessed ores or low‐grade materials. This term is used only with respect to materials at sites subject to remediation under title I of the Uranium Mill Tailings Radiation Control Act of 1978, as amended.

Site area emergency means events may occur, are in progress, or have occurred that could lead to a significant release of radioactive material and that could require a response by offsite response organizations to protect persons offsite.

Source Material means: (1) Uranium or thorium, or any combination thereof, in any physical or chemical form or (2) ores which contain by weight one‐twentieth of one percent (0.05%) or more of: (i) Uranium, (ii) thorium or (iii) any combination thereof. Source material does not include special nuclear material.

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Special nuclear material means: (1) Plutonium, uranium 233, uranium enriched in the isotope 233 or in the isotope 235, and any other material which the Commission, pursuant to the provisions of section 51 of the Act, determines to be special nuclear material; or (2) any material artificially enriched by any of the foregoing.

Transient shipment means a shipment of nuclear material, originating and terminating in foreign countries, on a vessel or aircraft that stops at a United States port.

United States, when used in a geographical sense, includes Puerto Rico and all territories and possessions of the United States.

Unrefined and unprocessed ore means ore in its natural form prior to any processing, such as grinding, roasting or beneficiating, or refining.

Uranium enrichment facility means:

(1) Any facility used for separating the isotopes of uranium or enriching uranium in the isotope 235, except laboratory scale facilities designed or used for experimental or analytical purposes only; or

(2) Any equipment or device, or important component part especially designed for such equipment or device, capable of separating the isotopes of uranium or enriching uranium in the isotope 235.

Uranium Milling means any activity that results in the production of byproduct material as defined in this part.

[26 FR 284, Jan. 14, 1961]

Editorial Note: For Federal Register citations affecting § 40.4, see the List of CFR Sections Affected in the Finding Aids section.

mmmmm) § 40.5 Communications.

(a) Unless otherwise specified or covered under the regional licensing program as provided in paragraph (b) of this section, any communication or report concerning the regulations in this part and any application filed under these regulations may be submitted to the Commission as follows:

(1) By mail addressed: ATTN: Document Control Desk, Director, Office of Nuclear Material Safety and Safeguards, or Director of Nuclear Security, Office of Nuclear Security and Incident Response, U.S. Nuclear Regulatory Commission, Washington, DC 20555‐0001.

(2) By hand delivery to the NRCʹs offices at 11555 Rockville Pike, Rockville, Maryland.

(3) Where practicable, by electronic submission, for example, via Electronic Information Exchange, or CD‐ROM. Electronic submissions must be made in a manner that enables the NRC to receive, read, authenticate, distribute, and archive the submission, and process and retrieve it a single page at a time. Detailed guidance on making electronic submissions can be obtained by visiting the NRCʹs Web site at http://www.nrc.gov/site‐help/eie.html, by calling (301) 415‐6030, by e‐mail to [email protected], or by writing the Office of the Chief Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 20555‐0001. The guidance discusses, among other topics, the formats the NRC can accept, the use of electronic signatures, and the treatment of nonpublic information.

(b) The Commission has delegated to the four Regional Administrators licensing authority for selected parts of its decentralized licensing program for nuclear materials as described in paragraph (b)(1) of this section. Any

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communication, report, or application covered under this licensing program must be submitted to the appropriate Regional Administrator. The administratorsʹ jurisdictions and mailing addresses are listed in paragraph (b)(2) of this section.

(1) The delegated licensing program includes authority to issue, renew, amend, cancel, modify, suspend, or revoke licenses for nuclear materials issued pursuant to 10 CFR parts 30 through 36, 39, 40, and 70 to all persons for academic, medical, and industrial uses, with the following exceptions:

(i) Activities in the fuel cycle and special nuclear material in quantities sufficient to constitute a critical mass in any room or area. This exception does not apply to license modifications relating to termination of special nuclear material licenses that authorize possession of larger quantities when the case is referred for action from NRCʹs Headquarters to the Regional Administrators.

(ii) Health and safety design review of sealed sources and devices and approval, for licensing purposes, of sealed sources and devices.

(iii) Processing of source material for extracting of metallic compounds (including Zirconium, Hafnium, Tantalum, Titanium, Niobium, etc.).

(iv) Distribution of products containing radioactive material to persons exempt pursuant to 10 CFR 32.11 through 32.26.

(v) New uses or techniques for use of byproduct, source, or special nuclear material.

(vi) Uranium enrichment facilities.

(2) Submission‐‐(i) Region I. The regional licensing program involves all Federal facilities in the region and non‐Federal licensees in the following Region I non‐Agreement States and the District of Columbia: Connecticut, Delaware, Maine, Massachusetts, New Jersey, Pennsylvania, and Vermont. All mailed or hand‐delivered inquiries, communications, and applications for a new license or an amendment or renewal of an existing license specified in paragraph (b)(1) of this section must use the following address: U.S. Nuclear Regulatory Commission, Region I, 475 Allendale Road, King of Prussia, Pennsylvania 19406‐1415; where e‐mail is appropriate it should be addressed to [email protected].

(ii) Region II. The regional licensing program involves all Federal facilities in the region and non‐Federal licensees in the following Region II non‐Agreement states and territories: Virginia, West Virginia, Puerto Rico, and the Virgin Islands. All mailed or hand‐delivered inquiries, communications, and applications for a new license or an amendment or renewal of an existing license specified in paragraph (b)(1) of this section must use the following address: U.S. Nuclear Regulatory Commission, Region II Material Licensing/Inspection Branch, Sam Nunn Atlanta Federal Center, Suite 23T85, 61 Forsyth Street, Atlanta, Georgia 30303‐8931; where e‐mail is appropriate it should be addressed to [email protected].

(iii) Region III. The regional licensing program involves all Federal facilities in the region and non‐Federal licensees in the following Region III non‐Agreement States: Indiana, Michigan, Minnesota, Missouri, Ohio, and Wisconsin. All mailed or hand‐delivered inquiries, communications, and applications for a new license or an amendment or renewal of an existing license specified in paragraph (b)(1) of this section must use the following address: U.S. Nuclear Regulatory Commission, Region III, Material Licensing Section, 801 Warrenville Road, Lisle, Illinois 60532‐4351; where e‐mail is appropriate it should be addressed to [email protected].

(iv) Region IV. The regional licensing program involves all Federal facilities in the region and non‐Federal licensees in the following Region IV non‐Agreement States and a territory: Alaska, Hawaii, Montana, Oklahoma, South Dakota,

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Wyoming, and Guam. All mailed or hand‐delivered inquiries, communications, and applications for a new license or an amendment or renewal of an existing license specified in paragraph (b)(1) of this section must use the following address: U.S. Nuclear Regulatory Commission, Region IV, Material Radiation Protection Section, 611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011‐4005; where e‐mail is appropriate it should be addressed to [email protected].

[48 FR 16031, Apr. 14, 1983, as amended at 49 FR 19631, May 9, 1984; 49 FR 47824, Dec. 7, 1984; 50 FR 14694, Apr. 15, 1985; 51 FR 36001, Oct. 8, 1986; 52 FR 8241, Mar. 17, 1987; 52 FR 38392, Oct. 16, 1987; 52 FR 48093, Dec. 18, 1987; 53 FR 3862, Feb. 10, 1988; 53 FR 43420, Oct. 27, 1988; 57 FR 18390, Apr. 30, 1992; 58 FR 7736, Feb. 9, 1993; 58 FR 64111, Dec. 6, 1993; 59 FR 17466, Apr. 13, 1994; 60 FR 24551, May 9, 1995; 62 FR 22880, Apr. 28, 1997; 68 FR 58806, Oct. 10, 2003]

nnnnn) § 40.6 Interpretations.

Except as specifically authorized by the Commission in writing, no interpretation of the meaning of the regulations in this part by any officer or employee of the Commission other than a written interpretation by the General Counsel will be recognized to be binding upon the Commission.

ooooo) § 40.7 Employee protection.

(a) Discrimination by a Commission licensee, an applicant for a Commission license, or a contractor or subcontractor of a Commission licensee or applicant against an employee for engaging in certain protected activities is prohibited. Discrimination includes discharge and other actions that relate to compensation, terms, conditions, or privileges of employment. The protected activities are established in section 211 of the Energy Reorganization Act of 1974, as amended, and in general are related to the administration or enforcement of a requirement imposed under the Atomic Energy Act or the Energy Reorganization Act.

(1) The protected activities include but are not limited to:

(i) Providing the Commission or his or her employer information about alleged violations of either of the statutes named in paragraph (a) introductory text of this section or possible violations of requirements imposed under either of those statutes;

(ii) Refusing to engage in any practice made unlawful under either of the statutes named in paragraph (a) introductory text or under these requirements if the employee has identified the alleged illegality to the employer;

(iii) Requesting the Commission to institute action against his or her employer for the administration or enforcement of these requirements;

(iv) Testifying in any Commission proceeding, or before Congress, or at any Federal or State proceeding regarding any provision (or proposed provision) of either of the statutes named in paragraph (a) introductory text.

(v) Assisting or participating in, or is about to assist or participate in, these activities.

(2) These activities are protected even if no formal proceeding is actually initiated as a result of the employee assistance or participation.

(3) This section has no application to any employee alleging discrimination prohibited by this section who, acting without direction from his or her employer (or the employerʹs agent), deliberately causes a violation of any requirement of the Energy Reorganization Act of 1974, as amended, or the Atomic Energy Act of 1954, as amended.

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(b) Any employee who believes that he or she has been discharged or otherwise discriminated against by any person for engaging in protected activities specified in paragraph (a)(1) of this section may seek a remedy for the discharge or discrimination through an administrative proceeding in the Department of Labor. The administrative proceeding must be initiated within 180 days after an alleged violation occurs. The employee may do this by filing a complaint alleging the violation with the Department of Labor, Employment Standards Administration, Wage and Hour Division. The Department of Labor may order reinstatement, back pay, and compensatory damages.

(c) A violation of paragraphs (a), (e), or (f) of this section by a Commission licensee, an applicant for a Commission license, or a contractor or subcontractor of a Commission licensee or applicant may be grounds for‐‐

(1) Denial, revocation, or suspension of the license.

(2) Imposition of a civil penalty on the licensee or applicant.

(3) Other enforcement action.

(d) Actions taken by an employer, or others, which adversely affect an employee may be predicated upon nondiscriminatory grounds. The prohibition applies when the adverse action occurs because the employee has engaged in protected activities. An employeeʹs engagement in protected activities does not automatically render him or her immune from discharge or discipline for legitimate reasons or from adverse action dictated by nonprohibited considerations.

(e)(1) Each specific licensee, each applicant for a specific license, and each general licensee subject to part 19 shall prominently post the revision of NRC Form 3, ʺNotice to Employeesʺ, referenced in 10 CFR 19.11(c).

(2) The posting of NRC Form 3 must be at locations sufficient to permit employees protected by this section to observe a copy on the way to or from their place of work. Premises must be posted not later than 30 days after an application is docketed and remain posted while the application is pending before the Commission, during the term of the license, and for 30 days following license termination.

(3) Copies of NRC Form 3 may be obtained by writing to the Regional Administrator of the appropriate U.S. Nuclear Regulatory Commission Regional Office listed in appendix D to part 20 of this chapter, by calling (301) 415‐5877, via e‐mail to [email protected], or by visiting the NRCʹs Web site at http://www.nrc.gov and selecting forms from the index found on the home page.

(f) No agreement affecting the compensation, terms, conditions, or privileges of employment, including an agreement to settle a complaint filed by an employee with the Department of Labor pursuant to section 211 of the Energy Reorganization Act of 1974, may contain any provision which would prohibit, restrict, or otherwise discourage an employee from participating in protected activity as defined in paragraph (a)(1) of this section including, but not limited to, providing information to the NRC or to his or her employer on potential violations or other matters within NRCʹs regulatory responsibilities.

[58 FR 52409, Oct. 8, 1993, as amended at 60 FR 24551, May 9, 1995; 61 FR 6765, Feb. 22, 1996; 68 FR 58806, Oct. 10, 2003]

ppppp) § 40.8 Information collection requirements: OMB approval.

(a) The Nuclear Regulatory Commission has submitted the information collection requirements contained in this part to the Office of Management and Budget (OMB) for approval as required by the Paperwork Reduction Act (44 U.S.C. 3501 et seq.). The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of

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information unless it displays a currently valid OMB control number. OMB has approved the information collection requirements contained in this part under control number 3150‐0020.

(b) The approved information collection requirements contained in this part appear in §§ 40.9, 40.23, 40.25, 40.26, 40.27, 40.31, 40.35, 40.36, 40.41, 40.42, 40.43, 40.44, 40.51, 40.60, 40.61, 40.64, 40.65, 40.66, 40.67, and appendix A to this part.

(c) This part contains information collection requirements in addition to those approved under the control number specified in paragraph (a) of this section. These information collection requirements and the control numbers under which they are approved are as follows:

(1) In §§ 40.31, 40.43, 40.44, and appendix A, NRC Form 313 is approved under control number 3150‐0120.

(2) In § 40.31, Form N‐71 is approved under control number 3150‐0056.

(3) In § 40.42, NRC Form 314 is approved under control number 3150‐0028.

(4) In § 40.64, DOE/NRC Form 741 is approved under control number 3150‐0003.

[49 FR 19626, May 9, 1984, as amended at 56 FR 40768, Aug. 16, 1991; 58 FR 68731, Dec. 29, 1993; 62 FR 52187, Oct. 6, 1997]

qqqqq) § 40.9 Completeness and accuracy of information.

(a) Information provided to the Commission by an applicant for a license or by a licensee or information required by statute or by the Commissionʹs regulations, orders, or license conditions to be maintained by the applicant or the licensee shall be complete and accurate in all material respects.

(b) Each applicant or licensee shall notify the Commission of information identified by the applicant or licensee as having for the regulated activity a significant implication for public health and safety or common defense and security. An applicant or licensee violates this paragraph only if the applicant or licensee fails to notify the Commission of information that the applicant or licensee has identified as having a significant implication for public health and safety or common defense and security. Notification shall be provided to the Administrator of the appropriate Regional Office within two working days of identifying the information. This requirement is not applicable to information which is already required to be provided to the Commission by other reporting or undating requirements.

[52 FR 49371, Dec. 31, 1987]

rrrrr) § 40.10 Deliberate misconduct.

(a) Any licensee, applicant for a license, employee of a licensee or applicant; or any contractor (including a supplier or consultant), subcontractor, employee of a contractor or subcontractor of any licensee or applicant for a license, who knowingly provides to any licensee, applicant, contractor, or subcontractor, any components, equipment, materials, or other goods or services that relate to a licenseeʹs or applicantʹs activities in this part, may not:

(1) Engage in deliberate misconduct that causes or would have caused, if not detected, a licensee or applicant to be in violation of any rule, regulation, or order; or any term, condition, or limitation of any license issued by the Commission; or

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(2) Deliberately submit to the NRC, a licensee, an applicant, or a licenseeʹs or applicantʹs contractor or subcontractor, information that the person submitting the information knows to be incomplete or inaccurate in some respect material to the NRC.

(b) A person who violates paragraph (a)(1) or (a)(2) of this section may be subject to enforcement action in accordance with the procedures in 10 CFR part 2, subpart B.

(c) For the purposes of paragraph (a)(1) of this section, deliberate misconduct by a person means an intentional act or omission that the person knows:

(1) Would cause a licensee or applicant to be in violation of any rule, regulation, or order; or any term, condition, or limitation, of any license issued by the Commission; or

(2) Constitutes a violation of a requirement, procedure, instruction, contract, purchase order, or policy of a licensee, applicant, contractor, or subcontractor.

[63 FR 1896, Jan. 13, 1998]

sssss) Exemptions ttttt) § 40.11 Persons using source material under certain Department of Energy and Nuclear Regulatory Commission contracts.

Except to the extent that Department facilities or activities of the types subject to licensing pursuant to section 202 of the Energy Reorganization Act of 1974 or the Uranium Mill Tailings Radiation Control Act of 1978 are involved, any prime contractor of the Department is exempt from the requirements for a license set forth in sections 62, 63, and 64 of the Act and from the regulations in this part to the extent that such contractor, under his prime contract with the Department, receives, possesses, uses, transfers or delivers source material for: (a) The performance of work for the Department at a United States Government‐owned or controlled site, including the transportation of source material to or from such site and the performance of contract services during temporary interruptions of such transportation; (b) research in, or development, manufacture, storage, testing or transportation of, atomic weapons or components thereof; or (c) the use or operation of nuclear reactors or other nuclear devices in a United States Government‐owned vehicle or vessel. In addition to the foregoing exemptions, and subject to the requirement for licensing of Department facilities and activities pursuant to section 202 of the Energy Reorganization Act of 1974 or the Uranium Mill Tailings Radiation Control Act of 1980, any prime contractor or subcontractor of the Department or the Commission is exempt from the requirements for a license set forth in sections 62, 63, and 64 of the Act and from the regulations in this part to the extent that such prime contractor or subcontractor receives, possesses, uses, transfers or delivers source material under his prime contract or subcontract when the Commission determines that the exemption of the prime contractor or subcontractor is authorized by law; and that, under the terms of the contract or subcontract, there is adequate assurance that the work thereunder can be accomplished without undue risk to the public health and safety.

[40 FR 8787, Mar. 3, 1975, as amended at 43 FR 6923, Feb. 17, 1978; 45 FR 65531, Oct. 3, 1980]

uuuuu) § 40.12 Carriers.

(a) Except as specified in paragraph (b) of this section, common and contract carriers, freight forwarders, warehousemen, and the U.S. Postal Service are exempt from the regulations in this part and the requirements for a license set forth in section 62 of the Act to the extent that they transport or store source material in the regular course of the carriage for another or storage incident thereto.

165

(b) The exemption in paragraph (a) of this section does not apply to a person who possesses a transient shipment (as defined in § 40.4(r)), an import shipment, or an export shipment of natural uranium in an amount exceeding 500 kilograms, unless the shipment is in the form of ore or ore residue.

[52 FR 9651, Mar. 26, 1987]

vvvvv) § 40.13 Unimportant quantities of source material.

(a) Any person is exempt from the regulations in this part and from the requirements for a license set forth in section 62 of the Act to the extent that such person receives, possesses, uses, transfers or delivers source material in any chemical mixture, compound, solution, or alloy in which the source material is by weight less than one‐twentieth of 1 percent (0.05 percent) of the mixture, compound, solution or alloy. The exemption contained in this paragraph does not include byproduct material as defined in this part.

(b) Any person is exempt from the regulations in this part and from the requirements for a license set forth in section 62 of the act to the extent that such person receives, possesses, uses, or transfers unrefined and unprocessed ore containing source material; provided, that, except as authorized in a specific license, such person shall not refine or process such ore.

(c) Any person is exempt from the regulation in this part and from the requirements for a license set forth in section 62 of the Act to the extent that such person receives, possesses, uses, or transfers:

(1) Any quantities of thorium contained in (i) incandescent gas mantles, (ii) vacuum tubes, (iii) welding rods, (iv) electric lamps for illuminating purposes: Provided, That each lamp does not contain more than 50 milligrams of thorium, (v) germicidal lamps, sunlamps, and lamps for outdoor or industrial lighting: Provided, That each lamp does not contain more than 2 grams of thorium, (vi) rare earth metals and compounds, mixtures, and products containing not more than 0.25 percent by weight thorium, uranium, or any combination of these, or (vii) personnel neutron dosimeters: Provided, That each dosimeter does not contain more than 50 milligrams of thorium.

(2) Source material contained in the following products:

(i) Glazed ceramic tableware, provided that the glaze contains not more than 20 percent by weight source material;

(ii) Piezoelectric ceramic containing not more than 2 percent by weight source material;

(iii) Glassware containing not more than 10 percent by weight source material; but not including commercially manufactured glass brick, pane glass, ceramic tile, or other glass or ceramic used in construction;

(iv) Glass enamel or glass enamel frit containing not more than 10 percent by weight source material imported or ordered for importation into the United States, or initially distributed by manufacturers in the United States, before July 25, 1983.1

(3) Photographic film, negatives, and prints containing uranium or thorium;

(4) Any finished product or part fabricated of, or containing tungsten or magnesium‐thorium alloys, provided that the thorium content of the alloy does not exceed 4 percent by weight and that the exemption contained in this subparagraph shall not be deemed to authorize the chemical, physical or metallurgical treatment or processing of any such product or part; and

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(5) Uranium contained in counterweights installed in aircraft, rockets, projectiles, and missiles, or stored or handled in connection with installation or removal of such counterweights: Provided, That:

(i) The counterweights are manufactured in accordance with a specific license issued by the Commission or the Atomic Energy Commission authorizing distribution by the licensee pursuant to this paragraph;

(ii) Each counterweight has been impressed with the following legend clearly legible through any plating or other covering: ʺDepleted Uraniumʺ;2

(iii) Each counterweight is durably and legibly labeled or marked with the identification of the manufacturer, and the statement: ʺUnauthorized Alterations Prohibitedʺ;2 and

(iv) The exemption contained in this paragraph shall not be deemed to authorize the chemical, physical, or metallurgical treatment or processing of any such counterweights other than repair or restoration of any plating or other covering.

(6) Natural or depleted uranium metal used as shielding constituting part of any shipping container: Provided, That:

(i) The shipping container is conspicuously and legibly impressed with the legend ʺCAUTION‐‐RADIOACTIVE SHIELDING‐‐URANIUMʺ; and

(ii) The uranium metal is encased in mild steel or equally fire resistant metal of minimum wall thickness of one‐eighth inch (3.2 mm).

(7) Thorium contained in finished optical lenses, provided that each lens does not contain more than 30 percent by weight of thorium; and that the exemption contained in this subparagraph shall not be deemed to authorize either:

(i) The shaping, grinding or polishing of such lens or manufacturing processes other than the assembly of such lens into optical systems and devices without any alteration of the lens; or

(ii) The receipt, possession, use, transfer, or of thorium contained in contact lenses, or in spectacles, or in eyepieces in binoculars or other optical instruments.

(8) Thorium contained in any finished aircraft engine part containing nickel‐thoria alloy, Provided, That:

(i) The thorium is dispersed in the nickel‐thoria alloy in the form of finely divided thoria (thorium dioxide); and

(ii) The thorium content in the nickel‐thoria alloy does not exceed 4 percent by weight.

(9) The exemptions in this paragraph (c) do not authorize the manufacture of any of the products described.

(d) Any person is exempt from the regulations in this part and from the requirements for a license set forth in section 62 of the Act to the extent that such person receives, possesses, uses, or transfers uranium contained in detector heads for use in fire detection units, provided that each detector head contains not more than 0.005 microcurie of uranium. The exemption in this paragraph does not authorize the manufacture of any detector head containing uranium.

[26 FR 284, Jan. 14, 1961]

Editorial Note: For Federal Register citations affecting § 40.13, see the List of CFR Sections Affected in the Finding Aids section.

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1 On July 25, 1983, the exemption of glass enamel or glass enamel frit was suspended. The exemption was eliminated on September 11, 1984.

2 The requirements specified in paragraphs (c)(5) (ii) and (iii) of this section need not be met by counterweights manufactured prior to Dec. 31, 1969: Provided, That such counterweights were manufactured under a specific license issued by the Atomic Energy Commission and were impressed with the legend required by § 40.13(c)(5)(ii) in effect on June 30, 1969.

wwwww) § 40.14 Specific exemptions.

(a) The Commission may, upon application of any interested person or upon its own initiative, grant such exemptions from the requirements of the regulation in this part as it determines are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest.

(b) [Reserved]

(c) The Department of Energy is exempt from the requirements of this part to the extent that its activities are subject to the requirements of part 60 or 63 of this chapter.

(d) Except as specifically provided in part 61 of this chapter any licensee is exempt from the requirements of this part to the extent that its activities are subject to the requirements of part 61 of this chapter.

[37 FR 5747, Mar. 21, 1972, as amended at 39 FR 26279, July 18, 1974; 40 FR 8787, Mar. 3, 1975; 45 FR 65531, Oct. 3, 1980; 46 FR 13979, Feb. 25, 1981; 47 FR 57481, Dec. 27, 1982; 66 FR 55790, Nov. 2, 2001]

xxxxx) General Licenses yyyyy) § 40.20 Types of licenses.

(a) Licenses for source material and byproduct material are of two types: general and specific. Licenses for long‐term care and custody of residual radioactive material at disposal sites are general licenses. The general licenses provided in this part are effective without the filing of applications with the Commission or the issuance of licensing documents to particular persons. Specific licenses are issued to named persons upon applications filed pursuant to the regulations in this part.

(b) Section 40.27 contains a general license applicable for custody and long‐term care of residual radioactive material at uranium mill tailings disposal sites remediated under title I of the Uranium Mill Tailings Radiation Control Act of 1978, as amended.

(c) Section 40.28 contains a general license applicable for custody and long‐term care of byproduct material at uranium or thorium mill tailings disposal sites under title II of the Uranium Mill Tailings Radiation Control Act of 1978, as amended.

[55 FR 45598, Oct. 30, 1990]

zzzzz) § 40.21 General license to receive title to source or byproduct material.

A general license is hereby issued authorizing the receipt of title to source or byproduct material, as defined in this part, without regard to quantity. This general license does not authorize any person to receive, possess, deliver, use, or transfer source or byproduct material.

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[45 FR 65531, Oct. 3, 1980]

aaaaaa) § 40.22 Small quantities of source material.

(a) A general license is hereby issued authorizing commercial and industrial firms, research, educational and medical institutions and Federal, State and local government agencies to use and transfer not more than fifteen (15) pounds of source material at any one time for research, development, educational, commercial or operational purposes. A person authorized to use or transfer source material, pursuant to this general license, may not receive more than a total of 150 pounds of source material in any one calendar year.

(b) Persons who receive, possess, use, or transfer source material pursuant to the general license issued in paragraph (a) of this section are exempt from the provisions of parts 19, 20, and 21, of this chapter to the extent that such receipt, possession, use or transfer are within the terms of such general license: Provided, however, That this exemption shall not be deemed to apply to any such person who is also in possession of source material under a specific license issued pursuant to this part.

(c) Persons who receive, possess, use or transfer source material pursuant to the general license in paragraph (a) of this section are prohibited from administering source material, or the radiation therefrom, either externally or internally, to human beings except as may be authorized by NRC in a specific license.

[26 FR 284, Jan. 14, 1961, as amended at 38 FR 22221, Aug. 17, 1973; 42 FR 28896, June 6, 1977; 45 FR 55420, Aug. 20, 1980]

bbbbbb) § 40.23 General license for carriers of transient shipments of natural uranium other than in the form of ore or ore residue.

(a) A general license is hereby issued to any person to possess a transient shipment of natural uranium, other than in the form of ore or ore residue, in amounts exceeding 500 kilograms.

(b)(1) Persons generally licensed under paragraph (a) of this section, who plan to carry a transient shipment with scheduled stops at a United States port, shall notify the Director, Division of Nuclear Security, Office of Nuclear Security and Incident Response, using an appropriate method listed in § 40.5. The notification must be in writing and must be received at least 10 days before transport of the shipment commences at the shipping facility.

(2) The notification must include the following information:

(i) Location of all scheduled stops in United States territory;

(ii) Arrival and departure times for all scheduled stops in United States territory;

(iii) The type of transport vehicle;

(iv) A physical description of the shipment;

(v) The numbers and types of containers;

(vi) The name and telephone number of the carrierʹs representatives at each stopover location in the United States territory;

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(vii) A listing of the modes of shipments, transfer points, and routes to be used;

(viii) The estimated date and time that shipment will commence and that each nation (other than the United States) along the route is scheduled to be entered;

(ix) For shipment between countries that are not party to the Convention on the Physical Protection of Nuclear Material (i.e., not listed in appendix F to part 73 of this chapter), a certification that arrangements have been made to notify the Director, Division of Nuclear Security when the shipment is received at the destination facility.

(c) Persons generally licensed under this section making unscheduled stops at United States ports, immediately after the decision to make an unscheduled stop, shall provide to the Director, Division of Nuclear Security the information required under paragraph (b) of this section.

(d) A licensee who needs to amend a notification may do so by telephoning the Division of Nuclear Security at (301) 415‐6828.

[52 FR 9651, Mar. 26, 1987, as amended at 53 FR 4110, Feb. 12, 1988; 60 FR 24551, May 9, 1995; 68 FR 58807, Oct. 10, 2003]

cccccc) § 40.24 [Reserved] dddddd) § 40.25 General license for use of certain industrial products or devices.

(a) A general license is hereby issued to receive, acquire, possess, use, or transfer, in accordance with the provisions of paragraphs (b), (c), (d), and (e) of this section, depleted uranium contained in industrial products or devices for the purpose of providing a concentrated mass in a small volume of the product or device.

(b) The general license in paragraph (a) of this section applies only to industrial products or devices which have been manufactured or initially transferred in accordance with a specific license issued pursuant to § 40.34 (a) of this part or in accordance with a specific license issued by an Agreement State which authorizes manufacture of the products or devices for distribution to persons generally licensed by the Agreement State.

(c)(1) Persons who receive, acquire, possess, or use depleted uranium pursuant to the general license established by paragraph (a) of this section shall file NRC Form 244, ʺRegistration Certificate‐‐Use of Depleted Uranium Under General License,ʺ with the Director of the NRCʹs Division of Industrial and Medical Nuclear Safety, by an appropriate method listed in § 40.5, with a copy to the appropriate NRC Regional Administrator. The form shall be submitted within 30 days after the first receipt or acquisition of such depleted uranium. The registrant shall furnish on Form NRC 244 the following information and such other information as may be required by that form:

(i) Name and address of the registrant;

(ii) A statement that the registrant has developed and will maintain procedures designed to establish physical control over the depleted uranium described in paragraph (a) of this section and designed to prevent transfer of such depleted uranium in any form, including metal scrap, to persons not authorized to receive the depleted uranium; and

(iii) Name and/or title, address, and telephone number of the individual duly authorized to act for and on behalf of the registrant in supervising the procedures identified in paragraph (c)(1)(ii) of this section.

(2) The registrant possessing or using depleted uranium under the general license established by paragraph (a) of this section shall report in writing to the Director, Division of Industrial and Medical Nuclear Safety, with a copy to the Regional Administrator of the appropriate U.S. Nuclear Regulatory Commission Regional Office listed in appendix D

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of part 20 of this chapter, any changes in information furnished by him in the Form NRC 244 ʺRegistration Certificate‐‐Use of Depleted Uranium Under General License.ʺ The report shall be submitted within 30 days after the effective date of such change.

(d) A person who receives, acquires, possesses, or uses depleted uranium pursuant to the general license established by paragraph (a) of this section:

(1) Shall not introduce such depleted uranium, in any form, into a chemical, physical, or metallurgical treatment or process, except a treatment or process for repair or restoration of any plating or other covering of the depleted uranium.

(2) Shall not abandon such depleted uranium.

(3) Shall transfer or dispose of such depleted uranium only by transfer in accordance with the provisions of § 40.51 of this part. In the case where the transferee receives the depleted uranium pursuant to the general license established by paragraph (a) of this section, the transferor shall furnish the transferee a copy of this section and a copy of Form NRC 244. In the case where the transferee receives the depleted uranium pursuant to a general license contained in an Agreement Stateʹs regulation equivalent to this section, the transferor shall furnish the transferee a copy of this section and a copy of Form NRC 244 accompanied by a note explaining that use of the product or device is regulated by the Agreement State under requirements substantially the same as those in this section.

(4) Within 30 days of any transfer, shall report in writing to the Director, Division of Industrial and Medical Nuclear Safety, with a copy to the Regional Administrator of the appropriate U.S. Nuclear Regulatory Commission Regional Office listed in appendix D of part 20 of this chapter, the name and address of the person receiving the source material pursuant to such transfer.

(e) Any person receiving, acquiring, possessing, using, or transferring depleted uranium pursuant to the general license established by paragraph (a) of this section is exempt from the requirements of parts 19, 20 and 21 of this chapter with respect to the depleted uranium covered by that general license.

[41 FR 53331, Dec. 6, 1976, as amended at 42 FR 28896, June 6, 1977; 43 FR 6923, Feb. 17, 1978; 43 FR 52202, Nov. 9, 1978; 52 FR 31611, Aug. 21, 1987; 60 FR 24551, May 9, 1995; 68 FR 58807, Oct. 10, 2003]

eeeeee) § 40.26 General license for possession and storage of byproduct material as defined in this part.

(a) A general license is hereby issued to receive title to, own, or possess byproduct material as defined in this part without regard to form or quantity.

(b) The general license in paragraph (a) of this section applies only: In the case of licensees of the Commission, where activities that result in the production of byproduct material are authorized under a specific license issued by the Commission pursuant to this part, to byproduct material possessed or stored at an authorized disposal containment area or transported incident to such authorized activity: Provided, That authority to receive title to, own, or possess byproduct material under this general license shall terminate when the specific license for source material expires, is renewed, or is amended to include a specific license for byproduct material as defined in this part.

(c) The general license in paragraph (a) of this section is subject to:

(1) The provisions of parts 19, 20, 21, and §§ 40.1, 40.2a, 40.3, 40.4, 40.5, 40.6, 40.41, 40.46, 40.60, 40.61, 40.62, 40.63, 40.65, 40.71, and 40.81 of part 40 of this chapter; and

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(2) The documentation of daily inspections of tailings or waste retention systems and the immediate notification of the appropriate NRC regional office as indicated in appendix D to 10 CFR part 20 of this chapter, or the Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555, of any failure in a tailings or waste retention system that results in a release of tailings or waste into unrestricted areas, or of any unusual conditions (conditions not contemplated in the design of the retention system) that if not corrected could lead to failure of the system and result in a release of tailings or waste into unrestricted areas; and any additional requirements the Commission may by order deem necessary. The licensee shall retain this documentation of each daily inspection as a record for three years after each inspection is documented.

(d) The general license in paragraph (a) of this section shall expire nine months from the effective date of this subparagraph unless an applicable licensee has submitted, pursuant to the provisions of § 40.31 of this part, an application for license renewal or amendment which includes a detailed program for meeting the technical and financial criteria contained in appendix A of this part.

[44 FR 50014, Aug. 24, 1979, as amended at 45 FR 12377, Feb. 26, 1980; 45 FR 65531, Oct. 3, 1980; 53 FR 19248, May 27, 1988; 56 FR 40768, Aug. 16, 1991]

ffffff) § 40.27 General license for custody and long‐term care of residual radioactive material disposal sites.

(a) A general license is issued for the custody of and long‐term care, including monitoring, maintenance, and emergency measures necessary to protect public health and safety and other actions necessary to comply with the standards promulgated under section 275(a) of the Atomic Energy Act of 1954, as amended, for disposal sites under title I of the Uranium Mill Tailings Radiation Control Act of 1978, as amended. The license is available only to the Department of Energy, or another Federal agency designated by the President to provide long‐term care. The purpose of this general license is to ensure that uranium mill tailings disposal sites will be cared for in such a manner as to protect the public health, safety, and the environment after remedial action has been completed.

(b) The general license in paragraph (a) of this section becomes effective when the Commission accepts a site Long‐Term Surveillance Plan (LTSP) that meets the requirements of this section, and when the Commission concurs with the Department of Energyʹs determination of completion of remedial action at each disposal site. There is no termination of this general license. The LTSP may incorporate by reference information contained in documents previously submitted to the Commission if the references to the individual incorporated documents are clear and specific. Each LTSP must include‐‐

(1) A legal description of the disposal site to be licensed, including documentation on whether land and interests are owned by the United States or an Indian tribe. If the site is on Indian land, then, as specified in the Uranium Mill Tailings Radiation Control Act of 1978, as amended, the Indian tribe and any person holding any interest in the land shall execute a waiver releasing the United States of any liability or claim by the Tribe or person concerning or arising from the remedial action and holding the United States harmless against any claim arising out of the performance of the remedial action;

(2) A detailed description, which can be in the form of a reference, of the final disposal site conditions, including existing ground water characterization and any necessary ground water protection activities or strategies. This description must be detailed enough so that future inspectors will have a baseline to determine changes to the site and when these changes are serious enough to require maintenance or repairs. If the disposal site has continuing aquifer restoration requirements, then the licensing process will be completed in two steps. The first step includes all items other than ground water restoration. Ground water monitoring, which would be addressed in the LTSP, may still be required in this first step to assess performance of the tailings disposal units. When the Commission concurs with the completion of ground water restoration, the licensee shall assess the need to modify the LTSP and report

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results to the Commission. If the proposed modifications meet the requirements of this section, the LTSP will be considered suitable to accommodate the second step.

(3) A description of the long‐term surveillance program, including proposed inspection frequency and reporting to the Commission (as specified in appendix A, criterion 12 of this part), frequency and extent of ground water monitoring if required, appropriate constituent concentration limits for ground water, inspection personnel qualifications, inspection procedures, recordkeeping and quality assurance procedures;

(4) The criteria for follow‐up inspections in response to observations from routine inspections or extreme natural events; and

(5) The criteria for instituting maintenance or emergency measures.

(c) The long‐term care agency under the general license established by paragraph (a) of this section shall‐‐

(1) Implement the LTSP as described in paragraph (b) of this section;

(2) Care for the disposal site in accordance with the provisions of the LTSP;

(3) Notify the Commission of any changes to the LTSP; the changes may not conflict with the requirements of this section;

(4) Guarantee permanent right‐of‐entry to Commission representatives for the purpose of periodic site inspections; and

(5) Notify the Commission prior to undertaking any significant construction, actions, or repairs related to the disposal site, even if the action is required by a State or another Federal agency.

(d) As specified in the Uranium Mill Tailings Radiation Control Act of 1978, as amended, the Secretary of the Interior, with the concurrence of the Secretary of Energy and the Commission, may sell or lease any subsurface mineral rights associated with land on which residual radioactive materials are disposed. In such cases, the Commission shall grant a license permitting use of the land if it finds that the use will not disturb the residual radioactive materials or that the residual radioactive materials will be restored to a safe and environmentally sound condition if they are disturbed by the use.

(e) The general license in paragraph (a) of this section is exempt from parts 19, 20, and 21 of this chapter, unless significant construction, actions, or repairs are required. If these types of actions are to be undertaken, the licensee shall explain to the Commission which requirements from these parts apply for the actions and comply with the appropriate requirements.

[55 FR 45598, Oct. 30, 1990]

gggggg) § 40.28 General license for custody and long‐term care of uranium or thorium byproduct materials disposal sites.

(a) A general license is issued for the custody of and long‐term care, including monitoring, maintenance, and emergency measures necessary to protect the public health and safety and other actions necessary to comply with the standards in this part for uranium or thorium mill tailings sites closed under title II of the Uranium Mill Tailings Radiation Control Act of 1978, as amended. The licensee will be the Department of Energy, another Federal agency designated by the President, or a State where the disposal site is located. The purpose of this general license is to

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ensure that uranium and thorium mill tailings disposal sites will be cared for in such a manner as to protect the public health, safety, and the environment after closure.

(b) The general license in paragraph (a) of this section becomes effective when the Commission terminates, or concurs in an Agreement Stateʹs termination of, the current specific license and a site Long‐Term Surveillance Plan (LTSP) meeting the requirements of this section has been accepted by the Commission. There is no termination of this general license. If the LTSP has not been formally received by the NRC prior to termination of the current specific license, the Commission may issue a specific order to the intended custodial agency to ensure continued control and surveillance of the disposal site to protect the public health, safety, and the environment. The Commission will not unnecessarily delay the termination of the specific license solely on the basis that an acceptable LTSP has not been received. The LTSP may incorporate by reference information contained in documents previously submitted to the Commission if the references to the individual incorporated documents are clear and specific. Each LTSP must include‐‐

(1) A legal description of the disposal site to be transferred (unless transfer is exempted under provisions of the Atomic Energy Act, § 83(b)(1)(A)) and licensed;

(2) A detailed description, which can be in the form of a reference of the final disposal site conditions, including existing ground water characterization. This description must be detailed enough so that future inspectors will have a baseline to determine changes to the site and when these changes are serious enough to require maintenance or repairs;

(3) A description of the long‐term surveillance program, including proposed inspection frequency and reporting to the Commission (as specified in appendix A, Criterion 12 of this part), frequency and extent of ground water monitoring if required, appropriate constituent concentration limits for ground water, inspection personnel qualifications, inspection procedures, recordkeeping and quality assurance procedures;

(4) The criteria for follow‐up inspections in response to observations from routine inspections or extreme natural events; and

(5) The criteria for instituting maintenance or emergency measures.

(c) The long‐term care agency who has a general license established by paragraph (a) of this section shall‐‐

(1) Implement the LTSP as described in paragraph (b) of this section;

(2) Care for the disposal site in accordance with the provisions of the LTSP;

(3) Notify the Commission of any changes to the LTSP; the changes may not conflict with the requirements of this section;

(4) Guarantee permanent right‐of‐entry to Commission representatives for the purpose of periodic site inspections; and

(5) Notify the Commission prior to undertaking any significant construction, actions, or repairs related to the disposal site, even if the action is required by a State or another Federal agency.

(d) Upon application, the Commission may issue a specific license, as specified in the Uranium Mill Tailings Radiation Control Act of 1978, as amended, permitting the use of surface and/or subsurface estates transferred to the United States or a State. Although an application may be received from any person, if permission is granted, the

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person who transferred the land to DOE or the State shall receive the right of first refusal with respect to this use of the land. The application must demonstrate that‐‐

(1) The proposed action does not endanger the public health, safety, welfare, or the environment;

(2) Whether the proposed action is of a temporary or permanent nature, the site would be maintained and/or restored to meet requirements in appendix A of this part for closed sites; and

(3) Adequate financial arrangements are in place to ensure that the byproduct materials will not be disturbed, or if disturbed that the applicant is able to restore the site to a safe and environmentally sound condition.

(e) The general license in paragraph (a) of this section is exempt from parts 19, 20, and 21 of this chapter, unless significant construction, actions, or repairs are required. If these types of actions are to be undertaken, the licensee shall explain to the Commission which requirements from these parts apply for the actions and comply with the appropriate requirements.

(f) In cases where the Commission determines that transfer of title of land used for disposal of any byproduct materials to the United States or any appropriate State is not necessary to protect the public health, safety or welfare or to minimize or eliminate danger to life or property (Atomic Energy Act, § 83(b)(1)(A)), the Commission will consider specific modifications of the custodial agencyʹs LTSP provisions on a case‐by‐case basis.

[55 FR 45599, Oct. 30, 1990]

hhhhhh) License Applications iiiiii) § 40.31 Application for specific licenses.

(a) A person may file an application for specific license on NRC Form 313, ʺApplication for Material License,ʺ in accordance with the instructions in § 40.5 of this chapter. Information contained in previous applications, statements or reports filed with the Commission may be incorporated by reference provided that the reference is clear and specific.

(b) The Commission may at any time after the filing of the original application, and before the expiration of the license, require further statements in order to enable the Commission to determine whether the application should be granted or denied or whether a license should be modified or revoked. All applications and statements shall be signed by the applicant or licensee or a person duly authorized to act for and on his behalf.

(c) Applications and documents submitted to the Commission in connection with applications will be made available for public inspection in accordance with the provisions of the regulations contained in parts 2 and 9 of this chapter.

(d) An application for a license filed pursuant to the regulations in this part will be considered also as an application for licenses authorizing other activities for which licenses are required by the Act: Provided, That the application specifies the additional activities for which licenses are requested and complies with regulations of the Commission as to applications for such licenses.

(e) Each application for a source material license, other than a license exempted from part 170 of this chapter, shall be accompanied by the fee prescribed in § 170.31 of this chapter. No fee will be required to accompany an application for renewal or amendment of a license, except as provided in § 170.31 of this chapter.

(f) An application for a license to possess and use source material for uranium milling, production of uranium hexafluoride, or for the conduct of any other activity which the Commission has determined pursuant to subpart A of

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part 51 of this chapter will significantly affect the quality of the environment shall be filed at least 9 months prior to commencement of construction of the plant or facility in which the activity will be conducted and shall be accompanied by any Environmental Report required pursuant to subpart A of part 51 of this chapter.

(g) In response to a written request by the Commission, an applicant for a license to possess and use source material in a uranium hexafluoride production plant or a fuel fabrication plant and any other applicant for a license to possess and use more than one effective kilogram of source material (except for ore processing, as defined in § 75.4(o) of this chapter) shall file with the Commission the installation information described in § 75.11 of this chapter, on Form N‐71. The applicant shall also permit verification of this installation information by the International Atomic Energy Agency and take other action as may be necessary to implement the US/IAEA Safeguards Agreement, in the manner set forth § 75.6 and §§ 75.11 through 75.14 of this chapter.

(h) An application for a license to receive, possess, and use source material for uranium or thorium milling or byproduct material, as defined in this part, at sites formerly associated with such milling shall contain proposed written specifications relating to milling operations and the disposition of the byproduct material to achieve the requirements and objectives set forth in appendix A of this part. Each application must clearly demonstrate how the requirements and objectives set forth in appendix A of this part have been addressed. Failure to clearly demonstrate how the requirements and objectives in appendix A have been addressed shall be grounds for refusing to accept an application.

(i) As provided by § 40.36, certain applications for specific licenses filed under this part must contain a proposed decommissioning funding plan or a certification of financial assurance for decommissioning. In the case of renewal applications submitted before July 27, 1990, this submittal may follow the renewal application but must be submitted on or before July 27, 1990.

(j)(1) Each application to possess uranium hexafluoride in excess of 50 kilograms in a single container or 1000 kilograms total must contain either:

(i) An evaluation showing that the maximum intake of uranium by a member of the public due to a release would not exceed 2 milligrams; or

(ii) An emergency plan for responding to the radiological hazards of an accidental release of source material and to any associated chemical hazards directly incident thereto.

(2) One or more of the following factors may be used to support an evaluation submitted under paragraph (j)(1)(i) of this section:

(i) All or part of the radioactive material is not subject to release during an accident because of the way it is stored or packaged;

(ii) Facility design or engineered safety features in the facility would reduce the amount of the release; or

(iii) Other factors appropriate for the specific facility.

(3) An emergency plan submitted under paragraph (j)(1)(ii) of this section must include the following:

(i) Facility description. A brief description of the licenseeʹs facility and area near the site.

(ii) Types of accidents. An identification of each type of accident for which protective actions may be needed.

(iii) Classification of accidents. A classification system for classifying accidents as alerts or site area emergencies.

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(iv) Detection of accidents. Identification of the means of detecting each type of radioactive materials accident in a timely manner.

(v) Mitigation of consequences. A brief description of the means and equipment for mitigating the consequences of each type of accident, including those provided to protect workers onsite, and a description of the program for maintaining the equipment.

(vi) Assessment of releases. A brief description of the methods and equipment to assess releases of radioactive materials.

(vii) Responsibilities. A brief description of the responsibilities of licensee personnel should an accident occur, including identification of personnel responsible for promptly notifying offsite response organizations and the NRC; also responsibilities for developing, maintaining, and updating the plan.

(viii) Notification and coordination. A commitment to and a brief description of the means to promptly notify offsite response organizations and request offsite assistance, including medical assistance for the treatment of contaminated injured onsite workers when appropriate. A control point must be established. The notification and coordination must be planned so that unavailability of some personnel, parts of the facility, and some equipment will not prevent the notification and coordination. The licensee shall also commit to notify the NRC operations center immediately after notification of the offsite response organizations and not later than one hour after the licensee declares an emergency.1

(ix) Information to be communicated. A brief description of the types of information on facility status, radioactive releases, and recommended protective actions, if necessary, to be given to offsite response organizations and to the NRC.

(x) Training. A brief description of the frequency, performance objectives and plans for the training that the licensee will provide workers on how to respond to an emergency including any special instructions and orientation tours the licensee would offer to fire, police, medical and other emergency personnel. The training shall familiarize personnel with site‐specific emergency procedures. Also, the training shall thoroughly prepare site personnel for their responsibilities in the event of accident scenarios postulated as most probable for the specific site, including the use of team training for such scenarios.

(xi) Safe shutdown. A brief description of the means of restoring the facility to a safe condition after an accident.

(xii) Exercises. Provisions for conducting quarterly communications checks with offsite response organizations and biennial onsite exercises to test response to simulated emergencies. Quarterly communications checks with offsite response organizations must include the check and update of all necessary telephone numbers. The licensee shall invite offsite response organizations to participate in the biennial exercises. Participation of offsite response organizations in biennial exercises although recommended is not required. Exercises must use accident scenarios postulated as most probable for the specific site and the scenarios shall not be known to most exercise participants. The licensee shall critique each exercise using individuals not having direct implementation responsibility for the plan. Critiques of exercises must evaluate the appropriateness of the plan, emergency procedures, facilities, equipment, training of personnel, and overall effectiveness of the response. Deficiencies found by the critiques must be corrected.

(xiii) Hazardous chemicals. A certification that the application has met its responsibilities under the Emergency Planning and Community Right‐to‐Know Act of 1986, title III, Pub. L. 99‐499, if applicable to the applicantʹs activities at the proposed place of the use of the source material.

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(4) The licensee shall allow the offsite response organizations expected to respond in case of an accident 60 days to comment on the licenseeʹs emergency plan before submitting it to the NRC. The licensee shall provide any comments received within the 60 days to the NRC with the emergency plan.

(k) A license application for a uranium enrichment facility must be accompanied by an Environmental Report required under subpart A of part 51 of this chapter.

(l) A license application that involves the use of source material in a uranium enrichment facility must include the applicantʹs provisions for liability insurance.

[26 FR 284, Jan. 14, 1961, as amended at 31 FR 4669, Mar. 19, 1966; 34 FR 19546, Dec. 11, 1969; 36 FR 145, Jan. 6, 1971; 37 FR 5748, Mar. 21, 1972; 46 FR 13497, Feb. 23, 1981; 49 FR 9403, Mar. 12, 1984; 49 FR 19626, May 9, 1984; 49 FR 21699, May 23, 1984; 49 FR 27924, July 9, 1984; 53 FR 24047, June 27, 1988; 54 FR 14061, Apr. 7, 1989; 57 FR 18390, Apr. 30, 1992; 68 FR 58807, Oct. 10, 2003]

1 These reporting requirements do not supersede or release licensees of complying with the requirements under the Emergency Planning and Community Right‐to‐Know Act of 1986, Title III. Pub. L. 99‐499 or other state or federal reporting requirements.

jjjjjj) § 40.32 General requirements for issuance of specific licenses.

An application for a specific license will be approved if:

(a) The application is for a purpose authorized by the Act; and

(b) The applicant is qualified by reason of training and experience to use the source material for the purpose requested in such manner as to protect health and minimize danger to life or property; and

(c) The applicantʹs proposed equipment, facilities and procedures are adequate to protect health and minimize danger to life or property; and

(d) The issuance of the license will not be inimical to the common defense and security or to the health and safety of the public; and

(e) In the case of an application for a license for a uranium enrichment facility, or for a license to possess and use source and byproduct material for uranium milling, production of uranium hexafluoride, or for the conduct of any other activity which the Commission determines will significantly affect the quality of the environment, the Director of Nuclear Material Safety and Safeguards or his designee, before commencement of construction of the plant or facility in which the activity will be conducted, on the basis of information filed and evaluations made pursuant to subpart A of part 51 of this chapter, has concluded, after weighing the environmental, economic, technical and other benefits against environmental costs and considering available alternatives, that the action called for is the issuance of the proposed license, with any appropriate conditions to protect environmental values. Commencement of construction prior to this conclusion is grounds for denial of a license to possess and use source and byproduct material in the plant or facility. As used in this paragraph, the term ʺcommencement of constructionʺ means any clearing of land, excavation, or other substantial action that would adversely affect the environment of a site. The term does not mean site exploration, roads necessary for site exploration, borings to determine foundation conditions, or other preconstruction monitoring or testing to establish background information related to the suitability of the site or the protection of environmental values.

(f) The applicant satisfies any applicable special requirements contained in § 40.34.

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(g) If the proposed activity involves use of source material in a uranium enrichment facility, the applicant has satisfied the applicable provisions of part 140 of this chapter.

[26 FR 284, Jan. 14, 1961, as amended at 36 FR 12731, July 7, 1971; 40 FR 8787, Mar. 3, 1975; 41 FR 53332, Dec. 6, 1976; 43 FR 6924, Feb. 17, 1978; 49 FR 9403, Mar. 12, 1984; 57 FR 18390, Apr. 30, 1992]

kkkkkk) § 40.33 Issuance of a license for a uranium enrichment facility.

(a) The Commission will hold a hearing pursuant to 10 CFR part 2, subparts A, G, and I, on each application with regard to the licensing of the construction and operation of a uranium enrichment facility. The Commission will publish public notice of the hearing in the Federal Register at least 30 days before the hearing.

(b) A license for a uranium enrichment facility may not be issued before the hearing is completed and a decision issued on the application.

[57 FR 18391, Apr. 30, 1992]

llllll) § 40.34 Special requirements for issuance of specific licenses.

(a) An application for a specific license to manufacture industrial products and devices containing depleted uranium, or to initially transfer such products or devices, for use pursuant to § 40.25 of this part or equivalent regulations of an Agreement State, will be approved if:

(1) The applicant satisfies the general requirements specified in § 40.32;

(2) The applicant submits sufficient information relating to the design, manufacture, prototype testing, quality control procedures, labeling or marking, proposed uses, and potential hazards of the industrial product or device to provide reasonable assurance that possession, use, or transfer of the depleted uranium in the product or device is not likely to cause any individual to receive in 1 year a radiation dose in excess of 10 percent of the annual limits specified in § 20.1201(a) of this chapter; and

(3) The applicant submits sufficient information regarding the industrial product or device and the presence of depleted uranium for a mass‐volume application in the product or device to provide reasonable assurance that unique benefits will accrue to the public because of the usefulness of the product or device.

(b) In the case of an industrial product or device whose unique benefits are questionable, the Commission will approve an application for a specific license under this paragraph only if the product or device is found to combine a high degree of utility and low probability of uncontrolled disposal and dispersal of significant quantities of depleted uranium into the environment.

(c) The Commission may deny an applicant for a specific license under this paragraph if the end uses of the industrial product or device cannot be reasonably foreseen.

[41 FR 53332, Dec. 6, 1976, as amended at 43 FR 6924, Feb. 17, 1978; 58 FR 67661, Dec. 22, 1993; 59 FR 41643, Aug. 15, 1994]

mmmmmm) § 40.35 Conditions of specific licenses issued pursuant to § 40.34.

Each person licensed pursuant to § 40.34 shall:

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(a) Maintain the level of quality control required by the license in the manufacture of the industrial product or device, and in the installation of the depleted uranium into the product or device;

(b) Label or mark each unit to: (1) Identify the manufacturer or initial transferor of the product or device and the number of the license under which the product or device was manufactured or initially transferred, the fact that the product or device contains depleted uranium, and the quantity of depleted uranium in each product or device; and (2) state that the receipt, possession, use, and transfer of the product or device are subject to a general license or the equivalent and the regulations of the U.S. NRC or of an Agreement State;

(c) Assure that the depleted uranium before being installed in each product or device has been impressed with the following legend clearly legible through any plating or other covering: ʺDepleted Uraniumʺ;

(d)(1) Furnish a copy of the general license contained in § 40.25 and a copy of Form NRC 244 to each person to whom he transfers source material in a product or device for use pursuant to the general license contained in § 40.25; or

(2) Furnish a copy of the general license contained in the Agreement Stateʹs regulation equivalent to § 40.25 and a copy of the Agreement Stateʹs certificate, or alternately, furnish a copy of the general license contained in § 40.25 and a copy of Form NRC 244 to each person to whom he transfers source material in a product or device for use pursuant to the general license of an Agreement State. If a copy of the general license in § 40.25 and a copy of Form NRC 244 are furnished to such person, they shall be accompanied by a note explaining that use of the product or device is regulated by the Agreement State under requirements substantially the same as those in § 40.25; and

(e)(1) Report to the Director of the Office of Nuclear Material Safety and Safeguards, by an appropriate method listed in § 40.5, all transfers of industrial products or devices to persons for use under the general license in § 40.25. Such report shall identify each general licensee by name and address, an individual by name and/or position who may constitute a point of contact between the Commission and the general licensee, the type and model number of device transferred, and the quantity of depleted uranium contained in the product or device. The report shall be submitted within 30 days after the end of each calendar quarter in which such a product or device is transferred to the generally licensed person. If no transfers have been made to persons generally licensed under § 40.25 during the reporting period, the report shall so indicate;

(2) Report to the responsible Agreement State Agency all transfers of industrial products or devices to persons for use under the general license in the Agreement Stateʹs regulation equivalent to § 40.25. Such report shall identify each general licensee by name and address, an individual by name and/or position who may constitute a point of contact between the Agency and the general licensee, the type and model number of device transferred, and the quantity of depleted uranium contained in the product or device. The report shall be submitted within 30 days after the end of each calendar quarter in which such product or device is transferred to the generally licensed person. If no transfers have been made to a particular Agreement State during the reporting period, this information shall be reported to the responsible Agreement State Agency;

(3) Keep records showing the name, address, and a point of contact for each general license to whom he or she transfers depleted uranium in industrial products or devices for use pursuant to the general license provided in § 40.25 or equivalent regulations of an Agreement State. The records must be retained for three years from the date of transfer and must show the date of each transfer, the quantity of depleted uranium in each product or device transferred, and compliance with the report requirements of this section.

(f) Licensees required to submit emergency plans by § 40.31(i) shall follow the emergency plan approved by the Commission. The licensee may change the plan without Commission approval if the changes do not decrease the effectiveness of the plan. The licensee shall furnish the change to the Director of the Office of Nuclear Material Safety and Safeguards, by an appropriate method listed in § 40.5, and to affected offsite response organizations, within six months after the change is made. Proposed changes that decrease the effectiveness of the approved emergency plan may not be implemented without application to and prior approval by the Commission.

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[41 FR 53332, Dec. 6, 1976, as amended at 43 FR 6924, Feb. 17, 1978; 52 FR 31611, Aug. 21, 1987; 53 FR 19248, May 27, 1988; 54 FR 14062, Apr. 7, 1989; 68 FR 58807, Oct. 10, 2003]

nnnnnn) § 40.36 Financial assurance and recordkeeping for decommissioning.

Except for licenses authorizing the receipt, possession, and use of source material for uranium or thorium milling, or byproduct material at sites formerly associated with such milling, for which financial assurance requirements are set forth in appendix A of this part, criteria for providing financial assurance for decommissioning are as follows:

(a) Each applicant for a specific license authorizing the possession and use of more than 100 mCi of source material in a readily dispersible form shall submit a decommissioning funding plan as described in paragraph (d) of this section.

(b) Each applicant for a specific license authorizing possession and use of quantities of source material greater than 10 mCi but less than or equal to 100 mCi in a readily dispersible form shall either‐‐

(1) Submit a decommissioning funding plan as described in paragraph (d) of this section; or

(2) Submit a certification that financial assurance for decommissioning has been provided in the amount of $225,000 by June 2, 2005 using one of the methods described in paragraph (e) of this section. For an applicant, this certification may state that the appropriate assurance will be obtained after the application has been approved and the license issued but before the receipt of licensed material. If the applicant defers execution of the financial instrument until after the license has been issued, a signed original of the financial instrument obtained to satisfy the requirements of paragraph (e) of this section must be submitted to NRC prior to receipt of licensed material. If the applicant does not defer execution of the financial instrument, the applicant shall submit to NRC, as part of the certification, a signed original of the financial instrument obtained to satisfy the requirements of paragraph (e) of this section.

(c)(1) Each holder of a specific license issued on or after July 27, 1990, which is covered by paragraph (a) or (b) of this section, shall provide financial assurance for decommissioning in accordance with the criteria set forth in this section.

(2) Each holder of a specific license issued before July 27, 1990, and of a type described in paragraph (a) of this section shall submit a decommissioning funding plan as described in paragraph (d) of this section or a certification of financial assurance for decommissioning in an amount at least equal to $1,125,000 in accordance with the criteria set forth in this section. If the licensee submits the certification of financial assurance rather than a decommissioning funding plan, the licensee shall include a decommissioning funding plan in any application for license renewal. Licensees required to submit the $1,125,000 amount must do so by December 2, 2004.

(3) Each holder of a specific license issued before July 27, 1990, and of a type described in paragraph (b) of this section shall submit, on or before July 27, 1990, a decommissioning funding plan, as described in paragraph (d) of this section, or a certification of financial assurance for decommissioning in accordance with the criteria set forth in this section.

(4) Any licensee who has submitted an application before July 27, 1990, for renewal of license in accordance with § 40.43 shall provide financial assurance for decommissioning in accordance with paragraphs (a) and (b) of this section. This assurance must be submitted when this rule becomes effective November 24, 1995.

(d) Each decommissioning funding plan must contain a cost estimate for decommissioning and a description of the method of assuring funds for decommissioning from paragraph (e) of this section, including means for adjusting cost estimates and associated funding levels periodically over the life of the facility. Cost estimates must be adjusted at intervals not to exceed 3 years. The decommissioning funding plan must also contain a certification by the licensee that financial assurance for decommissioning has been provided in the amount of the cost estimate for

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decommissioning and a signed original of the financial instrument obtained to satisfy the requirements of paragraph (e) of this section.

(e) Financial assurance for decommissioning must be provided by one or more of the following methods:

(1) Prepayment. Prepayment is the deposit prior to the start of operation into an account segregated from licensee assets and outside the licenseeʹs administrative control of cash or liquid assets such that the amount of funds would be sufficient to pay decommissioning costs. Prepayment may be in the form of a trust, escrow account, government fund, certificate of deposit, or deposit of government securities.

(2) A surety method, insurance, or other guarantee method. These methods guarantee that decommissioning costs will be paid. A surety method may be in the form of a surety bond, letter of credit, or line of credit. A parent company guarantee of funds for decommissioning costs based on a financial test may be used if the guarantee and test are as contained in appendix A to part 30. A parent company guarantee may not be used in combination with other financial methods to satisfy the requirements of this section. For commercial corporations that issue bonds, a guarantee of funds by the applicant or licensee for decommissioning costs based on a financial test may be used if the guarantee and test are as contained in appendix C to part 30. For commercial companies that do not issue bonds, a guarantee of funds by the applicant or licensee for decommissioning costs may be used if the guarantee and test are as contained in appendix D to part 30. For nonprofit entities, such as colleges, universities, and nonprofit hospitals, a guarantee of funds by the applicant or licensee may be used if the guarantee and test are as contained in appendix E to part 30. A guarantee by the applicant or licensee may not be used in combination with any other financial methods used to satisfy the requirements of this section or in any situation where the applicant or licensee has a parent company holding majority control of the voting stock of the company. Any surety method or insurance used to provide financial assurance for decommissioning must contain the following conditions:

(i) The surety method or insurance must be open‐ended or, if written for a specified term, such as five years, must be renewed automatically unless 90 days or more prior to the renewal date, the issuer notifies the Commission, the beneficiary, and the licensee of its intention not to renew. The surety method or insurance must also provide that the full face amount be paid to the beneficiary automatically prior to the expiration without proof of forfeiture if the licensee fails to provide a replacement acceptable to the Commission within 30 days after receipt of notification of cancellation.

(ii) The surety method or insurance must be payable to a trust established for decommissioning costs. The trustee and trust must be acceptable to the Commission. An acceptable trustee includes an appropriate State or Federal government agency or an entity which has the authority to act as a trustee and whose trust operations are regulated and examined by a Federal or State agency.

(iii) The surety method or insurance must remain in effect until the Commission has terminated the license.

(3) An external sinking fund in which deposits are made at least annually, coupled with a surety method or insurance, the value of which may decrease by the amount being accumulated in the sinking fund. An external sinking fund is a fund established and maintained by setting aside funds periodically in an account segregated from licensee assets and outside the licenseeʹs administrative control in which the total amount of funds would be sufficient to pay decommissioning costs at the time termination of operation is expected. An external sinking fund may be in the form of a trust, escrow account, government fund, certificate of deposit, or deposit of government securities. The surety or insurance provision must be as stated in paragraph (e)(2) of this section.

(4) In the case of Federal, State, or local government licensees, a statement of intent containing a cost estimate for decommissioning or an amount based on paragraph (b) of this section, and indicating that funds for decommissioning will be obtained when necessary.

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(5) When a governmental entity is assuming custody and ownership of a site, an arrangement that is deemed acceptable by such governmental entity.

(f) Each person licensed under this part shall keep records of information important to the decommissioning of a facility in an identified location until the site is released for unrestricted use. Before licensed activities are transferred or assigned in accordance with § 40.41(b) licensees shall transfer all records described in this paragraph to the new licensee. In this case, the new licensee will be responsible for maintaining these records until the license is terminated. If records important to the decommissioning of a facility are kept for other purposes, reference to these records and their locations may be used. Information the Commission considers important to decommissioning consists of‐‐

(1) Records of spills or other unusual occurrences involving the spread of contamination in and around the facility, equipment, or site. These records may be limited to instances when contamination remains after any cleanup procedures or when there is reasonable likelihood that contaminants may have spread to inaccessible areas as in the case of possible seepage into porous materials such as concrete. These records must include any known information on identification of involved nuclides, quantities, forms, and concentrations.

(2) As‐built drawings and modifications of structures and equipment in restricted areas where radioactive materials are used and/or stored, and of locations of possible inaccessible contamination such as buried pipes which may be subject to contamination. If required drawings are referenced, each relevant document need not be indexed individually. If drawings are not available, the licensee shall substitute appropriate records of available information concerning these areas and locations.

(3) Except for areas containing depleted uranium used only for shielding or as penetrators in unused munitions, a list contained in a single document and updated every 2 years, of the following:

(i) All areas designated and formerly designated as restricted areas as defined under 10 CFR 20.1003;

(ii) All areas outside of restricted areas that require documentation under § 40.36(f)(1);

(iii) All areas outside of restricted areas where current and previous wastes have been buried as documented under 10 CFR 20.2108; and

(iv) All areas outside of restricted areas that contain material such that, if the license expired, the licensee would be required to either decontaminate the area to meet the criteria for decommissioning in 10 CFR part 20, subpart E, or apply for approval for disposal under 10 CFR 20.2002.

(4) Records of the cost estimate performed for the decommissioning funding plan or of the amount certified for decommissioning, and records of the funding method used for assuring funds if either a funding plan or certification is used.

[53 FR 24047, June 27, 1988, as amended at 58 FR 39633, July 26, 1993; 58 FR 67661, Dec. 22, 1993; 58 FR 68731, Dec. 29, 1993; 59 FR 1618, Jan. 12, 1994; 60 FR 38238, July 26, 1995; 61 FR 24674, May 16, 1996; 62 FR 39090, July 21, 1997; 63 FR 29543, June 1, 1998; 68 FR 57336, Oct. 3, 2003]

oooooo) § 40.38 Ineligibility of certain applicants.

A license may not be issued to the Corporation if the Commission determines that:

(a) The Corporation is owned, controlled, or dominated by an alien, a foreign corporation, or a foreign government; or

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(b) The issuance of such a license would be inimical to‐‐

(1) The common defense and security of the United States; or

(2) The maintenance of a reliable and economical domestic source of enrichment services.

[62 FR 6669, Feb. 12, 1997]

pppppp) Licenses qqqqqq) § 40.41 Terms and conditions of licenses.

(a) Each license issued pursuant to the regulations in this part shall be subject to all the provisions of the act, now or hereafter in effect, and to all rules, regulations and orders of the Commission.

(b) Neither the license nor any right under the license shall be assigned or otherwise transferred in violation of the provisions of the Act.

(c) Each person licensed by the Commission pursuant to the regulations in this part shall confine his possession and use of source or byproduct material to the locations and purposes authorized in the license. Except as otherwise provided in the license, a license issued pursuant to the regulations in this part shall carry with it the right to receive, possess, and use source or byproduct material. Preparation for shipment and transport of source or byproduct material shall be in accordance with the provisions of part 71 of this chapter.

(d) Each license issued pursuant to the regulations in this part shall be deemed to contain the provisions set forth in sections 183b.‐d., of the Act, whether or not said provisions are expressly set forth in the license.

(e) The Commission may incorporate in any license at the time of issuance, or thereafter, by appropriate rule, regulation or order, such additional requirements and conditions with respect to the licenseeʹs receipt, possession, use, and transfer of source or byproduct material as it deems appropriate or necessary in order to:

(1) Promote the common defense and security;

(2) Protect health or to minimize danger of life or property;

(3) Protect restricted data;

(4) Require such reports and the keeping of such records, and to provide for such inspections of activities under the license as may be necessary or appropriate to effectuate the purposes of the act and regulations thereunder.

(f)(1) Each licensee shall notify the appropriate NRC Regional Administrator, in writing, immediately following the filing of a voluntary or involuntary petition for bankruptcy under any chapter of title 11 (Bankruptcy) of the United States Code by or against:

(i) The licensee;

(ii) An entity (as that term is defined in 11 U.S.C. 101(14)) controlling the licensee or listing the license or licensee as property of the estate; or

(iii) An affiliate (as that term is defined in 11 U.S.C. 101(2)) of the licensee.

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(2) This notification must indicate:

(i) The bankruptcy court in which the petition for bankruptcy was filed; and

(ii) The date of the filing of the petition.

(g) No person may commence operation of a uranium enrichment facility until the Commission verifies through inspection that the facility has been constructed in accordance with the requirements of the license. The Commission shall publish notice of the inspection results in the Federal Register.

[26 FR 284, Jan. 14, 1961, as amended at 31 FR 15145, Dec. 2, 1966; 45 FR 65531, Oct. 3, 1980; 48 FR 32328, July 15, 1983; 52 FR 1295, Jan. 12, 1987; 57 FR 18391, Apr. 30, 1992]

rrrrrr) § 40.42 Expiration and termination of licenses and decommissioning of sites and separate buildings or outdoor areas.

(a)(1) Except as provided in paragraph (a)(2) of this section, each specific license expires at the end of the day on the expiration date stated in the license unless the licensee has filed an application for renewal under § 40.43 not less than 30 days before the expiration date stated in the existing license (or, for those licenses subject to paragraph (a)(2) of this section, 30 days before the deemed expiration date in that paragraph). If an application for renewal has been filed at least 30 days before the expiration date stated in the existing license (or, for those licenses subject to paragraph (a)(2) of this section, 30 days before the deemed expiration date in that paragraph), the existing license expires at the end of the day on which the Commission makes a final determination to deny the renewal application or, if the determination states an expiration date, the expiration date stated in the determination.

(2) Each specific license that has an expiration date after July 1, 1995, and is not one of the licenses described in paragraph (a)(3) of this section, shall be deemed to have an expiration date that is five years after the expiration date stated in the current license.

(3) The following specific licenses are not subject to, or otherwise affected by, the provisions of paragraph (a)(2) of this section:

(i) Specific licenses for which, on February 15, 1996, an evaluation or an emergency plan is required in accordance with § 40.31(j);

(ii) Specific licenses whose holders are subject to the financial assurance requirements specified in 10 CFR 40.36, and on February 15, 1996, the holders either:

(A) Have not submitted a decommissioning funding plan nor certification of financial assurance for decommissioning; or

(B) Have not received written notice that the decommissioning funding plan or certification of financial assurance for decommissioning is acceptable;

(iii) Specific licenses whose holders are listed in the SDMP List published in NUREG 1444, Supplement 1 (November 1995);

(iv) Specific licenses whose issuance, amendment, or renewal, as of February 15, 1996, is not a categorical exclusion under 10 CFR 51.22(c)(14) and, therefore, need an environmental assessment or environmental impact statement pursuant to Subpart A of Part 51 of this chapter;

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(v) Specific licenses whose holders have not had at least one NRC inspection of licensed activities before February 15, 1996;

(vi) Specific licenses whose holders, as the result of the most recent NRC inspection of licensed activities conducted before February 15, 1996, have been:

(A) Cited for a Severity Level I, II, or III violation in a Notice of Violation;

(B) Subject to an Order issued by the NRC; or

(C) Subject to a CAL issued by the NRC.

(vii) Specific licenses with expiration dates before July 1, 1995, for which the holders have submitted applications for renewal under 10 CFR 40.43 of this part.

(b) Each specific license revoked by the Commission expires at the end of the day on the date of the Commissionʹs final determination to revoke the license, or on the expiration date stated in the determination, or as otherwise provided by Commission Order.

(c) Each specific license continues in effect, beyond the expiration date if necessary, with respect to possession of source material until the Commission notifies the licensee in writing that the license is terminated. During this time, the licensee shall‐‐

(1) Limit actions involving source material to those related to decommissioning; and

(2) Continue to control entry to restricted areas until they are suitable for release in accordance with NRC requirements;

(d) Within 60 days of the occurrence of any of the following, consistent with the administrative directions in § 40.5, each licensee shall provide notification to the NRC in writing and either begin decommissioning its site, or any separate building or outdoor area that contains residual radioactivity, so that the building or outdoor area is suitable for release in accordance with NRC requirements, or submit within 12 months of notification a decommissioning plan, if required by paragraph (g)(1) of this section, and begin decommissioning upon approval of that plan if‐‐

(1) The license has expired pursuant to paragraph (a) or (b) of this section; or

(2) The licensee has decided to permanently cease principal activities, as defined in this part, at the entire site or in any separate building or outdoor area; or

(3) No principal activities under the license have been conducted for a period of 24 months; or

(4) No principal activities have been conducted for a period of 24 months in any separate building or outdoor area that contains residual radioactivity such that the building or outdoor area is unsuitable for release in accordance with NRC requirements.

(e) Coincident with the notification required by paragraph (d) of this section, the licensee shall maintain in effect all decommissioning financial assurances established by the licensee pursuant to § 40.36 in conjunction with a license issuance or renewal or as required by this section. The amount of the financial assurance must be increased, or may be decreased, as appropriate, to cover the detailed cost estimate for decommissioning established pursuant to paragraph (g)(4)(v) of this section.

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(1) Any licensee who has not provided financial assurance to cover the detailed cost estimate submitted with the decommissioning plan shall do so when this rule becomes effective November 24, 1995.

(2) Following approval of the decommissioning plan, a licensee may reduce the amount of the financial assurance as decommissioning proceeds and radiological contamination is reduced at the site with the approval of the Commission.

(f) The Commission may grant a request to delay or postpone initiation of the decommissioning process if the Commission determines that such relief is not detrimental to the public health and safety and is otherwise in the public interest. The request must be submitted no later than 30 days before notification pursuant to paragraph (d) of this section. The schedule for decommissioning set forth in paragraph (d) of this section may not commence until the Commission has made a determination on the request.

(g)(1) A decommissioning plan must be submitted if required by license condition or if the procedures and activities necessary to carry out decommissioning of the site or separate building or outdoor area have not been previously approved by the Commission and these procedures could increase potential health and safety impacts to workers or to the public, such as in any of the following cases:

(i) Procedures would involve techniques not applied routinely during cleanup or maintenance operations;

(ii) Workers would be entering areas not normally occupied where surface contamination and radiation levels are significantly higher than routinely encountered during operation;

(iii) Procedures could result in significantly greater airborne concentrations of radioactive materials than are present during operation; or

(iv) Procedures could result in significantly greater releases of radioactive material to the environment than those associated with operation.

(2) The Commission may approve an alternate schedule for submittal of a decommissioning plan required pursuant to paragraph (d) of this section if the Commission determines that the alternative schedule is necessary to the effective conduct of decommissioning operations and presents no undue risk from radiation to the public health and safety and is otherwise in the public interest.

(3) The procedures listed in paragraph (g)(1) of this section may not be carried out prior to approval of the decommissioning plan.

(4) The proposed decommissioning plan for the site or separate building or outdoor area must include:

(i) A description of the conditions of the site or separate building or outdoor area sufficient to evaluate the acceptability of the plan;

(ii) A description of planned decommissioning activities;

(iii) A description of methods used to ensure protection of workers and the environment against radiation hazards during decommissioning;

(iv) A description of the planned final radiation survey; and

(v) An updated detailed cost estimate for decommissioning, comparison of that estimate with present funds set aside for decommissioning, and a plan for assuring the availability of adequate funds for completion of decommissioning.

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(vi) For decommissioning plans calling for completion of decommissioning later than 24 months after plan approval, a justification for the delay based on the criteria in paragraph (i) of this section.

(5) The proposed decommissioning plan will be approved by the Commission if the information therein demonstrates that the decommissioning will be completed as soon as practicable and that the health and safety of workers and the public will be adequately protected.

(h)(1) Except as provided in paragraph (i) of this section, licensees shall complete decommissioning of the site or separate building or outdoor area as soon as practicable but no later than 24 months following the initiation of decommissioning.

(2) Except as provided in paragraph (i) of this section, when decommissioning involves the entire site, the licensee shall request license termination as soon as practicable but no later than 24 months following the initiation of decommissioning.

(i) The Commission may approve a request for an alternate schedule for completion of decommissioning of the site or separate building or outdoor area, and license termination if appropriate, if the Commission determines that the alternative is warranted by consideration of the following:

(1) Whether it is technically feasible to complete decommissioning within the allotted 24‐month period;

(2) Whether sufficient waste disposal capacity is available to allow completion of decommissioning within the allotted 24‐month period;

(3) Whether a significant volume reduction in wastes requiring disposal will be achieved by allowing short‐lived radionuclides to decay;

(4) Whether a significant reduction in radiation exposure to workers can be achieved by allowing short‐lived radionuclides to decay; and

(5) Other site‐specific factors which the Commission may consider appropriate on a case‐by‐case basis, such as the regulatory requirements of other government agencies, lawsuits, ground‐water treatment activities, monitored natural ground‐water restoration, actions that could result in more environmental harm than deferred cleanup, and other factors beyond the control of the licensee.

(j) As the final step in decommissioning, the licensee shall‐‐

(1) Certify the disposition of all licensed material, including accumulated wastes, by submitting a completed NRC Form 314 or equivalent information; and

(2) Conduct a radiation survey of the premises where the licensed activities were carried out and submit a report of the results of this survey, unless the licensee demonstrates in some other manner that the premises are suitable for release in accordance with the criteria for decommissioning in 10 CFR part 20, subpart E or, for uranium milling (uranium and thorium recovery) facilities, Criterion 6(6) of Appendix A to this part. The licensee shall, as appropriate‐‐

(i) Report levels of gamma radiation in units of millisieverts (microroentgen) per hour at one meter from surfaces, and report levels of radioactivity, including alpha and beta, in units of megabecquerels (disintegrations per minute or microcuries) per 100 square centimeters removable and fixed for surfaces, megabecquerels (microcuries) per milliliter for water, and becquerels (picocuries) per gram for solids such as soils or concrete; and

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(ii) Specify the survey instrument(s) used and certify that each instrument is properly calibrated and tested.

(k) Specific licenses, including expired licenses, will be terminated by written notice to the licensee when the Commission determines that:

(1) Source material has been properly disposed;

(2) Reasonable effort has been made to eliminate residual radioactive contamination, if present; and

(3)(i) A radiation survey has been performed which demonstrates that the premises are suitable for release in accordance with the criteria for decommissioning in 10 CFR part 20, subpart E; or for uranium milling (uranium and thorium recovery) facilities, Criterion 6(6) of Appendix A to this part;

(ii) Other information submitted by the licensee is sufficient to demonstrate that the premises are suitable for release in accordance with the criteria for decommissioning in 10 CFR part 20, subpart E.

(4) Records required by § 40.61(d) and (f) have been received.

(l) Specific licenses for uranium and thorium milling are exempt from paragraphs (d)(4), (g) and (h) of this section with respect to reclamation of tailings impoundments and/or waste disposal areas.

[59 FR 36035, July 15, 1994, as amended at 60 FR 38239, July 26, 1995; 61 FR 1114, Jan. 16, 1996; 61 FR 24674, May 16, 1996; 61 FR 29637, June 12, 1996; 62 FR 39090, July 21, 1997; 66 FR 64738, Dec. 14, 2001; 68 FR 75390, Dec. 31, 2003]

ssssss) § 40.43 Renewal of licenses.

(a) Application for renewal of a specific license must be filed on NRC Form 313 and in accordance with § 40.31.

(b) If any licensee granted the extension described in 10 CFR 40.42(a)(2) has a currently pending renewal application for the extended license, that application will be considered to be withdrawn by the licensee and any renewal fees paid by the licensee for that application will be refunded.

[59 FR 36037, July 15, 1994, as amended at 61 FR 1114, Jan. 16, 1996; 62 FR 52187, Oct. 6, 1997]

tttttt) § 40.44 Amendment of licenses at request of licensee.

Applications for amendment of a license shall be filed on NRC Form 313 in accordance with § 40.31 and shall specify the respects in which the licensee desires the license to be amended and the grounds for such amendment.

[49 FR 19627, May 9, 1984, as amended at 56 FR 40768, Aug. 16, 1991]

uuuuuu) § 40.45 Commission action on applications to renew or amend.

In considering an application by a licensee to renew or amend his license the Commission will apply the applicable criteria set forth in § 40.32.

[26 FR 284, Jan. 14, 1961, as amended at 43 FR 6924, Feb. 17, 1978]

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vvvvvv) § 40.46 Inalienability of licenses.

No license issued or granted pursuant to the regulations in this part shall be transferred, assigned or in any manner disposed of, either voluntarily or involuntarily, directly or indirectly, through transfer of control of any license to any person, unless the Commission shall after securing full information, find that the transfer is in accordance with the provisions of this act, and shall give its consent in writing.

wwwwww) Transfer of Source Material xxxxxx) § 40.51 Transfer of source or byproduct material.

(a) No licensee shall transfer source or byproduct material except as authorized pursuant to this section.

(b) Except as otherwise provided in his license and subject to the provisions of paragraphs (c) and (d) of this section, any licensee may transfer source or byproduct material:

(1) To the Department of Energy;

(2) To the agency in any Agreement State which regulates radioactive materials pursuant to an agreement with the Commission or the Atomic Energy Commission under section 274 of the Act;

(3) To any person exempt from the licensing requirements of the Act and regulations in this part, to the extent permitted under such exemption;

(4) To any person in an Agreement State subject to the jurisdiction of that State who has been exempted from the licensing requirements and regulations of that State, to the extent permitted under such exemptions;

(5) To any person authorized to receive such source or byproduct material under terms of a specific license or a general license or their equivalents issued by the Commission or an Agreement State;

(6) To any person abroad pursuant to an export license issued under part 110 of this chapter; or

(7) As otherwise authorized by the commission in writing.

(c) Before transferring source or byproduct material to a specific licensee of the Commission or an Agreement State or to a general licensee who is required to register with the Commission or with an Agreement State prior to receipt of the source or byproduct material, the licensee transferring the material shall verify that the transfereeʹs license authorizes receipt of the type, form, and quantity of source or byproduct material to be transferred.

(d) The following methods for the verification required by paragraph (c) of this section are acceptable:

(1) The transferor may have in his possession, and read, a current copy of the transfereeʹs specific license or registration certificate;

(2) The transferor may have in his possession a written certification by the transferee that he is authorized by license or registration certificate to receive the type, form, and quantity of source or byproduct material to be transferred, specifying the license or registration certification number, issuing agency and expiration date;

(3) For emergency shipments the transferor may accept oral certification by the transferee that he is authorized by license or registration certificate to receive the type, form, and quantity of source or byproduct material to be

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transferred, specifying the license or registration certificate number, issuing agency and expiration date: Provided, That the oral certification is confirmed in writing within 10 days;

(4) The transferor may obtain other sources of information compiled by a reporting service from official records of the Commission or the licensing agency of an Agreement State as to the identity of licensees and the scope and expiration dates of licenses and registrations; or

(5) When none of the methods of verification described in paragraphs (d)(1) to (4) of this section are readily available or when a transferor desires to verify that information received by one of such methods is correct or up‐to‐date, the transferor may obtain and record confirmation from the Commission or the licensing agency of an Agreement State that the transferee is licensed to receive the source or byproduct material.

[45 FR 65532, Oct. 3, 1980]

yyyyyy) Records, Reports, and Inspections zzzzzz) § 40.60 Reporting requirements.

(a) Immediate report. Each licensee shall notify the NRC as soon as possible but not later than 4 hours after the discovery of an event that prevents immediate protective actions necessary to avoid exposures to radiation or radioactive materials that could exceed regulatory limits or releases of licensed material that could exceed regulatory limits (events may include fires, explosions, toxic gas releases, etc.).

(b) Twenty‐four hour report. Each licensee shall notify the NRC within 24 hours after the discovery of any of the following events involving licensed material:

(1) An unplanned contamination event that:

(i) Requires access to the contaminated area, by workers or the public, to be restricted for more than 24 hours by imposing additional radiological controls or by prohibiting entry into the area;

(ii) Involves a quantity of material greater than five times the lowest annual limit on intake specified in appendix B of §§ 20.1001‐20.2401 of 10 CFR part 20 for the material; and

(iii) Has access to the area restricted for a reason other than to allow isotopes with a half‐life of less than 24 hours to decay prior to decontamination.

(2) An event in which equipment is disabled or fails to function as designed when:

(i) The equipment is required by regulation or license condition to prevent releases exceeding regulatory limits, to prevent exposures to radiation and radioactive materials exceeding regulatory limits, or to mitigate the consequences of an accident;

(ii) The equipment is required to be available and operable when it is disabled or fails to function; and

(iii) No redundant equipment is available and operable to perform the required safety function.

(3) An event that requires unplanned medical treatment at a medical facility of an individual with spreadable radioactive contamination on the individualʹs clothing or body.

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(4) An unplanned fire or explosion damaging any licensed material or any device, container, or equipment containing licensed material when:

(i) The quantity of material involved is greater than five times the lowest annual limit on intake specified in appendix B of §§ 20.1001‐20.2401 of 10 CFR part 20 for the material; and

(ii) The damage affects the integrity of the licensed material or its container.

(c) Preparation and submission of reports. Reports made by licensees in response to the requirements of this section must be made as follows:

(1) Licensees shall make reports required by paragraphs (a) and (b) of this section by telephone to the NRC Operations Center.1 To the extent that the information is available at the time of notification, the information provided in these reports must include:

(i) The callerʹs name and call back telephone number;

(ii) A description of the event, including date and time;

(iii) The exact location of the event;

(iv) The isotopes, quantities, and chemical and physical form of the licensed material involved; and

(v) Any personnel radiation exposure data available.

(2) Written report. Each licensee who makes a report required by paragraph (a) or (b) of this section shall submit a written follow‐up report within 30 days of the initial report. Written reports prepared pursuant to other regulations may be submitted to fulfill this requirement if the reports contain all of the necessary information and the appropriate distribution is made. These written reports must be sent to the NRCʹs Document Control Desk by an appropriate method listed in § 40.5, with a copy to the appropriate NRC regional office listed in appendix D to part 20 of this chapter. The reports must include the following:

(i) A description of the event, including the probable cause and the manufacturer and model number (if applicable) of any equipment that failed or malfunctioned;

(ii) The exact location of the event;

(iii) The isotopes, quantities, and chemical and physical form of the licensed material involved;

(iv) Date and time of the event;

(v) Corrective actions taken or planned and the results of any evaluations or assessments; and

(vi) The extent of exposure of individuals to radiation or to radioactive materials without identification of individuals by name.

(3) The provisions of § 40.60 do not apply to licensees subject to the notification requirements in § 50.72. They do apply to those part 50 licensees possessing material licensed under part 40 who are not subject to the notification requirements in § 50.72.

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[56 FR 40768, Aug. 16, 1991, as amended at 59 FR 14086, Mar. 25, 1994; 68 FR 58807, Oct. 10, 2003]

1 The commercial telephone number for the NRC Operations Center is (301) 816‐5100.

aaaaaaa) § 40.61 Records.

(a) Each person who receives source or byproduct material pursuant to a license issued pursuant to the regulations in this part shall keep records showing the receipt, transfer, and disposal of this source or byproduct material as follows:

(1) The licensee shall retain each record of receipt of source or byproduct material as long as the material is possessed and for three years following transfer or disposition of the source or byproduct material.

(2) The licensee who transferred the material shall retain each record of transfer or source or byproduct material until the Commission terminates each license that authorizes the activity that is subject to the recordkeeping requirement.

(3) The licensee shall retain each record of disposal of source or byproduct material until the Commission terminates each license that authorizes the activity that is subject to the recordkeeping requirement.

(4) If source or byproduct material is combined or mixed with other licensed material and subsequently treated in a manner that makes direct correlation of a receipt record with a transfer, export, or disposition record impossible, the licensee may use evaluative techniques (such as first‐in‐first‐out), to make the records that are required by this Part account for 100 percent of the material received.

(b) The licensee shall retain each record that is required by the regulations in this part or by license condition for the period specified by the appropriate regulation or license condition. If a retention period is not otherwise specified by regulation or license condition, each record must be maintained until the Commission terminates the license that authorizes the activity that is subject to the recordkeeping requirement.

(c)(1) Records which must be maintained pursuant to this part may be the original or reproduced copy or microform if the reproduced copy or microform is duly authenticated by authorized personnel and the microform is capable of producing a clear and legible copy after storage for the period specified by Commission regulations. The record may also be stored in electronic media with the capability for producing legible, accurate, and complete records during the required retention period. Records such as letters, drawings, specifications, must include all pertinent information such as stamps, initials, and signatures. The licensee shall maintain adequate safeguards against tampering with and loss of records.

(2) If there is a conflict between the Commissionʹs regulations in this part, license condition, or other written Commission approval or authorization pertaining to the retention period for the same type of record, the retention period specified in the regulations in this part for such records shall apply unless the Commission, pursuant to § 40.14 of this part, has granted a specific exemption from the record retention requirements specified in the regulations in this part.

(d) Prior to license termination, each licensee authorized to possess source material, in an unsealed form, shall forward the following records to the appropriate NRC Regional Office:

(1) Records of disposal of licensed material made under § 20.2002 (including burials authorized before January 28, 1981(1)), 20.2003, 20.2004, 20.2005; and

(2) Records required by § 20.2103(b)(4).

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(e) If licensed activities are transferred or assigned in accordance with § 40.41(b), each licensee authorized to possess source material, in an unsealed form, shall transfer the following records to the new licensee and the new licensee will be responsible for maintaining these records until the license is terminated:

(1) Records of disposal of licensed material made under § 20.2002 (including burials authorized before January 28, 19811), 20.2003, 20.2004, 20.2005; and

(2) Records required by § 20.2103(b)(4).

(f) Prior to license termination, each licensee shall forward the records required by § 40.36(f) to the appropriate NRC Regional Office.

[45 FR 65532, Oct. 3, 1980, as amended at 53 FR 19248, May 27, 1988; 61 FR 24674, May 16, 1996]

1 A previous § 20.304 permitted burial of small quantities of licensed materials in soil before January 28, 1981, without specific Commission authorization. See § 20.304 contained in the 10 CFR, parts 0 to 199, edition revised as of January 1, 1981.

bbbbbbb) § 40.62 Inspections.

(a) Each licensee shall afford to the Commission at all reasonable times opportunity to inspect source or byproduct material and the premises and facilities wherein source or byproduct material is used or stored.

(b) Each licensee shall make available to the Commission for inspection, upon reasonable notice, records kept by him pursuant to the regulations in this chapter.

[45 FR 65532, Oct. 3, 1980]

ccccccc) § 40.63 Tests.

Each licensee shall perform, or permit the Commission to perform, such tests as the Commission deems appropriate or necessary for the administration of the regulations in this part, including tests of:

(a) Source or byproduct material;

(b) Facilities wherein source or byproduct material is utilized or stored;

(c) Radiation detection and monitoring instruments; and

(d) Other equipment and devices used in connection with the utilization and storage of source or byproduct material.

[45 FR 65533, Oct. 3, 1980]

ddddddd) § 40.64 Reports.

(a) Except as specified in paragraphs (d) and (e) of this section, each specific licensee who transfers, receives, or adjusts the inventory, in any manner, of uranium or thorium source material with foreign obligations by 1 kilogram or more or who imports or exports 1 kilogram of uranium or thorium source material shall complete a Nuclear Material Transaction Report in computer‐readable format in accordance with instructions (NUREG/BR‐0006 and

194

NMMSS Report D‐24, ʺPersonal Computer Data Input for NRC Licenseesʺ). Copies of the instructions may be obtained either by writing the U.S. Nuclear Regulatory Commission, Division of Nuclear Security, Office of Nuclear Security and Incident Response, Washington, DC 20555‐0001, by e‐mail to [email protected], or by calling (301) 415‐6828. Each licensee who transfers the material shall submit a Nuclear Material Transaction Report in computer‐readable format in accordance with instructions no later than the close of business the next working day. Each licensee who receives the material shall submit a Nuclear Material Transaction Report in computer‐readable format in accordance with instructions within ten (10) days after the material is received. The Commissionʹs copy of the report must be submitted to the address specified in the instructions. These prescribed computer‐readable forms replace the DOE/NRC Form 741 which has been previously submitted in paper form.

(b) Except as specified in paragraphs (d) and (e) of this section, each licensee authorized to possess at any one time and location more than 1,000 kilograms of uranium or thorium, or any combination of uranium or thorium, shall submit to the Commission within 30 days after September 30 of each year or with the licenseeʹs material status reports on special nuclear material filed under part 72 or 74, a statement of its source material inventory with foreign obligations as defined in this part. This statement must be submitted to the address specified in the reporting instructions (NUREG/BR‐0007), and include the Reporting Identification Symbol (RIS) assigned by the Commission to the licensee. Copies of the reporting instructions may be obtained either by writing to the U.S. Nuclear Regulatory Commission, Division of Nuclear Security, Office of Nuclear Security and Incident Response, Washington, DC 20555‐0001, by e‐mail to [email protected], or by calling (301) 415‐6828.

(c)(1) Except as specified in paragraph (d) of this section, each licensee who is authorized to possess uranium or thorium pursuant to a specific license shall notify the NRC Headquarters Operations Center by telephone, at the numbers listed in appendix A of part 73 of this chapter, of any incident in which an attempt has been made or is believed to have been made to commit a theft or unlawful diversion of more than 6.8 kilograms (kg) [15 pounds] of such material at any one time or more than 68 kg [150 pounds] of such material in any one calendar year.

(2) The licensee shall notify the NRC as soon as possible, but within 4 hours, of discovery of any incident in which an attempt has been made or is believed to have been made to commit a theft or unlawful diversion of such material. A copy of the written followup notification should also be made to the Director, Division of Nuclear Security, Office of Nuclear Security and Incident Response, by an appropriate method listed in § 40.5.

(3) The initial notification shall be followed within a period of sixty (60) days by a written followup notification submitted in accordance with § 40.5. A copy of the written followup notification shall also be sent to: ATTN: Document Control Desk, Director, Division of Nuclear Security, Office of Nuclear Security and Incident Response, U.S. Nuclear Regulatory Commission, Washington, DC 20555‐0001.

(4) Subsequent to the submission of the written followup notification required by this paragraph, the licensee shall promptly update the written followup notification, in accordance with this paragraph, with any substantive additional information, which becomes available to the licensee, concerning an attempted or apparent theft or unlawful diversion of source material.

(d) The reports described in paragraphs (a), (b), and (c) of this section are not required for:

(1) Processed ores containing less than five (5) percent of uranium or thorium, or any combination of uranium or thorium, by dry weight;

(2) Thorium contained in magnesium‐thorium and tungsten‐thorium alloys, if the thorium content in the alloys does not exceed 4 percent by weight;

(3) Chemical catalysts containing uranium depleted in the U‐235 isotope to 0.4 percent or less, if the uranium content of the catalyst does not exceed 15 percent by weight; or

195

(4) Any source material contained in non‐nuclear end use devices or components, including but not limited to permanently installed shielding, teletherapy, radiography, X‐ray, accelerator devices, or munitions.

(e) Any licensee who is required to submit inventory change reports and material status reports pursuant to part 75 of this chapter (pertaining to implementation of the US/IAEA Safeguards Agreement) shall prepare and submit such reports only as provided in §§ 75.34 and 75.35 of this chapter (instead of as provided in paragraphs (a) and (b) of this section).

[35 FR 12195, July 30, 1970, as amended at 36 FR 10938, June 5, 1971; 38 FR 1272, Jan. 11, 1973; 38 FR 2330, Jan. 24, 1973; 40 FR 8787, Mar. 3, 1975; 41 FR 16446, Apr. 19, 1976; 45 FR 50710, July 31, 1980; 49 FR 24707, June 15, 1984; 51 FR 9766, Mar. 21, 1986; 52 FR 31611, Aug. 21, 1987; 59 FR 35620, July 13, 1994; 68 FR 10364, Mar. 5, 2003; 68 FR 58807, Oct. 10, 2003]

eeeeeee) § 40.65 Effluent monitoring reporting requirements.

(a) Each licensee authorized to possess and use source material in uranium milling, in production of uranium hexafluoride, or in a uranium enrichment facility shall:

(1) Within 60 days after January 1, 1976 and July 1, 1976, and within 60 days after January 1 and July 1 of each year thereafter, submit a report to the Director of the Office of Nuclear Material Safety and Safeguards, using an appropriate method listed in § 40.5, with a copy to the appropriate NRC Regional Office shown in appendix D to part 20 of this chapter; which report must specify the quantity of each of the principal radionuclides released to unrestricted areas in liquid and in gaseous effluents during the previous six months of operation, and such other information as the Commission may require to estimate maximum potential annual radiation doses to the public resulting from effluent releases. If quantities of radioactive materials released during the reporting period are significantly above the licenseeʹs design objectives previously reviewed as part of the licensing action, the report shall cover this specifically. On the basis of such reports and any additional information the Commission may obtain from the licensee or others, the Commission may from time to time require the licensee to take such action as the Commission deems appropriate.

(2) [Reserved]

(b) [Reserved]

[40 FR 53230, Nov. 17, 1975, as amended at 41 FR 21627, May 27, 1976; 42 FR 25721, May 19, 1977; 52 FR 31611, Aug. 21, 1987; 57 FR 18391, Apr. 30, 1992; 68 FR 58807, Oct. 10, 2003]

fffffff) § 40.66 Requirements for advance notice of export shipments of natural uranium.

(a) Each licensee authorized to export natural uranium, other than in the form of ore or ore residue, in amounts exceeding 500 kilograms, shall notify the Director, Division of Nuclear Security, Office of Nuclear Security and Incident Response, by an appropriate method listed in § 40.5.

The notification must be in writing and must be received at least 10 days before transport of the shipment commences at the shipping facility.

(b) The notification must include the following information:

(1) The name(s), address(es), and telephone number(s) of the shipper, receiver, and carrier(s);

196

(2) A physical description of the shipment;

(3) A listing of the mode(s) of shipment, transfer points, and routes to be used;

(4) The estimated date and time that shipment will commence and that each nation (other than the United States) along the route is scheduled to be entered; and

(5) A certification that arrangements have been made to notify the Division of Industrial and Medical Nuclear Safety when the shipment is received at the receiving facility.

(c) A licensee who needs to amend a notification may do so by telephoning the Division of Industrial and Medical Nuclear Safety at (301) 415‐7197.


ggggggg) § 40.67 Requirement for advance notice for importation of natural uranium from countries that are not party to the Convention on the Physical Protection of Nuclear Material.

(a) Each licensee authorized to import natural uranium, other than in the form of ore or ore residue, in amounts exceeding 500 kilograms, from countries not party to the Convention on the Physical Protection of Nuclear Material (see appendix F to part 73 of this chapter) shall notify the Director, Division of Nuclear Security, Office of Nuclear Security and Incident Response, using an appropriate method listed in § 40.5. The notification must be in writing and must be received at least 10 days before transport of the shipment commences at the shipping facility.

(b) The notification must include the following information:

(1) The name(s), address(es), and telephone number(s) of the shipper, receiver, and carrier(s);

(2) A physical description of the shipment;

(3) A listing of the mode(s) of shipment, transfer points, and routes to be used;

(4) The estimated date and time that shipment will commence and that each nation along the route is scheduled to be entered.

(c) The licensee shall notify the Division of Industrial and Medical Nuclear Safety by telephone at (301) 415‐7197 when the shipment is received at the receiving facility.

(d) A licensee who needs to amend a notification may do so by telephoning the Division of Industrial and Medical Nuclear Safety at (301) 415‐7197.


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hhhhhhh) Modification and Revocation of Licenses iiiiiii) § 40.71 Modification and revocation of licenses.

(a) The terms and conditions of each license shall be subject to amendment, revision, or modification by reason of amendments to the Act, or by reason of rules, regulations, or orders issued in accordance with the Act.

(b) Any license may be revoked, suspended, or modified, in whole or in part, for any material false statement in the application or any statement of fact required under section 182 of the Act, or because of conditions revealed by such application or statement of fact or any report, record, or inspection or other means which would warrant the Commission to refuse to grant a license on an original application, or for violation of, or failure to observe any of, the terms and conditions of the Act, or the license, or of any rule, regulation or order of the Commission.

(c) Except in cases of willfulness or those in which the public health, interest or safety requires otherwise, no license shall be modified, suspended, or revoked unless, prior to the institution of proceedings therefor, facts or conduct which may warrant such action shall have been called to the attention of the licensee in writing and the licensee shall have been accorded opportunity to demonstrate or achieve compliance with all lawful requirements.

[26 FR 284, Jan. 14, 1961, as amended at 35 FR 11460, July 17, 1970; 48 FR 32328, July 15, 1983]

jjjjjjj) Enforcement kkkkkkk) § 40.81 Violations.







(i) Sections 53, 57, 62, 63, 81, 82, 101, 103, 104, 107, or 109 of the Atomic Energy Act of 1954, as amended;


(iii) Any rule, regulation, or order issued pursuant to the sections specified in paragraph (b)(1)(i) of this section;


(2) For any violation for which a license may be revoked under section 186 of the Atomic Energy Act of 1954, as amended.

[57 FR 55074, Nov. 24, 1992]

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lllllll) § 40.82 Criminal penalties.

(a) Section 223 of the Atomic Energy Act of 1954, as amended, provides for criminal sanctions for willful violation of, attempted violation of, or conspiracy to violate, any regulation issued under sections 161b, 161i, or 161o of the Act. For purposes of section 223, all the regulations in part 40 are issued under one or more of sections 161b, 161i, or 161o, except for the sections listed in paragraph (b) of this section.

(b) The regulations in part 40 that are not issued under sections 161b, 161i, or 161o for the purposes of section 223 are as follows: §§ 40.1, 40.2, 40.2a, 40.4, 40.5, 40.6, 40.8, 40.11, 40.12, 40.13, 40.14, 40.20, 40.21, 40.31, 40.32, 40.34, 40.43, 40.44, 40.45, 40.71, 40.81, and 40.82.

[57 FR 55075, Nov. 24, 1992]

mmmmmmm) Appendix A to Part 40‐‐Criteria Relating to the Operation of Uranium Mills and the Disposition of Tailings or Wastes Produced by the Extraction or Concentration of Source Material From Ores Processed Primarily for Their Source Material Content

Introduction. Every applicant for a license to possess and use source material in conjunction with uranium or thorium milling, or byproduct material at sites formerly associated with such milling, is required by the provisions of § 40.31(h) to include in a license application proposed specifications relating to milling operations and the disposition of tailings or wastes resulting from such milling activities. This appendix establishes technical, financial, ownership, and long‐term site surveillance criteria relating to the siting, operation, decontamination, decommissioning, and reclamation of mills and tailings or waste systems and sites at which such mills and systems are located. As used in this appendix, the term ʺas low as is reasonably achievableʺ has the same meaning as in § 20.1003 of this chapter.

In many cases, flexibility is provided in the criteria to allow achieving an optimum tailings disposal program on a site‐specific basis. However, in such cases the objectives, technical alternatives and concerns which must be taken into account in developing a tailings program are identified. As provided by the provisions of § 40.31(h) applications for licenses must clearly demonstrate how the criteria have been addressed.

The specifications must be developed considering the expected full capacity of tailings or waste systems and the lifetime of mill operations. Where later expansions of systems or operations may be likely (for example, where large quantities of ore now marginally uneconomical may be stockpiled), the amenability of the disposal system to accommodate increased capacities without degradation in long‐term stability and other performance factors must be evaluated.

Licensees or applicants may propose alternatives to the specific requirements in this appendix. The alternative proposals may take into account local or regional conditions, including geology, topography, hydrology, and meterology. The Commission may find that the proposed alternatives meet the Commissionʹs requirements if the alternatives will achieve a level of stabilization and containment of the sites concerned, and a level of protection for public health, safety, and the environment from radiological and nonradiological hazards associated with the sites, which is equivalent to, to the extent practicable, or more stringent than the level which would be achieved by the requirements of this Appendix and the standards promulgated by the Environmental Protection Agency in 40 CFR Part 192, Subparts D and E.

All site specific licensing decisions based on the criteria in this Appendix or alternatives proposed by licensees or applicants will take into account the risk to the public health and safety and the environment with due consideration to the economic costs involved and any other factors the Commission determines to be appropriate. In implementing this Appendix, the Commission will consider ʺpracticableʺ and ʺreasonably achievableʺ as equivalent terms.

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Decisions involved these terms will take into account the state of technology, and the economics of improvements in relation to benefits to the public health and safety, and other societal and socioeconomic considerations, and in relation to the utilization of atomic energy in the public interest.

The following definitions apply to the specified terms as used in this appendix:

Aquifer means a geologic formation, group of formations, or part of a formation capable of yielding a significant amount of ground water to wells or springs. Any saturated zone created by uranium or thorium recovery operations would not be considered an aquifer unless the zone is or potentially is (1) hydraulically interconnected to a natural aquifer, (2) capable of discharge to surface water, or (3) reasonably accessible because of migration beyond the vertical projection of the boundary of the land transferred for long‐term government ownership and care in accordance with Criterion 11 of this appendix.

As expeditiously as practicable considering technological feasibility, for the purposes of Criterion 6A, means as quickly as possible considering: the physical characteristics of the tailings and the site; the limits of available technology; the need for consistency with mandatory requirements of other regulatory programs; and factors beyond the control of the licensee. The phrase permits consideration of the cost of compliance only to the extent specifically provided for by use of the term available technology.

Available technology means technologies and methods for emplacing a final radon barrier on uranium mill tailings piles or impoundments. This term shall not be construed to include extraordinary measures or techniques that would impose costs that are grossly excessive as measured by practice within the industry (or one that is reasonably analogous), (such as, by way of illustration only, unreasonable overtime, staffing, or transportation requirements, etc., considering normal practice in the industry; laser fusion of soils, etc.), provided there is reasonable progress toward emplacement of the final radon barrier. To determine grossly excessive costs, the relevant baseline against which cost shall be compared is the cost estimate for tailings impoundment closure contained in the licenseeʹs approved reclamation plan, but costs beyond these estimates shall not automatically be considered grossly excessive.

Closure means the activities following operations to decontaminate and decommission the buildings and site used to produce byproduct materials and reclaim the tailings and/or waste disposal area.

Closure plan means the Commission approved plan to accomplish closure.

Compliance period begins when the Commission sets secondary ground‐water protection standards and ends when the owner or operatorʹs license is terminated and the site is transferred to the State or Federal agency for long‐term care.

Dike means an embankment or ridge of either natural or man‐made materials used to prevent the movement of liquids, sludges, solids or other materials.

Disposal area means the area containing byproduct materials to which the requirements of Criterion 6 apply.

Existing portion means that land surface area of an existing surface impoundment on which significant quantities of uranium or thorium byproduct materials had been placed prior to September 30, 1983.

Factors beyond the control of the licensee means factors proximately causing delay in meeting the schedule in the applicable reclamation plan for the timely emplacement of the final radon barrier notwithstanding the good faith efforts of the licensee to complete the barrier in compliance with paragraph (1) of Criterion 6A. These factors may include, but are not limited to:

(1) Physical conditions at the site;

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(2) Inclement weather or climatic conditions;

(3) An act of God;

(4) An act of war;

(5) A judicial or administrative order or decision, or change to the statutory, regulatory, or other legal requirements applicable to the licenseeʹs facility that would preclude or delay the performance of activities required for compliance;

(6) Labor disturbances;

(7) Any modifications, cessation or delay ordered by State, Federal, or local agencies;

(8) Delays beyond the time reasonably required in obtaining necessary government permits, licenses, approvals, or consent for activities described in the reclamation plan proposed by the licensee that result from agency failure to take final action after the licensee has made a good faith, timely effort to submit legally sufficient applications, responses to requests (including relevant data requested by the agencies), or other information, including approval of the reclamation plan; and

(9) An act or omission of any third party over whom the licensee has no control.

Final radon barrier means the earthen cover (or approved alternative cover) over tailings or waste constructed to comply with Criterion 6 of this appendix (excluding erosion protection features).

Ground water means water below the land surface in a zone of saturation. For purposes of this appendix, ground water is the water contained within an aquifer as defined above.

Leachate means any liquid, including any suspended or dissolved components in the liquid, that has percolated through or drained from the byproduct material.

Licensed site means the area contained within the boundary of a location under the control of persons generating or storing byproduct materials under a Commission license.

Liner means a continuous layer of natural or man‐made materials, beneath or on the sides of a surface impoundment which restricts the downward or lateral escape of byproduct material, hazardous constituents, or leachate.

Milestone means an action or event that is required to occur by an enforceable date.

Operation means that a uranium or thorium mill tailings pile or impoundment is being used for the continued placement of byproduct material or is in standby status for such placement. A pile or impoundment is in operation from the day that byproduct material is first placed in the pile or impoundment until the day final closure begins.

Point of compliance is the site specific location in the uppermost aquifer where the ground‐water protection standard must be met.

Reclamation plan, for the purposes of Criterion 6A, means the plan detailing activities to accomplish reclamation of the tailings or waste disposal area in accordance with the technical criteria of this appendix. The reclamation plan must include a schedule for reclamation milestones that are key to the completion of the final radon barrier including as appropriate, but not limited to, wind blown tailings retrieval and placement on the pile, interim stabilization

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(including dewatering or the removal of freestanding liquids and recontouring), and final radon barrier construction. (Reclamation of tailings must also be addressed in the closure plan; the detailed reclamation plan may be incorporated into the closure plan.)

Surface impoundment means a natural topographic depression, man‐made excavation, or diked area, which is designed to hold an accumulation of liquid wastes or wastes containing free liquids, and which is not an injection well.

Uppermost aquifer means the geologic formation nearest the natural ground surface that is an aquifer, as well as lower aquifers that are hydraulically interconnected with this aquifer within the facilityʹs property boundary.

I. Technical Criteria

Criterion 1‐‐The general goal or broad objective in siting and design decisions is permanent isolation of tailings and associated contaminants by minimizing disturbance and dispersion by natural forces, and to do so without ongoing maintenance. For practical reasons, specific siting decisions and design standards must involve finite times (e.g., the longevity design standard in Criterion 6). The following site features which will contribute to such a goal or objective must be considered in selecting among alternative tailings disposal sites or judging the adequacy of existing tailings sites:

Remoteness from populated areas;

Hydrologic and other natural conditions as they contribute to continued immobilization and isolation of contaminants from ground‐water sources; and

Potential for minimizing erosion, disturbance, and dispersion by natural forces over the long term.

The site selection process must be an optimization to the maximum extent reasonably achievable in terms of these features.

In the selection of disposal sites, primary emphasis must be given to isolation of tailings or wastes, a matter having long‐term impacts, as opposed to consideration only of short‐term convenience or benefits, such as minimization of transportation or land acquisition costs. While isolation of tailings will be a function of both site and engineering design, overriding consideration must be given to siting features given the long‐term nature of the tailings hazards.

Tailings should be disposed of in a manner that no active maintenance is required to preserve conditions of the site.

Criterion 2‐‐To avoid proliferation of small waste disposal sites and thereby reduce perpetual surveillance obligations, byproduct material from in situ extraction operations, such as residues from solution evaporation or contaminated control processes, and wastes from small remote above ground extraction operations must be disposed of at existing large mill tailings disposal sites; unless, considering the nature of the wastes, such as their volume and specific activity, and the costs and environmental impacts of transporting the wastes to a large disposal site, such offsite disposal is demonstrated to be impracticable or the advantages of onsite burial clearly outweigh the benefits of reducing the perpetual surveillance obligations.

Criterion 3‐‐The ʺprime optionʺ for disposal of tailings is placement below grade, either in mines or specially excavated pits (that is, where the need for any specially constructed retention structure is eliminated). The evaluation of alternative sites and disposal methods performed by mill operators in support of their proposed tailings disposal program (provided in applicantsʹ environmental reports) must reflect serious consideration of this disposal mode. In some instances, below grade disposal may not be the most environmentally sound approach, such as might be the case if a ground‐water formation is relatively close to the surface or not very well isolated by overlying soils and

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rock. Also, geologic and topographic conditions might make full below grade burial impracticable: For example, bedrock may be sufficiently near the surface that blasting would be required to excavate a disposal pit at excessive cost, and more suitable alternative sites are not available. Where full below grade burial is not practicable, the size of retention structures, and size and steepness of slopes associated exposed embankments must be minimized by excavation to the maximum extent reasonably achievable or appropriate given the geologic and hydrologic conditions at a site. In these cases, it must be demonstrated that an above grade disposal program will provide reasonably equivalent isolation of the tailings from natural erosional forces.

Criterion 4‐‐The following site and design criteria must be adhered to whether tailings or wastes are disposed of above or below grade.

(a) Upstream rainfall catchment areas must be minimized to decrease erosion potential and the size of the floods which could erode or wash out sections of the tailings disposal area.

(b) Topographic features should provide good wind protection.

(c) Embankment and cover slopes must be relatively flat after final stabilization to minimize erosion potential and to provide conservative factors of safety assuring long‐term stability. The broad objective should be to contour final slopes to grades which are as close as possible to those which would be provided if tailings were disposed of below grade; this could, for example, lead to slopes of about 10 horizontal to 1 vertical (10h:1v) or less steep. In general, slopes should not be steeper than about 5h:1v. Where steeper slopes are proposed, reasons why a slope less steep than 5h:1v would be impracticable should be provided, and compensating factors and conditions which make such slopes acceptable should be identified.

(d) A full self‐sustaining vegetative cover must be established or rock cover employed to reduce wind and water erosion to negligible levels.

Where a full vegetative cover is not likely to be self‐sustaining due to climatic or other conditions, such as in semi‐arid and arid regions, rock cover must be employed on slopes of the impoundment system. The NRC will consider relaxing this requirement for extremely gentle slopes such as those which may exist on the top of the pile.

The following factors must be considered in establishing the final rock cover design to avoid displacement of rock particles by human and animal traffic or by natural process, and to preclude undercutting and piping:

Shape, size, composition, and gradation of rock particles (excepting bedding material average particles size must be at least cobble size or greater);

Rock cover thickness and zoning of particles by size; and

Steepness of underlying slopes.

Individual rock fragments must be dense, sound, and resistant to abrasion, and must be free from cracks, seams, and other defects that would tend to unduly increase their destruction by water and frost actions. Weak, friable, or laminated aggregate may not be used.

Rock covering of slopes may be unnecessary where top covers are very thick ( or less); bulk cover materials have inherently favorable erosion resistance characteristics; and, there is negligible drainage catchment area upstream of the pile and good wind protection as described in points (a) and (b) of this Criterion.

Furthermore, all impoundment surfaces must be contoured to avoid areas of concentrated surface runoff or abrupt or sharp changes in slope gradient. In addition to rock cover on slopes, areas toward which surface runoff might be

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directed must be well protected with substantial rock cover (rip rap). In addition to providing for stability of the impoundment system itself, overall stability, erosion potential, and geomorphology of surrounding terrain must be evaluated to assure that there are not ongoing or potential processes, such as gully erosion, which would lead to impoundment instability.

(e) The impoundment may not be located near a capable fault that could cause a maximum credible earthquake larger than that which the impoundment could reasonably be expected to withstand. As used in this criterion, the term ʺcapable faultʺ has the same meaning as defined in section III(g) of Appendix A of 10 CFR Part 100. The term ʺmaximum credible earthquakeʺ means that earthquake which would cause the maximum vibratory ground motion based upon an evaluation of earthquake potential considering the regional and local geology and seismology and specific characteristics of local subsurface material.

(f) The impoundment, where feasible, should be designed to incorporate features which will promote deposition. For example, design features which promote deposition of sediment suspended in any runoff which flows into the impoundment area might be utilized; the object of such a design feature would be to enhance the thickness of cover over time.

Criterion 5‐‐Criteria 5A‐5D and new Criterion 13 incorporate the basic ground‐water protection standards imposed by the Environmental Protection Agency in 40 CFR Part 192, Subparts D and E (48 FR 45926; October 7, 1983) which apply during operations and prior to the end of closure. Ground‐water monitoring to comply with these standards is required by Criterion 7A.

5A(1)‐‐The primary ground‐water protection standard is a design standard for surface impoundments used to manage uranium and thorium byproduct material. Unless exempted under paragraph 5A(3) of this criterion, surface impoundments (except for an existing portion) must have a liner that is designed, constructed, and installed to prevent any migration of wastes out of the impoundment to the adjacent subsurface soil, ground water, or surface water at any time during the active life (including the closure period) of the impoundment. The liner may be constructed of materials that may allow wastes to migrate into the liner (but not into the adjacent subsurface soil, ground water, or surface water) during the active life of the facility, provided that impoundment closure includes removal or decontamination of all waste residues, contaminated containment system components (liners, etc.), contaminated subsoils, and structures and equipment contaminated with waste and leachate. For impoundments that will be closed with the liner material left in place, the liner must be constructed of materials that can prevent wastes from migrating into the liner during the active life of the facility.

5A(2)‐‐The liner required by paragraph 5A(1) above must be‐‐

(a) Constructed of materials that have appropriate chemical properties and sufficient strength and thickness to prevent failure due to pressure gradients (including static head and external hydrogeologic forces), physical contact with the waste or leachate to which they are exposed, climatic conditions, the stress of installation, and the stress of daily operation;

(b) Placed upon a foundation or base capable of providing support to the liner and resistance to pressure gradients above and below the liner to prevent failure of the liner due to settlement, compression, or uplift; and

(c) Installed to cover all surrounding earth likely to be in contact with the wastes or leachate.

5A(3)‐‐The applicant or licensee will be exempted from the requirements of paragraph 5A(1) of this criterion if the Commission finds, based on a demonstration by the applicant or licensee, that alternate design and operating practices, including the closure plan, together with site characteristics will prevent the migration of any hazardous constituents into ground water or surface water at any future time. In deciding whether to grant an exemption, the Commission will consider‐‐

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(a) The nature and quantity of the wastes;

(b) The proposed alternate design and operation;

(c) The hydrogeologic setting of the facility, including the attenuative capacity and thickness of the liners and soils present between the impoundment and ground water or surface water; and

(d) All other factors which would influence the quality and mobility of the leachate produced and the potential for it to migrate to ground water or surface water.

5A(4)‐‐A surface impoundment must be designed, constructed, maintained, and operated to prevent overtopping resulting from normal or abnormal operations, overfilling, wind and wave actions, rainfall, or run‐on; from malfunctions of level controllers, alarms, and other equipment; and from human error.

5A(5)‐‐When dikes are used to form the surface impoundment, the dikes must be designed, constructed, and maintained with sufficient structural integrity to prevent massive failure of the dikes. In ensuring structural integrity, it must not be presumed that the liner system will function without leakage during the active life of the impoundment.

5B(1)‐‐Uranium and thorium byproduct materials must be managed to conform to the following secondary ground‐water protection standard: Hazardous constituents entering the ground water from a licensed site must not exceed the specified concentration limits in the uppermost aquifer beyond the point of compliance during the compliance period. Hazardous constituents are those constituents identified by the Commission pursuant to paragraph 5B(2) of this criterion. Specified concentration limits are those limits established by the Commission as indicated in paragraph 5B(5) of this criterion. The Commission will also establish the point of compliance and compliance period on a site specific basis through license conditions and orders. The objective in selecting the point of compliance is to provide the earliest practicable warning that the impoundment is releasing hazardous constituents to the ground water. The point of compliance must be selected to provide prompt indication of ground‐water contamination on the hydraulically downgradient edge of the disposal area. The Commission shall identify hazardous constituents, establish concentration limits, set the compliance period, and may adjust the point of compliance if needed to accord with developed data and site information as to the flow of ground water or contaminants, when the detection monitoring established under Criterion 7A indicates leakage of hazardous constituents from the disposal area.

5B(2)‐‐A constituent becomes a hazardous constituent subject to paragraph 5B(5) only when the constituent meets all three of the following tests:

(a) The constituent is reasonably expected to be in or derived from the byproduct material in the disposal area;

(b) The constituent has been detected in the ground water in the uppermost aquifer; and

(c) The constituent is listed in Criterion 13 of this appendix.

5B(3)‐‐Even when constituents meet all three tests in paragraph 5B(2) of this criterion, the Commission may exclude a detected constituent from the set of hazardous constituents on a site specific basis if it finds that the constituent is not capable of posing a substantial present or potential hazard to human health or the environment. In deciding whether to exclude constituents, the Commission will consider the following:

(a) Potential adverse effects on ground‐water quality, considering‐‐

(i) The physical and chemical characteristics of the waste in the licensed site, including its potential for migration;

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(ii) The hydrogeological characteristics of the facility and surrounding land;

(iii) The quantity of ground water and the direction of ground‐water flow;

(iv) The proximity and withdrawal rates of ground‐water users;

(v) The current and future uses of ground water in the area;

(vi) The existing quality of ground water, including other sources of contamination and their cumulative impact on the ground‐water quality;

(vii) The potential for health risks caused by human exposure to waste constituents;

(viii) The potential damage to wildlife, crops, vegetation, and physical structures caused by exposure to waste constituents;

(ix) The persistence and permanence of the potential adverse effects.

(b) Potential adverse effects on hydraulically‐connected surface water quality, considering‐‐

(i) The volume and physical and chemical characteristics of the waste in the licensed site;


(iii) The quantity and quality of ground water, and the direction of ground‐water flow;

(iv) The patterns of rainfall in the region;

(v) The proximity of the licensed site to surface waters;

(vi) The current and future uses of surface waters in the area and any water quality standards established for those surface waters;

(vii) The existing quality of surface water, including other sources of contamination and the cumulative impact on surface‐water quality;

(viii) The potential for health risks caused by human exposure to waste constituents;

(ix) The potential damage to wildlife, crops, vegetation, and physical structures caused by exposure to waste constituents; and

(x) The persistence and permanence of the potential adverse effects.

5B(4)‐‐In making any determinations under paragraphs 5B(3) and 5B(6) of this criterion about the use of ground water in the area around the facility, the Commission will consider any identification of underground sources of drinking water and exempted aquifers made by the Environmental Protection Agency.

5B(5)‐‐At the point of compliance, the concentration of a hazardous constituent must not exceed‐‐

(a) The Commission approved background concentration of that constituent in the ground water;

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(b) The respective value given in the table in paragraph 5C if the constituent is listed in the table and if the background level of the constituent is below the value listed; or

(c) An alternate concentration limit established by the Commission.

5B(6)‐‐Conceptually, background concentrations pose no incremental hazards and the drinking water limits in paragraph 5C state acceptable hazards but these two options may not be practically achievable at a specific site. Alternate concentration limits that present no significant hazard may be proposed by licensees for Commission consideration. Licensees must provide the basis for any proposed limits including consideration of practicable corrective actions, that limits are as low as reasonably achievable, and information on the factors the Commission must consider. The Commission will establish a site specific alternate concentration limit for a hazardous constituent as provided in paragraph 5B(5) of this criterion if it finds that the proposed limit is as low as reasonably achievable, after considering practicable corrective actions, and that the constituent will not pose a substantial present or potential hazard to human health or the environment as long as the alternate concentration limit is not exceeded. In making the present and potential hazard finding, the Commission will consider the following factors:

(a) Potential adverse effects on ground‐water quality, considering‐‐

(i) The physical and chemical characteristics of the waste in the licensed site including its potential for migration;


(iii) The quantity of ground water and the direction of ground‐water flow;

(iv) The proximity and withdrawal rates of ground‐water users;

(v) The current and future uses of ground water in the area;

(vi) The existing quality of ground water, including other sources of contamination and their cumulative impact on the ground‐water quality;

(vii) The potential for health risks caused by human exposure to waste constituents;

(viii) The potential damage to wildlife, crops, vegetation, and physical structures caused by exposure to waste constituents;

(ix) The persistence and permanence of the potential adverse effects.

(b) Potential adverse effects on hydraulically‐connected surface water quality, considering‐‐

(i) The volume and physical and chemical characteristics of the waste in the licensed site;


(iii) The quantity and quality of ground water, and the direction of ground‐water flow;

(iv) The patterns of rainfall in the region;

(v) The proximity of the licensed site to surface waters; (vi) The current and future uses of surface waters in the area and any water quality standards established for those surface waters;

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(vii) The existing quality of surface water including other sources of contamination and the cumulative impact on surface water quality;

(viii) The potential for health risks caused by human exposure to waste constituents;

(ix) The potential damage to wildlife, crops, vegetation, and physical structures caused by exposure to waste constituents; and

(x) The persistence and permanence of the potential adverse effects.

5C‐Maximum Values for Ground‐Water Protection

Constituent or property Maximum concentration

Milligrams per liter:

Arsenic 0.05

Barium 1.0

Cadmium 0.01

Chromium 0.05

Lead 0.05

Mercury 0.002

Selenium 0.01

Silver 0.05

Endrin (1,2,3,4,10,10‐hexachloro‐1,7 ‐expoxy‐1,4,4a,5,6,7,8,9a‐octahydro‐1, 4‐endo, endo‐5, 8‐dimethano napthalene)

0.0002

Lindane (1,2,3,4,5,6‐hexachlorocyclohexane, gamma isomer) 0.004

Methoxychlor (1,1,1‐Trichloro‐2,2‐bis (p‐methoxyphenylethane) 0.1

Toxaphene (C10 H10 C16, Technical chlorinated camphene, 67‐69 percent chlorine) 0.005

2, 4‐D(2,4‐Dichlorophenoxyacetic acid) 0.1

2, 4,5‐TP Silvex (2,4,5‐Trichlorophenoxypropionic acid)

Picocuries per liter:

Combined radium‐226 and radium‐228 5

Gross alpha‐particle activity (excluding radon and uranium when producing uranium byproduct material or radon and thorium when producing thorium byproduct material)

15

5D‐If the ground‐water protection standards established under paragraph 5B(1) of this criterion are exceeded at a licensed site, a corrective action program must be put into operation as soon as is practicable, and in no event later than eighteen (18) months after the Commission finds that the standards have been exceeded. The licensee shall submit the proposed corrective action program and supporting rationale for Commission approval prior to putting the program into operation, unless otherwise agreed to by the Commission. The objective of the program is to return hazardous constituent concentration levels in ground water to the concentration levels set as standards. The licenseeʹs proposed program must address removing hazardous constituents that have entered the ground water at the point of

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compliance or treating them in place. The program must also address removing or treating any hazardous constituents that exceed concentration limits in ground water between the point of compliance and the downgradient facility property boundary. The licensee shall continue corrective action measures to the extent necessary to achieve and maintain compliance with the groundwater standard. The Commission will determine when the licensee may terminate corrective action measures based on data from the ground‐water monitoring program and other information that provide reasonable assurance that the ground‐water protection standard will not be exceeded.

5E‐In developing and conducting ground‐water protection programs, applicants and licensees shall also consider the following:

(1) Installation of bottom liners(Where synthetic liners are used, a leakage detection system must be installed immediately below the liner to ensure major failures are detected if they occur. This is in addition to the ground‐water monitoring program conducted as provided in Criterion 7. Where clay liners are proposed or relatively thin, in‐situ clay soils are to be relied upon for seepage control, tests must be conducted with representative tailings solutions and clay materials to confirm that no significant deterioration of permeability or stability properties will occur with continuous exposure of clay to tailings solutions. Tests must be run for a sufficient period of time to reveal any effects if they are going to occur (in some cases deterioration has been observed to occur rather rapidly after about nine months of exposure)).

(2) Mill process designs which provide the maximum practicable recycle of solutions and conservation of water to reduce the net input of liquid to the tailings impoundment.

(3) Dewatering of tailings by process devices and/or in‐situ drainage systems (At new sites, tailings must be dewatered by a drainage system installed at the bottom of the impoundment to lower the phreatic surface and reduce the driving head of seepage, unless tests show tailings are not amenable to such a system. Where in‐situ dewatering is to be conducted, the impoundment bottom must be graded to assure that the drains are at a low point. The drains must be protected by suitable filter materials to assure that drains remain free running. The drainage system must also be adequately sized to assure good drainage).

(4) Neutralization to promote immobilization of hazardous constituents.

5F‐‐Where ground‐water impacts are occurring at an existing site due to seepage, action must be taken to alleviate conditions that lead to excessive seepage impacts and restore ground‐water quality. The specific seepage control and ground‐water protection method, or combination of methods, to be used must be worked out on a site‐specific basis. Technical specifications must be prepared to control installation of seepage control systems. A quality assurance, testing, and inspection program, which includes supervision by a qualified engineer or scientist, must be established to assure the specifications are met.

5G‐‐In support of a tailings disposal system proposal, the applicant/operator shall supply information concerning the following:

(1) The chemical and radioactive characteristics of the waste solutions.

(2) The characteristics of the underlying soil and geologic formations particularly as they will control transport of contaminants and solutions. This includes detailed information concerning extent, thickness, uniformity, shape, and orientation of underlying strata. Hydraulic gradients and conductivities of the various formations must be determined. This information must be gathered from borings and field survey methods taken within the proposed impoundment area and in surrounding areas where contaminants might migrate to ground water. The information gathered on boreholes must include both geologic and geophysical logs in sufficient number and degree of sophistication to allow determining significant discontinuities, fractures, and channeled deposits of high hydraulic conductivity. If field survey methods are used, they should be in addition to and calibrated with borehole logging.

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Hydrologic parameters such as permeability may not be determined on the basis of laboratory analysis of samples alone; a sufficient amount of field testing (e.g., pump tests) must be conducted to assure actual field properties are adequately understood. Testing must be conducted to allow estimating chemi‐sorption attenuation properties of underlying soil and rock.

(3) Location, extent, quality, capacity and current uses of any ground water at and near the site.

5H‐‐Steps must be taken during stockpiling of ore to minimize penetration of radionuclides into underlying soils; suitable methods include lining and/or compaction of ore storage areas.

Criterion 6‐‐(1) In disposing of waste byproduct material, licensees shall place an earthen cover (or approved alternative) over tailings or wastes at the end of milling operations and shall close the waste disposal area in accordance with a design1 which provides reasonable assurance of control of radiological hazards to (i) be effective for 1,000 years, to the extent reasonably achievable, and, in any case, for at least 200 years, and (ii) limit releases of radon‐222 from uranium byproduct materials, and radon‐220 from thorium byproduct materials, to the atmosphere so as not to exceed an average2 release rate of 20 picocuries per square meter per second (pCi/m2s) to the extent practicable throughout the effective design life determined pursuant to (1)(i) of this Criterion. In computing required tailings cover thicknesses, moisture in soils in excess of amounts found normally in similar soils in similar circumstances may not be considered. Direct gamma exposure from the tailings or wastes should be reduced to background levels. The effects of any thin synthetic layer may not be taken into account in determining the calculated radon exhalation level. If non‐soil materials are proposed as cover materials, it must be demonstrated that these materials will not crack or degrade by differential settlement, weathering, or other mechanism, over long‐term intervals.

(2) As soon as reasonably achievable after emplacement of the final cover to limit releases of radon‐222 from uranium byproduct material and prior to placement of erosion protection barriers or other features necessary for long‐term control of the tailings, the licensee shall verify through appropriate testing and analysis that the design and construction of the final radon barrier is effective in limiting releases of radon‐222 to a level not exceeding 20 pCi/m2s averaged over the entire pile or impoundment using the procedures described in 40 CFR part 61, appendix B, Method 115, or another method of verification approved by the Commission as being at least as effective in demonstrating the effectiveness of the final radon barrier.

(3) When phased emplacement of the final radon barrier is included in the applicable reclamation plan, the verification of radon‐222 release rates required in paragraph (2) of this criterion must be conducted for each portion of the pile or impoundment as the final radon barrier for that portion is emplaced.

(4) Within ninety days of the completion of all testing and analysis relevant to the required verification in paragraphs (2) and (3) of this criterion, the uranium mill licensee shall report to the Commission the results detailing the actions taken to verify that levels of release of radon‐222 do not exceed 20 pCi/m2s when averaged over the entire pile or impoundment. The licensee shall maintain records until termination of the license documenting the source of input parameters including the results of all measurements on which they are based, the calculations and/or analytical methods used to derive values for input parameters, and the procedure used to determine compliance. These records shall be kept in a form suitable for transfer to the custodial agency at the time of transfer of the site to DOE or a State for long‐term care if requested.

(5) Near surface cover materials (i.e., within the top three meters) may not include waste or rock that contains elevated levels of radium; soils used for near surface cover must be essentially the same, as far as radioactivity is concerned, as that of surrounding surface soils. This is to ensure that surface radon exhalation is not significantly above background because of the cover material itself.

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(6) The design requirements in this criterion for longevity and control of radon releases apply to any portion of a licensed and/or disposal site unless such portion contains a concentration of radium in land, averaged over areas of 100 square meters, which, as a result of byproduct material, does not exceed the background level by more than: (i) 5 picocuries per gram (pCi/g) of radium‐226, or, in the case of thorium byproduct material, radium‐228, averaged over the first 15 centimeters (cm) below the surface, and (ii) 15 pCi/g of radium‐226, or, in the case of thorium byproduct material, radium‐228, averaged over 15‐cm thick layers more than 15 cm below the surface.

Byproduct material containing concentrations of radionuclides other than radium in soil, and surface activity on remaining structures, must not result in a total effective dose equivalent (TEDE) exceeding the dose from cleanup of radium contaminated soil to the above standard (benchmark dose), and must be at levels which are as low as is reasonably achievable. If more than one residual radionuclide is present in the same 100‐square‐meter area, the sum of the ratios for each radionuclide of concentration present to the concentration limit will not exceed ʺ1ʺ (unity). A calculation of the potential peak annual TEDE within 1000 years to the average member of the critical group that would result from applying the radium standard (not including radon) on the site must be submitted for approval. The use of decommissioning plans with benchmark doses which exceed 100 mrem/yr, before application of ALARA, requires the approval of the Commission after consideration of the recommendation of the NRC staff. This requirement for dose criteria does not apply to sites that have decommissioning plans for soil and structures approved before June 11, 1999.

(7) The licensee shall also address the nonradiological hazards associated with the wastes in planning and implementing closure. The licensee shall ensure that disposal areas are closed in a manner that minimizes the need for further maintenance. To the extent necessary to prevent threats to human health and the environment, the licensee shall control, minimize, or eliminate post‐closure escape of nonradiological hazardous constituents, leachate, contaminated rainwater, or waste decomposition products to the ground or surface waters or to the atmosphere.

Criterion 6A‐‐(1) For impoundments containing uranium byproduct materials, the final radon barrier must be completed as expeditiously as practicable considering technological feasibility after the pile or impoundment ceases operation in accordance with a written, Commission‐approved reclamation plan. (The term as expeditiously as practicable considering technological feasibility as specifically defined in the Introduction of this appendix includes factors beyond the control of the licensee.) Deadlines for completion of the final radon barrier and, if applicable, the following interim milestones must be established as a condition of the individual license: windblown tailings retrieval and placement on the pile and interim stabilization (including dewatering or the removal of freestanding liquids and recontouring). The placement of erosion protection barriers or other features necessary for long‐term control of the tailings must also be completed in a timely manner in accordance with a written, Commission‐approved reclamation plan.

(2) The Commission may approve a licenseeʹs request to extend the time for performance of milestones related to emplacement of the final radon barrier if, after providing an opportunity for public participation, the Commission finds that the licensee has adequately demonstrated in the manner required in paragraph (2) of Criterion 6 that releases of radon‐222 do not exceed an average of 20 pCi/m2s. If the delay is approved on the basis that the radon releases do not exceed 20 pCi/m2s, a verification of radon levels, as required by paragraph (2) of Criterion 6, must be made annually during the period of delay. In addition, once the Commission has established the date in the reclamation plan for the milestone for completion of the final radon barrier, the Commission may extend that date based on cost if, after providing an opportunity for public participation, the Commission finds that the licensee is making good faith efforts to emplace the final radon barrier, the delay is consistent with the definition of available technology, and the radon releases caused by the delay will not result in a significant incremental risk to the public health.

(3) The Commission may authorize by license amendment, upon licensee request, a portion of the impoundment to accept uranium byproduct material or such materials that are similar in physical, chemical, and radiological characteristics to the uranium mill tailings and associated wastes already in the pile or impoundment, from other sources, during the closure process. No such authorization will be made if it results in a delay or impediment to

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emplacement of the final radon barrier over the remainder of the impoundment in a manner that will achieve levels of radon‐222 releases not exceeding 20 pCi/m2s averaged over the entire impoundment. The verification required in paragraph (2) of Criterion 6 may be completed with a portion of the impoundment being used for further disposal if the Commission makes a final finding that the impoundment will continue to achieve a level of radon‐222 releases not exceeding 20 pCi/m2s averaged over the entire impoundment. In this case, after the final radon barrier is complete except for the continuing disposal area, (a) only byproduct material will be authorized for disposal, (b) the disposal will be limited to the specified existing disposal area, and (c) this authorization will only be made after providing opportunity for public participation. Reclamation of the disposal area, as appropriate, must be completed in a timely manner after disposal operations cease in accordance with paragraph (1) of Criterion 6; however, these actions are not required to be complete as part of meeting the deadline for final radon barrier construction.

Criterion 7‐‐At least one full year prior to any major site construction, a preoperational monitoring program must be conducted to provide complete baseline data on a milling site and its environs. Throughout the construction and operating phases of the mill, an operational monitoring program must be conducted to measure or evaluate compliance with applicable standards and regulations; to evaluate performance of control systems and procedures; to evaluate environmental impacts of operation; and to detect potential long‐term effects.

7A‐‐The licensee shall establish a detection monitoring program needed for the Commission to set the site‐specific ground‐water protection standards in paragraph 5B(1) of this appendix. For all monitoring under this paragraph the licensee or applicant will propose for Commission approval as license conditions which constituents are to be monitored on a site specific basis. A detection monitoring program has two purposes.The initial purpose of the program is to detect leakage of hazardous constituents from the disposal area so that the need to set ground‐water protection standards is monitored. If leakage is detected, the second purpose of the program is to generate data and information needed for the Commission to establish the standards under Criterion 5B. The data and information must provide a sufficient basis to identify those hazardous constituents which require concentration limit standards and to enable the Commission to set the limits for those constituents and the compliance period. They may also need to provide the basis for adjustments to the point of compliance. For licenses in effect September 30, 1983, the detection monitoring programs must have been in place by October 1, 1984. For licenses issued after September 30, 1983, the detection monitoring programs must be in place when specified by the Commission in orders or license conditions. Once ground‐water protection standards have been established pursuant to paragraph 5B(1), the licensee shall establish and implement a compliance monitoring program. The purpose of the compliance monitoring program is to determine that the hazardous constituent concentrations in ground water continue to comply with the standards set by the Commission. In conjunction with a corrective action program, the licensee shall establish and implement a corrective action monitoring program. The purpose of the corrective action monitoring program is to demonstrate the effectiveness of the corrective actions. Any monitoring program required by this paragraph may be based on existing monitoring programs to the extent the existing programs can meet the stated objective for the program.

Criterion 8‐‐Milling operations must be conducted so that all airborne effluent releases are reduced to levels as low as is reasonably achievable. The primary means of accomplishing this must be by means of emission controls. Institutional controls, such as extending the site boundary and exclusion area, may be employed to ensure that offsite exposure limits are met, but only after all practicable measures have been taken to control emissions at the source. Notwithstanding the existence of individual dose standards, strict control of emissions is necessary to assure that population exposures are reduced to the maximum extent reasonably achievable and to avoid site contamination. The greatest potential sources of offsite radiation exposure (aside from radon exposure) are dusting from dry surfaces of the tailings disposal area not covered by tailings solution and emissions from yellowcake drying and packaging operations. During operations and prior to closure, radiation doses from radon emissions from surface impoundments of uranium or thorium byproduct materials must be kept as low as is reasonably achievable.

Checks must be made and logged hourly of all parameters (e.g., differential pressures and scrubber water flow rates) that determine the efficiency of yellowcake stack emission control equipment operation. The licensee shall retain each log as a record for three years after the last entry in the log is made. It must be determined whether or not conditions

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are within a range prescribed to ensure that the equipment is operating consistently near peak efficiency; corrective action must be taken when performance is outside of prescribed ranges. Effluent control devices must be operative at all times during drying and packaging operations and whenever air is exhausting from the yellowcake stack. Drying and packaging operations must terminate when controls are inoperative. When checks indicate the equipment is not operating within the range prescribed for peak efficiency, actions must be taken to restore parameters to the prescribed range. When this cannot be done without shutdown and repairs, drying and packaging operations must cease as soon as practicable. Operations may not be restarted after cessation due to off‐normal performance until needed corrective actions have been identified and implemented. All these cessations, corrective actions, and restarts must be reported to the appropriate NRC regional office as indicated in Criterion 8A, in writing, within ten days of the subsequent restart.

To control dusting from tailings, that portion not covered by standing liquids must be wetted or chemically stabilized to prevent or minimize blowing and dusting to the maximum extent reasonably achievable. This requirement may be relaxed if tailings are effectively sheltered from wind, such as may be the case where they are disposed of below grade and the tailings surface is not exposed to wind. Consideration must be given in planning tailings disposal programs to methods which would allow phased covering and reclamation of tailings impoundments because this will help in controlling particulate and radon emissions during operation. To control dusting from diffuse sources, such as tailings and ore pads where automatic controls do not apply, operators shall develop written operating procedures specifying the methods of control which will be utilized.

Milling operations producing or involving thorium byproduct material must be conducted in such a manner as to provide reasonable assurance that the annual dose equivalent does not exceed 25 millirems to the whole body, 75 millirems to the thyroid, and 25 millirems to any other organ of any member of the public as a result of exposures to the planned discharge of radioactive materials, radon‐220 and its daughters excepted, to the general environment.

Uranium and thorium byproduct materials must be managed so as to conform to the applicable provisions of Title 40 of the Code of Federal Regulations, Part 440, ʺOre Mining and Dressing Point Source Category: Effluent Limitations Guidelines and New Source Performance Standards, Subpart C, Uranium, Radium, and Vanadium Ores Subcategory,ʺ as codified on January 1, 1983.

Criterion 8A‐‐Daily inspections of tailings or waste retention systems must be conducted by a qualified engineer or scientist and documented. The licensee shall retain the documentation for each daily inspection as a record for three years after the documentation is made. The appropriate NRC regional office as indicated in Appendix D to 10 CFR Part 20 of this chapter, or the Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC, 20555, must be immediately notified of any failure in a tailings or waste retention system that results in a release of tailings or waste into unrestricted areas, or of any unusual conditions (conditions not contemplated in the design of the retention system) that is not corrected could indicate the potential or lead to failure of the system and result in a release of tailings or waste into unrestricted areas.

II. Financial Criteria

Criterion 9‐‐Financial surety arrangements must be established by each mill operator prior to the commencement of operations to assure that sufficient funds will be available to carry out the decontamination and decommissioning of the mill and site and for the reclamation of any tailings or waste disposal areas. The amount of funds to be ensured by such surety arrangements must be based on Commission‐approved cost estimates in a Commission‐approved plan for (1) decontamination and decommissioning of mill buildings and the milling site to levels which allow unrestricted use of these areas upon decommissioning, and (2) the reclamation of tailings and/or waste areas in accordance with technical criteria delineated in Section I of this Appendix. The licensee shall submit this plan in conjunction with an environmental report that addresses the expected environmental impacts of the milling operation, decommissioning and tailings reclamation, and evaluates alternatives for mitigating these impacts. The surety must also cover the payment of the charge for long‐term surveillance and control required by Criterion 10. In establishing specific surety arrangements, the licenseeʹs cost estimates must take into account total costs that would

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be incurred if an independent contractor were hired to perform the decommissioning and reclamation work. In order to avoid unnecessary duplication and expense, the Commission may accept financial sureties that have been consolidated with financial or surety arrangements established to meet requirements of other Federal or state agencies and/or local governing bodies for such decommissioning, decontamination, reclamation, and long‐term site surveillance and control, provided such arrangements are considered adequate to satisfy these requirements and that the portion of the surety which covers the decommissioning and reclamation of the mill, mill tailings site and associated areas, and the long‐term funding charge is clearly identified and committed for use in accomplishing these activities. The licenseesʹs surety mechanism will be reviewed annually by the Commission to assure, that sufficient funds would be available for completion of the reclamation plan if the work had to be performed by an independent contractor. The amount of surety liability should be adjusted to recognize any increases or decreases resulting from inflation, changes in engineering plans, activities performed, and any other conditions affecting costs. Regardless of whether reclamation is phased through the life of the operation or takes place at the end of operations, an appropriate portion of surety liability must be retained until final compliance with the reclamation plan is determined.

This will yield a surety that is at least sufficient at all times to cover the costs of decommissioning and reclamation of the areas that are expected to be disturbed before the next license renewal. The term of the surety mechanism must be open ended, unless it can be demonstrated that another arrangement would provide an equivalent level of assurance. This assurance would be provided with a surety instrument which is written for a specified period of time (e.g., 5 years) yet which must be automatically renewed unless the surety notifies the beneficiary (the Commission or the State regulatory agency) and the principal (the licensee) some reasonable time (e.g., 90 days) prior to the renewal date of their intention not to renew. In such a situation the surety requirement still exists and the licensee would be required to submit an acceptable replacement surety within a brief period of time to allow at least 60 days for the regulatory agency to collect.

Proof of forfeiture must not be necessary to collect the surety so that in the event that the licensee could not provide an acceptable replacement surety within the required time, the surety shall be automatically collected prior to its expiration. The conditions described above would have to be clearly stated on any surety instrument which is not open‐ended, and must be agreed to by all parties. Financial surety arrangements generally acceptable to the Commission are:

(a) Surety bonds;

(b) Cash deposits;

(c) Certificates of deposits;

(d) Deposits of government securities;

(e) Irrevocable letters or lines of credit; and

(f) Combinations of the above or such other types of arrangements as may be approved by the Commission. However, self insurance, or any arrangement which essentially constitutes self insurance (e.g., a contract with a State or Federal agency), will not satisfy the surety requirement since this provides no additional assurance other than that which already exists through license requirements.

Criterion 10‐‐A minimum charge of $250,000 (1978 dollars) to cover the costs of long‐term surveillance must be paid by each mill operator to the general treasury of the United States or to an appropriate State agency prior to the termination of a uranium or thorium mill license.

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If site surveillance or control requirements at a particular site are determined, on the basis of a site‐specific evaluation, to be significantly greater than those specified in Criterion 12 (e.g., if fencing is determined to be necessary), variance in funding requirements may be specified by the Commission. In any case, the total charge to cover the costs of long‐term surveillance must be such that, with an assumed 1 percent annual real interest rate, the collected funds will yield interest in an amount sufficient to cover the annual costs of site surveillance. The total charge will be adjusted annually prior to actual payment to recognize inflation. The inflation rate to be used is that indicated by the change in the Consumer Price Index published by the U.S. Department of Labor, Bureau of Labor Statistics.

III. Site and Byproduct Material Ownership

Criterion 11‐‐A. These criteria relating to ownership of tailings and their disposal sites become effective on November 8, 1981, and apply to all licenses terminated, issued, or renewed after that date.

B. Any uranium or thorium milling license or tailings license must contain such terms and conditions as the Commission determines necessary to assure that prior to termination of the license, the licensee will comply with ownership requirements of this criterion for sites used for tailings disposal.

C. Title to the byproduct material licensed under this Part and land, including any interests therein (other than land owned by the United States or by a State) which is used for the disposal of any such byproduct material, or is essential to ensure the long term stability of such disposal site, must be transferred to the United States or the State in which such land is located, at the option of such State. In view of the fact that physical isolation must be the primary means of long‐term control, and Government land ownership is a desirable supplementary measure, ownership of certain severable subsurface interests (for example, mineral rights) may be determined to be unnecessary to protect the public health and safety and the environment. In any case, however, the applicant/operator must demonstrate a serious effort to obtain such subsurface rights, and must, in the event that certain rights cannot be obtained, provide notification in local public land records of the fact that the land is being used for the disposal of radioactive material and is subject to either an NRC general or specific license prohibiting the disruption and disturbance of the tailings. In some rare cases, such as may occur with deep burial where no ongoing site surveillance will be required, surface land ownership transfer requirements may be waived. For licenses issued before November 8, 1981, the Commission may take into account the status of the ownership of such land, and interests therein, and the ability of a licensee to transfer title and custody thereof to the United States or a State.

D. If the Commission subsequent to title transfer determines that use of the surface or subsurface estates, or both, of the land transferred to the United States or to a State will not endanger the public health, safety, welfare, or environment, the Commission may permit the use of the surface or subsurface estates, or both, of such land in a manner consistent with the provisions provided in these criteria. If the Commission permits such use of such land, it will provide the person who transferred such land with the right of first refusal with respect to such use of such land.

E. Material and land transferred to the United States or a State in accordance with this Criterion must be transferred without cost to the United States or a State other than administrative and legal costs incurred in carrying out such transfer.

F. The provisions of this part respecting transfer of title and custody to land and tailings and wastes do not apply in the case of lands held in trust by the United States for any Indian tribe or lands owned by such Indian tribe subject to a restriction against alienation imposed by the United States. In the case of such lands which are used for the disposal of byproduct material, as defined in this Part, the licensee shall enter into arrangements with the Commission as may be appropriate to assure the long‐term surveillance of such lands by the United States.

IV. Long‐Term Site Surveillance

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Criterion 12‐‐The final disposition of tailings, residual radioactive material, or wastes at milling sites should be such that ongoing active maintenance is not necessary to preserve isolation. As a minimum, annual site inspections must be conducted by the government agency responsible for long‐term care of the disposal site to confirm its integrity and to determine the need, if any, for maintenance and/or monitoring. Results of the inspections for all the sites under the licenseeʹs jurisdiction will be reported to the Commission annually within 90 days of the last site inspection in that calendar year. Any site where unusual damage or disruption is discovered during the inspection, however, will require a preliminary site inspection report to be submitted within 60 days. On the basis of a site specific evaluation, the Commission may require more frequent site inspections if necessary due to the features of a particular disposal site. In this case, a preliminary inspection report is required to be submitted within 60 days following each inspection.

1. In the case of thorium byproduct materials, the standard applies only to design. Monitoring for radon emissions from thorium byproduct materials after installation of an appropriately designed cover is not required.

2. This average applies to the entire surface of each disposal area over a period of a least one year, but a period short compared to 100 years. Radon will come from both byproduct materials and from covering materials. Radon emissions from covering materials should be estimated as part of developing a closure plan for each site. The standard, however, applies only to emissions from byproduct materials to the atmosphere.

3. The abbreviation N.O.S. (not otherwise specified) signifies those members of the general class not specifically listed by name in this list.

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Example Procedures

SEALED SOURCE SURVEYS AND LEAK TESTS

I. Scope This procedure covers performance of required surveys for sealed sources located on campus. These include leak tests and measurements of ambient radiation levels where applicable.

II. Purpose

It is necessary to leak test sealed sources that are in use to ensure that personnel are not exposed to radioactive contamination and to ensure that any measurements or calibrations made with these sources are accurate and correct. Federal regulations (10 CFR 35) and NRC Materials License No. 34‐000293‐02 require periodic leak testing.

III. References

10 CFR 20 (Standards for Protection Against Radiation) 10 CFR 35 (Medical Use of Byproduct Material) Regulatory Guide 10.8 (Guide for the Preparation of Applications for Medical Use

Programs) IV. Equipment

smear wipes 20‐ml liquid scintillation vials cotton swabs plastic sample vials protective gloves portable radiation detection meter specimen tongs NRC 3 forms

V. Precautions Follow approved procedures when handling any source. Always wear proper dosimetry and protective gloves. Be conscious of time, distance, and shielding. Sources may have high radiation levels on contact with the source. The

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following precautions shall be taken when surveying or leak testing high dose rate sources:

• Tongs and cotton swabs should be used for handling and smear wiping the sources whenever possible.

• Extremity and personal dosimetry shall be worn.

• Minimize the amount of time that the source is unshielded.

VI. Procedures

1. Contact RSO to arrange for testing and access to the source(s).

2. Ensure a current NRC Form 3 is posted in the source storage room ‐ replace if missing or obsolete

3. Verify all information listed on sealed source records is complete and accurate; correct or enter any incorrect, obsolete, or missing information.

• If it is not be possible to obtain the model or serial number of the source itself. In this case, the model and serial numbers of the device may be used.

• Records should also include the name, office, and telephone number of a person to contact with any questions which may arise regarding day‐to‐day use of the source.

4. Extend the source from the bottom of the nuclear gauge or open the gauge to expose the source or source holder.

5. Wipe outside of source or source capsule with dry cotton swab or smear wipe. Pay special attention to any seams or openings.

6. Survey the swab or smear wipe with a portable radiation detection instrument. If the count rate noted is in excess of 100 CPM above background, notify the RSO that the source may be leaking and is not to be used until further notice.

7. Place smear wipe or swab into LSV and break off tip (if swab used), then replace the lid on the vial. A gamma counter is also acceptable for counting smear wipes or cotton swabs if one is available and acceptable for the nuclide(s) in question. Cesium‐137, for example, could be counted in this manner.

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8. If the source is thought to be leaking, obtain smear wipes of the source storage container if possible and notify the RSO immediately.

9. Retract source into the gauge.

10. Count wipes on a liquid scintillation counter (or in gamma counter, if available and acceptable).

11. Record results on the appropriate record form. If the source is discovered to be leaking (or if leakage is confirmed) notify the RSO immediately. Notify the NRC as required by 10 CFR 35.59(e)(2).

12. Completed records shall be given to the RSO for review.

OPERATION AND CALIBRATION OF THE LUDLUM MODEL 12 SURVEY METER

1.0 Scope

This procedure sets forth the specific requirements to be used for the operation and calibration of the Ludlum Model 12 Survey Meter for use at the SRS.

2.0 Purpose

The purpose of this procedure is to provide instructions for the operation and calibration of the Ludlum Model 12 Survey Meter in conjunction with Model 44‐9 GM Pancake Probes, Model 43‐5 Alpha Scintillation Probes, and Model 44‐10 NaI Probes.

3.0 References, Definitions and Developmental Resources

3.1 References

3.1.1 Regulatory Guide 10.8, Rev. 2‐1987, Guide for the Preparation of Applications for Medical Use Programs

3.1.2 ANSI N3.1‐1987, Selection, Qualification and Training of Personnel

For Nuclear Power Plants

3.2 Definitions

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3.2.1 ACTIVITY ‐ The rate of disintegration (transformation) or decay of radioactive material. The units of activity for the purposes of this procedure are disintegration per minute(dpm) or micro‐Curies(FCi).

3.2.2 CALIBRATION CERTIFICATE ‐ Document certifying calibration.

It indicates the procedure used to calibrate the instrument and the record of data obtained prior to and during calibration. These documents are referred to as Instrument Service records.

3.2.3 CALIBRATION STICKER ‐ Sticker attached directly to instrument

which indicates calibration status of instrument.

3.2.4 CHECK SOURCE ‐ A sample of radioactive material in which the exact quantity of radioactive material is not known but the type and energy of the emission is known. These sources are used for field response checks of radiation detection instrumentation.

3.2.5 PERFORMANCE CHECK ‐ A check of a radiation detection

instrument in which the performance of the instrument is checked against a reference source with an acceptance value of ±10% of the reference value.

3.2.6 REFERENCE STANDARD ‐ A sample of radioactive material,

usually with a long half‐life, in which the number of radioactive atoms and the type of emission is known and is NIST traceable. These standards are used for calibration and performance checks of radiation detection instrumentation.

3.2.7 QUALIFIED USER ‐ An individual that has demonstrated to

management the skills and abilities to use instrumentation and is trained to perform specific program operations without supervision.

3.3 Developmental Resources

3.3.1 Instruction manual for the Ludlum Model 12 Survey Meter

3.3.2 ANSI N323‐1978, Instrumentation Test and Calibration

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4.0 Precautions, Limitations

4.1 Precautions

4.1.1 Take care not to puncture the thin mica window of the ʺpancakeʺ G‐M detector.

4.1.2 To prevent contamination of probes, avoid contact with the

person(s) or object(s) being surveyed.

4.1.3 Prior to returning an instrument for calibration, the instrument shall be surveyed for radioactive contamination if it was used in an area where the potential for contamination existed.

4.2 Limitations

4.2.1 The operation of the Model 12 depends on the condition of the

battery. Therefore, the battery check should be performed periodically to insure proper operation.

4.2.2 Calibration shall be performed annually, after maintenance is

performed, if instrument fails the performance test or if its proper operation is in question.

4.2.3 A daily performance test is required when the instrument is in use.

4.2.4 ʺPancakeʺ GM Detectors shall be considered 10% efficient unless

otherwise noted. 5.0 Responsibilities and Oualifications

5.1 Responsibilities

5.1.1 Health Physics Supervisor

5.1 .1.1 Implementation of this procedure.

5.1.1.2 Periodic review of the adherence of personnel to the requirements of this procedure.

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5.1.1.3 Performs periodic surveillance of the use and maintenance of the instrument.

5.1.2 Health Physics Technician

5.1.2.1 Ensures that the instrument is calibrated at specified

intervals.

5.1.2.2 Documentation of all records in this procedure.

5.1.2.3 Notification to Health Physics Supervision of any unsafe or unusual conditions observed during operation of the instrument.

5.2 Qualifications

5.2.1 Health Physics Technicians shall be qualified in accordance with

the requirements of ANSI 3.1 ‐ 1987 to operate this instrument for any of the following surveys: job coverage and unconditional releases.

5.2.2 Jr. Health Physics Technicians may operate this instrument under

supervision of a Health Physics Technician meeting the requirements of Section 5.2.1

6.0 Procedure

6.1 Operation

6.1.1 Verify that the instrument has a valid Calibration Sticker , and the

performance test has been performed and initialed on the Performance Test Signoff Sticker.

6.1.2 Examine the instrument for any obvious physical damage which

could interfere with its proper operation.

6.1.3 Perform a battery check on the instrument.

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6.1.4 Set the audio, response (fast or slow), and range selector to the appropriate settings.

6.1.5 Proceed with operation in accordance with the desired use.

6.2 Calibration

6.2.1 Instrument Calibration shall be performed by an approved instrument Vendor Company.

6.2.2 The Calibration Certificate shall include as a minimum the following:

6.2.2.1 The procedure used to calibrate the instrument.

6.2.2.2 The technician who performed the calibration.

6.2.2.3 Date calibrated and due date.

6.2.2.4 The instrument ʺAs Foundʺ and ʺAs Leftʺ data.

6.2.2.5 The instrument model and serial number.

6.3 Performance Test

6.3.1 Perform a performance test on the instrument and record all data on the Performance Test Log Sheet(HPF‐009) for that instrument.

Note: A performance test must be performed on each probe for that instrument. (i.e. GM Pancake and Alpha Scintillation Probe.)

6.3.2 Obtain the Performance Test source designated by the HPF‐009 for

the instrument and probe.

6.3.3 Record the information for each section of HPF‐009.

6.3.4 Examine the instrument for any obvious physical damage which could interfere with its proper operation.

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6.3.5 Verify that the instrument has a current Calibration Data Sticker ,

and Performance Test Signoff Sticker attached.

6.3.6 Perform a Battery Check to check that the battery is within the Batt OK range on the meter.

6.3.7 Expose the detector to the performance test source. If the response

is within the designated range for the source, proceed to step 6.3.9. If the instrument fails, record ʺFʺ for fail on HPF‐009 and remove the instrument from service for repair or calibration.

6.3.8 If the instrument fails any portion of the performance test, log the

instrument as failing on the Performance Test Log Sheet, remove from service, and notify the Health Physics Supervisor or designee. Tag the instrument out of service.

6.3.9 If the instrument passes the performance test, record ʺPʺ for pass on

HPF‐009, then initial the Performance Test Signoff Sticker on the instrument and initial the Performance Test Log Sheet.

6.4 Maintenance

6.4.1 No special storage requirements.

6.4.2 Electronic maintenance (except probe and cable replacements) shall

be performed by a vendor company in accordance with references. 7.0 Records

The following records will be generated as a result of using this procedure. Calibration records shall be kept on file until project completion, and then forwarded to the Branch Office. 7.1 Performance Test Log Sheet, HPF‐009

8.0 Exhibits (see attachments)

8.1 Exhibit 1, Performance Test Log Sheet, HPF‐009

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OPERATION AND CALIBRATION OF THE LUDLUM MODELS 12S and 19 MICRO‐R METERS

1.0 Scope

This procedure sets forth the specific requirements to be used for the operation and calibration of the Ludlum Models 12S and 19 micro‐R meters. This procedure applies to R & R International, Inc. personnel who may use this instrument at the SRS.

2.0 Purpose

The purpose of this procedure is to provide instructions for the operation and calibration of the Ludlum Models 12S and 19 micro‐R meters in accordance with the requirements specified in Reference 3.1.1.

3.0 References. Definitions and Developmental Resources

3.1 References

3.1.1 Regulatory Guide 10.8, Rev. 2‐1987, Guide for the Preparation of Application for Medical Programs

3.1.2 ANSI N3.1‐1987, Selection, Qualification, and Training of Personnel

for Nuclear Power Plants

3.2 Definitions

3.2.1 ACTIVITY ‐ The rate of disintegration (transformation) or decay of radioactive material. The units of activity for the purposes of this procedure are disintegration per minute (dpm) or micro‐Curies (μCi).

3.2.2 CALIBRATION CERTIFICATE ‐ Document certifying

calibration. It indicates the procedure used to calibrate the instrument and the record of data obtained prior to and during calibration. These documents are referred to as Instrument Service records.

3.2.3 CALIBRATION STICKER ‐ Sticker attached directly to

instrument which indicates calibration status of instrument.

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3.2.4 CHECK SOURCE ‐ A sample of radioactive material in which the exact quantity of radioactive material is not known but the type and energy of the emission is known. These sources are used for field response checks of radiation detection instrumentation.

3.2.5 PERFORMANCE CHECK ‐ A check of a radiation detection

instrument in which the performance of the instrument is checked against a reference source with an acceptance value of ±10% of the reference value.

3.2.6 REFERENCE STANDARD ‐ A sample of radioactive

material, usually with a long half‐life, in which the number of radioactive atoms and the type of emission is known and is NIST traceable. These standards are used for calibration and performance checks of radiation detection instrumentation.

3.2.7 QUALIFIED USER ‐ An individual that has demonstrated to

management the skills and abilities to use instrumentation and is trained to perform specific program operations without supervision.

3.3 Developmental Resources

3.3.1 LUDLUM Model 12S Micro‐R Meter Instruction Manual

3.3.2 LUDLUM Model 19 Micro‐R Meter Instruction Manual

3.3.3 ANSI‐N323‐1978, Radiation Protection Instrumentation Test and Calibration

4.0 Precautions and Limitations

4.1 Precautions

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4.1.1 Due to the very low ranges only the 5,000 μR scale (Model 19) or 3,000μR scale (Model 12S) can be calibrated to an actual source reading. All other scales will be calibrated to a pulse generator.

4.1.2 These detectors are not guaranteed light tight when outside of their

instrument cases.

4.1.3 Prior to returning an instrument for calibration, the instrument shall be surveyed for radioactive contamination if the instrument was used in an area where the potential for contamination existed.

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4.2 Limitations

4.2.1 Calibration shall be performed annually, after maintenance is

performed, if the instrument fails the performance test or if its proper operation is in question.

4.2.2 This instrument shall be performance tested daily when in use in

accordance with Section 6.3. 5.0 Responsibilities and Oualifications


5.l.1 Health Physics Supervisor

5.1 .1.1 Implementation of this procedure.

5.1.1.2 Periodic review of the adherence of personnel to the requirements of this procedure.

5.1.1.3 Performs periodic surveillance of the use and

maintenance of the instrument.

5.1.2 Health Physics Technician

5.1.2.1 Ensures that the instrument is calibrated at

specified intervals.

5.1.2.2 Documentation of all records in this procedure.

5.1.2.3 Notification to Health Physics Supervision of any unsafe or unusual conditions observed during operation of the instrument.

5.2 Qualifications

5.2.1 Health Physics Technicians shall be qualified in accordance with

the requirements of ANSI 3.1 ‐ 1987 to operate this instrument for

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any of the following surveys: job coverage, air sampling analysis and unconditional releases.

5.2.2 Jr. Health Physics Technicians may operate this instrument under

supervision of a Health Physics Technician meeting the requirements of Section 5.2.1

6.0 Procedure

6.1 Operation

6.1.1 Verify that the instrument has a valid Calibration Sticker and the daily performance test has been done and initialed on the Performance Test Signoff sticker. If the performance test has not been performed, see Section 6.3.

6.1.2 Examine the instrument for any obvious physical damage which

could interfere with its proper operation.

6.1.3 Perform a battery check by pressing the ʺBATʺ button on the Model 19 or by switching to the ʺBATʺ position on the Model 12S.

6.1.4 If the instrument fails any of the above checks, remove it from

service and notify the Health Physics Supervisor. 6.1.5 Set the range multiplier to an appropriate range for the activity

being investigated.

6.1.6 Read the meter after sufficient response time (i.e., the meter needle is relatively stable) changing ranges as necessary for the activity encountered. If the meter is used for an extended period of time, check the battery condition periodically to ensure proper operation.

6.1.7 Document all survey results in accordance with HP‐OP‐20.

6.2 Calibration

6.2.1 Instrument Calibration shall be performed by an approved

instrument Vendor Company.

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6.2.2 The Calibration Certificate shall include as a minimum the following items:

6.2.2.1 The procedure used to calibrate the instrument.

6.2.2.2 The technician who performed the calibration.

6.2.2.3 Date calibrated and due date.

6.2.2.4 The instrument ʺAs Foundʺ and ʺAs Leftʺ data.

6.2.2.5 The instrument model and serial number.

6.3 Performance Test

6.3.1 Perform a performance test on the instrument and record all data on HPF‐009, Performance Test Log Sheet.

6.3.2 Obtain the performance test source designated on HPF‐009 for the

instrument. Record the information for each Section of HPF‐009.

6.3.3 Examine the instrument for any obvious physical damage which could interfere with its proper operation.

6.3.4 Verify that the instrument has a current Calibration Data Sticker

and Performance Test Signoff Sticker.

6.3.5 Perform a battery check by turning the selector switch to BAT (Model 12S) or by turning the unit to 5000 μR/hr scale and depressing ʺBATʺ button (Model 19). If the unit does not read in BATTERY area, replace the batteries.

6.3.6 Expose the center of the detector to the designated source. If the

reading is within the range on HPF‐009, record ʺPʺ for ʺPASSʺ on HPF‐009. If the reading is outside the range, record ʺFʺ for ʺFAILʺ on HPF‐009 and remove the instrument from service.

6.3.7 If the instrument passes all sections of the performance test, initial

the Performance Test Signoff Sticker and complete HPF‐009.

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Note: Due to the extremely low ranges incorporated in these

instruments, only the high scales may be performance tested with an actual source reading.

6.4 Maintenance

6.4.1 No special storage requirements.

6.4.2 Electronic maintenance shall be performed by a Vendor company

in accordance with references.

6.4.3 Records of Maintenance shall be maintained with the calibration records.

6.4.4 All maintenance shall be performed in accordance with the

manufacturerʹs specifications.

6.4.5 When the instrument is repaired, the HP Supervisor or designee will review the nature of the repair and determine if a calibration is required.

6.4.6 If re‐calibration is not required, performance test the instrument (as

per Step 6.3) prior to returning the instrument to service. 7.0 Records

The following records will be generated as a result of using this procedure.

7.1 Performance Test Log Sheet, HPF‐009 8.0 Forms

8.1 HPF‐009, Performance Test Log Sheet ‐ See attachments DECONTAMINATION AND RELEASE OF MATERIAL

1.0 Scope

This procedure sets forth specific requirements, methods and techniques used for the removal of contamination from equipment and materials.

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2.0 Purpose

This procedure provides guidance for equipment and material decontamination and to minimize the potential for unintentionally releasing contaminated items from Restricted Areas.

3.0 References and Definitions

3.1 References

3.1.1 10 CFR 20 (1‐1‐92), Standard for Protection Against Radiation.

3.1.2 Reg. Guide 1.86 (June 1974), Guidelines for Decontamination

Facilities Equipment prior to Release for Unrestricted Use.

3.1.3 HP‐OP‐20 ‐ Performing Radiation and Contamination Surveys.

3.2 Definitions

Activity. The rate of disintegration or decay of radioactive material. The units of activity for the purposes of this procedure are disintegrations per minute in micro‐Curies for loose contamination, and disintegrations per minute, milliroentgen, or millirad for fixed contamination. These terms are defined in 10 CFR 20.

Contamination. Deposition of radioactive material in any place it is not

desired, particularly where its presence may be harmful. The harm may be actual

exposure to individuals or release of the material to the environment or general public. Contamination may be a result of the presence of radionuclides emitting alpha, beta, or gamma radiation.

Fixed Contamination. Radioactive contamination that is not readily removed from a surface by applying light to moderate pressure and wiping with a paper or cloth disc smear.

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Loose Contamination. Radioactive contamination that is readily removed from a surface by applying light to moderate pressure and wiping with a paper or cloth disc smear.

Minimum Detectable Activity (MDA). For purposes of this procedure,

MDA for removable radioactive contamination is defined as the smallest amount of

sample activity that will yield a net count with a 95% confidence level, based upon the background count rate of the counting system used.

Release for Unconditional Use. A level of radioactive material that is acceptable for use on property without license conditions or controls. Under normal circumstances, authorized limits for residual radioactive material are set to, or below, the values specified in USNRC Regulatory Guide 1.86.

Restricted Area ‐ An area, access to which is limited by the licensee for the

purposes of protecting individuals against undue risks from exposure to radiation and radioactive materials. Restricted Area does not include areas used as residential quarters, but separate rooms in a residential building may be set apart as a Restricted Area. 4.0 Precautions and Limitations

4.1 Precautions

4.1.1 Vacuums used for decontamination purposes shall have a HEPA filtration system, as necessary.

4.1.2 When using water to decontaminate nonporous surfaces,

waste water must be contained for sampling before being disposed of.

4.1.3 Care should be taken when choosing types of detergents,

agents, solvents and/or acids to prevent generating mixed waste.

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4.1.4 When using a sandblasting method of decontamination, local containment should be used to prevent the spread of contamination.

4.1.5 Surveys shall be performed in accordance with HP‐OP‐20.

4.1.6 Audible response shall be used on direct frisk surveys using

portable instrumentation.

4.2 Limitations

4.2.1 When using vacuums, the machine may become contaminated.

4.2.2 The maximum probe speed of instruments used during

direct frisk survey shall be 5 cm/sec. 5.0 Responsibilities and Qualifications


5.1.1 Health Physics Supervisor

5.1.1.1 Implementation of this procedure.

5.1.1.2 Initial qualification of technicians and periodic review of the adherence of personnel to the requirements of this procedure.

5.1.1.3 Signature approval to release material or

equipment, or to permit use of material or equipment with fixed activity above the limits specified in Section 6.1 of this procedure, but not above the values specified in Regulatory Guide 1.86.

5.1.2 Health Physics Technicians

5.1.2.1 Performing the requirements of this procedure.

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5.1.2.2 Adhere to other procedures referred to in this procedure.

5.1.3.3 Authorization to release materials and

equipment up to the ALARA limits specified in 6.1 of this procedure.

5.2 Qualifications

5.2.1 All personnel performing equipment and material

decontamination shall be qualified on this procedure prior to performing the task.

6.0 Procedure

6.1 ALARA Goals for Unconditional Release of Equipment and Materials:

EMISSION

TOTAL ACTIVITY

Alpha

< 20 dpm/100 cm2

Beta

< 100 dpm/100 cm2

6.2 Material and Equipment Release Limits (all limits in dpm/100 cm2)

NUCLIDE

AVERAGE

MAXIMUM

REMOVABLE

Unat, U‐235, U‐238 (alpha)

5.000

15,000

1,000

Transuranics, Ra‐226, Ra‐228, Th‐230, Th‐228, Pa‐231, Ac‐227, I‐125, I‐129

100

300

20

Th‐Nat, Th‐232, Sr‐90, Ra‐223, Ra‐224, U‐232, I‐126, I‐131, I‐133

1,000

3,000

200

Beta‐gamma emitters

5,000

15,000

1,000

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6.3 Decontamination Procedure

6.3.1 Make a reasonable effort to remove all physical signs of

contamination (e.g., visible dust, dirt, mud. or rocks).

6.3.2 Select an effective technique, such as vacuuming or using shovels, trowels, brushes, or brooms.

6.3.3 Use vacuum as the preferred method for porous surfaces.

6.3.4 Dispose of all material removed during decontamination

attempts as radiologically contaminated waste.

6.3.5 After each decontamination evolution, dry and re‐survey the equipment or material.

6.4 Survey Procedure

6.4.1 Perform a direct frisk of 100% of all accessible areas of the equipment or material, in accordance with the instrumentʹs operation procedure.

6.4.1.1 If the frisk indicates radioactive material on the

surface of the equipment or material at a level lower than the ALARA goals, then the equipment or material may be released.

6.4.1.2 If the frisk indicates radioactive material on the

surface of the equipment or material at a level greater than the ALARA goals, but below Regulatory limits, the material or equipment can be released with supervisory permission. If the frisk indicates a level greater than regulatory limits, the material may not be released.

6.4.2 If the equipment or materialʹs surface area is greater than 100

cm2 then perform a large area smear.

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6.4.2.1 If the presence of radioactive material is found

above background, the equipment shall be treated as contaminated until a detailed disc smear survey is performed.

6.4.3 Perform Disc smear of the effective accessible surface area.

6.4.4 Count and document the smears in accordance with

procedure HP‐OP‐20, Performing Radiation and Contamination Surveys.

6.4.4.1 Record smear data

6.4.4.2 If the smear results indicate transferable

activity above the ALARA goals, but below the values listed in Section 6.2 of this procedures, the material or equipment may be released with supervisory permission. However, if smear results are greater than regulatory limits, the material or equipment may not be released.

NOTE: Every effort shall be made to decontaminate the material and/or equipment to a level below the limits set forth in USNRC Reg. Guide 1.86, and preferably below the ALARA limits in this procedure.

6.4.5 Material or equipment released unconditionally shall be

documented on HPF‐002, Health Physics Survey Report, together with an associated survey log number.

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SURVEY OF RADIOACTIVE MATERIAL SHIPMENTS

1.0 Scope

This procedure applies to the shipment of radioactive material and waste at the SRS.

2.0 Purpose

This procedure describes the radiological survey requirements for outgoing shipments of radioactive materials and waste.

3.0 References and Definitions

3.1 References

3.1.1 10 CFR 20, Standards for Protection Against Radiation

3.1.2 10 CFR 71, Transportation of Radioactive Material

3.2 Definitions

3.2.1 Radioactive Material ‐ For the purposes of this procedure, radioactive material includes any material, equipment, or system component determined to be contaminated or suspected of being contaminated. Items located in known or suspected Contamination or Airborne Radioactivity Areas are considered radioactive material. Radioactive material also includes activated material, sealed and unsealed sources, and materials that emit radiation.

3.2.2 Transport Index ‐ The Transport Index (TI) is the maximum

measured dose rate at one (1) meter from the surface of the package, rounded to the nearest 0.1 mrem/hr. The TI is required to be listed on the shipping label per U.S. Department of Transportation (DOT) regulations.

4.0 Precautions and Limitations

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4.1 The instruments selected to conduct the radiological surveys should be appropriate for the type of radiation to be detected (i.e., alpha frisker for alpha contamination, etc).

4.2 All materials that are transferred out of radiological areas are subject to

the release survey requirements that are detailed in HP‐OP‐21.

4.3 All waste generated in Contamination and Airborne Radioactivity areas shall be considered radioactive waste.

4.4 All radioactive waste containers shall be properly packaged and labeled

according to the requirements of Reference 3.1.2.

4.5 The instrument Minimum Detectable Activities (MDAs) should be equal to or less than the removable contamination limits shown in Attachment 1.

4.6 Intake air filters shall be treated as non‐radioactive waste if the direct

survey of the surfaces with portable instrumentation, after 1 week of decay time, is less than the instrument MDA.

4.7 Smear wipes shall be treated as non‐radioactive waste if the direct survey

of the surfaces with portable instrumentation is less than the instrument MDA.

4.8 The transport vehicle shall be surveyed prior to loading to verify that it

complies with the established contamination limits. 5.0 Responsibilities

5.1 The Health Physics Supervisor shall be responsible for:

5.1.1 The implementation and administration of this procedure.

5.1.2 Designation of personnel necessary to carry out the requirement of the procedure.

5.1.3 Review of documents generated during the performance of this

procedure.

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5.2 Health Physics Technicians shall be responsible for the performance of the requirements set forth in this procedure.

6.0 Procedure

6.1 External Radiological Surveys of Outgoing Shipments

6.1.1 Visually inspect the shipping container and note any damage on the radiological survey record.

6.1.2 Perform an external radiation survey to determine the maximum

dose rate at 1 meter from the shipping container. Survey at 1 meter from all sides, including the top and bottom of the container, as appropriate.

6.1.3 Perform a removable contamination survey of the exterior surfaces

of the container to be shipped.

6.1.3.1 Wipe the surface of the container with a masslin cloth or other absorbent material.

6.1.3.2 Survey the swipe with the appropriate survey

instrument(s).

6.1.4 If any removable contamination is detected, the container should be decontaminated and step 6.1.3 repeated.

6.1.5 If further decontamination of the container is impractical, a

quantitative removable contamination survey must be performed. Smears shall be taken over 100 cm2 areas. Determine the maximum contamination level in dpm/cm2 (see note 2 to Attachment 1).

6.1.6 Record the results of the radiation and contamination surveys on

the radiological Survey Form

6.2 Transport Vehicle Surveys for Shipment of Radioactive Materials

6.2.1 For non‐exclusive use vehicles transporting single packages with low dose rates, this survey is not necessary.

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dose rate at any occupied location in the vehicle cab.

6.2.3 Survey the surfaces of the transport vehicle to determine the maximum dose rate present.


dose rate at two (2) meters from the surface of the transport vehicle (or if a flatbed trailer, at a point two meters from the vertical planes at the edges of the trailer).

6.2.5 Record the results of the survey on a Health Physics Survey Report

(HPF‐002). 7.0 Records

7.1 Health Physics Survey Report, HPF‐002

7.2 Radioactive Shipment Records (Copy to be kept with Survey Report) 8.0 Exhibits

8.1 Exhibit 1, Contamination Limits for Shipments of Radioactive Material

8.2 Exhibit 2, External Radiation Limits for Shipments of Radioactive Material

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University of Rochester Radiation Safety Unit

Standard Operating Procedure

Procedure 11 Radiation and Contamination Surveys Revision 0 12/17/98 Approved by:__________________________

RSO, University of Rochester 1 Purpose

1.1 The purpose of this procedure is to define how to conduct radioactive contamination surveys. The performance of periodic surveys and the way in which they are performed are required under sections of the New York Code of Rules and Regulations. This procedure is based on the model radiological survey procedure developed by the State of New York. In this procedure, steps which are mandated by New York law are italicized.

2 Scope

2.1 This procedure applies to all research laboratories at the University of Rochester and Strong Medical Center.

3 References

3.1 NYS DOH, Bureau of Environmental Protection Radiation Guide Series 10.1 rev. 2, Appendix G

3.2 NYS DOH, Bureau of Environmental Protection Radiation Guide Series 10.2 rev. 1, Appendix I

3.3 New York State Sanitary Code Chapter I, Part 16 4 Equipment

4.1 one inch (2.5 cm) diameter filter paper smear wipes (or equivalent) 4.2 impermeable gloves 4.3 radiation detection instrument (liquid scintillation counter or hand‐held

meter)

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5 Precautions 5.1 Protective gloves should be worn when smear wiping potentially‐

contaminated surfaces 5.2 Wipes should be separated to avoid cross‐contamination 5.3 Do not use a single smear wipe to survey multiple 100 cm2 locations or to

survey areas significantly larger than 100 cm2 (for example, one smear wipe may not be used to survey an entire laboratory bench top or to survey several 100 cm2 locations on a work bench)

6 Procedure for conducting smear wipe surveys for radioactive contamination

6.1 Research laboratories where unsealed sources of radioactive material are used shall be surveyed monthly.

6.2 Pay special attention to posted work areas, hoods, waste disposal areas,

storage areas, floor surfaces. Also check non‐use areas in labs such as desks, trash containers, phones and areas where possible cross contamination might occur.

6.3 Conduct wipe tests to measure contamination levels from H‐3, C‐14, S‐35,

Ni‐63 and/or other beta‐emitting radionuclides with decay energies less than 300 KeV. Wipe tests may be used to detect other beta‐emitting radionuclides (e.g. P‐32) if desired. Wipe tests are performed by wiping a piece of dry filter paper or equivalent over an area of 100 square centimeters.

6.4 Count smear wipes for removable contamination in an appropriate

counting device (i.e. liquid scintillation counter or beta counting system) 6.5 Record results as described below

7 Procedure for conducting a meter survey for radioactive contamination

7.1 Radiation meter surveys may only be performed for isotopes emitting gamma radiation or beta particles with an energy greater than 300 KeV.

7.2 Prior to conducting a radiation meter survey, the following checks shall be

performed:

7.2.1 All radiation detection instruments must be calibrated annually. Verify the meter to be used is in calibration

7.2.2 Verify proper battery operation by taking the main switch to the

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“Battery Test” position (or equivalent) and observing the needle deflection to the “Battery Test” (or equivalent) position

7.2.3 Verify the physical condition of the instrument is satisfactory 7.2.4 Verify the instrument cable is intact, in good physical condition,

and does not have any cuts or tears in the insulation 7.2.5 Set the audible response switch to the “On” position and set the

response switch to the “F” (fast response) position 7.2.6 Verify the meter has been checked for proper response against a source of

known strength on the day of use.

7.3 Hold the radiation detector between 0.5 and 1 cm from the surface to be surveyed and move at a rate of 3‐5 cm per second. NOTE: holding the probe at an excessive distance or moving the probe too rapidly may result in not detecting radioactive contamination. Holding the probe too close to the surface surveyed may result in contamination of the probe.

7.4 Record the highest net count rate reading noted on the survey map as

noted. 7.4.1 To determine the net count rate, subtract background count rate

from the instrument reading. For example, if you have 50 cpm from background radiation (measured outside the laboratory) and the instrument reads 300 cpm, your net count rate is 250 cpm.

7.4.2 Convert this count rate to a disintegration rate using the meter efficiency

for the isotope in use. For example, if a count rate of 250 counts per minute is noted for P‐32 and the meter efficiency for P‐32 is 50%, the disintegration rate is 500 disintegrations per minute (dpm).

8 Recording survey results

8.1 Permanent records will be kept of all survey results, including negative results. These records shall be maintained in the radiation safety record binder maintained by each radiation permit holder.

8.2 The record will include location, date, serial numbers of instruments used, results

of the daily instrument response check (see 7.2.6, above), and the name of the person conducting survey. Record background count rate in units of counts per minute if a direct radiation meter survey was used instead of smear wipes.

8.3 The survey record will include drawings of areas surveyed, identifying

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relevant fixtures such as active storage areas, waste, and work areas. 8.4 Results shall be noted in units of disintegrations per minute per 100 cm2. To

convert instrument count rates to disintegration rates, divide the count rate by the instrument efficiency for the specific nuclide detected.

8.5 The RSO will be immediately notified if contamination levels exceed 500

dpm per 100 cm2. Any areas containing removable contamination in excess of 200 dpm per 100 cm2 shall be decontaminated to less than these levels.

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Running an Effective Health Physics Program

Introduction Radiation safety is a requirement for any facility that uses radiation or radioactivity. Even small hospitals have x‐ray and fluoroscopy machines, and larger facilities also have CT units, nuclear medicine, and radiation oncology as well. Regardless of the nature and size of a hospital’s radiation safety program, some issues are constant – policies and procedures must be developed and enforced, personnel must be issued and required to wear dosimetry, regulators must be appeased, and so forth. On the other hand, running a large medical health physics program also has some fundamental differences compared to a smaller program; there may be on‐going research, many more modes of radiation and radioactive materials use, the presence of a Radiation Safety Committee, and more. My experience is in running a large research and medical radiation safety program, and this chapter is written from that perspective. Where possible, I will note differences between large and small programs, and many of these differences will be obvious even without explicit mention. I also depart from convention in this chapter by writing much of it in the first person; it is based on my experiences, and writing in the third person would be awkward at best, and unreadable at worst. Finally, I assume that any RSO is conversant in the regulatory and technical aspects of his or her job, and many of these form the basis of other chapters in this text. In addition, many of these issues are addressed quite nicely by existing references; in particular, I have found Miller’s book (Miller 1992) and NCRP Report 127 (1998) to be particularly helpful. In addition, Operational Radiation Safety contains a plethora of helpful and informative papers on a regular basis. In a similar vein, there are many issues related to managing the radiation safety staff, but the majority of these matters are little different from other management challenges and are addressed quite nicely in business‐ and management‐related literature and courses. Accordingly, I will not dwell on regulatory, management, and technical issues but, rather, on what I have sometimes called the art of being an RSO. My priorities as RSO were to keep people safe from harm, to comply with regulatory and policy requirements, and to provide the highest level of service possible to those at our facility who depended on radiation and radioactivity for their work. On a daily basis, being a service provider was what was most noticed, and our goal was to be perceived as an organization that helped, not hindered research and medical care.

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Licensing and Device Registration Issues The most common licensing and registration issues that arise are those involving initial applications, renewal applications, and amendments. Both initial and renewal applications are fairly similar, so they can be treated similarly.

Don’t tie yourself down with specifics The rule of thumb I always tried to follow with regards to licensing matters is that it is better to be generally correct than to be specifically wrong. In other words, there is not often a need to tie yourself down with very specific promises or statements when a more general statement will work as well. As one example, I worked for a licensee that had promised in its license that biological waste (e.g. rat carcasses) would be packed in lime, frozen, double‐bagged, and placed in 33‐gallon drums with absorbent materials for ultimate disposal. This was exactly what the licensee was doing at the time of the license renewal, so the RSO was simply recording the current practice. Unfortunately, a new waste broker and changing practices led to a change in practice – at the time I worked for this licensee, we were simply double‐bagging and freezing the biologics, and packing them with absorbents into 55‐gallon drums for disposal. The over‐specificity in our license forced us to be in non‐compliance with either our license conditions, or with waste acceptance criteria. When we renewed our license, this was amended to say that we would simply package and dispose of biologics in accordance with waste acceptance criteria set forth by our waste broker and disposal contractors. It is often tempting to go overboard in a license application or amendment – to promise to do more than required, to be extra‐careful, to be seen to exceed expectations so that the request will be granted more quickly and without the anticipated back‐and‐forth that often accompanies any license‐related requests. Such an approach may, indeed, save time up front, but it is almost certain to cost more time down the road, and to create a license that is more complicated than need be the case. In applying for a license or for a license amendment, I tend to prefer to work with regulators from the start. By this I mean that, before I begin drafting the document(s), I call our license reviewer to talk with them about the request. I want to tell them what we want to do and why, how we intend to proceed, and to ask their thoughts on the matter – what we might be forgetting, whether or not the request is likely to be approved, and what we might need to do to speed up the process. And, before sending off the request, I often e‐mail a draft version to the license reviewer to again ask for a quick read‐through and comments before officially submitting the request. This process

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takes more work than simply drafting a request and mailing it off, but it also helps ensure that, by the time a license or license amendment request is submitted, it has a nearly 100% chance of being accepted fairly rapidly because the people responsible for approving the documents have already, in effect, signed off on it. And, at times, this process has helped me to understand that a request is premature – in one instance (trying to extend our decay‐in‐storage half‐life limitation from 90 days to 270 days), I was told that our waste storage area needed upgrading before any such request could be approved – this not only gave me impetus to hasten these upgrades, but also saved me the time and trouble of trying to get an amendment that had no chance of success.

Minimize attachments One other caution comes to mind with respect to submitting license or license amendment applications; the amount of supporting documents should be restricted to the minimum needed. For example, many licensees will submit copies of their procedures, instrument inventory lists, policies, lesson plans, and so forth to show their regulators exactly what they intend to do. Unfortunately, one submitted, these documents may be incorporated into a license, leaving the licensee unable to modify them without a license amendment. On the one hand, it is nice to have a regulatory stamp of approval on these important documents. On the other hand, it is also nice to be able to edit and revise important internal documents without going through the rigmarole of a license amendment. Such documents should not be submitted as part of a license or amendment application package unless your intent is that these documents become part of your license.

Working with Regulators Regulators, especially during inspections, try to find things wrong with our programs, which they use to embarrass us, put us to work, and maybe even fine us. Since they are trying to dig up stuff to use against us, we should not tell or show them anything except exactly what they ask for. In the world of regulatory compliance, it’s us licensees against them, the regulators. At least, this is how many RSOs view the regulator/regulated relationship, and I think this approach is dead wrong. I have had much more luck working with regulators than trying to work against them, and I have found that this approach is less stressful and more productive than the alternate. What this means from a practical standpoint is that I think it makes sense to work with regulators on things such as informing them of problems (even when this may not be required), cooperating with them during inspections, and asking their advice when it makes sense.

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Keeping regulators informed Informing regulators of problems does not mean running to them with every minor setback, or tattling on yourself. It means, instead, realizing two facts about your regulators: first, they are going to find out about major problems anyhow, so why not involve them at the start and; second, your regulators are a resource who, in many cases, have seen more variety of problems than many RSOs. When a problem does arise, I think it makes sense to first stabilize the situation (if appropriate), then to develop an approach to address the problem, and finally to give my inspector a phone call (which I prefer over an e‐mail) to discuss the problem and my proposed course of action so that the regulator has the opportunity to think about whether or not my approach makes sense, or if I might have overlooked something. For example, we had a fairly extensive spill of P‐32 in one of our laboratories, which my staff and I immediately went to work to characterize and clean up. As soon as the situation was under control and a recovery plan developed, I called our inspector of the situation and what we intended to do. I made a point of noting that I was calling voluntarily, not because I was required to do so, and that I would appreciate her thoughts and comments. She felt that we should perform urine bioassay to check for uptake, which was something I had overlooked. I kept our inspector up‐to‐date with recovery actions, and she continued to advise me when she felt I had overlooked something. Could I have addressed this problem without involving our regulators? Certainly. But by involving our regulators from the start, I was able to obtain another perspective, to gain some valuable advice, and (most importantly) our regulators realized that I was not trying to hide a potentially embarrassing situation but, rather, that I simply wanted to address the problem as effectively as possible.

Surviving inspections Inspections are another area in which I think it makes sense to work with, rather than against regulators. And, to me, it makes sense to try to use an inspection to help your program rather than simply viewing it as a burden. First and foremost, I strongly feel that a radiation safety program should be “inspection ready” at all times. Since most inspections are unannounced, you will likely not have the opportunity to prepare for one in advance. Paperwork should be organized, inspections and instrument calibrations should be up‐to‐date, waste should be processed, and so forth on a continuing basis so that, whenever you are inspected, the inspectors will find that your day‐to‐day operations are acceptable to them. If you are

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continually putting off something important until “just before the inspection”, you are in danger of having an unexpected inspection pre‐empt your planned work. When the inspectors arrive, they will likely wish to speak with the RSO in an entrance interview. This gives them a chance to tell you their purpose in appearing on your doorstep and to ask you about your view of the program. This can be your opportunity to make sure the inspectors know of all the improvements you have made since the last inspection, as well as to point out to them areas you know are weak (within reason) and your plans to address these weaknesses. One year, for example, I told our inspectors that I felt I needed more staff and asked them to mention this in their exit interview and inspection report if, after their inspection, they agreed. Another year, I noted that our Radiation Safety Committee was uncomfortable in overseeing the use of medical devices (which were registered with the state, but not included as radioactive materials on our license), and I asked what other similar institutions did. In both cases, the inspectors made comments in their reports that supported my position, making it easier to gain approval for these changes. During inspections, I would ensure that our inspectors were never left alone; at all times (except during restroom visits), a member of my staff was on hand to escort our inspectors, to keep track of the progress of the inspection, and to answer questions and address concerns immediately. For example, there is to be no food storage in radiological posted laboratories, but many labs keep table salt and powdered milk as laboratory reagents. A new inspector may not realize this, which could lead to noting a potential violation. On the other hand, if escorted by a knowledgeable health physicist or technician, the inspector can be informed that the powdered milk is for laboratory use only, and to (hopefully) show the inspector that the box is labeled to show this. In this way, many potential violations can be avoided. In the event there is a violation found, it may be possible for the Radiation Safety staff member to correct it on the spot or, at the very least, to be able to report to the RSO first‐hand what was found. When potential violations are noted, I will make an effort to correct them before the inspection concludes so that, at the exit interview, we can report on corrective actions. Some cannot be addressed so readily, of course, in which case I will try to sketch out a corrective action plan to present to our regulators before they depart.

Performing Inspections Larger licensees will be required to perform periodic inspections of the various aspects of their radiation safety programs. This may include analytical and research laboratories, rooms assigned to nuclear medicine or radiation oncology, radioactive

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waste storage areas, and similar areas in which radioactive work is conducted. Even small licensees should conduct periodic inspections; examples include checking personnel dosimetry use among radiologists, confirming the use of good radiation safety practices among radiology technologists, or reviewing annual refresher training records among cardiologists who use fluoroscopy. The primary goal of inspections is to confirm that personnel are following regulations and site‐specific policies. A secondary goal is to talk with radiation workers, to find out what their concerns are, to give them a chance to ask questions, and to provide Radiation Safety presence on a regular basis. The purpose of inspections is NOT to find fault, although problems that are identified should be noted and corrected – this is discussed in greater detail in the following section. Rather, the purpose of an inspection is to obtain a snapshot of the quality of radiation safety in a particular place at a given time – good, bad, or indifferent. Radiation Safety has an undeniable regulatory function, especially in a broad‐scope licensee. However, Radiation Safety works best when it works in cooperation with radiation workers; if Radiation Safety staff come across as the radiation police, if an adversarial atmosphere is allowed (or encouraged) to develop, the radiation workers will cease working with Radiation Safety and will, instead, work against them. Such a situation is clearly not in the best interests of good radiation safety. My preference was for the inspector to be a conduit of information between the laboratory and Radiation Safety, and to be seen as an asset and a resource rather than as an opponent.

Conduct of inspections Inspections must be seen as fair and unbiased by those being inspected; otherwise accusations of favoritism or bias may be raised. It is inevitable that you will like some radiation workers better than others; this cannot be permitted to affect the inspection program. To that end, I have found it best to develop an inspection checklist and to distribute copies of the checklist to all who may be inspected. By so doing, the radiation workers understand what is expected of them, and they are assured that they will be held to a consistent set of standards, regardless of personal feelings. Inspections should be thorough, but not annoyingly so unless the laboratory in question has a history of poor behavior or is recovering from a previous bad inspection. It is essential to set aside sufficient time to conduct a thorough review of the relevant paperwork (inventory records, training records, survey records, dosimetry records, etc.), to perform both radiation and contamination surveys, to observe radiological work practices, and to check to ensure that radiation workers exhibit an acceptable level of knowledge of radiation safety. It is also important to look at good radiological

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housekeeping, radioactive waste storage, and so forth. Depending on the size of a laboratory, an inspection may take as little as a half hour, or as much as 2‐3 hours. By its nature, any inspection will cause some degree of interruption in a laboratory’s routine, but no inspection should bring work to a halt or interfere significantly with the daily operation of a laboratory – unless significant safety or regulatory violations are found. During the course of an inspection, it is possible that deficiencies will be noted, and the manner in which these are addressed can have a significant impact on the relations between Radiation Safety and the radiation workers. Violations should be addressed in a professional, matter‐of‐fact manner at the time they are noted, the Authorized User or laboratory supervisor should be informed of the violations, and notified of the significance of the violations found. The inspector may consider waiving minor violations, especially if they are corrected on the spot – this was my preference as RSO. This and other corrective action options will be discussed in further detail in following sections. Feedback must be provided following an inspection, even if no deficiencies were noted. This should be in the form of a letter; e‐mail is acceptable, but a hard copy letter is preferable. If violations are given, the letter should be signed by the RSO, after reviewing and agreeing that it is appropriate to issue these violations.

Enforcing Compliance with Regulations and Policies Rules must be followed, whether they are regulations or internal policies. If radiation workers are found breaking established rules, they must be informed of their transgression, informed of the requirements, and possibly subjected to corrective or disciplinary actions if such actions are warranted. It is only fair that radiation workers be aware of the regulations and policies to which they are expected to adhere. This should be communicated in initial and refresher radiation safety training, via periodic communications from Radiation Safety to radiation workers, via e‐mail, and other modes as applicable. It is helpful to distribute printed copies of your Radiation Safety Manual (or equivalent) to all laboratories and departments, and to have copies of this available on‐line as well. It is also helpful to provide a listing of the most common major and minor violations and a listing of potential actions associated with these violations. These corrective or disciplinary actions should be consistent and should reflect the severity of the violation or violations noted. For example, a minor paperwork violation

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does not warrant the same response as, for example, pipetting by mouth. Similarly, a single violation does not warrant the same response as multiple violations. Although I do not like “keeping score”, I have found it helpful to have a point system. For example, a minor violation is worth a half point and a major violation is worth a full point. Laboratories accumulating 4 points in any six‐month period would be subject to formal action by the Radiation Safety Committee, which could range from requiring refresher training to suspension of a responsible individual to suspension of a laboratory’s radioactive materials permit. Severe penalties (e.g. suspension of an individual or an entire laboratory) should be approved or issued by the Radiation Safety Committee (if the licensee has an RSC) or by the RSO’s supervisor if possible. This not only gives the RSO a second check to confirm that the violation(s) warrant some degree of penalty, but also assures those penalized that the corrective or disciplinary actions have been reviewed by more senior management and found to be appropriate. Although deciding whether or not to take actions can be relatively simple, deciding which action(s) to take is more complex. For example, if a single person is responsible for a string of violations, it may not make sense to penalize an entire laboratory. However, before doing this, it’s necessary to make sure that you have identified the right person – taking action against the wrong person not only serves no corrective action, but can actually be counter‐productive by generating contempt of Radiation Safety. On the other hand, if there are multiple problems caused by general negligence, if the Authorized User has actively or passively encouraged lax radiation safety practices, or if a single responsible party cannot be identified then it may be appropriate to take actions against the entire laboratory or department. In the following table, I list the corrective actions for a number of violations from incidents at a former licensee for which I was RSO.

Violation(s) Corrective / Disciplinary Actions Widespread contamination, failure to perform required surveys, food storage in posted lab

Entire laboratory required to attend special radiation safety refresher training given by RSO

Refusal to wear dosimetry for 6 months, refusal to complete radiation worker refresher training for 3 months

Temporary suspension of radiation worker status pending completion of refresher training

Widespread contamination of laboratory due to single worker following poor work practices

Worker suspended 2 weeks, required to attend radiation worker initial training, required to attend interview with RSO and to write report on causes of contamination

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incident I‐125 contamination stemming from poor work practices and failure to use correct detector, resulted in minor uptake by 6 people

Responsible individual suspended for 3 weeks, required to attend radiation worker initial training, required to attend interview with RSO

Widespread P‐32 contamination, use of radioactive materials by unauthorized personnel, deliberate use of radioactive materials in unposted laboratory, failure to inform Radiation Safety of contamination

Laboratory shut down for 1 week, Authorized User required to meet with RSO and RSC Chair, refresher training required for all laboratory personnel

Radiation survey meters out‐of‐calibration, minor amounts of contamination

Violations waived, laboratory required to have instruments calibrated

P‐32 contamination found in laboratory used by multiple permit holders; culprit not identified

All personnel using lab required to attend refresher radiation safety training presented by RSO, one major violation assigned to each permit holder using P‐32 in this room

For those permit holders who reported problems (typically spills or skin contamination) my policy was to first concentrate on correcting the problem, leaving the question of corrective or disciplinary actions until later. In particular, I was reluctant to penalize radiation workers for reporting a problem to us, preferring to encourage this sort of self‐reporting; my concern was that radiation workers would be less likely to come forward to ask for help if they feared being penalized for this candor.

Working with Committees Committee work is a necessary part, some would say a necessary evil, of working in any medical or research environment. While the RSO may be asked to participate in committees in a variety of areas, there are several committees that are directly relevant to Radiation Safety work and with which the RSO (or a representative) will be required to participate. These include the Radiation Safety Committee, the Quality Assurance Committee, the Human Use of Radionuclides Committee, and the Radioactive Drug Research Committee. Not all licensees will have all of these committees; this will depend on the extent of radioactive materials use at a particular licensee. The RSO may also be offered the opportunity to participate in other committees; this may be required, strongly recommended, or entirely voluntary depending on the particular licensee, guidance provided by the RSO’s supervisor, and the preference of the RSO. In deciding

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whether or not to participate in a committee directly, indirectly, or not at all, the RSO should weigh the impact of the committee’s work on the activities of Radiation Safety, the contribution Radiation Safety can make to the committee, the amount of time required for committee work, and other factors that may be relevant. For example, if the RSO is a faculty member, committee work may be expected for those faculty applying for tenure.

Radiation Safety Committee Large licensees will be required to have a Radiation Safety Committee (RSC) to oversee the operations of Radiation Safety and the RSO. Ideally, the RSC should be comprised of a wide variety of persons who represent those involved in the use of radiation and radioactivity at your facility. This should include representatives from all medical departments that use radiation and/or radioactivity, a representative from nursing if nurses work with nuclear medicine patients, representatives of your facility’s management, and (if applicable) researchers from the various disciplines using radiation and radioactivity in their work. This helps ensure that all radioactive materials permit holders will have one or more peers on the RSC who can understand the particulars of their work, and gives the RSO knowledgeable personnel to use as a resource when establishing new policies and when determining corrective or disciplinary actions. The RSC should be large enough to provide this cross‐section, but not so large as to become unwieldy. It is often difficult to persuade busy physicians and researchers to take time to attend RSC meetings; a committee that is unnecessarily large may suffer from the inability to achieve a quorum. Although the primary purpose of the RSC is that of oversight, they also represent a tremendous intellectual asset to the RSO. For example, most RSOs are not physicians, and even those that are, are usually authorized to practice in only one specialty area. Having access to physicians in other specialty areas helps the RSO to better understand the work of, for example, a nuclear pharmacy or a radiation oncology linear accelerator. This, in turn, can help the RSO to understand the impact suggested policy changes on these departments, or the effect of suggested corrective or disciplinary actions on those departments. The RSC can also serve the function of providing “backup” for the RSO. Disciplinary actions, for example, that are approved or issued by the RSC are less personal than those issued by the RSO, and the affected physicians or researchers understand that the these actions were approved by their peers on the RSC; that they are not the wild idea of an ignorant RSO. Similarly, new policies that are reviewed by the RSC are more likely to be well‐reasoned and appropriate to all those who are impacted, and are more likely to be accepted if they are perceived as having the approval of the RSC than if

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they, too, are seen as the idea of an RSO who may be seen as being ignorant of the realities of running a research or medical laboratory. Having a wide variety of competent professionals review policies, procedures, or disciplinary actions can only help to ensure that these are consistent not only with good radiological work practices, but also with efficient and effective medical and research practices. Yet another function of the RSC will be to approve researchers and possibly physicians for the use of radioactive materials. For some licensees, this comprises the bulk of the RSC work while, for others, this is minor. In most cases, either state regulations or internal documents will specify the requirements to be an authorized radioactive materials user and to prescribe radioactive materials for medical diagnosis or treatment. The role of the RSC, then, is to review the qualifications of the proposed user (physician or researcher) to determine if these qualifications have been met. The review may be performed by the entire RSC, by a subcommittee of the RSC, or by an individual RSC member (usually the RSO or RSC Chair). In the latter two cases, approval by the full RSC will typically be required following a recommendation by the subcommittee or individual RSC member. No matter how the approval process is administered, it is important that this process be taken seriously – approving an unqualified physician, for example, can severely damage a hospital’s reputation; especially if that physician comes under regulatory or media scrutiny. Although researchers are less likely to wind up in the headlines, approving an unqualified scientist as a radioactive materials user can bring unwanted regulatory attention if the approval process is flawed. In either case, it behooves the RSC to take this approval process seriously and to take all reasonable steps to make sure that the person in question meets the regulatory and internal requirements and that their qualifications are legitimate. To do so, it may be necessary to contact previous employers, preceptors, and other relevant persons to confirm references, experience, coursework, and other claims. Most RSOs will be required to submit an annual report on their activities to the RSC. Although the primary purpose of this report is to keep the RSC apprised of the RSO’s work during the year, it can also serve as a reminder to the entire radiation worker community of the nature and scope of work carried out by the Radiation Safety staff. I used this as an opportunity to summarize all of the work we performed, if only to disabuse the medical and research communities of the impression that we were simply the radiological equivalent of janitors. Once approved by the RSC, the annual reports were converted to PDF format and posted on the Radiation Safety web site for public access. I was frequently surprised when researchers or physicians told me they had read the report.

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Quality Assurance Committee If your facility uses machines for human diagnosis or therapy, some sort of Quality Assurance Committee will be required for the purpose of overseeing the proper maintenance of these machines. The QAC will typically meet to review periodic quality‐related reports, including gamma camera tests, x‐ray machine QA checks, and other factors related to image quality. In many cases, the QAC is set up such that it is virtually toothless, often because it lacks the authority to issue penalties or to enforce decisions and recommendations. This can be addressed, in part, by requiring the QAC to report to a group that does have this authority; this can be the Radiation Safety Committee, to a senior hospital administrator, or to another hospital committee with this level of authority over medical departments. In the absence of this level of authority, QA Committees often find they lack the ability to effect change.

Human Use of Radionuclides and Radioactive Drug Research Committees Some facilities conduct research involving the administration of radioactive materials to human subjects. While this is sometimes the subject of muck‐raking journalism, such research provides valuable insights into the functioning of the human body (for example, using positron‐emitting nuclides to elucidate brain function, using tracer levels of radionuclides to investigate pharmaceutical biokinetics, or testing new nuclear medicine modalities, to name a few). However, any human research involving the administration of radioactive materials is sensitive and requires a higher level of oversight and review than non‐isotopic research. For this reason, regulations typically call for review by not only a Research Subjects Review Board (or equivalent), but also by one or more separate committees that are asked to look only at the radiological aspects of the research. These committees, often called some variation of Human Use of Radionuclides and Radioactive Drug Research Committees, are comprised of representatives of Radiation Safety, the research community, and the medical community. At least one member will be the RSO, and it is usually required that another member be a licensed pharmacist. The work of these committees is to review proposed research, to determine if the research requires the use of radionuclides, if the researcher is competent to perform the research, if the radioactivity to be administered is appropriate for the task, if the radiation dose is within acceptable limits, if the informed consent form is clear and accurate, and if there are other non‐isotopic methods available to accomplish the same end.

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In many institutions, the HURC and/or RDRC will meet infrequently and their workload may be low. This does not lessen these committees’ importance – when their work is needed, it must be of high quality. These committees are typically required by regulations, and the proposed work they review is typically both important and potentially controversial. It is essential that the review process be thorough, thoughtful, and insightful so that any projects approved will stand up to both professional and public scrutiny.

Working with Physicians Physicians are, of course, a necessary part of any hospital. They can also be difficult to work with, especially from the standpoint of enforcing regulations and policies. Many physicians are no more difficult to work with than any other radiation workers. Unfortunately, there are many physicians who either feel that complying with time‐consuming or onerous regulations is somehow beneath their dignity as physicians, or who feel that their time and work is too valuable to be consumed with “minor” regulatory matters. I have had physicians tell me that they are “too busy saving lives”, “too highly paid”, “too highly educated”, and simply “too busy” to “waste time” on regulatory matters such as undergoing annual refresher training, wearing dosimetry, undergoing fluoroscopy credentialing, and so forth. This can cause difficulties, particularly if the department chief also opposes these requirements or, indeed, is one of the offenders. While I was never able to develop a universal response to such challenges, I was able to develop some responses that, on a case‐by‐case basis were able to defuse many of these situations. One advantage I had was that my direct supervisor was a physician and a respected dean of the medical school. As he commented once, my Ph.D. helped many physicians view me with “slightly less contempt” than might otherwise have been the case. However, it did not earn the respect of the most recalcitrant of the physicians, who respected only other medical doctors. This is where my boss helped tremendously – most of my communications to medical practitioners or to their departments were countersigned by him, at least until I was sufficiently entrenched to earn, if not respect, at least some degree of compliance. My boss was also helpful in one‐on‐one meetings with difficult physicians, if only as a tacit reminder that my request, requirements, or recommendations came with the imprimatur of another physician, who had presumably screened out the chaff. Along similar lines, another aid in working with physicians was the Radiation Safety Committee, two members of which were also physicians.

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Having physicians back me up, however, was no guarantee that all of our physicians would follow all of the requirements – especially those such as annual refresher training or turning in dosimetry, which are often seen as being more objectionable than is really the case. In many cases, we actually had physicians spend more time trying to avoid a distasteful requirements than compliance would have required. One tactic I used with some degree of success was commiseration – I would whole‐heartedly agree that, for example, annual refresher training was a waste of time for a radiologist. I would then point out that, nonetheless, it was a regulatory requirement because the idiots in Albany simply didn’t appreciate this. I would further note that, waste of time or not, this was a requirement and that we could lose our ability to function as a hospital if we were found in non‐compliance with regulatory requirements. Finally, I would point out that we went to great lengths to make the refresher training as painless as our regulatory burden would permit, and that I would hope that 15 minutes could be found to help keep the medical center out of regulatory trouble. This approach worked more often or not. When it failed, my next recourse was to involve my boss and, if necessary, the chief of the appropriate department to help rein in the transgressor. Such requests were made in the form of a letter to the offending physician, copied to both my boss and the appropriate department. In these letters, I clearly spelled out the regulatory requirement, summarized efforts made to obtain compliance from the physician in question, and setting forth the possible penalties for continued non‐compliance. These penalties always included the possibility of suspension from work with or around radiation until full compliance was achieved. This approach worked in all but one instance during my five years as RSO. The one instance in which this approach failed was an older physician who received his medical training in Germany, and who was adamant that he would not “waste time on such trivial matters.” Following three months of repeated letters, e‐mails, telephone calls, and other entreaties to complete his annual refresher training and to wear his dosimetry, I finally felt it necessary to suspend his privileges to work around radiation‐generating equipment until these matters were resolved. This was done in consultation with my boss, the Chair of the RSC, and the Chief of Radiology (who, sadly, was not a supporter of Radiation Safety). We also kept hospital management involved in this process throughout, as well as the Quality Assurance Committee. Although the physician in question was livid when he was suspended, he finally acquiesced and completed the requirements, at which time his radiation worker privileges were reinstated. I can also relate that, following this incident, we had no further problems

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with Radiology, or with physicians from other departments for my remaining time as RSO. As satisfying as suspension can be, it is a step I recommend only cautiously. This physician was the only person at our hospital qualified to interpret certain radiologic procedures – his suspension meant that these patients were referred to other hospitals until he was reinstated two weeks later. This, in turn, caused some degree of embarrassment and a loss of revenue to our hospital and forced me to justify my actions to our management. Had I not already spent a great deal of time justifying this action to my boss and to myself, I may have been at a loss, and I would certainly have lost a great deal of credibility with my boss, the RSC, hospital management, and the medical community. In the particular case, my boss was quite insistent that I justify to him why a failure to wear dosimetry and to complete refresher training did warrant such a punishment and inconvenience to our patients. My reply was that, after repeated (more than 10) attempts over a 3‐month period to achieve compliance on these matters, there could be no question that the physicians fully understood the regulatory requirements, their basis in state law, the possible consequences of non‐compliance, and exactly what was required to achieve compliance. Accordingly, his refusal to meet these requirements had become a willful disregard for regulatory requirements and, as such, deserved a more drastic response than mere non‐compliance with these particular regulations. At that point, I felt I had no choice but to carry through the threat of suspension, if only so that this particular case could serve as an example to others that any refusal to comply with regulatory and policy requirements would be dealt with. My supervisor was not enthusiastic in his agreement, but did support me in this action. In all honesty, although I remain convinced that I took the only reasonable action available, I suspect my supervisor continued to have reservations about this particular matter. Having said this, I should also point out that the majority of our physicians complied, albeit grudgingly, with our regulatory requirements. In some cases, we were able to work radiation safety refresher training into their routine in‐service and continuing education program, saving us all valuable time. Having on‐line refresher training helped as well, giving them the option of accomplishing this task from home if they desired. It was only a small fraction who were determined to make things difficult, and most of them fell into line when, instead of threatening them, I commiserated and pointed out that we were all subject to the same state regulations. And, of these, there was only once in five years that I was forced to take actual disciplinary actions. On balance, I think that I spent more time on a per‐person basis trying to convince physicians of the need to comply with regulations than I did with our non‐physician

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medical and our research personnel. However, this work paid off in that I took fewer disciplinary measures with our physicians than with the other populations.

Summary and Concluding Thoughts Managing a medical radiation safety program can be complex; there is certainly more to it than what I have described in this chapter. The successful RSO will have to do his or her best to please many communities, most of whose interests are not aligned with each other or with radiation safety. To many medical personnel, radiation safety is seen as a time‐consuming intrusion into their time that serves only to add (to them) another layer of meaningless requirements that has no relationship to saving lives – to them, I would point out that loss of our license would also reduce their ability to save lives. To hospital administration, radiation safety is a cost center that does not produce revenue – to them, I would point out that Radiation Safety made possible the revenue from Nuclear Medicine, Radiation Oncology, our research laboratories, and (to a minor extent) Radiology. While this did not eliminate the grumbling, it at least helped keep it in check to some extent. Other communities had other priorities, not to mention the occasional personal agendas, feuds, and petty bickering and positioning that occurs everywhere. There is sometimes a temptation to take on a project that may have only a peripheral relationship to Radiation Safety; often at the request of someone else. However well‐intentioned or necessary such projects may appear, I found it best to view them with a great deal of skepticism, and I learned to avoid such projects unless they met certain criteria. These included:

1. Did the project relate to radiation safety? 2. Was the project interesting to me? 3. Did I see a need for the work to be done that justified investing my time? 4. Was the project something that I could actually help with, or was I being

recruited only for the benefits of my reputation or perceived approval? 5. Was the project likely to succeed or to effect the desired changes? Would my

participation increase the chances of success? 6. Would working on the project detract from my credibility or from that of

Radiation Safety? In general, I preferred not to work on projects that seemed to have little chance of success or that might negatively impair my credibility or that of Radiation Safety. My approach as RSO was to first build my credibility by successfully tackling obvious problems in a reasonable manner, and explaining my reasoning to the radiation worker

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community so that they could understand the problem‐solving approach. My feeling was that this helped me build the credibility I needed to later take on less obvious problems, or to propose potentially controversial solutions. It also let me, on occasion, take on problems that might be outside my immediate purview as RSO or that may not be as amenable to easy solution. To me, my credibility as RSO, and the credibility of Radiation Safety as an organization were resources to be built as possible, and to be used as necessary; and never to be squandered on hopeless and/or unimportant matters. References Miller, Kenneth. CRC Handbook of Management of Radiation Protection Programs, Second Edition. CRC Press, Boca Raton FL. 1992 National Council on Radiation Protection and Measurements. Report #127 (Operational Radiation Safety Program). NCRP, Bethesda MD. 1998

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Examples of forms and reports

Quality Assurance For Medical Uses of Radiation

Meeting date: 15-Nov-01

TEST: QUARTER: 2000 2001

FILM PROCESSORS (TOTAL) 4th 1st 2nd 3rd 4th 1st 2nd 3rd

Standard 780 798 793 786 784 792 792 660

No. of processors checked 752 800 787 755 735 803 776 655

No. out of service 0 0 0 1 0 1 0 2

No. out of control 0 2 2 0 1 10 0 4

No. of repaired within 24hrs. 0 0 2 1 1 11 0 4

Percent out of control (<2.0%) 0 0.3 0.3 0 0.1 1.2 0 0.6

RADIOLOGY FILM PROCESSORS (daily, MWF, weekly)

Standard 636 642 642 639 637 637 637 508





Percent out of control (<1.0%) 0 0 0.3 0 0 0.6 0 0

RADIATION ONCOLOGY FILM PROCESSOR (T,TH)

Standard 26 26 26 24 25 26 26 25





Percent out of control (<12%) 0 8.3 0 0 4 23.1 0 16

DENTAL FILM PROCESSOR (daily)

Standard 118 130 125 123 122 129 129 126





Percent out of control (<5.0%) 0 0 0 0 0 0 0 0

HEALTH PHYSICS SURVEY (annual)

Standard 21 20 16 21 20 17 15 30

No. of X-ray tubes surveyed 17 17 19 29 19 16 14 30

No. out of Service 1 0 0 0 0 0 0 0

No. of tubes -- failed test(s) 5 0 3 3 1 1 0 0

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CLINICAL IMAGING SURVEY (annual)

Standard 9 12 14 6 11 11 7 9

No. of X-ray Tubes Surveyed 9 12 14 6 11 11 7 9

COLLIMATION CHECKS (quarterly)

Standard 45 47 54 46 49 58 40 33

No. of collimators checked 45 47 48 46 49 58 40 33


No. of collimators failed 0 0 0 0 0 0 0 0

Percent failure (<10%) 0 0 0 4.3 0 0 0 0

DARKROOM FOG TESTING (semi-annual)

Standard (No. of Darkrooms) 3 0 8 0 7 0 7 0

No. of darkrooms evaluated 3 0 8 0 7 0 7 0

No. failing fog test 0 0 0 0 0 0 0 0

No. retested satisfactory NA NA NA NA NA NA NA NA

Percent failure (=0%) 0 0 0 0 0 0 0 0

SAFETY DEVICE CHECKS (semi-annual)

Standard (number issued) 0 0 0 0 0 0 *

No. returned 0 0 0 0 0 0

No. of rooms with problems noted 0 0 0 0 0 0

No. corrected 0 0 0 0 0 0

Percent returned (=100%) 0 0 0 0 0 0

FILM VIEWBOX RELAMPING (annual as of 4/1/99)*

Standard (No. of viewboxes) 10 10 10 10 10 0 10 10


No. of viewboxes relamped 11 0 5 0 14 0 18 8

No. outside acceptance limits NA NA NA NA NA NA NA NA

No. repaired NA NA NA NA NA NA NA NA

No. rechecked satisfactory NA NA NA NA NA NA NA NA

Percent failure (<30%) NA NA NA NA NA NA NA NA

REPEAT/REJECT ANALYSIS (% repeat)

Overall rate (excluding CT) (<4.5%) 3.7 4.2 4 5.8 3.8

ACCEPTANCE TESTING (as needed)

No. of reports due 0 2 0 1 0 8 1 0

No. of reports received 0 2 0 0 0 7 0 0

Percent received (=100%) 0 100 0 0 0 87.5 0 0

FILM SCREEN CONTACT (as needed)

Total No. of cassettes 13 14 12 12 12 **

No. of cassettes tested 13 14 12 0 0

No. of cassettes failing test 0 0 0 0 0

No. of cassttes repaired or replaced NA NA NA NA NA

% tested when placed in service (=100%) NA NA NA NA NA

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PROTECTIVE DEVICES (semi-annual**)

No. of aprons 133 0 148 0 0 200-230 ~300

No. of aprons tested 133 0 148 0 0 88 80

No. of aprons failing test 0 0 0 0 0 0 0

No. of aprons removed from service 0 0 0 0 0 0 0

Percent test (>80%) 80 0 80 0 0 40 27

HYPO RETENTION (semi-annual)

No. required 21 1 19 1 18 1 18 1

No. performed 21 1 19 1 18 1 18 1

No. outside acceptance limits 0 0 0 0 0 0 0 0

Percent Failure (=0%) 0 0 0 0 0 0 0 0

CT QUALITY ASSURANCE (weekly)

No. required (4 units as of 10/01/00) 39 39 39 39 52 54 52 53

No. performed 39 39 26 11 41 54 51 52

Number outside acceptance limits 0 0 0 0 0 0 0 0

Percent failure (=0%) 0 0 0 0 0 0 0 0

MAMMOGRAPHY QUALITY ASSURANCE (weekly)*

No. Required (Phantom Images) 17 13 13 13 13 13 13 13

No. performed 17 21 15 13 13 14 14 13

No. outside acceptance limits 0 0 0 0 0 0 0 0

Percent failure 0 0 0 0 0 0 0 0

No. of defects (=0) 0 0 0 0 0 0 0 0 *Weekly testing required as of April 28 1999

NUC. MED. SCINT. CAMERA RESOLUTION (weekly)

No. Required ( 6 units) 78 73 78 78 78 79 78 77

No. Performed 78 73 78 78 78 82 77 77

No. Outside Acceptance Limits 8 2 6 1 3 6 9 11

Percent Failure 10.2 2.7 7.7 1.3 3.8 7.3 11.6 14.3

NUC. MED. SCINT. CAMERA UNIFORMITY (daily)

No. Required (6 units) 378 390 383 378 378 378 384 378

No. Performed 377 389 381 378 376 377 380 374


No. Repaired 34 23 15 3 2 5 5 12

Percent Failure 9 5.9 3.9 0.8 0.5 3.7 10.5 13

NUC. CARD. SCINT CAMERA RESOLUTION (weekly)


No. Performed 22 30 26 27 25 28 26 28

No. Outside acceptance Limits 0 0 0 0 0 1 0 0

No. Repaired 0 0 0 0 0 0 0

Percent Failure 0 0 0 0 0 3.6 0 0

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NUC. CARD. SCINT. CAMERA UNIFORMITY (daily)


No. Performed 122 133 128 128 122 128 128 126


No. Repaired 0 0 1 0 0 0 0 0

Percent Failure 0 0 0.8 0 0 0 0 0

NUC. CARD. DOSE CALIBRATOR CONSTANCY (daily)

No. Required (1 unit) 61 66 64 64 61 64 64 63

No. Performed 60 66 64 64 62 64 64 63


No. Repaired 1 0 0 0 0 0 0 0

Percent Failure 1.7 0 0 0 0 0 0 0

NUC. CARD. DOSE CALIBRATOR ACCUR. (annual)

No. Required 61 66 64 1 0 0 0 1

No. Performed 60 66 64 1 0 0 0 0


No. Repaired 1 0 0 0 0 0 0 0

Percent Failure 1.7 0 0 0 0 0 0

NUC. CARD. DOSE CALIBRATOR LINEARITY (qtrly)

No. Required 1 1 1 1 1 1 1 1

No. Performed 1 1 1 1 0 1 1 0


No. Repaired 0 0 0 0 0 0 0 0

Percent Failure 0 0 0 0 0 0 0

* RADIOACTIVE MATERIALS *

NUCLEAR MEDICINE THYROID MONITORING (qtrly)

No.of Weekly Monitoring events Required 15 34 36 51 61 40+ 66

No.of Weekly Monitoring events performed 21 27 24 29 42 47 55

Per Cent failing to be monitored (=0%) 42.8 45.5 45.8 33.3 48.6 0 23

No. of Thyroid Uptakes Above Accepable Levels 0 0 0 0 0 0 0

RADIOACTIVE PHARMACEUTICALS (daily)

No. Prepared Nuclear Medicine 1686 1854 1687 1711 1698 1797 1959

No. Properly Administered 1686 1854 1687 1711 1698 1794 1959

No. of Misadministrations* (=0) 0 0 0 0 0 0 0 0

No. Prepared Nuclear Cardiology 1077 1325 1237 1286 1382 1298 949

No. Properly Administered 1077 1325 1237 1286 1382 1298 949

No. of Misadministrations* (=0) 0 0 0 0 0 0 0 0 *As defined by BERP, Section 16.25, 10NYCRR

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NUC. MED. DOSE CALIBRATOR CONSTANCY (daily)

No. Required *(2 units as of 1/1/00) 89 159* 150 153 154 154 159

No. Performed 89 155 150 153 154 154 159


No Repaired 0 0 0 0 0 0 0 0

Percent Failure 0 0 0 0 0 0 0 0

NUC. MED. DOSE CALIBRATOR ACCURACY(semi-ann)

No. Required 1 2 0 2 0 0 2 0

No. Performed 1 2 0 2 0 0 2 0


No Repaired 0 0 0 0 0 0 0 0


NUC.MED DOSE CALIBRATOR LINEARITY (qtrly)

No. Required 1 2 2 2 2 2 2 2

No. Performed 1 2 2 2 2 2 2 2


No. Repaired 0 0 0 0 0 0 0 0


RADIATION SAFETY TRAINING

Initial Training 29 43 39 59 30 22 47 44

Refresher Training 5 3 31 270 143 3 0 0

* Safety device checklist were performed for research units and underway for Radiology. ** The CR cassettes were evaluated.

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Example laboratory inspection checklist Date___________ Permit Holder__________________________ Permit #_______ Room(s) inspected________________________ Name of inspector_______________ Administrative items: Review of Preliminary Inspection Report (PIR) by RSU prior to inspection ______ Condition of door postingʹs ______ Isotopes listed in agreement with permit? ______ Are all rooms/ storage areas of radioactivity labeled or properly posted? ______ Inventory (by nuclide Inventory verification acceptable? ______ Monthly decay corrections performed (not required for H‐3 or C‐14)? ______ Are all isotopes present authorized for use in this lab? ______

Survey records Dates of lab surveys during past quarter ___________ ___________ Contamination levels found, entered in DPM ___________ Surveys include all work areas and random locations ___________ Survey meter calibration due date(s) ___________ Survey instruments appropriate for nuclide(s) used? ___________ If “No use” indicated, do radioactive material waste logs support this? ___________

Training records Do training certificates exist for all personnel working with radioactive materials?____ Have all personnel completed annual refresher training for this calendar year? ______ Have all badged personnel been properly trained? ___________

Laboratory safety and radiological work practices Personnel wearing badges if required ___________ Badges not in use stored in central location away from RAM ___________ Any eating, drinking, or application of cosmetics noted in lab ___________ Proper lab attire (lab coat, gloves, no open‐toed shoes, no shorts/skirts) ___________ All contaminated equipment (pipettors, centrifuges, etc.) labeled ___________ Radioactive materials secured (doors or freezers locked if appropriate) ___________ Use of S35 done in safe manner to minimize contamination and exposure ___________ Shielding available and in use (I‐125, P‐32, and other appropriate nuclides)___________

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Any areas with contamination in excess of 200 dpm/100 cm2 ___________ Are fume hoods in use? ____ If yes, date of last flow check ___________ List highest radiation level noted (must be less than 0.5 mr/hr at 30 cm) ___________ List highest contamination levels noted (must be less than 200 dpm/100 cm2)__________ Have thyroid bioassays been performed as required? (24‐72 hours following any use of more than 1 mCi of I‐125 or I‐131) – may not apply to all labs ___________ Radioactive waste Number of waste containers present: liquid ________ solid __________ Waste containers overly full liquid ________ solid __________ (containers are overly full if solid waste is at the top of the container or if liquid waste is < 4” from container top) Each waste container has inventory sheet showing nuclide(s), activity present, and stock vial ID number (RS number) for each discharge ___________ Waste segregation (short/long‐lived, incinerable/non‐incinerable, etc.) ___________ Biohazardous, mixed waste, sharps segregated and marked? ___________ Liquid waste segregated and labeled (aqueous/organic, specific chemicals) ___________ Secondary containment provided for liquid waste? ___________ Waste properly shielded and stored away from high‐occupancy locations ___________ Radioactive labels completely removed or defaced for all materials in short‐lived waste container? ___________ No inappropriate materials in waste containers (lead pigs, sharps, etc.) ___________ Any personnel receiving less than 70% on verbal examination? ___________ Attachments: ____RSU survey result ____ RSU inspection checklist ____Violation letter Verbal examination questions (ask at least 10) – indicate questions asked and/or missed 1. What is meant by the term “half‐life”? 2. What isotopes are used in your laboratory? 3. What radiation is emitted by each isotope used in your laboratory? 4. What is the proper shielding for this kind of radiation? 5. What agency regulates the use of radioactive materials in the State of New York? 6. Who is the Radiation Safety Officer at the University of Rochester? 7. What actions should you take in the event of a spill of radioactive materials? 8. What is the maximum allowable level for radioactive contamination?

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9. When do you notify Radiation Safety in the event of a radiological emergency? 10. Name three kinds of radiological emergencies. 11. How far from the surface do you hold a radiation detector when performing a

contamination survey? 12. How quickly should you move the detector across the surface during a survey? 13. How frequently should a radiation meter be calibrated? 14. What checks do you need to perform before each use of a radiation or contamination

survey? 15. What is the appropriate detector for the isotopes in use in your laboratory? 16. What actions do you take if you splash a small amount of radioactive liquid on your

skin? On your clothes? 17. What is the maximum amount of radiation you can be exposed to at work in a year? 18. What is bremsstrahlung? 19. What is the most likely effect of receiving an occupational radiation dose of about

100 mrem/yr for 40 years? 20. Who do you contact in the event of a radiological emergency after normal working

hours?

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APPLICATION FOR AUTHORIZATION TO USE RADIOACTIVE MATERIALS

1. Principal Investigator:_______________ 2. Date:_____/______/______ 3. Department:______________________ 4. Office Extension:____________ 5. Lab Extension:________-____________ 6. E-mail Address:____________________ 7. SS#:_______-_______-____________ 8. Date of Birth:____/_____/_______ 9. Room(s) where isotopes are used /stored _____________

Radionuclide

Source Limit (per stock vial)

Activity Limit (Maximum possession amount)

Physical and Chemical Form

(Requisition of the above isotope(s) must be placed under the user’s name). 10. PERSONNEL (Please attach to this form a completed RADIATION WORKER

INFOMATION FORM for each radiation worker and for the PI ): Do you or your lab staff require dosimetry?____

11. FACILITIES: please provide a map of the laboratory spaces with the following:

a. Area(s ) where radioisotope is used b. Area(s) where radioisotope is stored c. Show any shielding, hoods, or other safety features that are applicable d. Show location of waste containers

Note: Liquid containers require secondary containment 12. CONTAMINATION / RADIATION MONITORING EQUIPEMENT List all contamination and radiation detection equipment below indicate shared equipment INSTRUMENT MANUFACTURER

INSTRUMENT MODEL AND SERIAL NUMBERS

TYPE OF PROBE MODEL AND SERIAL NUMBERS

INSTRUMENT CALIBRATION DATES

LIST ANY EXTERNAL/INTERNAL STANDARDS USED WITH L.S.C. - ISOTOPE, ACTIVITY, ASSAY DATE ___________________________________________________________________________ HOODS (Note: A hood is not necessarily required for all types of radioactive material usage) a. are radioactive materials used in a hood?______ b. Stored in hood? _______ c. Indicate the date of the last flow test was performed and the fpm reading__/__/__ f.p.m.___ (Note: The Principle Investigator is responsible for strict inventory control of all Radioactive compounds. If the applicant wishes to delegate this responsibility, please indicate responsible Authorized User ____________________

I have read the Radiation Safety Manual and agree to follow all regulations and policies. Principle Investigators Signature ______________________Dated ____/____/____ Approval______________ Dated ____/____/__ Approval ___________Dated ___/_____/____ Radiation Safety Officer Chair, Radiation Safety Committee

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PART B PROCEDURE SUMMARY FORM 1. Principle Investigator________________________ 2. Date________________ 3. Department_______________________________ 4. Lab. #_______________ 5. Radionuclide(s)____________________________ 5a. Intramural Box ______ 6. If this part is not accompanied by Part A, supply the following: (a) Maximum possession amount:___ mCi (b) Physical and Chemical Form:_____ 7. USE IN ANIMALS

(a) Species of animals:___________________ (b) Where housed:____________ (c) Number of animals per day, week, or month:____________________________ (d) Max. injection dose in uCi per animal:___________ (e) Biological half -life:_____ (f) Special animal handling procedures to be used:__________________________

8. IN VITRO OR OTHER USE

(a) Unit of use (test tube, culture, etc.):____________________________________ (b) Number of above units per day, week, or month:__________________________ (c) Maximum concentration in microcuries/unit:_____________________________

9. HUMAN USE: ( Approval of Human Use Committee is required)

(a) Number of patients per month:________________________________________ (b) Maximum injection dose in microcuries/patient___________________________

10. Describe procedures used in the experiment (brief protocol, purpose, counting

method, and any special isotope handling or storage procedures): 11. RADIOACTIVE WASTE MATERIAL ESTIMATE: Describe and estimate the

amount of waste that will generated from the use described above.:

(a) Non-burnable solid:___________________________________ cu./ft./month (b) Burnable solid:_______________________________________ cu./ft./month (c) Animal carcasses: __________number month ________microcuries/carcass (d) Tissue (describe):_______________________________________________ (e) Liquid:____________ gal/month_________ % water________ % flammable (f) Other (describe)_________________________________________________

Principle Investigators Signature ______________________Dated ____/____/____ Approval________________________ Dated ____/____/__ Radiation Safety Officer

272

COURSE SLIDES

273

Slide 1

Radiation Safety OfficerTraining Class

Andrew Karam, Ph.D., [email protected]

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Slide 2 Course outline

• Monday - Scientific and technical basis• Tuesday - Instruments• Wednesday – Regulatory stuff• Thursday – The art of being an RSO• Friday – Case studies and wrap-up

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Slide 3

A historical perspective

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Slide 4 Historical perspective

• Radiation discovered in 1895 by Roentgen• Radioactivity discovered by Bequerel in 1896• Radiation injury recognized by 1900• X-rays and radioactivity (mostly radium) used

for legitimate and quack medical procedures for several decades

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Slide 5 Historical perspective (cont.)

• Manhattan Project gave radiation studies impetus

• Atomic weapons testing and reactor accidents have had biggest impact on public’s current view of radiation and radioactivity

• Other uses include medical, research, and industrial

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Slide 6 What this means….

• We have over a century of experience working with radiation and radioactivity

• We know more about the health effects of radiation than about virtually any other harmful substance

• Anyone who says otherwise is itching for a fight!

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Slide 7

How and where we use radiation and radioactivity

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Slide 8 Medical uses

• Cancer therapy• X-ray and CT for diagnostic purposes• X-ray and fluoroscopy in the OR• Radiopharmaceuticals for medical

diagnosis• Radioactive immunoassay (RIA)

measurements to diagnosis disease

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Slide 9 Research uses

• Radioactively-tagged molecules used as tracers of biological activity

• Gene sequencing• Irradiation of cells or organisms for

research purposes• Toxicological studies (biokinetics of toxins

and drugs)

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Slide 10 Industrial uses

• Industrial linear accelerators• Radiography devices• Gauges (tank levels, thickness, etc)• Electron and ion beams (welding, ion

implantation, etc.)• Well logging and soil density gauges• Smoke detectors, self-illuminating signs,

and other consumer products

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Slide 11 Nuclear reactors

• The US generates about 20% of its electricity with about 100 nuclear power plants

• World-wide, nearly 500 nuclear reactors produce about 16% of the world’s electrical power

• Nuclear reactors are also used to produce isotopes for medical, industrial, research, and military uses

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Slide 12

Background information

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Slide 13 Atomic structure

• Atoms have a nucleus in the center of an outer cloud of electrons

• The nucleus contains both protons and neutrons

• The number of electrons is generally equal to the number of protons

• The number of protons is what determines the chemical properties of an atom

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Slide 14 Protons and neutrons

• Protons carry a positive charge, and repel each other

• To assemble an atom, it is necessary to have a force to overcome the electrostatic repulsion of the protons

• Neutrons carry the strong nuclear force, and they act as the “glue” that holds the protons together

• There is an optimal neutron:proton ratio to make an atom stable

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Slide 15 In general…

• Atoms want to have the right ratio of protons to neutrons to be stable

• As we add or subtract neutrons, atoms become unstable (radioactive)

• As the neutron excess (or deficit) increases, the half-life decreases and the decay energy increases

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Slide 16

decay energy versus half-life for beta-gamma emitters

1.E-07

1.E-05

1.E-03

1.E-01

1.E+01

1.E+03

1.E+05

1.E+07

1.E+09

1.E+11

0 1000 2000 3000 4000 5000 6000 7000

decay energy (KeV)

half-

life

(yrs

)

half-life versus decay energy, alpha emitters

1100

100001E+061E+081E+101E+12

3.5 4 4.5 5 5.5 6

energy (MeV)

half-

life

(yrs

)

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Slide 17 What’s an isotope/nuclide

• An atomic nucleus contains protons and neutrons

• The number of protons determines which chemical element is present

• A nucleus may have different numbers of neutrons

• Each different number of neutrons with the same number of protons is a different isotope (or nuclide) of that element

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Slide 18 Examples

• C-12• C-14• P-32• Co-60• Cs-137• U-235• U-238

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Slide 19

http://www.nndc.bnl.gov/chart/reZoom.jsp?newZoom=1

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Slide 20

Radioactive decay

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Slide 21 Conservation laws

• Nucleon number – (# of protons + neutrons)

• Electrical charge• Mass• Energy

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Slide 22 Radioactive decay

• Some atomic nuclei have too much energy• They shed this energy by emitting radiation• Each isotope has a specific rate at which it

decays and emits a specific radiation energy • Half-life is the time in which half of a given

amount of isotope will decay

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Slide 23 Radioactive decay equation

12

ln2

tt o

A N

t

A A e λ

λ

λ

−

=

=

=

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Slide 24 Units

• Curie (Ci) – the amount of an isotope that has a decay rate of 37 billion decays per second (dps)

• Becquerel (Bq) – the amount of an isotope that has a decay rate of 1 dps

• Use the “typical” modifiers (m, k, M, μ, etc)

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Slide 25 Example problem

1

160

(0.1315 25 )25

0.693 0.13155.27

500 18.67

Co

yr yrsyrs

yryrs

A mCi e mCi

λ

−

−−

− ×

= =

= × =

• Assume you have a 500 mCi Co-60 source that is 25 years old

• Co-60 half-life is 5.27 years• Make sure units are the same in the exponent• So, after 25 years (almost 5 half-lives):

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Slide 26 Types of radioactive decay

• Alpha decay• Beta decay• Electron capture• Isomeric transition• Internal conversion• Spontaneous fission

• The “purpose” of radioactive decay is to either adjust the neutron : proton ratio or to help the nucleus to shed excess energy

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Slide 27 Radioactive decay series

• U-235, U-238, and Th-232 all decay to stability via a series of radioactive isotopes

• These include both alpha and beta emitters

• For example, U-238 decays to Pb-206 via 14 series nuclides, including Ra-226, Rn-222, and Po-210

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Slide 28

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Slide 29 Alpha decay/radiation

• Alpha emitters are heavy atoms• Alpha particles are helium nuclei• Can shield alphas with a piece of paper• Alphas are very damaging to DNA• Alphas are emitted with a fixed energy• Example: 4238 234

292 90U Thα α⎯⎯→ +

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Slide 30 Beta decay/radiation

• All elements have at least one isotope that decays via beta emission

• Betas can have + or – charge• Betas are emitted with a range of energies• Positively-charged betas (positrons) are

anti-matter and also emit high-energy gamma radiation (2 x 511 keV photons)

• Example: 234 23490 91Th Paβ β

− −⎯⎯→ +

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Slide 31 Electron capture

• Converts a proton into a neutron by “capturing” an inner electron

• Nucleus may emit a gamma ray, or the nuclide produced may be radioactive

• Example: 22221011

ECNa Ne γ⎯⎯→ +

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Slide 32 Isomeric transition

• Following beta decay, the daughter nuclide will sometimes be in an excited (unstable) state, even if the atom is stable

• The nucleus de-excites by emitting a photon

• Example:

137 137137 m ITCs Ba Baβ β γ− −⎯⎯→ + ⎯⎯→ +

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Slide 33 Internal conversion

• Sometimes an excited nucleus will emit a photon that’s absorbed by an inner electron

• This adds enough energy to eject the electron from the atom

• The electrons will be emitted with a fixed energy– Sometimes an x-ray will also be emitted

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Slide 34 Spontaneous fission

• Very large atomic nuclei are larger than the range of the strong nuclear force

• These atoms sometimes break in two parts (fission) spontaneously

• This leaves two fission fragments, neutrons, and some gammas

• Example: 137238 995592 37 2SFU Cs Rb n γ⎯⎯→ + + +

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Slide 35 Types of radiation and their properties

hydrogen5-20WB01Neutron

Lead1WB00Gamma

Plastic1< 1cm+/- 10.0005Beta

none20<5 μm+24 amuAlpha

ShieldQFRangeChargeMassRadiation Type

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Slide 36 Bremsstrahlung

• X-ray radiation emitted when beta particles pass an atom

• Heavy materials (such as lead) create more brems. radiation

• Best to use plastic to shield beta emitters

• Some hunters in Georgia (nation) recently died of bremsstrahlung radiation burns

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Slide 37

Radiation dose and shielding calculations

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Slide 38 Radiation dose

• A measure of energy deposited in an absorber by ionizing radiation

• 1 rad = 100 ergs/gram of absorber– = 6.242x107 MeV/gm

• 1 Gray (Gy) = 1 Joule/kg– =100 rads

• Rem = rad x QF• Sievert (Sv) = Gy x QF

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Slide 39 Distance effects on radiation dose

• Dose rate drops off as the inverse square of the distance to the source

• So doubling your distance reduces dose rate by a factor of 4

21

2 1 22

rDR DRr

=

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Slide 40 Example distance problem

50 meters10 meters5 rem/hr

? rem/hr

210 5? 5 0.250 25

mrem remhr hrm

⎛ ⎞= × = =⎜ ⎟⎝ ⎠

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Slide 41 Radiation Shielding

• Used to reduce radiation dose to protect personnel from over-exposure

• Shielding equation:

• x is the shield thickness• μ is the attenuation coefficient in units of

cm2/gm

xsh unshDR DR e μ−= ×

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Slide 42 Buildup

• In most real-life cases, the radiation beam is broad, and some gammas will scatter within the shield back into the primary beam

• This is radiation that would not otherwise have caused radiation dose

• Buildup causes the radiation dose to be higher than what would be calculated with just the shield

• Buildup factors vary according to the gamma energy and shield material – you must use a reference to find them

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Slide 43 The full shielding equation

xsh unshDR B DR e μ−= × ×

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Slide 44 Sample problem

3 metersDR = 15 rem/hrat 1 foot

6” Pb

μ= 0.02 cm2/gmρ= 11.7 gm/cm3

6” = 15 cm

What is the dose rate to the worker?

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Slide 45 Use distance and shielding

2

23015 / 0.15 /300

cmDR rem hr rem hrcm

⎛ ⎞= × =⎜ ⎟⎝ ⎠

( )230.02 11.7 15

150 / 4.48 /gmcm cmgm cm

shDR mrem hr e mrem hr− × ×

= × =

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Slide 46 Design a shield to reduce dose to 100 mrem/hr

1 meter

DR = 15 rem/hr

μ= 0.02 cm2/gmρ= 11.7 gm/cm3

6” = 15 cm

? cm

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Slide 47 Solve equation for x and solve numerically:

( ) ( )

2

12 3

0.1 /ln ln15 /

0.02 / 11.7 /

21.4 8.45

DR rem hrDR rem hrx

cm gm gm cm

cm in

μρ

⎛ ⎞ ⎛ ⎞⎜ ⎟ ⎜ ⎟⎝ ⎠ ⎝ ⎠= =− − ×

= =

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Slide 48 Using the gamma constant

• Every gamma-emitting isotope will have an associated gamma constant (which can be referenced)

• DR=ΓA at one meter where Γ is the gamma constant in units of r/hr per Ci at a distance of 1 meter (or equivalent SI units)

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Slide 49 Gamma constant example

• ΓCo-60 = 1.33 r/hr per Ci at 1 meter• So a 1000 Ci source will give a dose rate

of 1330 r/hr at 1 meter, or about 330 r/hr at 2 meters

• ΓCs-137 = 0.332 r/hr per Ci at 1 meter• So – calculate the dose from a 500 Ci Cs-

137 source that is 2 meters away

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Slide 50 Interactions of photons with matter

• Photo-electric effect– Photon absorbed by electron, knocks electron

from atom• Compton scattering

– Photon absorbed by electron, re-emitted with lower energy in random direction

• Pair production– High-energy photon spontaneously forms

electron/positron pair when passing near atomic nucleus

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Slide 51

Photon energy

Photo-electric effect

Pair production

Compton scattering

1.022 MeV

Atom

ic #

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Slide 52

Some sample problems

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Slide 53 Sample problem

• Calculate radiation dose to a person spending 2 hours at a distance of 5 meters from a 1500 CiCo-60 source.– Γ = 1.33 r/hr per Ci at 1 meter

• What stay time would you give a person at this distance?

• Now calculate the radiation dose from this source in 33 years– T1/2 = 5.27 yrs

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Slide 54 Same source….

• What is the radiation dose rate at 1 meter from this source (when new) with 4 inches of lead shielding? – 2 inches of lead will reduce dose rate by a

factor of 10 for Co-60

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Slide 55 Contamination surveys

• You are counting a C-14 sample with a GM pancake probe. The count rate is about 100,000 cpm and counting efficiency is 5%. How many micro-Curies of activity do you have?– 1 μCi = 2.22 million dpm

• How many cpm do you expect from the same activity in a P-32 source?– Counting efficiency = 35%

• How many cpm do you expect with an NaIprobe?

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Slide 56

Background radiation

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Slide 57 Sources

• Radon – 200 mrem/yr• Rocks and soils – 28 mrem/yr• Biochemistry – 40 mrem/yr• Cosmic radiation – 27 mrem/yr• Medical procedures – 53 mrem/yr• Consumer products – 10 mrem/yr• Other – 2 mrem/yr• Total – 360 mrem/yr (average)

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Slide 58 Variations

• Cosmic radiation dose is higher at high elevations and near the poles

• Dose from geology and radon are higher in areas with more igneous rocks, coals, and black shales at the surface

• Can’t change dose from biochemistry• Background radiation levels are up to 25

rem/yr (Ramsar, Iran)

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Slide 59 Ramsar

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Slide 60

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Slide 61

Biological effects of radiation exposure

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Slide 62 Interactions in the cell

• Radiation causes ionizations in the cell• These can create free radicals that attack

DNA, causing many kinds of damage• Some of this damage can be carginogenic• In addition, radiation can directly cause

single- or double-stranded DNA breaks

• However, radiation is considered a weak carcinogen

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Slide 63 Cellular response

• Cells are capable of repairing DNA damage from radiation and other mutagens

• A mutation only occurs if DNA damage is passed on to a “daughter” cell

• Cells that repair damage correctly or that die do not cause mutations

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Slide 64 DNA damage repair

• Three options:– Damage is properly repaired (no mutation)– Damage is not repaired properly (or not

repaired) and cell dies (no mutation)– Damage is not repaired properly and cell does

not die (mutation)

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Slide 65 Mutations can be:

• Harmful – birth defects, cancer, etc.• Beneficial – evolution• Neutral – most common

– Much of genome seems to be non-coding– Not all genes are active in any given cell– Many bases can be changed slightly without

changing the amino acid or protein

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Slide 66 Factors affecting cells’ radiation sensitivity

• These factors make a cell more sensitive to radiation– Less specialized– Faster rate of division– Longer reproductive lifetime– Higher level of oxygen

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Slide 67 Two types of radiation exposure

• Acute radiation exposure is exposure to high levels of radiation in a short period of time– Typical of radiation accident victims

• Chronic exposure is exposure to low levels of radiation for prolonged periods of time– Typical of radiation workers and HBRA

inhabitants

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Slide 68 Acute whole-body exposure

• ~1 rem – chromosomal damage• ~25 rem – changes in blood cell counts• ~100 rem – radiation sickness• ~450 rem – LD50

• ~1000 rem – LD100

• At higher levels of exposure, see some syndromes– GI, CNS, hematopoetic, etc.

• High exposure to parts of body may not result in death of patient (e.g. cancer therapy, XRD, etc)

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Slide 69 Chronic exposure

• Less well-defined• Most regulations based on Linear, No-

Threshold (LNT) hypothesis that all radiation above background is potentially harmful and risk is proportional to dose

• However, there is some evidence that there is a threshold, below which radiation is either harmless or beneficial (hormesis)

• Subject of much study and debate for many years

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Slide 70

Radiation dose

Cancerrisk

LNT

Hormesis

Threshold

Supra-linear

Linear quadratic

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Slide 71

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Slide 72 Studies show:

• Residents of HBRA are not affected by high natural radiation levels

• Radiologists are not affected by occupational radiation exposure

• Nuclear workers are not affected by occupational radiation exposure

• Some Hiroshima, Nagasaki, and Chernobyl survivors show no increase (in some cases a small drop) in cancer rates

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Slide 73 Risks

• Using LNT, the risk of getting cancer from exposure to 100-200 mrem/yr above background is about 1 in 10,000

• Risk of dying in an auto accident is about 1 in 7,000 each year (about 1 in 100 over a lifetime)

• Background cancer risk is about 16 in 100

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Slide 74 HPS and ICRP

• HPS advises against calculating risk for any exposures of less than 10 rem because of high uncertainties in numbers

• ICRP recommends against calculating cancer deaths in a population when most-exposed individual receives a very small dose

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Slide 75

Radiation detectors and personnel monitoring devices

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Slide 76 Types of detectors

• Gas-filled detectors– Geiger-Müeller (GM) tubes– Ion chambers – Proportional counters

• Scintillation-type detectors– Sodium-iodide (NaI) crystals– Zinc sulfide (ZnS) crystals– Liquid scintillation counters

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Slide 77

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Slide 78 Pulse height versus voltage

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Slide 79 GM tubes

• Apply high voltage to gas• Radiation causes complete ionization of gas,

giving a count on the meter• There is short “dead time” after each count,

so tube can saturate• Three types of probe – “pancake”, “hot dog”,

and energy-compensated• Used for contamination and radiation

surveys

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Slide 80 GM limitations

• Not energy-sensitive (except for energy-compensated GM)

• Dead time means that very high count rates can saturate tube

• Not sensitive to low-energy betas and gammas

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Slide 81 GM benefits

• Relatively simple and inexpensive• Relatively rugged• Relatively constant counting efficiency

across a variety of radiation energies

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Slide 82 GM uses

• Pancake probe very good for contamination surveys

• Hot dog probe useful for radiation surveys, provided properly calibrated

• Energy-compensated GM good for radiation level surveys

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Slide 83 Energy-dependence effect

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Slide 84 Ion chambers

• Radiation causes ionizations in chamber• Ionizations create electrical current, and

strength of current is related to radiation dose rate

• Used to measure radiation levels• Sensitivity does not depend on radiation

energy• Pressurized ion chamber more sensitive

and respond faster, but can develop leaks

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Slide 85 Ion chamber pressure and

temperature correction factor

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Slide 86 Proportional counters

• Radiation interactions cause spike in voltage reading

• Size of spike is proportional to radiation energy

• Since α particles are more energetic than betas, can use same detector to count and β radiation simultaneously

• However, need constant flow of gas (propane) to function properly

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Slide 87 Some proportional counters

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Slide 88

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Slide 89 NaI scintillation detectors

• Crystal emits photons during interactions with gamma rays

• Photons travel to photo-multiplier tube, which amplifies signal

• Sensitive to gamma radiation• Used for contamination and radiation surveys for

gamma emitters• Also used for nuclide ID (when used with multi-

channel analyzer)

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Slide 90 NaI detector sizes

• Thin-crystal (1” x 1mm) – used for low-energy gammas (10-100 keV)

• Thick-crystal (1”x1”) – used for high-energy gammas (100+ keV)

• Thicker crystals available (2”x2”, 3”x3”, and larger sizes)

• Background counts increase as crystal size increases – up to 14,000 cpm for a 3”x3”)

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Slide 91 NaI limitations

• Not sensitive to alpha or beta radiation• Often have low efficiency • Usually have higher background counts

than GM tubes (harder to detect low levels of contamination)

• Crystals are fragile and thermally sensitive

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Slide 92 NaI benefits

• Can have very large crystals to increase sensitivity

• Can be used to identify isotopes by looking for a specific gamma energy

• Can be used to measure radiation levels because # photons is proportional to gamma energy

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Slide 93 NaI uses

• Nuclide ID and gamma spectroscopy• Radiation levels (usually as part of a

micro-R meter)• Gamma contamination measurements• Can be used to measure beta emitters via

bremsstrahlung emitted from betas interacting in crystal, but not reliable

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Slide 94

Ideal versus actual

gamma spectrum

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Slide 95 Gamma spec features• X-ray peaks at around 100 keV. These are generated by photoelectric

absorption within the material surrounding the crystal. • Backscatter peaks at around 250 keV. These are a result of the backward

scattering of rays from outside of the detector. • An annihilation peak at 511 keV, from the detection of an annihilation photon

from pair production outside the Ge crystal. • Single and double escape peaks at energies of 511 keV and 1022 keV below

the photopeak energy. At energies greater than twice the rest mass of the electron (1.022 MeV), pair production is a possibility. The positron may then recombine with an electron and decay back to two 511keV rays which may be detected in the annihilation peak, but if one or more of these escape from the detector without any further interactions the escape peaks will be detected at energies of - 511 keV and - 1022 keV.

• The Compton edge and Compton continuum are caused by Compton scattering. The Compton edge corresponds to a maximum energy transfer when θ = 180. The continuum arises from all other scattering angles.

• The Co photopeaks at 1173 and 1333 keV, correspond to all of the γ-ray energy from the interaction being deposited in the detector.

• The photopeaks are of most interest. To -ray spectroscopists, the majority of interest lies in the region 0.2 - 2 MeV, and so the need to reduce the other features becomes a priority.

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Slide 96 ZnS crystals

• Only used to measure alpha radiation• Crystals are sensitive to shock• Efficiency is typically low

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Slide 97 Some alpha detectors

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Slide 98 Liquid scintillation counters

• Used to count β emitters• Sensitive to very wide range of beta

energies• Can be self-correcting for counting

efficiencies (give DPM readout)• Can be tricked by static electricity and

chemical luminescence• Can be expensive ($40 K or more)

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Slide 99 Other detectors

• Neutron detectors (fast and thermal neutrons)

• HPGe gamma spectroscopy units• Alpha spectroscopy (multiple types)• Radon detectors• High-pressure pressurized ion chambers

(up to 25 atmospheres)• Whole-body counters• More….

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Slide 100 Counting geometry

• Proportional counters can give nearly 4 π geometry

• “4 π” means that the sample is completely surrounded by the counting medium– Comes from the surface area of a

sphere, which is equal to 4 π steradians• Counting on a flat surface is 2 π

geometry

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Slide 101 When selecting instruments for response to an incident:

• Choose appropriate detectors for expected types of radiation

• Take at least 1 backup instrument• If situation is unknown, take GM, NaI, and

ion chamber or micro-R meter

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Slide 102

Statistics

Not very exciting, but important anyhow

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Slide 103 Basic terms

• Mean• Mode• Median• Standard Deviation• Confidence Levels

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Slide 104 Mean, median, and mode

• Mean - the arithmetic average of a collection of measurements.

• Mode - the most likely value in a set of measurements.

• Median - the center value in a set of measurements.

• In a “normal” distribution, these are the same value

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Slide 105 Gaussian (“normal”) distribution

• Centered around the mean

• Area beneath curve is equal to 1

• Asymptotically approaches X-axis

• The goal of statistics (in D&D, for example) is to decide if a given high count-rate point is actual contamination, or just a high outlier in the background distribution

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Slide 106 Some more terminology

• Accuracy – how close is the data to the “true” value?

• Precision – how closely clustered are individual measurements?

• Uncertainty – how much do the data vary from the “true” value.

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Slide 107 Accuracy versus precision

• Accuracy

• Precision

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Slide 108 Sampling a population

• Statistical methods attempt to infer information about an entire population from a sample of that population

• Population - the entire group of possible measurements - population parameters are the true values.

• Sample - a small group of the population -sample parameters are the estimate of the true value made from the sample results.

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Slide 109 Standard deviation

• Standard Deviation – a measure of the variability in the data.

• Mathematically noted as multiples of σ– 68% of members of a population fall within

the 1 σ error bars– 95% fall within the 2 σ error bars– 99% fall within the 3 σ error bars– So, if a paper reports results to the 95%

confidence interval, they are reporting results to the 2 σ level

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Slide 110 Case study #1

• In a remediation plan, one licensee claimed a pile of soil was well-characterized and could be released for disposal– The release limit was 35 pCi/g– Ave. contamination levels were 33 pCi/g– Standard deviation was 34.5 pCi/g

• Should the licensee’s claim be believed?

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Slide 111 Case study #2

• A licensee’s I-131 emissions suddenly skyrocketed, leading to cessation of I-131 use for several months. However, 3 months after use stopped, I-131 was still counted in the sample filters.

• The reason for this is that, with 1-minute sample and background counts, random fluctuations in background showed up as “real” counts and, subsequently, as discharges

• Solution was to increase counting times and to use a single-channel analyzer to reduce background counts in energy regions of interest

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Slide 112 Case study #2 data

trial counts trial counts count rate1 407 1 4365 4372 432 2 4492 4493 407 3 4449 4454 414 4 4581 4585 428 5 4362 4366 431 6 4417 4427 469 7 4370 4378 448 8 4263 4269 423 9 4332 43310 452 10 4364 43611 426

Descriptive statistics Descriptive statistics

Mean 430.6364 Mean 439.95Median 428 Median 436.75Mode 407 Mode #N/AStandard Deviation 19.26797 Standard Deviation 8.954856Sample Variance 371.2545 Sample Variance 80.18944Confidence Level(95.0%) 12.9444 Confidence Level(95.0%) 6.405923

I minute background 10 minute background

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Slide 113 Control charts

• Used to identify instrument trends, even before an instrument goes “out of specification”

• Obtain a number of sample counts for a source and background.

• Determine the mean and standard deviation of both sets of data.

• Plot daily or weekly checks against mean and pre-determined error bars and examine for trends

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Slide 114 Expected response – variations around the mean

Alpha Background Control Chart

-0.10

-0.05

0.00

0.05

0.10

0.15

0.20

0.25

0.30

6/8/2001 6/13/2001 6/18/2001 6/23/2001 6/28/2001 7/3/2001 7/8/2001 7/13/2001 7/18/2001

Date

Alp

ha B

ackg

roun

d C

ount

Rat

e (c

pm)

Alpha Background-2 Sigma+2 SigmaAverage

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Slide 115 Possible problem – dropping trend in beta readings

Beta Source Response Control Chart

3500

3600

3700

3800

3900

4000

4100

4200

4300

4400

4500

6/8/2001 6/13/2001 6/18/2001 6/23/2001 6/28/2001 7/3/2001 7/8/2001 7/13/2001 7/18/2001

Date

Bet

a S

ourc

e Re

spon

se (c

pm)

Beta Response+10%-10%Average

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Slide 116

Dosimetry

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Slide 117 Types of dosimetry

• Film badge• Thermo-luminescent dosimeters (TLD)• Optically-stimulated dosimeters (OSL)• Neutron badges (various types)• Electronic dosimeters• Self-reading dosimeters (usually not used

as dosimeter-of-record)

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Slide 118 Film badge

• Radiation exposes film, so higher dose gives darker film

• Inexpensive, simple, cheap• Oldest dosimetry technology• Can be read multiple times• However, can be sensitive to environment• Lowest reading about 10 mrem

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Slide 119 TLD

• Radiation exposure causes electrons to jump to higher stable energy level

• Heating crystal releases e-, photon emitted• Rugged, accurate, mature technology• Sensitive to wide range of energies• Crystals can crack, can only be read once• Lowest reading about 10 mrem

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Slide 120 OSL

• Dose read by scanning with laser beam instead of heating

• Rugged, sensitive, can be re-read • More expensive than other badges• Sensitivity and performance sometimes

erratic (personal observation)• Only offered by Landauer

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Slide 121 Neutron

• Variety of types• Most are only sensitive to high OR low-

energy neutrons, not both• Need to know neutron energies in order to

use correct badge

• Neutrons are hard to measure!

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Slide 122 Electronic dosimeters

• Many designs• Can be used in specific situations (e.g.

entry into high radiation areas)• Can also be used (sometimes) as

dosimeter-of-record• Can have interference from outside

electrical signals• Usually rugged and reliable• Continuing to evolve at a rapid rate

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Slide 123 Pocket ion chambers (also called self-reading, or “pen”

dosimeters)

• Usually used only for entry into high radiation areas

• Usually used only for worker information• Some recent models can be used as

dosimeter-of-record

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Slide 124 Dosimeter regulations

• Required for anyone expected to receive more than 125 mrem in a calendar quarter

• Required for anyone entering a radiation area or high radiation area

• In some cases, multiple dosimetry is required– Ring badges for people handling isotopes– Arm, collar, fetal, and other specialty badges

as appropriate

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Slide 125 Importance of dose records

• The single most important document to protect your company in the event of a radiation injury lawsuit is a set of dosimetry records showing that the worker did not exceed radiation exposure limits while working for your company

• Dose records must be retained for 30 years after an employee leaves

• May be asked to supply dose records to new employers for a departing employee

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Slide 126

Air sampling

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Slide 127 Why perform air sampling?

• Measure environmental releases• Measure breathing zone airborne nuclide

concentrations• Measure results of accidental release• Routine environmental monitoring (inside

and outside facility)

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Slide 128 Typical setup

• Sample point• Sample lines• Filter• Pump

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Slide 129 Sample point

• Should be towards the center of the airflow, in the air to be sampled– i.e. downstream of exhaust filter for

environmental sampling• Should be in a straight section of duct

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Slide 130 Sample lines

• Should be as straight as possible and bends should be large radius

• Should be as short as possible• Should slope downhill from sample

point to filter if possible• Should be material that will not react

chemically with isotopes• Should be smooth and warm enough

to keep volatile nuclides from condensing

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Slide 131 Filter

• Should be suitable to capture isotopes of interest

• Must be able to collect for entire sampling period (e.g. one week) without saturating or releasing materials

• Should be able to be removed and counted easily and efficiently

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Slide 132 Sample pump

• Must be reliable – sample times may be as much as several months

• Must draw enough air to maintain isokinetic sampling conditions

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Slide 133 Isokinetic sampling

• Ideally, the flow velocity in the sample tube should be equal to the flow velocity in the exhaust stream

• Different flow rates can lead to over- or under-representing various particle sizes– Ex: faster flow rate in sample tube can skew

results towards larger particles

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Slide 134 Isokinetic sampling equations

2

FlowrateVelocityrπ

=

2

2stack stack

isosample

FR rFlowrater

×=

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Slide 135

A little on bioassays

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Slide 136 Bioassay programs

• Required to perform bioassay whenever there is an uptake, a potential uptake, or when isotope use meets certain criteria

• These criteria are specified in NUREG 8.9 and NUREG 4884 and ANSI standards

• Examples – your policy may call for bioassay following use of 1 mCi of I-131 or I-125, 40 mCi of H-3, or 20 mCi of C-14 or S-35

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Slide 137 Uses for bioassay

• Confirm (or reject) uptake of radioactive materials

• Quantify uptake for subsequent internal dosimetry

• Track progress of elimination either through natural means or following decorporation therapy

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Slide 138 Types of bioassay

• In vivo – bioassay performed on the person (e.g. whole-body counting, thyroid bioassay)

• In vitro – bioassay performed on sample removed from the person (e.g. urine, fecal, blood samples)

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Slide 139 In vitro bioassays

• Typically urine sample, although other media can be used

• You will normally collect the sample and then pipette 1 ml into a liquid scintillation vial (or other appropriate medium) for counting

• May use fluorometry for gross U, gamma counting for Cs-137, LSC for beta-emitters

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Slide 140 Quantifying results

• For urine bioassay, it may be possible to relate the amount of isotope in the urine to the amount in the whole body– For example, tritium, C-14, and Cs-137

typically distribute through the entire body• However, it may be necessary to

understand the biokinetics of the isotope to be able to relate bioassay amounts to uptake

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Slide 141 One example – Am-241

6.0E+002.0E+005.4E-031.5E-031006.1E+002.2E+006.6E-031.6E-03906.2E+002.4E+008.2E-031.7E-03806.3E+002.6E+001.0E-021.8E-03706.4E+002.8E+001.3E-021.9E-03606.6E+003.1E+001.7E-022.0E-03506.8E+003.4E+002.1E-022.3E-03407.1E+003.8E+002.8E-022.6E-03307.4E+004.3E+003.7E-023.3E-03207.6E+004.6E+004.2E-023.9E-03157.9E+005.0E+005.7E-024.9E-03108.2E+005.2E+002.3E-015.8E-0379.1E+005.3E+001.3E+007.2E-0351.5E+015.5E+008.0E+001.3E-0232.6E+015.6E+001.5E+012.3E-0225.0E+015.8E+001.1E+011.8E-011

Retained in bodyRetained in lungs24-h fecal excretion24-h urine excretionDay after intake

Table Am-3. Reference values for retention and excretion of 241Am (% of intake) as a function of time after acute inhalation ofmoderately soluble form (Type M) by an adult; particle size = 5 μm (AMAD)

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Slide 142 More on Am-241

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Slide 143 In vivo bioassay

• Most common in a non-reactor setting is a thyroid count

• May also perform whole-body counting for gamma emitters

• If isotope is unknown, or if there are multiple isotopes, may need to also perform gamma spectroscopy to ID isotopes and to permit quantification

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Slide 144 Interpreting in vivo bioassay

• Will first need some sort of efficiency measurements for your probe under the conditions of use

• For example, with a thyroid bioassay, will have to use a neck phantom and an NIST-traceable source to determine counting efficiency

• May need other phantoms for other parts of body (e.g. lungs, stomach, etc.)

• Once you know your counting efficiency, it is fairly easy to determine the body (or organ) burden) from your count

• This can then be used to determine amount of uptake, using your understanding of the biokinetics

• From there, may be able to use information in FGR 11 or other sources to determine body and organ dose

• I prefer to do the calculations once, then set up a spreadsheet

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Slide 145

Professional resources

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Slide 146 Professional organizations

• Health Physics Society (www.hps.org)– May consider joining HPS RSO Section – need not be

HPS member• American Academy of Health Physics

(www.hps1.org/aahp) • Nuclear Regulatory Commission (www.nrc.gov)• State regulators (usually Health Department,

although varies by state)• American Association of Physicists in Medicine

(www.aapm.org) • Each of these web pages has many links to

information or to other organizations

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Slide 147 References

• Health Physics and Radiological Health Handbook

• EPA Federal Guidance Document #11 (dose from uptake of radioactivity)

• NCRP reports (www.ncrp.com)• ICRP reports (www.icrp.org)• UNSCEAR reports (published by UN)• BEIR reports (published by NAS)• IAEA techdocs (www.iaea.or.at)

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Slide 148 Other agencies and resources

• State regulatory agencies usually have their own web pages

• DOE (www.doe.gov)• EPA (www.epa.gov)• DOT (www.dot.gov) • FDA (www.fda.gov) • ORISE HP resource CD-ROM

(www.orau.gov/orise.htm)

• HPS has an “Ask the Expert” feature on the HPS web page

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Slide 149

Regulations

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Slide 150 Regulatory agencies

• NRC/state agencies regulate most uses of radioactivity (10 CFR 20 and others)

• DOT regulates transportation of radioactive materials (49 CFR 171-173)

• EPA regulates some NORM and some environmental discharges or contamination

• States regulate most NORM/TENORM and radiation-generating machinery

• FDA regulates medical devices that emit radiation

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Slide 151 Regulations cover

• Dose to workers (10 CFR 20.1201)• Training requirements (10 CFR 19.12)• Badging requirements (10 CFR 20.1502)• Records (10 CFR 20.2101)• RSO qualifications (NUREG 1556)• RAM transportation (49 CFR 171-173)• Reports to regulators (10 CFR 20.2201)• Posting and labeling areas (10 CFR 20.1902)• Radioactive waste disposal (10 CFR 20.2001)

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Slide 152 To be an RSO…

• Training and experience requirements will vary according to size and complexity of program

• Small programs may have part-time RSO as a collateral duty

• Large programs will need a full-time RSO• RSO is responsible for entire program• RSO must have authority from licensee to

carry out duties properly

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Slide 153 Controlling radiological areas

• Some areas require administrative and/or engineering controls

• One form of administrative controls is posting a room or area as a radiologically controlled area

• Other areas require audible and visual alarms and other controls

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Slide 154 Posting requirements

• Radiation area – can exceed 5 mrem in 1 hr• High radiation area – can exceed 100 mrem in 1 hr• Contamination area – more than 1000 dpm/100

cm2 of removable contamination (up to 15,000 dpmtotal)

• Radioactive materials storage area – contains more than 10 times exemption limits of isotopes (vary by isotope)– “Sum of fractions” rule applies

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Slide 155 Other dose control features

• HRAs must be locked and have entry controls• Irradiator rooms must have audible and visual

alarms• Doors must be interlocked to retract source if

opened• Must have visual confirmation room is empty

prior to exposing source or energizing machine

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Slide 156 Reports to regulators

• Loss of radioactive materials• Personnel exceeding dose limits• Some instances of skin contamination or

uptake of radioactive materials• Other incidents or emergencies involving

radioactive materials (for example, a fire in a RAM storage area)

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Slide 157 Transportation

• RAM must be properly packaged according to isotope(s) and activity present

• Package and vehicle external rad levels determine mode of shipment, vehicle placarding, and other requirements

• Package rad levels and amount of activity present determine labeling on package

• Must attend DOT training every 3 years

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Slide 158 More transportation stuff

• May need shipping papers and/or manifest that must be easily accessible by driver

• All RAM must be properly blocked and braced in vehicle

• Vehicle should be exclusive-use if possible (easier with company vehicle)

• RAM must be secured against unauthorized removal at all time

• Vehicle may require placarding, depending on activity and radiation levels

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Slide 159 Receiving radioactive packages

• Covered in 10 CFR 20.1906• Must have procedure for receiving rad

packages• Must survey (radiation and possibly

contamination) within 3 hours of receipt• Must inspect package for damage• Notify carrier (e.g. FedEx) if package is

damaged or contaminated)

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Slide 160 Dose limits

• 5 rem/yr – Radiation workers (WB)• 0.1 rem/yr – everyone else• 50 rem/yr – any internal organ• 15 rem/yr – lens of eye• 50 rem/yr – skin and extremities• Can exceed dose limits for Planned

Special Exposure – several requirements in 10 CFR 20.1206

• Minors are limited to 10% of adult doses

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Slide 161 Pregnant workers

• Pregnancy must be declared voluntarily and in writing to the RSO

• Worker must be given fetal badge (unless she works with radiation that won’t be detected)

• Fetal dose limits are 500 mrem during the entire pregnancy, 50 mrem in any month

• Worker may un-declare pregnancy at any time

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Slide 162 ALARA

• As Low As Reasonably Achievable• Should take all reasonable measures to

reduce radiation dose as much as possible• Need not take unreasonable measures• All employees must be trained in ALARA• May need to put together an ALARA plan

for your company, depending on type of work performed

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Slide 163 Training requirements

• All radiation workers require initial training before working with radiation or radioactivity

• All radiation workers must receive refresher training annually

• Ancillary workers may require training, depending on work duties– Maintenance, housekeeping, etc.

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Slide 164 Records to keep• Training records• Dose and exposure reports• Radiation surveys• Instrument calibration sheets• Maintenance and repair for all radiation

equipment• Sealed source inventory and leak tests• Internal audit and external inspection reports• RAM inventories• Shipping documents• Waste disposal• Incident reports

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Slide 165 Training is important!

• Make sure everyone receives the training required by regulations

• Keep training records for all attendees• Ensure content includes regulatory

requirements• Go over ALARA, pregnant worker

program, dose limits, and site-specific information

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Slide 166 Radioactive waste

• Short-lived waste can be stored for decay (usually <90 day half-life)

• Long-lived solid waste must be shipped for disposal to a licensed facility

• Liquid wastes can sometimes be disposed of into sanitary sewer system

• Some wastes (animal carcasses and liquid scintillation vials with H-3 and C-14) can be treated as non-radioactive – called “de minimis”

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Slide 167 Waste storage area

• Must be capable of safely storing waste through DIS period or until shipping

• Must be dry and secure• Should have sprinkler system in case of fire• Should be inspected periodically to ensure

waste package integrity• Should also survey periodically for radiation

and contamination levels

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Slide 168 Sewer disposal

• Only for authorized liquids (nothing hazardous)

• Must calculate average activity concentrations, based on flow of water through sewer at facility

• Average activity concentrations must be less than those in CFRs or state regulations for each isotope

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Slide 169

ALI and DAC calculations

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Slide 170 Airborne radioactivity

• ALI – allowable limit for intake – the amount of radioactivity that will give a person 5 rem WB or 50 rem to an organ

• DAC – derived air concentration –radioactivity concentration that will give an unprotected worker 1 ALI if breathed for 2000 hours (listed in FGR 11)

• Can use ALI to determine dose from uptake of radioactivity

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Slide 171 DAC example

• A worker who works in an atmosphere of 20 DAC for 10 hours will receive 200 DAC-hrs of exposure

• Workers are allowed 2000 DAC-hrs in a year, which gives 5 rem WB or 50 rem organ dose

• So 200 DAC-hrs will give a worker 10% of allowable dose for the year

• OR working 2000 hrs in 1 DAC atmosphere while wearing a respirator with a protection factor of 10 will also give 200 DAC-hrs of exposure

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Slide 172 ALI example

• Ingesting 1 ALI will give a worker 5 rem WB or 50 rem organ dose

• So a worker who ingests 20% of the ALI for a particular isotope will receive 10% of their allowable dose for the year

• ALI and DAC values are given in FGR 11 (an EPA document)

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Slide 173 ALI example

• A worker accidentally ingests 10 μCi of Cs-137 in powdered form

• ALI for Cs-137 is 40 μCi with no target organ

• Worker ingestion is 0.25 ALI• Worker receives dose of 1.25 rem (25% of

5 rem WB limit)

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Slide 174

Common tasks

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Slide 175 Surveys

• Periodic radiation and/or contamination surveys– Daily, weekly, or monthly, depending on

operating circumstances– Frequency depends on level of use

• “Special” surveys– Post-work surveys if handling isotopes– Post-maintenance– After any significant change or work that could

affect shielding or other characteristics

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Slide 176 Pre-survey checks

• Confirm proper detector for survey• Verify meter calibrated within last year• Verify physical integrity of meter and cable• Perform battery check• Perform response check against source of

known strength (+/-20% of ave. counts)• Verify switch positions if appropriate

(audible “on” and response “f” or “fast”

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Slide 177 Contamination surveys

• Perform pre-survey meter checks• Hold detector no more than ½ inch from

surface to be surveyed• Move detector at no more than 2”/second• Survey 100% of surface if possible• Watch probe to ensure proper survey while

listening to detector – if you hear an increase in count rate, see if area is contaminated

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Slide 178 Contamination surveys (con’t)

• Record any area in which count rate increases appreciably (e.g. 50 cpm for a GM, 100 – 200 for thin-xtal NaI)

• Must record net cpm and dpm on survey record– Cpm – bkg = net cpm– Cpm/meter efficiency = dpm

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Slide 179 Example

• Say background levels are 50 cpm with a GM probe

• At a certain location, you read 130 cpm• 130-50 = 80 net cpm• If meter efficiency is 40%:

– dpm = 80 cpm/0.40 = 200 dpm

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Slide 180 Radiation dose rate surveys

• Use appropriate meter– Ion chamber, micro-R meter, etc.

• For general area surveys, hold meter at waist level (about 1 m above ground) and walk slowly through area being surveyed

• “General area” is 30 cm (1 foot) from any surface

• “On contact” used to measure hot spots –usually taken to be 1” from surface or point surveyed

• Record highest readings as well as area readings

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Slide 181 Some trouble-shooting: common problems

• Elevated counts, no contamination• Spurious high and/or low counts or erratic

readings• No readings (including background)

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Slide 182 Elevated count rate – possible causes

• Light leak (scintillation probes only)– Cover probe, then uncover; compare counts– Light leak indicated if counts ↑ when probe exposed

to light• Contaminated probe

– Move probe a meter or so from survey location– If high counts remain, probe may be contaminated

• High radiation levels– Hold probe near body– If readings drop, high counts may be due to high

radiation levels

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Slide 183 Erratic readings

• Most likely cause is a short or open circuit in the cable (very common)

• If high or low counts can be reproduced by jiggling cable or holding probe in a certain position relative to meter, you should suspect a cable problem

• Use another meter to survey location OR change cable on problematic meter

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Slide 184 No readings

• Most likely cause is dead batteries– Perform battery check– Replace batteries

• Also possible that detector is broken or disconnected– If batteries are OK, use check source to

test meter response– Examine detector and detector window

for breakage, replace if necessary

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Slide 185 If all else fails….

• Change cable and probe OR• Send meter for re-calibration• Send meter for repair• Buy a new meter

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Slide 186 Smear wipe surveys

• May have to perform smear wipe survey for isotopes with low meter efficiency (such as H-3, C-14, S-35)

• Smear wipes are dry filter paper, wiped over 100 cm2

• Want to avoid cross-contamination of wipes• Want to avoid cross-contaminating surfaces• Usually counted in proportional counter,

gamma counter, or scintillation counter

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Slide 187 Sealed source checks

• Required to inventory sealed sources either quarterly or semi-annually

• Required to leak test sealed sources with same periodicity– Some sources do not require leak testing –

check your license or regulations• Performing a leak test does not

necessarily count as an inventory check –must record both events separately

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Slide 188 Importance of surveys

• If you do not have dosimetry records, your survey records are the only means you have to combat unjustified radiation injury claims

• For this reason, properly performing complete surveys and retaining survey records is crucial

• Equally important is ensuring your survey equipment is properly maintained and calibrated

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Slide 189 Leak testing

• Must report significant source leakage (check your regulations for amount)

• Must be able to show that test methodology will detect at or below contamination limit

• Several acceptable methods of leak testing and inventory

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Slide 190 Leak test methods

• Best to directly wipe 100% of source capsule– May wish to consider holding source and wipe

with tongs to reduce finger dose• Can also wipe area around source if

source is not directly accessible• Can also wipe outside of source holder

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Slide 191 Inventory methods

• Best to visually sight the source• Can also verify source in sealed container

that has been sealed since last inventory• Can also verify operability of device that

relies on presence of source• Other methods may be acceptable –

check with the source/device manufacturer and regulators

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Slide 192

Radiation safety program management

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Slide 193 Priorities• Keep people safe and healthy• Violate as few regulations as possible• Provide HP support services to users• Remember that your priorities are not

necessarily those of your management or users– Authorized users are most concerned about their

research– A university is most concerned about grant

dollars, publications, and (hopefully) education

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Slide 194 My primary goals as RSO

• Make sure that we keep our license• Keep our program out of the news• Make sure our physicians and researchers

are relatively unaware of the existence of Radiation Safety

• Unfortunately, our job is like people who do lighting in theater – nobody knows we’re there unless we screw up

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Slide 195 Policies and procedures• Policies give the big picture as to how radiation

and radioactivity are used at our facility– Ex: pregnant worker policy

• Procedures are the little picture items that tell us how to perform each task to comply with the policies– Ex: where a fetal dosimeter is worn

• Both are necessary, and both should be reviewed and approved by management and/or RSC

• Both should be in writing and easily available to all radiation workers and authorized users as the RSM and SOPs

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Slide 196 Radiation Safety’s role….

• Propose (and draft) new policies and procedures as necessary

• Enforce compliance with policies and procedures via inspection

• Determine appropriate corrective and/or disciplinary actions in cases of non-compliance

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Slide 197 Purpose of license

• Gives the licensee permission to use radioactive materials

• Sets conditions for that use• May list specific users and/or specific

sources– For example – irradiator sources will probably

be listed by serial number and activity

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Slide 198 Incidents and emergencies

• Need to train workers in how to respond to unexpected events– Also need to train Security

• Should maintain records of all radiological incidents

• May need to notify regulators, depending on reporting requirements– I will often inform regulators of incidents even if a

formal report is not required – helps build good working relationship, and regulators may be able to suggest actions I haven’t thought of

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Slide 199 Incidents and emergencies (II)

• Encourage reporting incidents to RS– We don’t penalize honest mistakes that lead to spills,

skin contamination, etc.– Should have policy as to what level of incident your

rad workers can address on their own (e.g. “major”vs. “minor” spills)

• Train people in immediate actions to limit incident (i.e. “SWIM” for spills)

• Should investigate serious or potentially serious incidents to find out what happened, why, and what can be learned

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Slide 200 Working with the RSC• RSC has three primary functions

– Oversee activities of RS and RSO– Support RSO with advice in areas beyond

RSO’s experience– Determine need for disciplinary or corrective

actions if necessary• RSO should keep RSC informed of

incidents as they occur• RSC can be a good sounding board for

policy changes, SOPs, etc.

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Slide 201 Working with management• Need management support to have an effective radiation

safety program– Controversial or unpopular policies are more palatable if

supported by management– May need management to support disciplinary actions

• So management needs to be kept in the loop!• It’s best to consult with management early, especially if

decisions involve added cost, or researchers with lots of funding

• Should be able to explain to management the reasons for any controversial actions– Regulatory requirements that were violated (or need to be met)– Possible litigation issues– JCAHO (if a medical center)

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Slide 202 Working with regulators• Two approaches to working with

regulators – adversarial or cooperative• Usually best to try to work with regulators• Can sometimes get useful support from

regulators if nothing else is working– Need for extra staff– Needed policy changes– Proper role of RS or RSU

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Slide 203 Cultural differences: Management

• Being cynical, I assume that management only wants things to run smoothly at low cost with no lawsuits or bad press

• This means that you will have to justify policy changes, extra expenses (equipment, staff...) in terms of management’s concerns

• Similarly, if enforcement actions will affect a highly funded researcher or a medical department, you should be prepared to justify them in terms management will relate to (JCAHO, fines, litigation prevention, etc.)

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Slide 204 My general approach to being RSO• Assume people want to do things correctly, but

will sometimes make mistakes• If disciplinary actions are required, they should

be appropriate• Keep people informed as much as possible• Work with people in person instead of via phone,

e-mail, or letters– Get out and about to visit labs and meet PIs – Do some of the work (inspections, training, etc.)

instead of just attending meetings and writing policies and procedures

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Slide 205

Inspecting in your facility

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Slide 206 Laboratory inspections – a general approach

• We are not the radiation Gestapo - we support research and medical care

• Inspections are necessary to ensure compliance with regulations, policies, and good practices

• We make every effort to make our inspections objective and fair

• Minor violations may be corrected on the spot without being cited

• Everything must be documented

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Slide 207 U of R inspection stats

• Compliance rate is about 90%• About 75% of violations we find are minor

and are not cited• We conduct about 650 inspections

annually and find about 100 violations among all permit holders

• Most cited and uncited violations are directly or indirectly due to failing to perform surveys

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Slide 208 Inspection cycle• “Good” labs (those with no violations in

past 12 months) are inspected semi-annually

• Everyone else (including new permit holders) are inspected quarterly

• Semi-annual inspections are more thorough and include a formal inventory verification

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Slide 209 Old system• RS was responsible for monthly contamination

surveys, which doubled as “inspection”• Due to under-staffing, other duties and large

number of labs, most monthly surveys were missed

• When inspections were performed, they were cursory at best

• An external audit found violations in 70-80% of labs – mostly failure to perform needed surveys and eating in posted areas

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Slide 210 Changes we made• Labs became responsible for all monthly

surveys• Began quarterly inspection cycle• Developed inspection checklist and gave copies

to all permit holders• Began intensive training on radiation safety

policies during initial and refresher training, along with scheduling meetings with all labs to explain policy changes

• Gradually ratcheted down on labs over a year to get them used to the new system and expected performance standards

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Slide 211 Results of policy changes• Increased staff from 4 to 6 tech staff• Compliance changed from 20%→90%• No violations in last 3 regulatory inspections• Freed up enough staff time to clean up

legacy waste, close out 3 un-needed waste rooms, re-write RSM, and make other needed changes

• More thorough inspections also gives more time to talk with rad workers during visits –improves relations

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Slide 212 UR’s “Chief Inspector”• One person performs most of our inspections• Our inspector is very professional, calm, thorough,

fair, and can’t be talked into looking the other way– Former Air Force Master Sergeant

• Users respect our inspector’s detailed knowledge of radiation safety and his willingness to take time to work with them during inspections

• Having the same person do virtually all inspections is a plus because of the heightened familiarity with rad workers

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Slide 213 Inspection preparation• Pull file on permit holder

– Review past inspection reports– What isotopes do they work with?– Are there any permit conditions or

restrictions?– How active and how large is their program?

• Assemble and check equipment– Appropriate meter for survey(s)– Smear wipes (if appropriate)

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Slide 214 During inspection• Review paperwork• Observe work practices• Quiz rad workers on basic “level of

knowledge” questions• Perform confirmatory contam. survey• Perform radiation level survey

– Use this to confirm dosimetry not needed, as long as rad levels are less than 0.2 mr/hr in an accessible area

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Slide 215 Paperwork review

• Inventory verification– Waste container inventory sheets– Stock vial record sheets

• Monthly survey review• Check for training and refresher training

certificates• Review dosimetry reports (if badges

issued)

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Slide 216 Work practices

• Are workers properly garbed (lab coats, gloves, no bare legs or feet)?

• Are surveys performed properly?• Check for eating, drinking, or food storage

in posted areas• Do workers survey hands frequently

during work with isotope?• Look at radiological and general

housekeeping

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Slide 217 What every radiation worker needs to know:

• Laboratory isotope use information– Isotope(s), half-lives, survey methods

• Dose limits– Rad workers, pregnant workers, public

• Work practices– Survey technique, bremsstrahlung, PPE

• General information– Regulatory agency, who is RSO, etc.

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Slide 218 Surveys

• Inspection contamination surveys are confirmatory in nature– They do not substitute for the lab’s required

surveys• Radiation surveys are performed only by

RSU– Used to justify not badging most of our

research rad workers– Rad levels can be used as proxy for personal

dosimeters

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Slide 219 Violations• Two categories – major and minor• Minor violations (meter out-of-cal, minor

paperwork problems, etc.) can often be corrected on the spot

• Major violations (inappropriate disposal of radwaste, failure to survey, etc.) are cited in report (with concurrence of RSO)

• Permit holders can appeal violations to RSC• Need to reply in writing in two weeks with

corrective actions taken

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Slide 220 Enforcement actions

• Any researcher who receives more than 4 major violations in any continuous six-month period is subject to disciplinary or corrective actions

• Actions are recommended by RSO and approved by RSC

• Failure to correct problems or to reply in a timely manner is considered a major violation

• We warn permit holders when they reach 3 violations or greater

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Slide 221 What corrective or disciplinary actions are available?

• Mandatory refresher training for lab• Special meeting of lab staff with RSO• 1-on-1 training with individual rad workers• Temporary suspension of a worker• Temporary permit suspension• Temporary suspension of ability to order

isotope• Permanent suspension of individual

worker or of permit

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Slide 222 After the inspection - RS

• We maintain records of compliance rate and number of violations

• Violations are tracked by area– Most common violations are failure to survey,

contamination in lab, meters out of calibration, and radioactive material security

• Summary statistics are presented to RSC annually

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Slide 223 After the inspection – rad worker

• Inspection report sent within 1 week of inspection

• Worker has two weeks to reply to any violations with corrective actions or to contest violation

• May have follow-up inspection, depending on number of violations

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Slide 224 Summary

• Managing a radiation safety program is not simple, but it need not be all-consuming

• Important to be able to understand and work with a wide variety of people, each with their own priorities

• Inspections are a vital part of any good RS program

• Inspections need not be adversarial

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Slide 225

Regulatory Inspections

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Slide 226 Regulatory inspections

• Most licensees will be inspected annually• Inspections may last from a half day to several

days• 1 or more inspectors may show up at your door• Inspectors have the right to see any part of your

radiation safety program and to speak with any worker

• You have the right to accompany the inspectors during their visit

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Slide 227 What they’ll look at

• Training records• Dosimetry records• Survey records• Inventory records• Shipping papers (if appropriate)• Waste records (if appropriate)• Work practices• Meter calibration and maintenance records• Equipment service records (if appropriate)

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Slide 228 Other things that might happen

• Interviews with management to confirm support for Radiation Safety

• Interviews with non-rad workers• General housekeeping inspection• Discussion of overall program with RSO• Review of Radiation Safety Committee

records• Entrance and exit interviews

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Slide 229 If you get a violation…

• “Apparent violations” noted during exit interview – must be confirmed by formal inspection report

• Try to correct violation during inspection and notify inspectors of corrective actions

• Take all violations seriously – they are serious to the inspectors

• You may contest any violations, but should limit this to those for which you have good documentation

• You must reply in writing, accepting violation and noting corrective actions taken to resolve

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Slide 230

About your license….

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Slide 231 Licensing

• Whether writing a new license, a license amendment, or a license renewal, try to stay as simple, as general, and as “clean”as possible

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Slide 232 Simplicity

• Try to use the simplest language, procedures, and the simplest approach possible when addressing specific issues in the license application

• If you inherit a needlessly complex license, consider the degree to which you can simplify it during renewal

• For example, it may be better to put everyone on monthly (or quarterly) doismeter reads rather than having quarterly for some, 6-month for others, and monthly for the rest

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Slide 233 Generality

• It’s better to be generally right than specifically wrong

• Try not to tie yourself down with needless specificity when a broader generality will accomplish the same thing

• For example – say that you will package biological wastes in accordance with waste acceptance criteria instead of promising to pack in lime, freeze, double-bag, and surround with absorbant materials

• Or say that you will have alpha, beta, gamma, and neutron survey meters on-hand; but don’t list your current instrument inventory

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Slide 234 A “clean” license

• Try to minimize attachments and referenced materials – the license should be as “stand-alone” as possible

• For example, avoid attaching your site-specific procedures, radiation safety manual, training book, etc. to your license– Whatever materials you attach to the license

become part of your application, and you can’t change them without regulatory approval

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Slide 235

Responding to radiological incidents and emergencies

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Slide 236 Spills

• Stop spill• Warn others• Isolate the area• Minimize exposure• Stop ventilation if appropriate and possible

• Start to clean up after immediate actions completed

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Slide 237 Stop spill

• Pick up container (if possible) and place into bucket or deep tray

• Place absorbent materials over spilled liquid (or damp paper towels or rags over spilled powder)

• Idea is to stop adding more material to spill and to limit contaminated area

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Slide 238 Warn others

• Call RSO or others who can assist• Inform others nearby

• Idea is to keep unprepared people out of spill area and to get help in cleaning spill

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Slide 239 Isolate area

• Put up physical boundary around spill• Spill boundary should be at least 1-2 meters

from farthest splash if possible• May simply close and lock door to room• ALWAYS put sign on spill boundary• Nobody should enter spill area unless dressed in

proper PPE• Nobody should leave spill area until surveyed by

RSO or trained rad worker (including personnel in area when spill occurred)

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Slide 240 Minimize exposure

• Take a short time to think through situation• Make sure you are taking proper and

reasonable actions• Make sure you know how to deal with

situation• Most radiological incidents are not life-

endangering – you have the luxury of taking a few minutes to make sure you’re doing the right thing

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Slide 241 Stop ventilation

• If spilled material is powdered or volatile, stopping ventilation can help to reduce spread of contamination

• Whether or not to try to stop ventilation is a judgment call on the part of the RSO or person at the scene

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Slide 242 Clean-up

• Work from outside of spill area towards the center

• Work from top to bottom (if appropriate)• Can usually use commercial cleaners

(Formula 409, Windex, Easy-Off, etc.)• Use Radiac Wash and similar products if

you have radioactive metals• Monitor area periodically to ensure

cleanup efforts are effective

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Slide 243 Skin contamination

• Contact RSO immediately• Get good count rate on contaminated area

and write it down for future reference• Begin washing with mild soap and cool to

warm water• Count area every few washes to monitor

clean-up progress• May need to notify state, calculate skin

dose, and/or monitor for internal exposure

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Slide 244 Skin contamination

• The three Cs– Contact the Health and Safety Officer– Count the contaminated area with a Geiger

counter and record the count rate– Clean the contaminated area

• In this case, rapid decontamination is important – should go to the nearest sink to clean up

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Slide 245 Skin decontamination

• Don’t do anything that will be painful or uncomfortable

• Use mild soap and cool to warm water• Get a good count rate every few washes and

record the information to make sure that decontamination efforts are effective

• Keep cleaning until Radiation Safety says it’s OK to stop, or until all elevated counts are gone

• Remember – some isotopes are absorbed through the skin, so dropping counts may also indicate absorption and not decontamination

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Slide 246 Injured personnel in radiological incidents or emergencies

• Must treat each on a case-by-case basis• Always make emergency responders or ER

personnel aware of radiological concerns• Even highly-contaminated patients pose very little

or no risk to emergency responders or to ER staff• Take care of most pressing problems first• Examples:

– Life-threatening injuries must be treated immediately– Mild injuries can sometimes wait for decontamination– Move injured personnel from dangerous radiation levels

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Slide 247 High radiation levels

• Get good dose reading in area and in other “populated” areas (offices, etc.) – evacuate area if necessary

• Try to determine source of radiation• Try to stop/shield source

– Machines can be turned off– Sources must be either retracted or covered

with shielding such as lead

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Slide 248

How to respond to a radiological emergency

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Slide 249 Priorities

• Stay alive and healthy• Take care of victims• Put out fires and other emergency

response work

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Slide 250 Protect yourself

• Wear proper PPE – including respiratory protection

• Wear your dosimetry (at least one person per team)

• Use radiation detectors (if available)• Use Universal Precautions when working

with contaminated victims• Limit stay time in dangerous areas

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Slide 251 Caring for victims• Even highly-contaminated patients pose

very little or no risk to emergency responders or to ER staff

• Take care of most pressing problems first• Examples:

– Life-threatening injuries must be treated immediately

– Mild injuries can sometimes wait for decontamination

– Move injured personnel from dangerous radiation levels

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Slide 252 Contamination control

• If time and victim’s condition permit:– Remove outer clothing– Wrap in blanket or sheet– Cover with “bunny suit”– Wipe down with damp sponge or rag– Shower or spray with water to remove

contamination• Take whatever actions you can without

placing the patient at greater risk

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Slide 253

Radioactive sources

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Slide 254 What sources look like

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Slide 255 Irradiators and radiography sources

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Slide 256

Radiological terrorism

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Slide 257

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Slide 258 Scenario 1: RDD attack

• A terrorist group sets off an RDD during some big event

• ~50 people are killed by blast• About 10,000 people are directly

contaminated• Plume drifts across downtown and

surrounding neighborhoods

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Slide 259 Some questions:

• What are the likely radiation effects on those in the area?

• On emergency responders?

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Slide 260 So – is it safe?

• 1000 Ci of Co-60 will give a dose rate of about 8 rem/hr

• Co-60 is not a huge inhalation hazard, and plume should settle fairly quickly

• Contamination can be controlled by wearing turnout gear

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Slide 261 What about risks to the public?

• Those at the scene may have heavy contamination and possible inhalation –they are at the greatest risk.

• However, many will probably have higher risk from injuries – and serious injuries must be treated first

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Slide 262 And the emergency responders?

• Heat and fatigue are probably going to keep an emergency responder from being in an area long enough to develop radiation sickness– It will take up to 12 hours in the area in this

scenario to start to get radiation sickness• So if responders wear their PPE, they

should not be at risk

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Slide 263 Scenario 2: Irradiator attack

• A terrorist group plops a 1000 Ci Co-60 irradiator in a movie theater (say for a Lord of the Rings re-run)

• During the movie, some people exit the theater vomiting, and others complain that there are sick and unconscious people in the theater

• An usher found a canister beneath a seat

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Slide 264 Same questions:

• What are the likely radiation effects on those in the area? On emergency responders?

• What actions should the public take?

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Slide 265 Radiation levels

• 1 meter from source, rad levels will be fatal in about 45 minutes

• 2 meters from source, rad levels will fatal in 3 hours

• 10 meters from source, rad levels will take about 10 hours to make you sick

• So you can work here – just wear your dosimeter

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Slide 266 Risk to the public

• A 1000 Ci source in a movie theater may kill up to a dozen or so people – assuming that everyone stays for the entire movie

• Up to 100 people may develop mild to severe radiation sickness over the next several weeks, again assuming they stay for the entire movie

• There is NO risk to anyone else because irradiated people do not become radioactive

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Slide 267 Risk to emergency responders

• There is NO risk to emergency responders, provided they heed dose limits

• Will probably have to use several teams to recover bodies of those nearest the source

• Recovering the source is potentially hazardous, but can be done safely by radiation safety professionals

• At worst, will have to close theater for a few days

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Slide 268 When to be worried

• If radiation dose rate is greater than 10 rem/hr

• If contamination levels are more than 100,000 cpm

• If your dosimeter reads more than 3 rem (if you are a radiation worker) or more than 75 mr (if you’re not a radiation worker)

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Slide 269 When to leave the area right away

• If radiation levels are more than 100 rem/hr or are off-scale

• If contamination levels are too high to register on the highest scale of the meter

• If you see health physicists looking scared

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Slide 270

Nuclear Terrorism

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Slide 271 How Likely is Nuclear Terror?

• Nobody believes any longer that “it takes a Manhattan Project” to build a nuclear device– Given 235U in sufficient quantity building a gun-

assembled device is quite straight-forward• Few-man portable; mass probably < 200 kg• Yield ~ 100 tons to 1 kT

– Given plutonium, it is possible to build simpler implosion devices than “Fat Man”

• Mass < 1000 kg• Yield ~ 1 kT to 10 kT

– Weapons can be (have been?) stolen.

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Slide 272 Primitive and Simple Nuclear Designs are Well Known and Generally Understood

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Slide 273 Design and Construction

• Gun-assembled devices– Extremely simple– Target bolted to gun; can’t miss– Builders will figure out the tricks along the way.

• Implosion devices– Won’t use “Fatman-style lenses”– Will be very unpredictable/unreliable– Yield could be as high as several kT– Very, very dirty from Pu contamination

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Slide 274 Nuclear devices

• Hiroshima– Weapon size about 15 kT (similar to IND)– Total inhabitants, 320,081– Deaths, 122,358– Injured, 79,130– Uninjured, 118,613

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Slide 275

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Slide 276 Comparison with terrorist device

• No way to predict yield – IND may be small, while stolen weapon may be higher

• Hiroshima was an airburst, while IND would likely be a surface burst– Surface burst will be “dirtier” due to inclusion

of surface materials in blast• Site of detonation will influence effects

– One likely scenario would be in port facilities, although small (suitcase) device, or device in cargo container may be taken almost anywhere

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Slide 277 Effects

• Blast• Radiation and fallout – can be fatal up to a

few miles away in the plume• Thermal• Firestorm – as in Dresden and Tokyo, may

be more destructive than initial blast

• All depend strongly on yield of device and location of detonation

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Slide 278 Making or getting nuclear weapons

• It’s thought that up to 40 former Soviet “suitcase nukes” are not accounted for

• There are a number of research reactors still fueled with HEU

• Pakistan offered nuclear technology to many nations, including N. Korea, Iraq, and Libya – may have sold to others, too

• May be able to purchase weapons, Pu, or HEU on nuclear black market

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Slide 279 The bottom line

• It is likely that terrorist organizations are trying to obtain nuclear weapons or fissionable materials

• There are many opportunities to obtain such weapons

• Nuclear weapons effects, even for a “fizzle” or low-yield device, would be horrific

• Nuclear weapons can be smuggled in cargo containers or suitcases – detection is possible, but only if we are looking for something

• However, it is unlikely that such devices are now in the possession of terrorists

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Slide 280 Resources

• Health Physics Society (www.hps.org)– Including Ask the Experts feature

• CDC (http://www.bt.cdc.gov/)• AFRRI (http://www.afrri.usuhs.mil/)• REAC/TS (http://www.orau.gov/reacts/)• Radiation Information Network

(http://www.physics.isu.edu/radinf/)

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Slide 281

A few more odds and ends…

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Slide 282 Radiological triage flowchart

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Slide 283 Radiological area entry and exit point

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Slide 284

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Slide 285 One way to assess dose

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Slide 286 Prognosis tool

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Slide 287 Expected progression of symptoms

Symptoms0 4 8 12 16 20 24 1 2 3 4 5 6 7 1 2 3 4 5 6

NauseaVomiting/retchingAnorexiaDiarrhea/crampsFatigueWeaknessHypotensionDizzinessDisorientationBleedingFeverInfectionUlcerationFluid loss/electrolyte imbalanceHeadacheFaintingProstrationDeathMedical treatmentClinical remarks

Hours Days Weeks

Dose of 0-0.75 Gy (0-75 rad) in airTime post-exposure

Reassurance, counselingPossible anxiety, possible mild lymphocyte depression in 24 hrs

0-5% mild

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Slide 288 Symptoms

0 4 8 12 16 20 24 1 2 3 4 5 6 7 1 2 3 4 5 6NauseaVomiting/retchingAnorexiaDiarrhea/crampsFatigueWeaknessHypotensionDizzinessDisorientationBleedingFeverInfectionUlcerationFluid loss/electrolyte imbalanceHeadacheFaintingProstrationDeathMedical treatmentClinical remarks

B. Increased susceptibility to non-opportunistic pathogens

Dose of 0.75-1.5 Gy (75-150 rad) in air

Debridement and primary closure of wounds, no surgery delayA. Moderate drop in lymphocyte, platelet, granulocyte counts

B

5-20% mild

15-50% mild

A A

5-30% mild

Time post-exposureHours Days Weeks

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Slide 289 Symptoms

0 4 8 12 16 20 24 1 2 3 4 5 6 7 1 2 3 4 5 6NauseaVomiting/retchingAnorexiaDiarrhea/crampsFatigueWeaknessHypotensionDizzinessDisorientationBleedingFeverInfectionUlcerationFluid loss/electrolyte imbalanceHeadacheFaintingProstrationDeathMedical treatmentClinical remarks

B. Drop in granulocytes from 6 to 2.0-4.5 x 103 per mm3

C. Drop in lymphocytes from 3 to 1.0-2.0 x103 per mm3

C

Hours Days Weeks

A. Drop in platelets from 3 to 0.8-1.8 x105 per mm3

mildmild

A 10% mild

B 10-50%

<5%

30-60% mild-moderate30-60% mild-moderate

Dose of 1.5 - 3.0 Gy (150-300 rad) in airTime post-exposure

30-70% mild-moderate20-70% mild-moderate

50-90%

Fluid, electrolytes for GI losses, cytokines for immune compromised

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Slide 290 Symptoms

0 4 8 12 16 20 24 1 2 3 4 5 6 7 1 2 3 4 5 6NauseaVomiting/retchingAnorexiaDiarrhea/cramps 10% moderateFatigueWeaknessHypotensionDizzinessDisorientationBleedingFeverInfectionUlcerationFluid loss/electrolyte imbalanceHeadacheFaintingProstrationDeathMedical treatmentClinical remarks

B. Drop in granulocytes from 6 to 0.5-2.0 x 103 per mm3

C. Drop in lymphocytes from 3 to 0.4-1.0 x103 per mm3

D. Possible epilation

Dose of 3.0 to 5.3 Gy (300 to 530 rad) in airTime post-exposure

Hours Days Weeks

Fluid, electrolytes for GI losses, cytokines, specific antibioticsA. Drop in platelets from 3 to 0.1-0.8 x105 per mm3

70-90% moderate0-80% moderate

90-100% severe 60%40-60%

5-50%

60-90% moderate mild moderatemoderatemild60-90% moderate

A 0-50% moderateB

C 80% moderateD 30% mod.

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Slide 291 Symptoms

0 4 8 12 16 20 24 1 2 3 4 5 6 7 1 2 3 4 5 6NauseaVomiting/retchingAnorexiaDiarrhea/cramps 10% moderate to severeFatigueWeaknessHypotensionDizzinessDisorientationBleeding A 50-100% mod-sevFeverInfectionUlceration D 50% mild-modFluid loss/electrolyte imbalanceHeadacheFaintingProstrationDeathMedical treatment Tertiary level intensive care, cytokines, fluids, antibiotics, GI deconClinical remarks

B. Severe granulocyte drop to 0.0-0.5 x103 per mm3

Complete surgery 36-48 C. Severe lymphocyte drop 0.0-0.1 x105 per mm3

hrs, or wait for 6 weeks D. Epilation E. Mild intestinal damage

Dose of 5.3 to 8.3 Gray (530-830 rad) in airTime post-exposure

Hours Days Weeks

90-100% mod - severe 60-100%severe80-100% mod-severe

100%60-100% severe

100%

50% mild-moderate 50%

90-100% moderate to severe90-100% moderate to severe

60% moderate60% moderate

B 60-100% mod-severeC

40% mild to moderate E 30%

50%60%

50-99%

A. Severe platelet drop to 0.0-0.1 x105 per mm3

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Slide 292 Symptoms

0 4 8 12 16 20 24 1 2 3 4 5 6 7 1 2 3 4 5 6NauseaVomiting/retchingAnorexiaDiarrhea/cramps 10% moderate to severe 100% moderate to severeFatigueWeaknessHypotension 100% severeDizziness 100% severeDisorientation 100% severeBleedingFeverInfectionUlcerationFluid loss/electrolyte imbalanceHeadacheFaintingProstrationDeathMedical treatmentClinical remarks

B. Granulocyte count drops to nearly 0Bone marrow totally C. Lymphocyte count drops to nearly 0depleted D. Epilation E. Moderate intestinal damage

Weeks

Supportive therapy, aggressive therapy if evidence of responseA. Platelet count drops to nearly 0

100% moderate to severe100% moderate to severe 100%

Dose of 8.3 to 11 Gy (830-1100 rad) in airTime post-exposure

Hours Days

100%100% 100%

100% severe100% severe

100%

A 100% severeB 100% severe

CD 100% severe

80% moderateE 80% severe

80% moderate 100% severe

80% mod to severe

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Slide 293 Symptoms

0 4 8 12 16 20 24 1 2 3 4 5 6 7 1 2 3 4 5 6Nausea 100% moderate-severeVomiting/retching 100% mod-severeAnorexiaDiarrhea/cramps 10% severeFatigueWeaknessHypotensionDizziness 100% severeDisorientation 100% severeBleeding B 100% severeFeverInfectionUlcerationFluid loss/electrolyte imbalanceHeadache 100% moderate to severeFaintingProstrationDeathMedical treatmentClinical remarks

B. Platelet, lymphocyte, granulocyte counts drop to 0C. Epilation D. Moderate to severe intestinal damage

Dose 11-15 Gray (1100-1500 rad) in airTime post-exposure

Hours Days Weeks

D 100% severe

Bone marrow totally depleted, tertiary care may help somewhatA. Blood press drops 25%, temp increases to 102 F


100% severe

B 100% severeB

C 100% severe100% moderate to

severe

100% severe100% severe100% severe


A 80% mild100% severe

100% severe

100% severe70% mod-severe70% mod-severe

100%

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Slide 294 Symptoms

0 4 8 12 16 20 24 1 2 3 4 5 6 7 1 2 3 4 5 6NauseaVomiting/retchingAnorexiaDiarrhea/cramps 20% severeFatigueWeaknessHypotensionDizzinessDisorientationBleeding A 100% severeFeverInfectionUlcerationFluid loss/electrolyte imbalanceHeadacheFaintingProstrationDeathMedical treatmentClinical remarks

B Severe intestinal damageBone marrow completely C Renal failuredepleted within days.

Supportive therapyA Platelet, granulocyte, lymphocyte counts drop to 0

Dose of 15-30 Gray (1500-3000 rad) in airTime post-exposure

Hours Days Weeks



100% severe

100% severe100% severe100% severe

100% severe100% moderate to severe


C 100%

45-80% mod-severe

B 100% moderate to severe100% severe 80% severe


A 100% severeA

100% severe

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Slide 295 Lymphocyte depletion

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Slide 296

Case studies

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Slide 297 Problem 1: Spill

• Technician calls to tell you about a spill. He claims to have readings of 20 mr/hr on a GM counter a meter above a spill of Tc-99m (10 mCi).

• Is this reading reasonable?• How do you confirm your answer above?• What equipment do you bring to the spill?• How do you deal with the incident?

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Slide 298 Problem 2: Refractory brick

• A load of brick from your furnace trips a landfill radiation alarm.

• Why does this happen?• What do you do with the brick?• What sort of risk do your employees face

from the radiation?• Do you have to get a license and train

your employees as radiation workers?

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Slide 299 Problem 3: Landfill alarm

• A load of residential waste sets off a landfill radiation portal monitor. The “hot”waste is traced to a nuclear medicine patient from your hospital.

• What do you do with the waste?• What instruments would you take to the

landfill?• Are you responsible for the patient’s

waste?

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Slide 300 Problem 4: High badge readings

• An employees badge reads 53 rem one month. The employee did nothing different compared to other months.

• How can you confirm or reject this reading?

• What actions do you need to take?

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Slide 301

Licensing exercise

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Slide 302 For the exercise

• Break into groups of 5 or so• Draw up a brief description of a license

application that addresses the questions on the following slide

• Include sketches as necessary and appropriate

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Slide 303 Irradiator license application

• You want to get a license for a 500 Ci Co-60 research irradiator.

• What instrumentation do you need for surveys?• Where in your building should the irradiator be

placed?• What topics should you include in your training?• Who should receive dosimetry? Do you need

any other dosimeters?• How do you leak test this source?• How do you return a spent source to the

vendor?

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Slide 304

Closing thoughts

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Slide 305 Health risks

• At levels allowed by regulations, radiation is not harmful

• In most cases, radiological risks can be readily managed and reduced

• Many activities (e.g. driving, some sports) are much more dangerous

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Slide 306 Legal risks

• Nevertheless, it is possible to be sued over radiation-induced illness, even if the probability of causation is very low

• For this reason, personnel monitoring and radiological survey records (and instrument maintenance records) are essential documents

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Slide 307 RSO’s role

• The RSO is the single person most responsible for maintaining a safe and legal radiation safety program

• Most people in the organization will look to the RSO for information and advice

• Don’t give out erroneous information – if you don’t know the answer, look it up or ask someone for assistance

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Slide 308 If you have any questions:

• E-mail me at [email protected]

• Check out the HPS web site atwww.hps.org

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Slide 309

Bonus Materials

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Slide 310

How uranium enrichment works

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Slide 311 Uranium enrichments

• Natural U – 99.2% U-238, 0.72% U-235• Enriched U - >1% U-235• Reactors – 3-8%• Research reactors – 20%• Weapons – 90+%

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Slide 312

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Slide 313

Communicating with the Media and the Public

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Slide 314 There is a tendency to assume that reporters either can’t understand technical issues, is not interested in technical issues, or has already made up its mind. This is not always the case, and we should generally treat them as willing and able to learn about issues of importance. Anyone who picks up the phone to call has already made an effort to learn something – we must respect that effort.

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Slide 315 Questions to consider• What prompted the communication in

the first place?• Who are you talking to?• What information is the other person

looking for?• What information do you want to

convey?• Is this an opportunity to get any other

information across?

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Slide 316 What prompted the communication in the first place?

• News story– Nuclear power– Radioactive waste– Radiological terrorism– Nuclear proliferation

• Exposure to radiation (to the person calling or to a family member)

• Medical problem thought to be linked to previous radiation exposure

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Slide 317 Who are you talking with?

• An experienced science writer• A junior reporter

– What “beat”? Science, environment, political, city, national, etc.?

• Someone trained in science, or some other field?

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Slide 318 What information do you want to convey?• Salient points about radiation

– High levels of radiation can be harmful– The risks of exposure to low levels of radiation are

typically over-stated– Radiation is a weak carcinogen and very poor at

inducing reproductive effects• Salient points about the matter at hand

– Evaluate dose calculations (if applicable)– Compare calculated dose with those required to

produce effects– Is dose less than 10 rem? If so, refer to HPS position

paper regarding low-dose risks

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Slide 319 Is this an opportunity to get any other information across?

• Radiation is a natural part of the environment and varies widely around the world

• HPS and ICRP have both advised against calculating risk from low doses of radiation

• We have over a century of experience working with radiation and know a great deal about its biological and medical effects

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Slide 320 Don’t try to sugarcoat the issues; if there are negative aspects to the story or question, acknowledge this. But don’t be afraid to point out that the negative is not as bad as many would lead us to believe. We “win” if the end result is someone with accurate information and the inclination to call us back in the future. We lose if we are seen as being as biased as the “antis”.

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Slide 321 What is the media after?

• Accurate information• The most important relevant facts• A good story that educates readers• Increased readership, viewers, listeners• Balance• Background information (if time permits)

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Slide 322 General rules• Only talk about things you know about• Don’t be patronizing or condescending• Leave yourself enough time to review

background information – don’t rush the interview

• Assume anything you say may be quoted• Be ready to refer reporter on to other HPs

who may know more about the subject matter

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Slide 323 Time is always an issue• All reporters have deadlines to meet• Reporters’ deadlines are usually fixed

and unchangeable• If a reporter can’t talk with you and meet

their deadline, they’ll go to press without your information

• Non-reporters (i.e. the curious public) is important, too – don’t make them wait either. Something prompted their call, and they deserve a timely response.

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Slide 324

Don’t be afraid to say “I don’t know” if appropriate.

Hopefully this will be followed by “I know who can help

you.” or “Let me find out and I’ll get back to you.”

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Slide 325 Accurate information

• In my experience, the vast majority of reporters want accurate information

• Only a few reporters are out to confirm their preconceptions

• Each reporter can reach thousands or even millions of people with their stories – by talking with reporters, you are talking to the public too

• We do everyone a disservice if the information we convey is not accurate

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Slide 326 A good story:• Sheds light on a topic of interest or

importance• Helps the public make informed decisions

about some matter• Helps decision-makers better understand

the issues at stake• Helps sell the media to the public (i.e. more

newspapers, more viewers, more listeners, etc.)

• May provide a different way of looking at an issue

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Slide 327 A balanced story

• Has information from all sides of an issue• May not give equal weight to all sides, but

tries to give appropriate weight to all sides• Makes an effort to let the readers decide,

based on the facts presented

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Slide 328

There is a tendency to be suspicious of the media and to assume they will distort information to support their pre-conceived notions or to sell more papers. This is sometimes the case, but most reporters honestly want to write a good, accurate story.

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Slide 329 Breaking news:

• When something is unfolding, there may be only a little time to try to work with either the public or the media

• Examples:– Tokaimura– Arrest of dirty bomb suspect– Kursk sinking– Any actual future radiological or nuclear

attack, incident, or emergency

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Slide 330 In a breaking story:

• Reporters generally have a higher sense of urgency

• There may not be time to go into much background information

• It’s best to stick to the most pertinent facts regarding the story and radiation in general

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Slide 331 I take the approach that the reporter is doing me a favor by giving me the opportunity to present the public with good information. That means that I expect to work with the interviewer to return the favor by helping set up interview times, locations, or settings that will make the interview convenient and valuable for the crew.

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Slide 332 Example: Dirty bombs• Background information

– Radiological versus nuclear weapons– LNT debate– Natural background exposure levels

• Pertinent information– Immediate actions – go inside, not to car– Not radiologically dangerous– Contaminated patients no risk to doctors

• Ancillary information– Contamination vs. dose– HPS position paper on low-dose risk– Detecting radiation and contamination– Results of calculations (if you have any)

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Slide 333 Example: Fetal radiation exposure

• Background information– Where fetal exposure comes from– Regulatory dose limits– We have a century of information about radiation

health effects• Pertinent information

– It takes at least 5 rem to harm a fetus– Unless there are multiple CT scans or a lot of

fluoroscopy, chances are that there will be no harm

• Ancillary information– There is a background rate of birth defects and

miscarriage for everyone

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Slide 334 Example: Nuclear war in South Asia

• Background information– What is a nuclear weapon and how does it work?– What effects have been noted in the past?

• Pertinent information– US not likely to have any health effects from

nuclear war in Asia– Limited nuclear war, although horrible, will not

end civilization• Ancillary information

– How can we detect nuclear weapons or weapons testing?

– How can someone make a nuclear weapon?

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Slide 335 Talking on radio• Try to speak slowly and distinctly so it’s easy to

understand you• Use analogies if possible and appropriate

– Try to create a picture in the listeners’ minds• Avoid jargon and technical terminology except

when absolutely necessary (and then explain the terms)

• Simplify as much as possible, but not to the point of being patronizing, condescending, or being unable to support your main points

• Don’t be afraid to ask the interviewer for tips on how to make sure their audience will get what you’re trying to say

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Slide 336 TV interviews• Dress comfortably and professionally• Don’t move too quickly• Look wherever the interviewer or camera

person tells you to look – usually NOT at the camera

• If you give demonstrations, keep them simple, easy to follow, and easy to understand

• Ask about props (i.e. meters) or preferred locations for filming

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Slide 337 Some common comments and rebuttals

• You can’t put a value on human life.– Do you have a life insurance policy? Medical insurance?

• Risks from driving and working are voluntary and controllable.– How do you get to the store? Can you control the

drunk/teenage/distracted idiot in the car next to yours?• We don’t know enough about the effects of

radiation– We’ve been working with radiation for over a century –

far longer than almost any other harmful agent. We know a LOT about the effects of radiation, especially compared to most toxins or carcinogens

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Slide 338 More snappy answers to common comments

• Cancer rates are going up because of radiation– 100 years ago the average person lived for 40

years. How many people 40 and younger get cancer?

• There is no “safe” level of radiation exposure.– Define “safe” – low levels of radiation are no

more dangerous than driving, working, or eating fast food

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Slide 339 Summary• We have an obligation to share accurate

information with the public, either directly or through members of the media

• We have an obligation to make sure our information is correct

• This need not be a stressful or adversarial process

• Our profession can gain much from this process, especially if we do it responsibly

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Appendix A: Glossary of Terminology Activity: the number of nuclear transformations occurring in a given amount of material per unit time

A= λ m N(a) W

A = activity level in disintegrations/sec λ = decay constant (units = /sec) = ln(2)/t1/2 m = mass of material present (units = grams) N(a) = Avogadroʹs number = 6.023x1023 W = atomic weight (units = atomic mass units) Base: one of the four fundamental building blocks of DNA (A = adenine, C = cytosine, G = guanine, T = thymine); bases on opposite sides of the DNA molecule always pair up so that A is across from T and C across from G – this combination is called a base pair BEIR: Biological Effects of Ionizing Radiation; a series of reports issued by a committee of the National Academy of Sciences Bq (Becquerel): the amount of material that gives a disintegration rate of 1 disintegration per second ‐ the Becquerel is the SI unit for activity level Ci (Curie): the amount of material which gives a disintegration rate of 3.7x1010 disintegrations per second‐a measure of activity; the Curie is the American unit for activity level

1 Ci = 3.7x1010 Bq Codon: a group of three base pairs that codes for a single amino acid Decay Chain (series): a series of isotopes resulting from the decay of a parent nuclide and its subsequent radioactive daughters to an ultimate stable form Decay Constant: the fraction of the number of atoms that will decay in a unit interval of time λ= ln(2)

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t1/2 t1/2 = isotope half‐life DNA: deoxyribonucleic acid, the primary genetic information molecule for all known life except for RNA‐based viruses Dose (absorbed): the energy imparted to matter by ionizing radiation per unit mass of irradiated material at the place of interest D = dr x t dr = dose rate in mr hr‐1 t = time in hrs Units of Dose: 1 rad = 100 ergs/gram in any material = 6.242x107 MeV g‐1 1 roentgen = 2.58 x 10‐4 Coulomb kg‐1 of air 1 gray = 100 rad = 10,000 ergs g‐1 in any material Dose Equivalent: the biological damage caused by the absorbed dose

DE = D x QF QF = quality factor = 1 (beta and gamma radiation) = 3 (thermal neutron radiation) = 10 (fast neutron radiation) = 20 (alpha radiation) Dose Rate: absorbed dose delivered per unit time (mr/hr) Gy (Gray – radiation dose): The SI unit of radiation dose; 1 Gray (Gy) = 1 Joule kg‐1 energy deposition from ionizing radiation Half‐Life (effective): the amount of time required for radioactive material subject to multiple loss terms to have its activity reduced by 50% by a combination of radioactive decay and other losses. t1/2(eff) = t1/2(1) x t1/2(rad) t1/2(1) + t1/2(rad) Half‐Life (radiological): the amount of time that is required for a radioactive substance to lose 1/2 of its activity

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Half‐Value Layer (HVL): That amount of a material that is required to reduce the dose rate from a radiation source by a factor of 2 Health Physics: the profession devoted to the safe use of radiation and radioactivity Hormesis: the presence of beneficial effects from exposure to an agent in small quantities, even if that agent may be harmful in large quantities IAEA: International Atomic Energy Agency ICRP: International Council on Radiation Protection (an international advisory body) Isotope: atom of the same atomic number (containing the same number of protons) but with a different number of neutrons in the nucleus (different atomic mass) ‐ can be stable or radioactive LNT: Linear, No‐Threshold; a hypothesis for radiation dose‐response that suggests that all exposure to radiation is potentially harmful and the risk increases linearly with dose NCRP: National Council on Radiation Protection and Measurements (a US governmental advisory body) Nuclide: an atom characterized by the number of protons in its nucleus AND its energy level (ex: Tc‐99m is a different nuclide than Tc‐99, exhibiting a different half‐life and different decay energies due to its existing in a different nuclear excitation (metastable) state Nuclide (parent): the nuclide that exists prior to radioactive decay, decaying to form the progeny nuclide ex: Xe‐138 will B‐decay to form Cs‐138 therefore, Xe‐138 is the parent nuclide and Cs‐138, the progeny nuclide Nuclide (progeny): the nuclide resulting from the radioactive decay of a parent nuclide (current terminology has replaced the term “daughter” with “progeny”) Radiation: the emission and propagation of energy through space and/or through a material medium in the form of waves or, by extension, corpuscular emissions such as α or β particles

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Rad: the deposition in any absorber of 100 ergs g‐1 due to the absorption of ionizing radiation Rem: exposure to that amount of ionizing radiation causing the biological damage equivalent to the deposition of 100 ergs g‐1 in body tissue; the rem is the US unit for dose equivalent Radioactivity: the property of certain nuclides of spontaneously emitting particles or gamma radiation following orbital electron capture, electron emission, isomeric transition, nuclear rearrangement, or spontaneous fission RNA: ribonucleic acid; an information transfer molecule used in most cells, RNA is also the primary information‐carrying molecule for RNA viruses Sievert (Sv): exposure to that amount of an ionizing radiation causing the biological damage equivalent to the deposition of 1 J kg‐1 in body tissue; the Sv is the SI unit for dose equivalent UNSCEAR: United Nations Science Committee on the Effects of Atomic Radiation

6. UV: ultraviolet light; that portion of the electromagnetic spectrum that has a higher frequency and shorter wavelength than violet light; UV has three recognized wavelength bands, UVA (320-400 nm), UVB (290-320 nm), and UVC (<290 nm)

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Appendix B: How Nuclear Reactors Work Contents 2) Introduction 3) The Nuclear Fuel Cycle 4) Kinds of Nuclear Reactors 5) Environmental Issues and Sustainability 6) Conclusions

Glossary of Terms Activation products – Atoms that become radioactive after being bombarded with neutron radiation (see neutron activation) AVLIS – atomic vapor laser isotopic separation; a method of using carefully tuned laser beams to increase the amount of 235U to make reactor fuel Becquerel (Bq) – a unit of radioactivity equal to one radioactive disintegration per second. Metric system multiples are used to describe large amounts of activity so that a kBq is 1000 Bq, an MBq is 106 Bq, a GBq is 109 Bq, and a TBq is 1012 Bq Boiling water reactor – a type of nuclear reactor in which water in the reactor core boils, generating steam to produce power Control rods – neutron‐absorbing rods (often made of cadmium, hafnium, silver, indium, or some other neutron‐absorbing metal) that are used to control reactor power; inserting control rods into the reactor core brings the criticality to a halt Depleted uranium – uranium from which 235U has been removed (in order to make enriched uranium); depleted uranium has less than 0.72% 235U Desalinization – the process of removing salt from sea water to make it drinkable or useable in agriculture Emergency cooling system – a system designed to remove waste heat from a reactor’s core in the event an emergency renders the reactor coolant pumps inoperable Enriched uranium – uranium in which the amount of 235U has been increased to a level greater than 0.72% in order to make reactor fuel Fission products (also called fission fragments) – radioactive atoms formed from an atom that fissions; fission of 235U leads to the formation of fission fragments with atomic masses around 100 and 135 atomic mass units. Fuel – the fissionable material (usually enriched uranium) that makes it possible to sustain a critical chain reaction; the uranium fuel is loaded into fuel rods which, in turn, are assembled into fuel assemblies Gas centrifuge – a method of uranium enrichment in which centrifugal force is used to help separate molecules containing the lighter atoms of 235U from the uranium feed

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Gas‐cooled reactor – a type of nuclear reactor in which gas (usually helium) is used to transfer the heat of nuclear fission from the reactor core to a steam generator or turbine Gaseous diffusion – a method of enriching uranium in which the slightly lighter molecules gas molecules containing 235U are more likely to penetrate (or diffuse through) a permeable barrier Graphite‐moderated reactor – a type of nuclear reactor in which graphite is used to moderate (or slow down) neutrons, making fission possible Mill and mine tailings – the waste materials left over after digging uranium ore from the ground and removing the uranium from it for further processing Moderation – the process of slowing down fast neutrons released from nuclear fission by causing them to collide, and exchange energy with atoms of the moderator Neutron activation – bombarding stable atoms with neutron radiation will result in some atoms absorbing neutrons; this process makes the atoms radioactive Nuclear criticality – the process by which a nuclear reactor produces power; in a critical reactor, the number of neutrons (a measure of reactor power production) remains constant over time – note that all nuclear reactors are critical when operating Nuclear fission – when some atoms absorb neutrons they will split into two or more parts and will emit neutrons; this process is called nuclear fission Nuclear reactor – a device that is designed to produce power by allowing nuclear fission to proceed in a controlled manner for prolonged periods of time Ore – a geologic body in which it is economically feasible to recover minerals or metals for industrial or commercial use Pebble bed modular reactor – a type of nuclear reactor in which small spheres of moderator and fuel are loosely stacked; helium is circulated through the spheres as a coolant Person‐Sv (or person‐rem) – radiation exposure to a group of people, defined as the summation of dose to every person measured; for example, a dose of 0.1 Sv to a group of 10,000 people will give a collective dose of 1000 person‐Sv Pressurized water reactor – a type of nuclear reactor in which the coolant (water) is pressurized to keep it from boiling as it transfers heat from the core to a steam generator Primary plant (systems) – the parts of a nuclear reactor that come in direct contact with coolant that has passed through the reactor core Radiation – the transfer of energy from one place to another via an intermediary; in particular, ionizing radiation uses alpha, beta, or gamma radiation to transfer excitation energy from an unstable atomic nucleus to an absorber Radiation dose – a measure of energy deposited in an absorber by ionizing radiation; 1 Gray (Gy) is the dose resulting from the absorption of 1 Joule of energy per kg of absorber. The Sievert (Sv) is a measure of dose equivalence and accounts for the fact that some kinds of radiation are more damaging than others

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Reactor coolant pumps – pumps used to circulate reactor coolant (usually water) through the reactor core SCRAM – an emergency reactor shutdown; scrams can be initiated automatically or manually Secondary plant (systems) – in a pressurized water reactor, the systems that do not have direct contact with water that has passed through the reactor core Sievert – exposure to ionizing radiation that produces the biological damage equivalent to depositing 1 Joule of gamma radiation per kilogram of body tissue – the US unit is the rem Steam generator – a piece of equipment in which hot gas or water passes through a heat exchanger and is used to boil water to produce steam; the steam is then used to turn a turbine to produce electrical power TRU – an acronym for trans‐uranic; any element heavier than uranium Turbine – a piece of equipment in which a hot gas such as steam causes the turbine to turn; the other end of the turbine is connected to an electrical generator to produce electricity Uranium enrichment – the process of increasing the amount of 235U from 0.72% (which is found in natural uranium) to a higher percentage needed for most nuclear reactors to operate 7. Summary

Nuclear power plants have many applications in our every day world. A significant fraction of the global electricity supply is produced by the world’s 400+ (as of June, 2001) nuclear power plants. Other nuclear reactors produce radioactive isotopes for research and medical treatment or generate radiations used in other scientific research. Nuclear energy also promises to help reduce the emissions of greenhouse gases, and some have noted that nuclear energy may be the best way to supply the growing demand for electrical energy without further contributions to global warming. However, nuclear energy is not an unmixed blessing. Nuclear power plants have also been decried as being costly, unsafe, and environmentally unfriendly. Several accidents have tarnished nuclear power’s image, and a few of these have resulted in fatalities. Nuclear reactors generate both high‐level and low‐level radioactive waste, and the ultimate disposal of these wastes is subject to a great deal of public scrutiny and generates considerable concern. The fears of the public and the environmentalists are further heightened by widespread fear of radiation and its potential long‐term effects on the public health. Nuclear power cannot be

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ignored as a potential source of global energy because of its undeniable benefits; neither can it be embraced unquestionably, for the reasons noted above. In this article, many of the issues surrounding nuclear energy will be discussed. This discussion will include a brief description of the manner in which nuclear reactors are currently used, a more detailed description of the nuclear fuel cycle and the theory underlying nuclear reactor operations and design, and a discussion of the scientific basis for many of the public health concerns raised by nuclear power. Finally, we will discuss some of the political and environmental issues surrounding nuclear energy, comparing nuclear reactors to other energy sources in an effort to provide an unbiased comparison of the benefits and drawbacks they provide. 8. Introduction a) A brief history of nuclear reactors and their uses The first man‐made nuclear reactor was constructed at the University of Chicago by

a team led by Enrico Fermi in 1943. Built as part of the Manhattan Project, this nuclear reactor was designed and constructed explicitly for research leading to the eventual construction of the world’s first nuclear weapons. Although much scientific research suggested that uranium could be used to generate a self‐sustaining nuclear chain reaction, until Fermi’s reactor achieved criticality, this had not been demonstrated in practice. Using knowledge from this reactor, other scientists were able to develop not only nuclear weapons, but also built the first plutonium production reactors, used to create fuel for other atomic bombs. Thus, from the very start, nuclear reactors became intimately associated with nuclear weapons; an association that has since haunted all discussion of nuclear energy.

Following the end of World War II, the United States, the Soviet Union, Canada, and Great Britain were the first nations to continue exploring the potential of nuclear energy for both civilian and military purposes. In particular, the United States and the Soviet Union, spurred on by the Cold War, led the way in designing and building increasingly sophisticated designs for both nuclear reactors and nuclear weapons, as well as devising an increasing number of uses for nuclear power in other settings. This led to development and construction of small reactors for use in research, somewhat larger reactors used to produce radio‐labeled compounds for research and medical purposes, and the investigation of “portable” nuclear reactors for generating power in near‐combat or remote locations. In addition, nuclear reactors were utilized by several navies, where they revolutionized submarine warfare.

The world’s first commercial nuclear power plant went into operation in late June, 1954 in the Russian city of Obninsk, near Moscow. This was followed in 1956 by the British plant in Calder Hill, and in 1957, the first American commercial nuclear power plant went on‐line in Shippingport, Pennsylvania. The 1950s, and to a lesser

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extent the 1960s, was an age of nuclear optimism, particularly in the US and the Soviet Union. The widespread use of nuclear power was seen as a nearly unlimited source of inexpensive, reliable power that would help to lift much of the world out of poverty while simultaneously providing fresh water via desalinization plants, new drugs from research using radioisotopes, both at reduced environmental impact from reduced emissions. However, the continued development of nuclear weapons, their testing in the atmosphere, and the growing awareness of the potential for radiation injury became concerns. In the 1960s, with the growing strength of the global environmental movement, these concerns were voiced to governments with increasing volume. One milestone along this path was Linus Pauling’s successful campaign to halt atmospheric nuclear weapons testing, which was given enhanced visibility by his subsequent Nobel Peace Prize for his efforts. However, it was not until 1979, with the accident at the US nuclear power plant at Three Mile Island (TMI) that the anti‐nuclear power movement really took off.

Although the TMI accident resulted in exceedingly low radiation exposure to the general public, the reactor core was destroyed, and the perception was that it represented a narrowly‐averted disaster. Coming on the heels of the successful (although technically inaccurate) movie “The China Syndrome”, the accident was an unmitigated disaster for the US nuclear power industry. Seven years later, much more serious accident at the Soviet (now Ukranian) Chernobyl nuclear reactor gained global notoriety, the political results of which are still felt today.

As of this writing (June, 2001), the global outlook for nuclear energy is mixed. Japan, France, and to a lesser extent Russia, Canada, and Great Britain seem to have mature and relatively politically secure nuclear power capabilities. Several European nations, however (including Sweden and Germany) have announced plans to eliminate nuclear power plants, although the source of alternate energy has not yet been announced. Still other nations (particularly China and, to a lesser extent, Iran) are embarking on large programs to increase their dependence on nuclear energy, and the US stance remains mixed and undecided. b) Nuclear reactor theory and operations – a general description Nuclear reactors generate energy by fissioning (splitting) atoms of uranium. This

simple statement hides a great deal of physics and engineering. The physics describes why splitting atoms produces energy and how this fissioning can be maintained for prolonged periods of time, and the engineering is necessary if this energy is to serve any useful purpose.

It is not immediately obvious that simply splitting a uranium atom should release energy. After all, splitting a log, or a stone, or any other object we are familiar with requires energy – swinging an ax is hard work. Similarly, there is no obvious reason that fissioning one atom should result in a second atom splitting, just as it is not readily

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apparent why uranium must be used instead of, say, lead or iron. A brief foray into nuclear structure is necessary to understand this, but the description is neither mathematical nor abstract.

All atoms are composed of a central nucleus surrounded by a cloud of electrons. The electrons do not concern us for the purposes of this discussion. The nucleus, in turn, is made up of protons with a positive electrical charge and neutrons with no charge, all confined to a very small space. Similar electrical charges repel one another, and the protons in the nucleus are subject to strong forces that try to force the nucleus apart. What holds atoms together is a force, called the strong nuclear force, and this force is carried by the neutrons. The neutrons are the duct tape that helps to hold the protons together. However the strong force only works over very short distances, so as atomic nuclei become larger, the strong force loses its ability to hold onto all of the protons. This means that, in general, large, heavy atoms are inherently less stable than small, light atoms. Another way to look at it, using the duct tape analogy, is that a piece of tape has a finite length. If we use a 30 cm piece of tape to hold together a few sticks, it will serve quite well. However, as the group of sticks grows, the tape is less able to wrap around to hold them all, and the entire bundle becomes less stable and easier to tear or fall apart.

Uranium is the largest atom that exists in abundance on Earth. There are small amounts of plutonium that are present naturally, and large amounts of plutonium and even heavier elements are formed in stellar explosions, but they are not long‐lived and are uncommon on Earth. This means that uranium is also the atom most likely to fall apart on its own (called spontaneous fission) or to be forced apart (induced fission) by adding a neutron to the atomic nucleus. The addition of an extra neutron to the atomic nucleus also adds energy to the nucleus, making it vibrate. If an atom is teetering on the edge of stability, as is the case with the uranium atom, this added energy and vibration can cause it to fly apart, or fission.

Uranium also comes in several “flavors”, or isotopes. The chemical properties of an atom are determined by the number of protons in the nucleus. Every atom with 82 protons (lead) is chemically identical, as is every atom with 92 protons (uranium). However, atoms of the same element can have different numbers of neutrons present, giving them a variety of atomic weights and different atomic properties. In the case of uranium, 99.2 percent of the uranium in the world has 92 protons and 146 neutrons, giving it an atomic weight of 238 (written as 238U, or U‐238). About 0.72 percent has three fewer neutrons; 235U. In spite of having the same chemical properties, U‐235 and U‐238 have different nuclear properties, and U‐235 is more likely to absorb passing neutrons than is U‐238. When this happens the strong nuclear force, already stretched thin by the sheer size of the nucleus, can no longer hold the atom together and it falls apart. As the atom fissions, it produces two fission fragments (which are radioactive), two to three neutrons, gamma rays, and energy. The energy released is what we

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harness to make electricity, the neutrons go on to cause other fissions, and the fission fragments become radioactive waste that, in most cases, remains locked within the fuel.

As noted above, between two and three neutrons are released from each fission. Some of these neutrons go on to be absorbed by the water, steel, and lead of the reactor plant and are lost to the plant. Others are absorbed by uranium atoms, but do not cause fission, and still others escape the reactor altogether. The entire secret of nuclear reactor design is to arrange the fuel in such a way that, for each atom of uranium that fissions, exactly one neutron is produced that goes on to create another fission. When this happens, the total number of fissions in the reactor at any time is constant, so the production of energy is constant. Making this happen requires a certain mass of uranium arranged in a certain configuration, called the “critical mass” and “critical geometry”. When you achieve such conditions, the nuclear reactor is said to be “critical”. Put another way, all nuclear reactors are critical when they are operating, and nuclear criticality in an operating nuclear reactor is hardly an emergency (as a corollary, those who understand this fact are usually amused by television shows or movies in which someone announces in a panic‐stricken voice that “the reactor is critical” – this just indicates that the writer is not terribly knowledgeable about nuclear reactors, and suggests that their writing should be viewed with some degree of skepticism).

On an atom‐by‐atom basis, nuclear fission releases a tremendous amount of energy. Splitting one uranium atom produces about 100 times as much energy as burning one molecule of gasoline, so nuclear energy can produce much higher energy densities than can plants that rely on chemical reactions (such as combustion). However, this energy is useless unless it can be harnessed in a usable form; we cannot simply pump through wires. In the case of nuclear reactors, the energy generated by fission turns into heat, which heats the reactor fuel. The fuel is surrounded by a coolant, usually water, and the heat energy is transferred into the water. The hot water, in turn, is used to produce steam, which turns turbines, which generate electricity. Although the process sounds somewhat laborious, it is no more so than many other forms of electricity generation, and the efficiency of most nuclear power plants (i.e. the ratio of electrical energy to thermal energy) is higher than many competing forms of energy generation.

Technical note – in reality, energy cannot be created, it can only be changed from one form to another. Nuclear reactors release energy already present in an atomic nucleus, turn it into heat energy, and the heat energy is transformed into electrical energy. The total amount of energy contained in the power lines coming out of a nuclear reactor plant is the same as the total amount of energy originally present in the uranium atoms that were fissioned.

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c) Uses of nuclear reactors Although the most visible and obvious uses to which nuclear reactors have been

put, are the generation of electrical energy and the production of materials for nuclear weapons. They have found many other uses in the half‐century or so they have been in use. These uses include the production of radioactive isotopes for medical diagnosis and treatment, production of isotopes for research, and desalinating seawater for drinking and industrial purposes. Generating power has already been discussed above, and nuclear weapons are beyond the scope of this chapter, so the next few paragraphs will discuss the use of nuclear reactors for military (non weapons‐related), research, and medical purposes.

i) Military, non‐weapons use The first non‐weapons use to which nuclear energy was applied was in nuclear

submarines. Every major naval power immediately understood that nuclear energy offered the promise of creating a submarine force that could operate submerged for prolonged periods of time, virtually undetectable. Earlier submarines were hybrid machines – they ran on diesel engines on the surface and on batteries while submerged. Since their batteries could only give a limited amount of service before recharging, diesel submarines were designed to operate on the surface of the ocean, submerging only when necessary to attack or to hide.

Nuclear reactors, unlike diesel engines, do not need air or oxygen to produce power. A nuclear submarine could operate at full power, completely submerged, almost indefinitely. The reactors actually produce more than enough energy to meet the ship’s needs, leaving additional energy for distilling fresh water, purifying the atmosphere, and more. In addition, freed from the constraints of a large battery for underwater attacks or evasions (and the diesels to recharge it), much more of the volume could be devoted to carrying weapons, electronics, and crew. The development of nuclear submarines was the most significant revolution in the history of submarines and was one of the biggest innovations in the history of modern naval warfare. Although nuclear power has also been put in use on surface combatants by the US and the Soviet Union (now Russia), its advantages on the ocean’s surface are not nearly as pronounced as they are beneath the waves.

In addition to the naval use of nuclear reactors, some nations experimented with “portable” nuclear power plants that could be used to supply energy to military headquarters in remote locations. Another proposed use was in a nuclear airplane, a project begun but abandoned by the US in the 1960s. Others have used nuclear reactors in space. This latter use should not be confused with radio‐isotopic thermal generators (or RTGs) which are used on most deep‐space missions to the outer solar system. RTGs make use of heat released by radioactive decay, but they do not use nuclear fission for

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this purpose, they do not generate radioactive fission products or high levels of neutrons.

ii) Medical and research uses Although only a few isotopes will fission, most can be induced to capture

neutrons, protons, or other atomic particles under the appropriate conditions. When this happens, the resulting atom will become radioactive. One example of this is the formation of radioactive carbon in the atmosphere. In this reaction, a cosmic ray neutron will strike a nitrogen atom, ejecting a proton from the nucleus and turning it into a carbon atom. This reaction is written 14 14N n C p+ = + This reaction is the one that creates the carbon‐14 used to date archeological artifacts, tree rings, and many other objects. Similar reactions can be used to create many other isotopes that are widely used in research, to diagnose medical conditions, or to treat cancer and some other diseases. Research performed with the aid of nuclear reactor‐generated isotopes includes genetic sequencing, investigation of basic biological functions, the development and understanding of new drugs, and better understanding of brain functions.

Nuclear reactors are also used as a source of neutrons for other research purposes. Neutrons can be used to probe the structure of matter and to investigate the chemical composition of geologic specimens (also called rocks). Cell cultures can be exposed to radiation from nuclear reactors to learn more about how DNA is damaged and repaired, nuclear reactor‐produced neutrons have also been used to help treat some forms of cancer, and chemical compounds containing radioactive atoms created in nuclear reactors are used in research and in the diagnosis or treatment of disease. d) Desalinating seawater

The primary product of nuclear fission is heat, and this heat can be used for purposes other than generating electricity. Nuclear desalinization plants use this heat to boil salt water because the steam that is produced can then be condensed to form fresh water for drinking. Nuclear desalinization is not a new technology; nuclear submarines and surface ships do this routinely to produce drinking water for their crews and fresh water for the engineering plant. However, this technology is not used for making public drinking water because of public response against nuclear reactors. The International Atomic Energy Agency has given this subject much study and has written several excellent fact sheets that are worth reading. Nuclear energy is not the

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least expensive method available for making fresh water, but it is one of the only technologies that can produce millions of gallons of fresh water daily, in virtually any part of the world that is near an ocean. 9. The Nuclear Fuel Cycle

As mentioned above, natural uranium cannot sustain a nuclear chain reaction; for

this to occur, the fraction of U‐235 present in the uranium must be increased from 0.72 percent to at least one percent of the total number of uranium atoms present. In reality, the uranium must be enriched further yet, because commercial nuclear reactors use fuel containing from three to six percent U‐235. Some nuclear reactors, primarily those used for research, make use of fuel enriched to 20 percent U‐235, and some military nuclear reactors use fuel that is nearly pure U‐235 (similar to the concentrations used in some older nuclear weapons). As an aside, this means that commercial and research nuclear reactors cannot explode like nuclear bombs – it is physically impossible for them to do so because of the relatively low concentrations of fissionable uranium. This is not to say that nuclear reactors cannot experience accidents; the Chernobyl accident spread large amounts of radioactive contamination over a large area, and the reactor core at Three Mile Island was destroyed in its accident (although the release of radioactivity at TMI was minor). However, stories of commercial nuclear reactors exploding like atomic bombs are wildly inaccurate.

The process of making fuel for nuclear reactors begins when the uranium ore is mined and processed, continues through the process of uranium enrichment, and culminates with fabrication of the nuclear reactor fuel. Eventually, the U‐235 in the reactor fuel is fissioned to the point at which nuclear reactions no longer occur, and the reactor is refueled so it may continue operating. The spent fuel is then either stored on site, sent for disposal, or recycled. This whole process is called the nuclear fuel cycle and is the subject of this section. e) Uranium mining

The first step in the nuclear fuel cycle is locating a uranium ore deposit and bringing the ore to the earth’s surface. Virtually all rocks, soils, and most waters contain trace amounts of uranium. However, the uranium is present in low quantities, and it is not possible to recover the uranium at a reasonable cost. The term “ore” is an economic term, not a scientific one; ore is the presence of a material (usually metal) in a place and chemical concentration that makes it possible to mine at a profit. For example, deep‐sea manganese nodules contain large amounts of very high‐purity manganese, but their location on the seafloor makes the metal horrendously expensive to recover. These nodules of nearly pure metal are not considered ore because of this.

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On the other hand, large mining operations in the AmericaN West make a lot of money mining ore that contains small amounts of manganese, simply because it is found near the earth’s surface and the mining is not expensive.

In the case of uranium, a large number of ore deposits have been found on virtually every continent. Among the most famous are those in the American west, Australia, Canada, and several places in Africa. Some of these ore deposits are near the earth’s surface, and strip mining operations are sufficient to recover the ore, while others are deeply buried, requiring mine shafts and tunnels. Regardless of how the ore is mined, it is brought to the surface along with tens of thousands (or even millions) of cubic meters of other rock. This other rock (mine tailings) is often mildly radioactive (because of elevated, but not economic levels of uranium); it is generated in the process of digging to the ore body. Most mines generate large piles of tailings, but in the case of uranium mining, the tailings are often regulated and must be treated as radioactive waste. This can create problems, because the piles can leach uranium into ground or surface waters, they emit radioactive radon gas, and many nations’ laws will not permit their disposal into landfills or even into the mine shafts from which they were excavated. This forces companies to find some way to contain the tailings, with varying degrees of expense and success. f) Processing the ore

After removal from the ground, the uranium ore is shipped to a processing facility. Here, the rock is crushed and chemically processed to remove as much of the uranium as possible. The uranium may then be chemically treated to turn it into uranium hexafluoride (UF6), making it suitable for enrichment via gaseous diffusion (described in the following section). If another enrichment process is used, the uranium may be placed in another chemical form more conducive to that particular process.

Following removal of uranium from the rock, there is, again, a great deal of waste material. Some uranium ores contain less than one percent uranium by weight, and far less by volume. This means that virtually all of the ore brought to the uranium mill is discarded as waste, generating huge piles of mill tailings. As with the mine tailings, this waste is mildly radioactive and emits radon. This, too, must be properly handled to adhere to regulatory requirements and to minimize environmental and health effects. It must be pointed out that, in the case of both mine and mill tailings, the risk from radioactivity present in the waste materials is exceptionally low, and there have been no documented human health effects from this radioactivity. It must also be noted that the radiation levels from the ore and the UF6 containers are also too low to create a health risk to the public.

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g) Uranium enrichment As of this writing, there are three primary methods of uranium enrichment used

by most nations with domestic uranium processing. These are gaseous diffusion, gas centrifuge, and atomic vapor laser isotopic separation (AVLIS). The first two of these make use of the very slight difference in weight between 235UF6 and 238UF6 and AVLIS takes advantage of small atomic differences that can be used to “sort” uranium atoms to concentrate the 235U to whatever degree is desired. h) Reactor fuel fabrication

Finally, the enriched uranium is ready to be made into fuel for a nuclear reactor. To do this, the UF6 must be chemically converted into uranium oxide, which is more chemically stable and less hazardous than the uranium hexafluoride. Once this oxide conversion is completed, the enriched uranium oxide powder is mixed with zirconium or some other metal and pressed into small pellets, each about 1 cm in diameter and a few cm long. These fuel pellets are loaded into stainless steel rods, and the rods are then assembled into bundles called fuel assemblies. It is these fuel assemblies that are loaded into the reactor core. i) Reactor re‐fueling

After a period of time, usually 12‐18 months, the reactor fuel can no longer sustain a nuclear chain reaction efficiently because so much of the U‐235 has been fissioned. During the time the fuel is in the reactor it is periodically moved from place to place within the reactor core to help maximize the life of the fuel assembly while maintaining desired characteristics within the core itself. For example, new reactor fuel has a higher concentration of U‐235 so it will produce more power than an older fuel bundle, or it will produce power in a less neutron‐rich environment. Such bundles are often placed near the edge of the reactor core, where there are fewer neutrons, because this helps to balance power production in the reactor core. As the fuel burns out, it will be shifted closer to the center of the core, experiencing higher neutron levels and producing the same amount of energy. After a sufficiently long time the fuel will no longer produce enough energy to warrant keeping it in the core, so it is removed and more new fuel is added.

The process of moving and replacing reactor fuel is complex and carries with it the potential for very high radiation doses to workers. For this reason, reactor re‐fueling is done as infrequently as possible, and is never done unless necessary.

In addition, the reactor must be shut down to perform the refueling, removing it from a nation’s electrical power generation system. Shutting the reactor down means that workers can enter the reactor compartment to perform needed maintenance, but it also means that the reactor is not generating power or income. Because of this, these “outages” are tightly scripted to get the most work possible done with the least amount

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of expense and in the shortest time possible. In the US, most planned reactor outages are scheduled for the spring or autumn, when neither air conditioning nor heating is necessary, reducing electrical power demands as much as possible. j) Reactor fuel reprocessing

After removal from the reactor core, some nations send the spent fuel to a large chemical processing facility to recover as much unfissioned uranium as possible. Although the spent fuel rods are dangerously radioactive when they are first removed from the core, this radioactivity fades relatively quickly because most of the fission products that make the fuel radioactive have very short half‐lives, do not remain radioactive for long, and can be safely handled after several months.

In the reprocessing facility, the fuel pellets are removed from their cladding and are dissolved in nitric acid. After much chemical processing, the facility is left with uranium, plutonium, and waste. The plutonium and waste will be discussed in other sections of this chapter. The uranium is recycled, enriched again, and made into new reactor fuel.

The United States reprocessed reactor fuel until the mid‐1970s, at which time the practice was halted. The reason the US stopped reprocessing reactor fuel is because of the plutonium; President Carter felt that the chance for this plutonium to be obtained by terrorists or by nations trying to develop nuclear weapons was too great. At present, the US does not recycle reactor fuel but, instead, considers it all to be waste. The fate of spent reactor fuel and other radioactive wastes is discussed in the following section. k) Radioactive waste

Radioactive waste is any radioactive material that serves no useful purpose and that is not in its natural state. For example, uranium that is found naturally in soil is not radioactive, nor is it considered waste when it is producing energy in a nuclear reactor. However, once removed from the reactor, when it is no longer useful, the remaining uranium contained within a fuel rod is considered radioactive waste and must be properly disposed of. Similarly, items made radioactive by bombardment with neutrons become radioactive waste when they are removed from the system, and items contaminated with radioactivity are also considered radioactive waste.

Contrary to popular beliefs, the majority of radioactive waste is not glowing green deadly liquid. In fact, most radioactive waste is gloves, tools, paper towels, rags, and other items that are mildly radioactive or mildly contaminated. There are, of course, wastes that are dangerously radioactive, but they constitute the minority of the volume that is sent to radioactive waste disposal facilities around the world.

Most nations have strict regulations regarding the packaging, shipping, disposal, and accountability of radioactive wastes. Typically, radioactive wastes are placed into a large metal drum or box, and trained technicians keep accurate and detailed records of

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the exact contents of each waste container. These records are used to determine the appropriate way to ship the wastes, and are kept on hand by the driver until the waste is delivered to the final disposal facility. Along the way, the waste may be sent for processing to reduce its volume, liquid wastes may be solidified by mixing them with concrete, or other actions may be taken to make final waste disposal as safe and economical as possible. Waste treatment and disposal depends on many factors, including the physical form of the waste (solid or liquid) and the level of radioactivity present (high‐level or low‐level radioactive waste). Some of these factors are described in the following sections.

i) Waste treatment and processing Many options exist for treating or processing radioactive wastes prior to ultimate

disposal. For example, radioactive liquids may be mixed with concrete to solidify them, making leakage less likely. Some materials, including many solvents, paper, plastics, and wood can be incinerated and the ashes shipped for disposal – this reduces the volume of the waste by a factor of ten or more. Metals may be melted, and the ingots can then be disposed of in a more compact form, and some materials are vitrified (turned into glass) and disposed. Some materials may be chemically processed to remove the radioactivity, and other materials are simply compacted by enormous hydraulic presses to reduce the final volume as much as possible before burial. Reducing the volume of radioactive material is a prime consideration in many of these treatment techniques because no landfill has unlimited space available, and radioactive waste landfills are even more restricted than many others.

ii) Waste disposal – low level radioactive waste (LLRW) Low‐level radioactive waste is generally waste that poses little or no health risk

due to its radioactivity. It may or may pose other hazards due to its chemical or physical properties; for example, dropping a 1000 kg container of LLRW on a person will hurt that person, even if they receive virtually no radiation dose from it!

In general, most nations have at least one specially designed repository for the disposal of LLRW. In most cases, these repositories must adhere to strict siting guidelines that are intended to minimize the chance for radioactivity to escape into the environment. Such sites are also typically ringed with monitoring stations designed to detect even the slightest leakage of radioactivity, and they are inspected frequently by regulatory agencies. In the early years of LLRW disposal some sites did experience the release of minor amounts of radioactivity into the environment. However, in all cases (except for a few in the former Soviet Union) these releases were not sufficient to cause lasting harm to the environment, to workers, or nearby residents.

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In some nations, LLRW disposal facilities are not unlike other properly designed landfills. In such facilities, soil is excavated and the resulting pit is lined with compacted clay to isolate it from the groundwater. Clay is used because it is relatively inexpensive, abundant, can deform to accommodate shifts in the ground, and is relatively impermeable to liquids. Beneath this clay there may be monitoring lines, to detect leakage through the clay if it occurs. When the waste containers are placed in the pit, it is capped with more compacted clay and sometimes a polymer sheet designed to keep rain out of the pit.

In other nations, LLRW facilities are constructed above ground. In these facilities, concrete vaults are built atop a compacted clay foundation. These vaults are filled with an orderly array of waste containers. As with the below‐ground waste facilities, the environment is monitored on a continuing basis, including looking for evidence of radioactivity in the groundwater. The vaults may be covered with soil when they are filled, or they may be left uncovered to facilitate inspecting for damage over the years. In any event, both forms of waste disposal have generally proven to be effective at isolating radioactive waste from the environment safely.

iii) High‐level radioactive waste, TRU, and spent reactor fuel Aside from low‐level radioactive wastes, nuclear reactors also produce high‐level

radioactive waste (HLRW), trans‐uranic (TRU) waste, and spent reactor fuel. These are all treated differently than LLRW because of the unique hazards each poses.

HLRW is radioactive waste that is dangerously radioactive and may remain so for many years. It can include resin from the water purifying units used to ensure water purity in the reactor, or nuclear reactor components that have become radioactive after years or decades of neutron bombardment. HLRW is often placed in special casks that provide additional shielding. It may be stored in this manner at the site where it was generated, or it can be shipped for disposal in a special HLRW facility. In most nations, HLRW is disposed of in deep geologic repositories; tunnels and chambers drilled into solid rock, often hundreds of meters deep. These are thought to be capable of safely holding the wastes until they are no longer radioactive.

By neutron capture, some uranium atoms will be transmuted into plutonium, americium, and other elements beyond uranium on the periodic table. These transuranic elements are often very radioactive and many are chemically toxic as well. In addition, some isotopes of plutonium are fissionable and can be made into either nuclear reactor fuel or nuclear weapons. Although Pu, contrary to popular belief, is not the “most toxic substance known to man”, it is both radioactive and toxic and it must be treated with care. However, the greatest risk posed by Pu is the potential for it to be seized by a terrorist group or a rogue nation, made into a nuclear weapon, and that weapon used in an act of terrorism or war. This is one of the reasons the US stopped

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reprocessing spent nuclear reactor fuel; separating Pu from spent reactor fuel is difficult, dangerous, and expensive, and it is thought that few nations have developed the ability to do so at this time. In the US, TRU wastes are now disposed of at the Waste Isolation Pilot Plant (WIPP), located near Carlsbad New Mexico in the US desert Southwest. WIPP is located within a deep underground layer of salt that, over time, will engulf the waste containers, keeping them safe for millennia.

In many nations, spent reactor fuel is reprocessed, and the resulting waste is disposed of as HLRW. The US and some other nations, however, dispose of spent reactor fuel, which can remain highly radioactive for several years after it is removed from the nuclear reactor core.

The US is in the process of developing a spent reactor fuel disposal facility inside of a mountain in the Nevada desert, not far from Las Vegas. Until that time, spent reactor fuel is stored at the nuclear reactor plant where it was generated. When first removed from the reactor core, radioactive decay of fission products generates so much heat that the spent fuel is kept in a large tank of water, called a spent fuel pool. The water helps to keep the fuel cool and also shields the radiation coming from the fuel assemblies. However, due to the long delay in opening the Yucca Mountain facility, many US reactor plants have filled their spent fuel pools, so they have begun moving old, “colder” spent fuel to land‐based storage containers. Here, the spent fuel can be safely maintained for many years, until it can be sent to Yucca Mountain as a final repository. Other nations have a variety of methods for dealing with spent fuel, including some or all of the options mentioned above. l) Summary The nuclear fuel cycle is a long and complex series of events that involves every

facet of mining and processing uranium to make fuel for nuclear reactors and dealing with the wastes at all steps of this process. Although the details vary between countries, the overall scheme remains the same – uranium must be mined and enriched, fuel must be manufactured and installed into nuclear reactors, and the inevitable waste products must be treated and disposed of in a safe manner that is protective of the environment. 10. Kinds of Nuclear Reactors The basic physics behind uranium fission is the same, regardless of the manner in

which it happens. Each atom of U‐235 is virtually identical, each requires the same conditions to fission, and each behaves roughly the same when it fissions. There are some differences between atoms due to quantum mechanical uncertainties, but generally speaking, the process of fissioning an atom of U‐235 is the same for all U‐235 atoms. However, the manner in which fission is initiated, controlled, and the energy

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utilized varies between types of nuclear reactors. In this section, we will look at some of the most common nuclear reactor designs in terms of how they work, their advantages and disadvantages, and how many are in use in the world. First, however, we will briefly discuss the safety features present in virtually all nuclear reactors plants because it is these safety features that keep a nuclear reactor accident from becoming a catastrophe. It must be noted that, in the case of the Chernobyl nuclear reactor, some of these safety features were not in place or were bypassed; major factors that led to the severity of that accident. It must also be noted that the accident at the US Three Mile Island plant, while very damaging to the reactor core, resulted in very low radiation dose to the workers and surrounding population, a fact that is often not appreciated.

There have been many lawsuits, virtually all of which were found to be without scientific or legal merit. Several studies have concluded, based on readings from dosimeters in place around the facility before the accident occurred, that the highest dose to any person outside the facility was less than 10 mrem. This is less than the dose from an x‐ray and is less than the variation in radiation dose to someone moving from a low elevation to the mountains. The official dosimetry studies performed by the NRC were quite complete and the results were reviewed by a number of independent scientists and found to not have been manipulated. Also noteworthy is the city of Ramsar, Iran in which the average radiation exposure to residents is 100 – 1000 times higher than what any TMI residents received. Overall health in Ramsar (as well as in the vicinity of Three Mile Island) is no different than in nearby areas. a) Nuclear Reactor Safety Features

Virtually all nuclear reactors require safety features to ensure the following:

• The reactor will automatically shut down if the core integrity is threatened • The reactor can be manually shut down if necessary • The reactor core will remain cool • Major releases of radioactivity will be isolated from the environment

Reactors are shut down by inserting neutron‐absorbing control rods into the core. In

virtually all nuclear reactors, the control rods are withdrawn from the reactor core by energizing a motor that pulls them out to the required position. As the rods are pulled, a powerful spring, the scram spring is compressed. Once in position, powerful latches are activated electrically and they hold the control rods in place against the pressure of the scram spring. The control rods are typically powered from the same source of electricity as the other important reactor plant systems. If electrical power is lost, the electromagnet holding the control rod latches will lose power and the latches will spring open. With nothing to hold them in place, the control rods are forced into the core by the scram spring, shutting down the reactor. The automatic shutdown system is

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also designed to activate in the event that certain reactor plant parameters become serious. For example, if the core is producing too much power, if the core temperature is too hot, or if flow through the reactor core drops below a certain level, the safety systems will scram the reactor to keep it from damaging itself or endangering workers, the environment, or the public. This same system can also be activated manually if an operator turns a scram switch.

Even shutdown, the decay of radioactive fission products will produce large amounts of heat that can damage the reactor core if not removed. For this reason, most nuclear reactors have a few systems to make sure that the reactor core is kept full of coolant (usually water) and that this coolant is circulated through a heat exchanger to transfer this heat away from the reactor. In most cases, these systems are designed so that they will automatically activate if needed. In newer reactors, they are also designed to take advantage of basic laws of physics so they will operate properly even in the total absence of electrical power. For example, a tank to keep the reactor plant full of water will be located high above the reactor so that gravity will ensure water flows into the reactor. Other systems use natural circulation and convection to circulate water through the reactor core and into heat exchangers, taking advantage of the fact that hot water, being less dense than cold water, will rise into a heat exchanger located above the reactor and the cooled water will descend through another pipe back into the reactor core. Together, these emergency systems help to ensure that temperatures in the reactor core are kept low enough that the fuel is not severely damaged and the radioactive fission products are safely contained.

The main reason for all of these systems is to make sure that the reactor fuel keeps its integrity so that the fission products are kept safely within the fuel matrix. However, it is always possible for the core to be damaged in spite of these systems, so a final defense is to isolate the fission products from the environment by using multiple layers to contain them. The first layer of containment is the reactor fuel itself, and as long as the fuel is intact, the fission products are safe. If the fuel is breached, as happened at Three Mile Island, the fission products will flow into the reactor plant, and the piping itself acts as the second layer of containment. As long as the reactor system piping retains its integrity, the fission products will remain isolated from the environment. However, this too can fail, and there have been instances of broken reactor system components that release reactor coolant into the reactor compartment. Thus, the entire building in which the reactor is located constitutes yet another layer of containment, the final layer. In the US, the reactor containment buildings are designed to stringent criteria, ensuring their ability to withstand tornados, earthquakes, terrorist attacks, and even airplane crashes. As long as the outmost containment remains intact, the worst of the fission products will remain safely isolated from the environment. Typical reactor containment buildings consist of a reinforced concrete building that is kept at a slight negative pressure so that radioactivity will not leak out. Doors into the containment are

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airlocked, and at least one door must be shut at all times. Other penetrations (cables, pipes, ventilation ducts) are sealed to prevent leakage along the outside of the pipes or cables, and the pipes and ducts themselves are equipped with valves that automatically shut if high radiation levels, excessive pressures, high temperatures, or other conditions exist that are indicative of a reactor plant problem. In fact, at Three Mile Island, virtually the entire reactor core was destroyed and the highly radioactive coolant leaked from the reactor plant into the containment building. However, because of the containment design, the radioactivity largely stayed put, and very little was inadvertently released to the environment. By comparison, the Chernobyl plant lacked a containment building and the accident there released enormous amounts of radioactivity that were tracked throughout the Northern Hemisphere. b) Types of nuclear reactors There are a number of ways to use a self‐sustaining chain reaction to generate

electrical energy, and there are several competing designs for nuclear reactors. However, all commercial nuclear power plants have some points of similarity. In this section, we will first look at these similarities, followed by a brief description of the most important nuclear reactor plant designs, their strong and weak points, important design characteristics, and their “popularity” in the world nuclear energy picture.

i) General nuclear reactor plant design Regardless of details of engineering, all commercial nuclear reactor plants share a

commonality of purpose; to produce electrical energy as effectively and efficiently as possible. To do so, they must accomplish the following tasks:

1. Maintain a self‐sustaining nuclear chain reaction for prolonged periods of time 2. Contain fuel and fission products safely 3. Provide adequate margins of safety against accidents 4. Remove heat from the core 5. Convert the thermal energy of the reactor core into electrical energy

Of these, we have already discussed the first three in preceding sections and they will not be further discussed here.

The fission process generates a tremendous amount of energy and this energy is deposited in the fuel in the form of heat. Left uncooled, fuel temperature would easily rise to the point of melting the fuel, destroying the reactor core as happened at Three Mile Island. However, this heat is the reason for building a nuclear reactor in the first place. The trick is to safely transfer the heat of nuclear fission to a place where it can be

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used to generate electricity. All commercial nuclear reactors do this by circulating some cooling fluid through the nuclear reactor core to transfer the heat of nuclear fission to a turbine, which spins to generate electricity. The fluid varies, and different reactor designs use water, liquid metal, liquid salt, and helium. The method of generating electricity varies as well; some reactors boil water directly to make steam which turns the turbine, some use the expansion of hot gas in the reactor core, some pass hot liquid through a heat exchanger to generate steam. In spite of these differences, though, all commercial nuclear reactors need to accomplish the same goal – to use the heat produced by nuclear fission to cause a turbine to spin and generate electricity. (Military nuclear reactors produce steam for propulsion and other reactors produce neutrons for research, radioactive isotopes for medical or research use, or plutonium and tritium for military purposes). There are seven types of reactors which can be used commercially. Pressurized water reactors (PWR), boiling water reactors (BWR), gas cooled reactors (GCR), liquid metal fast breeder reactors (LMFBR), heavy water reactors (HWR), graphite moderated reactors (LWGR), and pebble bed moderator reactors (PBMR). The following table describes some of the relevant operating parameters and environmental discharges of each of these, and the reference note cites where additional information can be found. Reactor type

Coolant Temp‐erature (C)

Pressure (atm) 1

# in use 2

# on order

GWe generated

Annual Emissions (TBq)

Population dose (person‐Sv)

PWR Light water 325 150‐155 256 33 167.7 196 14.7 BWR Light water 290 70‐75 92 6 61.6 42.8 58 GCR Helium 740 45‐50 32 0 9.2 32.7 23.2 LMFBR Sodium 535 1 2 3 0.44 0.57 0.17 HWR Heavy

water3 310 100‐105 43 9 12.4 3940 53.3

LWGR Light water 280 60‐65 13 1 7.8 105 7.86 PBMR Helium 870 0 2 0 4 4

Totals N/A N/A N/A 438 54 259.2 4212 157.2 5

1 One atmosphere is equal to 14.5 psi, 101.325 kPa, or 760 torr 2 As of July, 2001 3 Heavy water consists of water containing molecules of deuterium (which has a

proton and a neutron in the nucleus) instead of hydrogen (which has only a proton)

4 The first PBMRs are on order, but are not yet completed. Accordingly, the energy output of these reactors cannot yet be reported.

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5 The dose reported is spread across the entire exposed population which is several hundred million people. Accordingly, the dose to any single person is very small.

Nuclear News World List of Nuclear Reactors, March 2001 (Reactor plant statistics) UNSCEAR 2000 Report to the UN General Assembly (emissions) A Guidebook to Nuclear Reactors (operating parameters) Nuclear Engineering, Ronald Knief (operating parameters) 11. Environmental Issues

Any discussion of the merits of nuclear power as a source of energy invariably

includes a discussion of its environmental impact. On the one hand, anti‐nuclear extremists claim that nuclear power plants and their waste are causing irreversible damage to the earth’s environment. On the other hand, pro‐nuclear extremists claim that nuclear energy may be the best way to reduce environmental degradation and even to undo some environmental damage. There is some merit to arguments made by both sides, as well as some fallacies. However, rather than attempt to resolve this debate, this section will simply attempt to lay out the scientific evidence in as unbiased a manner as possible. In particular, we will examine the environmental impacts of uranium mining, nuclear reactor operations, the biological risks of exposure to radiation, and radioactive waste. With respect to accidents during reactor plant operations, it should suffice to note that, in about 50 years of nuclear reactor plant operations, there has been only one serious accident, Chernobyl, that has released damaging amounts of radiation to the environment, and the circumstances that made this accident so damaging are not likely to be repeated. a) Uranium mining As noted in Section 3d, uranium is mined in several locations around the world, and

uranium mining often results in the generation of large amounts of mine tailings. What was not mentioned is that uranium is a naturally occurring element that is found in, quite literally, virtually every bit of rock and soil on Earth. Uranium ore is simply a rock formation in which the uranium has become concentrated to an unusual degree. However, there is no doubt that mining uranium produces a great deal of waste rock, just as happens when mining copper, molybdenum, gold, and any other metal. With uranium mine tailings, the primary concern is the emission of radon, a radioactive gas, from the mine tailings, and many people worry that the radon can damage the health of nearby residents. For this reason, many US mines are required to take measures to reduce radon emissions from their uranium mine tailings. However, it must also be noted that the radon is rapidly diluted in the air, and concentrations are reduced to

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levels that cannot be distinguished from naturally occurring levels at distances of a kilometer or so from the tailings.

Another concern with using uranium as a fuel is that it might run out. For this reason, the International Atomic Energy Agency has performed periodic assessments of the economically recoverable uranium reserves and they have determined that they are adequate to meet projected energy requirements for at least the next 50‐100 years. By comparison, it is thought that petroleum reserves may run out in 50‐60 years, natural gas in 70‐90 years, and coal in 250‐300 years at current rates of consumption and estimated global reserves for each of these energy sources. Some potential sources of energy, such as solar, wind, geothermal, and hydroelectric energy will not run out, but these forms of energy are likely to remain niche providers for some time for many reasons that are beyond the scope of this chapter. It is possible that new technologies will, at some point, make these options viable for large‐scale energy production, just as it is possible that hydrogen fusion will become economically viable, but there is no way today to guess when that day will arrive. b) Nuclear reactor plant operations Nuclear reactors unavoidably emit radiation and radioactivity when they operate.

This radiation and radioactivity enters the environment and results in slight increases in the amount of radiation to which nearby residents and the environment are exposed. This section will address the amount of radioactivity released by nuclear reactors as compared to other methods of energy production and the next section will discuss the biological and environmental effects of this exposure.

It is inevitable that minor amounts of fission products will enter the reactor coolant, even in the absence of defects in the fuel cladding. Trace amounts of uranium are almost always present in the clad, and this uranium can fission too, and the fission products can enter the reactor coolant. In addition, the high levels of neutron radiation in the reactor core cause gases dissolved in the reactor coolant to become radioactive. Then, when the coolant is removed from the reactor plant for any reason, these radioactive gases can escape into the environment. However, this is expected, the potential release points are continually monitored, and the radiation exposure to the public is very small. In fact, in the US, dose to the public is limited to 0.25 mSv annually, which is less one tenth the exposure from natural sources of radiation. According to the United Nations Science Committee on the Effects of Atomic Radiation (UNSCEAR) in their 2000 report to the General Assembly, the total radiation dose from nuclear reactors worldwide is 157.2 person Sv per year and the total amount of radioactivity released by the world’s commercial nuclear reactors is about 4212 TBq per year.

Although coal, petroleum, and natural gas are not nuclear, all of these sources of energy contain naturally occurring radioactive materials, and these are released to the

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environment when they are burned to generate electricity. According to the US Environmental Protection Agency (USEPA), the total radiation dose to members of the US population from these three sources of energy is greater than the dose from nuclear power plants. A summary of these exposures is contained in the following table, and is compared to the total population dose from natural background radiation to the population of the US. As this table shows, nuclear energy exposes people to no more radiation, even on a per‐megawatt basis, than do other forms of energy. It is also apparent that nuclear energy releases smaller amounts of greenhouse gases than does burning fossil fuels, and that uranium reserves are likely to last longer than those of other sources of energy currently available. Form of energy

TW – yr produced 1, 2

Total population radiation dose (person‐Sv)3, 4

Radiation exposure (person‐Sv per TW‐yr)

Global reserves (years)1

Greenhouse gas emissions (Gtons of carbon)1, 5

Nuclear 0.60 258 430 50‐100 0 Coal 2.89 14,161 4900 250‐300 638 Petroleum 4.06 6.61 1.63 50‐60 110 Natural gas

2.25 ~500 6 ~200 70‐90 72

1 From International Energy Energy Annual, DOE 2001 2 1 TW‐yr is equal to 1012 watts of energy production produced over the course of 1

year 3 By comparison, radiation dose from exposure to natural background radiation is

estimated to be 144 x 106 person‐Sv annually (UNSCEAR 2000) across the entire human population

4 Values for radiation dose from fossil fuel plants are calculated from values given in NCRP Report #95

5 1 Gton is equal to 109 tons, or 1012 kg of carbon emissions 6 The majority of radiation dose from natural gas combustion comes from home use

for cooking and heating; power generation produces very little radiation exposure. The values given are for electrical power generation only; other uses of natural gas yield about 2000 person‐Sv per year across the world’s population, and disposal of radium‐containing wastes adds a further, unknown dose

c) Health Effects The effects of exposure to high levels of radiation are well‐known and unequivocal.

Studies of the survivors of Hiroshima, Nagasaki, workers at Chernobyl, and people

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involved in accidents involving radiation or radioactivity show conclusively that exposure to high levels of radiation will cause cancer or radiation sickness, depending on the amount of radiation received. In general, after a single person’s exposure to more than 1 Sv (100 rem) of radiation exposure, people will start to feel ill and their risk of cancer will increase by a factor of 2 or more. After exposure to more than about 4 Sv (400 rem), about half of those exposed will die of radiation sickness, and 100 percent of those exposed to doses of 9‐10 Sv (900 – 1000 rem) will die.

The effects of radiation exposure to individuals (described in the preceeding paragraph) should not be confused with the effects of collective radiation exposure, such as reported in preceeding tables. Collective radiation exposure is a measure of radiation exposure to an entire population from a given activity. For example, if a group of 100,000 people each receive a dose of 0.01 Sv (the annual dose limit to the general public from all sources of man‐made radiation combined) then the collective dose will be 1000 person‐Sv. This dose is far in excess of that which will cause a fatality to a single person, yet no individual is expected to fall ill from their individual low exposure.

What is not known as precisely are the effects of exposure to lower levels of radiation, such as those in the vicinity of operating nuclear power plants. The reason these effects are not as well known is that the effects, if any, are very small so they are very difficult to measure with accuracy. In fact, for every report that concludes that this level of radiation exposure is harmful, there is another report showing there are no effects, and even some reports showing possible beneficial effects from exposure to low levels of radiation. The manner in which we respond to low levels of radiation exposure will likely not be known for many years.

Most governments set their radiation regulations under the assumption that all exposure to radiation is potentially harmful, and they further assume that the risk of getting cancer from radiation is directly proportional to the amount of dose received. Called the Linear, No‐Threshold (LNT) model, it is the most conservative of the major hypotheses in use because it assumes the greatest risk from radiation. Under the LNT model, a person receiving 1 mSv of radiation exposure per year in excess of background levels will have one chance in 10,000 of developing cancer as a result of that exposure. By comparison, the overall average cancer rate in the US is about 1600 in 10,000. Other models, which suggest there may be a threshold level below which radiation exposure is not harmful, also suggest this risk may be even lower. There are a few conclusions that can be drawn from this discussion:

• Nuclear power plants produce less public radiation exposure than fossil fuel plants, so the cancer risk from fossil fuel plants is higher than from nuclear power plants

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• Neither fossil fuel plants nor nuclear power plants emit enough radiation to constitute a cancer threat to the public, even under the most conservative scenario

• We all receive more radiation from natural sources (including naturally occurring radioactive potassium in our bodies) than from either nuclear or fossil fuel power plants

d) Radioactive waste Another area of some controversy is the safe disposal of radioactive waste, in

particular, spent reactor fuel. Many anti‐nuclear activists are concerned that radioactive waste cannot be safely isolated from the environment and that it poses a long‐term risk to the environment and to nearby residents. They are also concerned that spent reactor fuel, which is often highly radioactive, poses a risk to the population during transport and subsequent burial. On the other hand, the nuclear power industry feels it is quite possible to safely sequester both “routine” radioactive wastes and spent reactor fuel for prolonged periods of time, during which the radioactivity present will decay to safe levels.

Those opposed to nuclear energy are particularly concerned about the possibility that radioactive waste disposal sites will leak, letting radioactivity enter groundwater systems, the atmosphere, farms, and so forth, and that this radioactivity will both pollute the environment and cause lasting health effects to nearby residents. In support of their claims, they note the difficulty of designing a landfill to successfully isolate waste for a period of decades, and that many items of radioactive waste remain radioactive for centuries or longer. These arguments gain added strength when discussing the disposition of spent nuclear reactor fuel, which contains plutonium as well as some radioactive elements that remain radioactive for millennia.

On the other hand, advocates of nuclear energy point out that virtually all radioactive waste sites are required to have elaborate containment and monitoring systems that will catch any radioactive leakage long before it can become a health concern. In addition, they note that the great majority of radioactive waste has a relatively short half‐life and will decay within a few decades or, at most, a few centuries, so that most of the radioactivity that is put into a radioactive waste disposal facility will vanish long before it can escape into the environment. With respect to spent reactor fuel, although it will remain radioactive for thousands of years, the majority of radioactivity is again gone within a few centuries, so the fuel will be dangerous for only a relatively short period of time, and there is little doubt that it can be safely stored for that time. Finally, the plutonium found in spent fuel, although dangerous, is no more dangerous (and no more toxic) than many items found in chemistry labs, and does not pose any extraordinary hazard.

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Most nations that operate nuclear power plants have built repositories for both radioactive waste and spent reactor fuel. Virtually all radioactive waste facilities are designed to allow continuous monitoring for leaks as well as environmental monitoring for accidental releases. Most, too, have strict waste acceptance criteria, to ensure that the waste received will remain safely within its package for decades or longer. And, in the case of spent reactor fuel or very highly radioactive waste, many nations have built or are in the process of constructing disposal facilities buried deeply underground in geologically stable areas that fully capable of safely containing the waste for millions of years. Finally, many scientists take some solace in noting that radioactive waste at the Oklo natural nuclear reactor (discovered in 1972) has remained stable for almost 2 billion years, in spite of being located in porous sandstone that has frequently been below the water table. 12. Conclusions For better or for worse, nuclear reactors provide much of the world’s energy needs,

and provide an important role in powering naval vessels in several of the world’s most powerful nations. Like any source of energy, nuclear power plants have both good and bad, and we can only hope that the benefits we derive from the continued use of nuclear energy outweigh the negative. In many instances, this seems to be the case, but like any technology, some risks are unavoidable and will always remain. There are several points in favor of continued use of nuclear energy: 1) Nuclear power plants release no more radioactivity (and possibly less) than

comparable fossil fueled power plants 2) Nuclear power plants release far fewer greenhouse gas emissions than do fossil fuel

plants 3) The world’s uranium reserves are likely to last for a much longer period of time than

fossil fuel reserves 4) Nuclear energy is more dependable than alternate sources of energy and relies on

proven technology, rather than hoped‐for future breakthroughs 5) Nuclear power plants are no more dangerous than are fossil fuel power plants and

have, in fact, suffered fewer accidents than have other sectors of the power industry However, there are also some drawbacks to nuclear energy as a source of power: 1) Nuclear power plants are often more expensive to build than are other types of

power plants

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2) Spent reactor fuel can provide a source of plutonium for nuclear weapons production

3) Disposal of radioactive waste is neither simple nor inexpensive 4) In some nations, public opposition to nuclear power is organized and vociferous,

making political support of nuclear energy problematic at times

Balancing these issues is difficult, even under the best of circumstances. It is reasonable to point out that nuclear power has been in use for over a half century, and (counting military nuclear reactors), nearly 1000 nuclear reactors have operated at one time or another for the past 50 years. In addition to the energy produced by these nuclear reactors, millions have benefited from the isotopes produced which are used in medicine and research. Against that, we have the indisputable tragedy of Chernobyl and the expense (even in the absence of risk) of Three Mile Island. It seems likely that nuclear energy can play a safe and important role in meeting the world’s increasing energy needs, in spite of the drawbacks noted above, but this is a decision that must be made by the world’s citizens and their governments. Bibliography: DOE; International Energy Outlook 2001; March 2001 (available on the World Wide

Web at www.eia.doe.gov/oiaf/ieo/index.html) DOE; International Energy Annual 1999; February 2001 (available on the World Wide

Web at www.wia.doe.gov/iea) Eisenbud and Gesell, Environmental Radioactivity from Natural, Industrial, and

Military Sources, Fourth Edition, Academic Press, 1997 Knief, RA; Nuclear Engineering, Second Edition, Hemisphere Publishing Company,

1992 NCRP, Report No. 95, Radiation Exposure of the US Population from Consumer

Products and Miscellaneous Sources, 1987 NCRP, Report No. 93, Ionizing Radiation Exposure of the Population of the United

States, 1987 Nero, AV; A Guidebook to Nuclear Reactors, University of California Press, 1979 Nuclear News; Third Annual Reference Issue and World List of Nuclear Power Plants,

March 2001 Ohio State University Fact Sheets on Radioactive Waste (found on the World Wide Web

at http://ohioline.osu.edu/lines/ennr.html and at http://www.ag.ohio‐state.edu/~rer/index.html)

UNSCEAR, Report to the General Assembly, Volumes 1 and 2, United Nations Science Committee on the Effects of Atomic Radiation, 2000

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APPENDIX C: THE NATURAL NUCLEAR REACTOR AT OKLO: A COMPARISON WITH MODERN NUCLEAR REACTORS

All rights are reserved by the Author. Contact the author for permission to use this article for any purpose other than educational.

Editors note: Despite some claims, there is no evidence or even credible theory that the Oklo nuclear reactor was anything but a natural phenomenon. The 6 reactor zones are spread over a huge area that was a uranium mine during the time it was first discovered. The reactor zones were the result of natural physical processes, active for thousands of years. It should also be noted that the possibility of natural nuclear reactors was first postulated by P. K. Kuroda (1956).

i) Abstract

Uranium contains only one naturally occurring isotope, 235U, which will sustain a nuclear chain reaction using normal water to moderate and reflect neutrons. At present, this isotope is present in low abundance (0.72%), requiring enrichment to 3% or greater for effective use in commercial nuclear reactors. Two billion years ago, however, the natural abundance of 235U was approximately 3%. Evidence indicates that a rich uranium deposit in Gabon, West Africa achieved nuclear criticality and operated for tens of thousands of years or longer. Comparing the geometric and nuclear characteristics of the Gabon reactor with those of modern, artificial nuclear reactors supports this possibility. An examination of rare earth elements and 235U abundance in the rocks that comprise the reactor zone confirm that a nuclear reactor did operate at this site about 2 billion year ago (Ga), using surface and ground waters to moderate and reflect fission neutrons in order to sustain the chain reaction. Finally, it is apparent that 239Pu was produced in measurable quantities, suggesting that uranium is not the heaviest naturally occurring element known.

ii) Introduction

Natural uranium is composed of three major isotopes, 238U (abundance = 99.2745%), 235U (abundance = 0.7200%), and 234U (abundance = 0.0055%). The isotopic composition of uranium is thought to be homogeneous globally (Faure, 1986). However, uranium in a rich deposit located in Gabon, West Africa, was found to have 235U abundances as low as 0.440%.

Subsequent investigations indicated the presence of isotopes of neodymium and other elements which, in conjunction with the lower 235U abundance, suggest that a natural nuclear reactor existed in the past. Other zones contained slightly elevated abundances of 235U. These are thought to represent the decay products of 239Pu, formed by neutron capture of 238U during reactor operation. Similar reactions occur in modern nuclear reactors and, indeed, form the basis of plutonium production. This reaction (including the subsequent decay of 239Pu, to = 2.411 x 104 a) is:

The absence of 236U (t1/2 = 2.342 x l07a) indicates that induced fission stopped at least 108 years ago. Dating of the strata in which the reactor is found indicates an age of approximately 1.8 Gy (Cowan, 1976). At that time, 235U had an abundance of approximately 3%, owing to its shorter half‐life relative to 233U (t1/2=7.04 x l08 yrs and 4.468 x l09 yrs, respectively). This paper discusses briefly the geologic setting of the Gabon reactor and compares its nuclear characteristics with those of modern, manmade nuclear reactors in terms of fuel loading, geometry, neutron flux, power, and uranium enrichment.

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iii) The Gabon Reactor

The Gabon reactor consists of several mineralized zones of uranium minerals in sandstone and conglomerate. The uranium probably originated in nearby igneous deposits, dissolved in oxygenated surface waters, and was deposited at an oxidation‐reduction front (similar to the roll‐front deposits in Texas). The mobilization, therefore, probably did not commence until there was sufficient free oxygen in the earthʹs atmosphere to allow oxygenation of surface waters (about 2 Ga). The formation in which the deposits reside was deposited about 1.74 (+0.20) billion years ago (Ga). At this time, 235U had a relative abundance of approximately 3%.

According to Neuilly et al, (1972), ʺthe deposit is stratiform. It is located within the sandstone which forms the basis of the Francevillian (sedimentary basin). The sedimentological characteristics of these sandstones suggest fluviodeltaic deposit conditions. Formed essentially of detrital quartz with some accessory feldspars, they have a cement consisting of secondary silica, phyllite, and organic matter (of asphaltic type). The uraniferous mineralization consists primarily of oxides (uraninite, pitchblende). It occurs in the cement which it may almost totally replace in zones bearing the highest grade uranium ore. It is associated with a few sulfides (pyrite, galena, etc).

ʺThe thickness of the mineralized layer ranges from 5 to 8 meters. The layer is locally affected by a tectonic movement whose activity served to uplift the sandstone formations bordering on an elongated depression in the sole. The uraniferous mineralization and any redistributions it may have undergone as a result of groundwater flow, appears almost contemporaneous with the sedimentation of the Francevillian.ʺ

A total of 16 reactor zones have been found in the primary location, and another reactor zone identified about 20 km away (the Bangombé reactor). Current thinking is that a zone of petroleum formation produced an oxygen‐free zone because oxygen in groundwater would have been removed by chemical reactions with the petroleum. Uranium, which is fairly highly soluble in oxygenated waters, is insoluble in anoxic waters, so the uranium precipitated out of solution as it crossed the oxidation/reduction front near the petroleum deposits (Janeszek 1999). This is what formed the reactor zones.

An illustration of the reactor is shown in Figure 1 (from Cowan, 1976).

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iv) Artificial Nuclear Reactors

A diagram of a typical nuclear reactor fuel assembly is shown in Figure 2 (from Knief, 1990).

Nuclear reactors produce power by controlled induced splitting (fission) of fissionable atoms. Typically these are 235U or 239Pu (although the latter is forbidden for use as reactor fuel in the US). The neutron capture process, however, creates other fissionable nuclides such as 241Pu and 243Pu which, by the end of core life, can contribute significantly to a reactorʹs power production (Knief, 1992). The nuclear fission process is more efficient with low energy (thermal) neutrons, although fast fission of both 235U and 238U does occur in the fuel.

Since fission neutrons are produced with high energies, they must be slowed down (thermalized) in a moderator in order to be effective at inducing fission. This moderator is typically normal water, although graphite is also used, particularly in reactors intended for plutonium production. Submerging the fuel in the moderator increases reactor efficiency (and decreases required fuel loading) by reflecting escaping neutrons back into the core, allowing them to participate in the fission process.

Due to the low abundance of 235U at present, natural uranium does not undergo a sustained nuclear chain reaction with natural water as a moderator. This is due to the relatively high thermal neutron cross section of hydrogen, which causes the absorption of neutrons that would otherwise cause fissions. Light water reactors, such as the Russian RBMK (graphite moderated) or the heavy water moderated CANDU (CANadian Deuterium Uranium) reactor can maintain self‐sustaining chain reactions with natural uranium. Most nuclear reactors operate with uranium fuel which has been enriched in 235U, typically between 3% and 5% (Knief, 1992), although research reactors have 235U enrichments of up to 20%. Military reactors may have enrichments in excess of 90% in order to achieve the power densities necessary for their effective use.

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v) Reactor Physics

Nuclear reactors operate and produce power by maintaining a nuclear fission chain reaction. In order for the reactor to maintain a constant power level, as many neutrons must be produced in each generation as are absorbed or escape from the core.

Of those neutrons which do not escape the core, many are absorbed, but do not cause fissions. Absorbers can be metal non‐fuel components of the core, hydrogen in the moderator, fission products that have a high neutron capture cross section, ʺcontrol poisonsʺ which are placed into the core to shape the neutron flux and to control reactor power, or uranium atoms which do not fission. When the neutron population of the core remains constant the reactor is said to be critical. A growing neutron population is characteristic of a supercritical reactor while a shrinking neutron population makes a reactor subcritical. Therefore, all reactors, when operating at constant power, are critical.

Water is used to cool the reactor core and also to help control reactor power. Neutrons from fission will collide with the atoms in water, losing energy with each collision – this is called “moderation”. This slows the neutrons to the point at which they can cause fission more efficiently. Water will also reflect escaping neutrons back into the reactor fuel, reducing the loss of neutrons from the fuel‐bearing region. This effectively increases the neutron population in the core and helps to reduce the amount of fuel needed for a critical assembly. If water is lost from the core, the amount of moderation and reflection is greatly reduced and the reactor shuts down.

The term keff refers to the relative size of subsequent neutron populations. In a critical reactor, keff is equal to 1. The value of keff is determined by the ʺsix factor formulaʺ given below:

in which:

• (the fast fission factor) refers to the ratio of total fissions to those produced by thermal neutrons,

• Lf (the fast nonleakage factor) is the fraction of fast neutrons that do not leak from the core,

• P (the resonance escape probability) gives the fraction of neutrons which are not resonantly absorbed in the core,

• Lth (the thermal nonleakage probability) is the fraction of thermal neutron which do not leak from the core,

• f (thermal utilization factor) is the fraction of neutrons which are absorbed by the fuel, and

• (the fission yield) is the average number of neutrons produced for each thermal neutron captured by fuel material.

The factors Lf and Lth, are included in the ʺbucklingʺ term (B2); a factor based on core geometry which describes the loss of neutrons from the core through core boundaries. A sphere has the smallest buckling term which is given by:

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with R being the radius of the sphere. The presence of a reflector reduces the geometric buckling, as does increasing the size of the fuel bearing region.

As shown by the six factor formula and the buckling equation, there are several ways to increase the probability that a reactor will be able to ʺgo criticalʺ. These are:

• Increase the size of the fuel bearing region in order to reduce the buckling (Lf and Lth, terms)

• Increase the ʺfuel to waterʺ ratio to increase the probability that thermal neutrons will be absorbed by fissionable material

• Minimize the number of non‐fissionable atoms with high neutron capture cross sections (poisons)

• Provide a reflector to further reduce buckling • Use a moderator which is close to the mass of a neutron in order to have the

maximum energy transfer per collision, reducing the chance for resonance absorption of the neutrons

vi) Reactor Characteristics

Nuclear power reactors typically operate at pressures and temperatures in excess of 1500 psi and 500 °F, although most research reactors operate at atmospheric temperatures and pressures. Fuel loading varies greatly, from tens of kilograms to hundreds of metric tons. Neutron fluxes, too, vary based on the core characteristics, but are typically between 1013 and 1015 n cm‐2 sec‐1. A commercial nuclear reactor typically has several hundred fuel assemblies, each containing 50 or 60 fuel pins that are approximately 20 feet long and arranged in a circle approximately 20 feet in diameter (Nero, 1979). Military reactors, due to their higher fuel enrichment, are much smaller, as are research reactors, due to their higher 235U enrichment and smaller power output.

Critical mass experiments have shown that a uranium sphere with a 3% enrichment that is fully water reflected will have a minimum critical mass of approximately 2 kg in a heterogeneous reactor and nearly 3 kg in a homogeneous reactor. This minimum critical volume will be just large enough to go critical one time and will not sustain a chain reaction for a prolonged period of time. A larger reactor will have either a higher power output, will be able to operate longer, or both.

vii) Fission Products

Each fission produces approximately 200 MeV in the form of gamma radiation and the kinetic energy of two fission fragments that are produced. This energy is dissipated in the fuel material as the fission fragments interact with surrounding atoms and slow down. The fission fragments are nearly always of uneven mass with varying probabilities of production. Fission yields are highest for nuclides with masses of 95 and 140 amu (when fissioning 235U). The distribution of fission products is characteristic for each fissionable nuclide and can be used to identify the fuel.

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Some fission products have high thermal neutron capture cross sections. These can accumulate and poison the chain reaction resulting in either lower power or increased neutron flux to compensate. They are included in the thermal utilization factor in the six factor formula. The final ʺmixʺ of fission products, therefore, will reflect the fission yield of each atomic mass number, the half‐life and neutron absorption cross section of members of each fission product chain, and the neutron flux in the reactor.

viii) Nuclear Characteristics of the Gabon Reactor

At the time that the Gabon reactor went critical, the abundance of 235U was 3%, similar to that in current commercial nuclear reactors. The approximate shape of the reactor zones is that of a compact mass of uranium oxide surrounded by porous rocks, which were presumably hydraulically connected to surface or ground water, allowing moderation and reflection of the neutrons produced by spontaneous fission or cosmic ray induced fission.

The relatively large size and spherical shape of the uranium bearing region reduced buckling. When the surrounding porous rocks were saturated with water, the subsequent moderation and reflection allowed the reactor to achieve criticality. It is likely that criticality was not continuous. As the reactor power increased, the water moderator would heat, reducing its density and its effectiveness as a moderator and reflector. This process, known as a negative temperature coefficient, helps to control power during transient conditions in manmade nuclear reactors.

If sufficient power was produced the reactor would have lost moderation and reflection, resulting in a shutdown. Until short lived fission product poisons decayed away, even immediate re‐saturation with water may not have resulted in restarting the nuclear chain reactions. Therefore, the reactor probably did not operate continuously, but at discrete intervals with the operating time determined by the power output, water supply pressure and temperature, and water flow through the reactor. The duration of the shutdown periods would have been determined by the buildup of fission product poisons and the length of time required to replace the moderator (if it boiled away) or to cool it sufficiently to resume the reaction.

In fact, a recent paper (Meshik et al, 2004) looked at the operation of the Oklo reactor. The authors deduced that the reactor likely operated cyclically, operating for a half hour until accumulated heat boiled away the water, then shutting down for up to 2.5 hours until the rocks cooled sufficiently to allow water saturation again. The authors also note that the majority of fission products from these nuclear reactions have remained in place for nearly 2 billion years, in spite of their location in fractured, porous, and water‐saturated sandstone for most of that time.

In all, the Oklo reactor is thought to have operated for a period in excess of 150,000 years, based on the quantity of fission products present. The total neutron fluence is thought to have been about 1021 neutrons per square cm over the life of the reactor, producing a total of about 15 GW yr of thermal energy. During this time it consumed an estimated 5‐6 metric tons of 235U, and producing an equal mass of fission products (de Laeter, et al, 1980). Meshick estimates an average operating power of about 100 kW, similar to that of modern research reactors.

Large core size, a low fuel/water ratio, and the presence of fission product poisons would require a large neutron flux with respect to reactor power in order to maintain reactor criticality. Neuilly et al (1972) estimated that the Gabon reactor had a thermal neutron flux of at least 109 neutrons cm‐2 sec‐1 and a total fluence of 1021 neutrons per square cm. By comparison, the complete fission of one kg of 235U in a nuclear detonation would release approximately 1026 neutrons in approximately one microsecond through an area of approximately 200 cm2 (Serber, 1992), giving a neutron flux of about 1030 neutrons cm‐2 sec‐1. Current nuclear reactors have neutron fluxes on the order of 1013 to 1014 neutrons cm‐2 sec‐1. The estimates for neutron flux are, therefore, not unreasonable, given the reactor characteristics of low power and large core size.

The rocks of the Gabon reactor indicate the presence of fission product nuclides in abundances which roughly tally with those expected of 235U induced fission. Most interesting are the isotopes of neodymium, which show enrichment in those mass numbers that are characteristic of uranium fission and depletion in those mass numbers which have the

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highest neutron capture cross sections. The distribution of these nuclides is described in Cowan (1976) and in Neuilly et al (1972). Also convincing is the enrichment in 235U and the presence of 232Th noted in some sections of the reactor zone. The 235U is thought to represent areas in which neutron capture by 238U produced 239Pu which subsequently decayed to 235U. Thorium is thought to result from the reaction:

Other compelling evidence of induced uranium fission is the negative correlation of uranium content versus 235U abundance through this deposit. This indicates that the areas of highest uranium content underwent the greatest depletion of fissionable material, which would be expected in a nuclear reactor.

ix) Conclusions

The reactor zones found in Gabon have the requisite physical and nuclear characteristics to form a self‐sustaining chain reaction, given the abundance of 235U present nearly 2 Ga. The compact mass of the reactor zones would have been conducive to minimizing buckling and maximizing thermal neutron utilization in the uranium while the surrounding sandstone and conglomerate would provide ample water to moderate and reflect neutrons, as is the case with artificial reactors today. The inventory of fission products and the slight excess of 235U noted in some zones support the conclusion that induced fission of 235U and neutron capture by 238U occurred in this area.

It also seems likely that other natural reactors were operational in the past. Other parts of the world have large, high assay deposits of uranium mineralization in sedimentary strata, so the circumstances which led to the formation of the Gabon reactor may not have been unique. It seems safe to assume that this process may have taken place throughout the history of the earth. Indeed, there are hints that a natural reactor was operational in the Colorado Plateau, based on a slight depletion of 235U in ore specimens there (Cowan, 1976). It may be that our knowledge of natural nuclear reactors is limited primarily by our explorations to date.

Finally, although the existence of transuranic elements was predicted by Goldschmidt, the common conception is that uranium is the heaviest natural element. Any nuclear reactor, however, produces transuranic elements via the neutron capture reactions mentioned above. Therefore, uranium may need to relinquish this position to plutonium or heavier elements.

b) Acknowledgements c) Many thanks to the following persons:

• Gunter Faure, my MS advisor for whose class this paper was originally written • Bruce Busby, for translating this into HTML and maintaining the posting on‐line • Jaroslav Franta, for providing feedback, suggestions, and corrections to the original

(as well as the drawing of an actual fuel assembly) • Graham Cowan, for helpful comments and corrections to the original

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References Cowan, A Natural Fission Reactor, Scientific American, 7/76: 3647

de Laeter, J.R., Rosman, K.J.R, Smith, C.L., The Oklo Natural Reactor: Cumulative Fission Yields and Retentivity of the Symmetric Mass Region Fission Products, Earth and Planetary Science Letters, 50(1980) 238246

Faure, G., Principles of Isotope Geology, John Wiley and Sons, 1986

Janueczek, J. Mineralogy and Geochemistry of Natural Fission Reactors in Gabon; Chapter 7 in Reviews in Mineralogy vol. 38, Uranium: Mineralogy, Geochemistry, and the Environment, Burns PC and Finch R (eds); Mineralogical Society of America; pp 321‐392. 1999

Knief, R. A., Nuclear Engineering: Theory and Technology of Commercial Nuclear Power, Hemisphere Publishing Corporation, 1992

Kuroda PK; On the Nuclear Physical Stability of the Uranium Minerals. Journal of Chemical Physics vol 25, pp 781‐782. 1956

Meshik, A.P. et al., Record of Cycling of the Natural Nuclear Reactor in the Oklo/Okelobondo Area in Gabon. Physical Review Letters vol 93, No. 18. 2004

Meshik AP et al., Anomalous xenon in zone 13 Okelobondo. Geochemica et Cosmochimica Acta vol. 64 no. 9, p 1651‐1661. 2000

Nero, A.V., A Guidebook to Nuclear Reactors, University of California Press, 1979

Neuilly, M., et al, Evidence of Early Spontaneous Chain Reaction found in Gabon Mine, excerpts from press conference regarding Geological and Mineral Documentation published by Commissariat a lʹEnergie Atomique, 1972

Paxton, H.C. and Pruvost, N.L., Critical Dimensions of Systems Containing 235U, 239PU, and 233U, Los Alamos National Laboratory, 1986

Serber, R., The Los Alamos Primer: The First Lectures on how to Build and Atomic Bomb, University of California Press, 1992

This document was originally written in 1996 as a paper for a graduate class in environmental isotope geology, taught by Gunter Faure. Several e‐mail correspondents have noted errors in the original paper; this revision accounts for their comments as well as for research papers published between 1996 and April 2005.

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Appendix D: Regulatory Guide 8.36 - Radiation Dose to the Embryo/Fetus

July 1992

d) A. INTRODUCTION

Section 20.1208 of 10 CFR Part 20, ʺStandards for Protection Against Radiation,ʺ requires that each licensee ensure that the dose to an embryo/fetus during the entire pregnancy, from occupational exposure of a declared pregnant woman, does not exceed 0.5 rem (5 mSv). Paragraph 20.1208(b) requires the licensee to make efforts to avoid substantial variation above a uniform monthly exposure rate to a declared pregnant woman that would satisfy the 0.5 rem (5 mSv) limit. The dose to the embryo/fetus is to be the sum of (1) the deep‐dose equivalent to the declared pregnant woman (10 CFR 20.1208(c)(1)) and (2) the dose to the embryo/fetus from radionuclides in the embryo/fetus and radionuclides in the declared pregnant woman (10 CFR 20.1208(c)(2)).

This guide is being developed to provide guidance on calculating the radiation dose to the embryo/fetus. Regulatory Guide 8.13, ʺInstruction Concerning Prenatal Radiation Exposure,ʺ provides instructions concerning the risks associated with prenatal radiation exposure.

Any information collection activities mentioned in this regulatory guide are contained as requirements in 10 CFR Part 20, which provides the regulatory basis for this guide. The information collection requirements in 10 CFR Part 20 have been cleared under OMB Clearance No. 3150‐0014.

e) B. DISCUSSION

Calculating the radiation dose to the embryo/fetus from internally deposited radionuclides requires quantitative information about maternal radionuclide intake, placental transfer and kinetics, and resulting embryo/fetus radionuclide concentrations. Intakes of radioactive material occurring prior to the pregnancy may also be important if these materials remain in the pregnant woman during all or part of the gestation period. Transfer kinetics from the mother to the embryo/fetus are modeled as a function of stage of pregnancy, route of intake by the pregnant woman, and time after intake. The stage of gestation (or fetal development) is an important parameter in estimating radionuclide concentrations in the embryo/fetus. The geometry of the embryo/fetus (i.e., size and weight) affects the radionuclide dosimetry.

It is recognized that calculation of prenatal radiation doses from internally deposited radionuclides has many associated difficulties, including a lack of quantitative information about prenatal radionuclide concentrations and transfer across the placenta. The International Commission on Radiological Protection (ICRP) in Publication 56 (Ref. 1) states that, for most radionuclides, preliminary estimates from dosimetric and biokinetic models indicate that the dose to the embryo can be approximated by the dose to the uterus. The dose to the fetus is dependent upon the activity present in both fetal and maternal tissues. ICRP Publication 56 (Ref. 1) also states that, for most radionuclides, the dose to fetal tissue will be similar to or less than the dose to the corresponding maternal tissues.

The current methods available for assessing the radiation dose to the human embryo/fetus from internally deposited radioactive materials in the pregnant woman are subject to a number of uncertainties. Revison 1 to NUREG/CR‐5631, ʺContribution of Maternal Radionuclide Burdens to Prenatal Radiation Doses‐‐Interim Recommendationsʺ (Ref. 2), provides recommendations and methods for estimating the radiation doses to the embryo/fetus from internal radionuclides. In Revision 1 to NUREG/CR‐5631, a number of radionuclides were evaluated. To expedite efforts, the initial evaluation was directed to those radionuclides that were expected to be of greatest significance for prenatal

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exposure in the work environment. The radionuclides that were identified and included were 3H, 14C, 57Co, 58Co, 60Co, 89Sr, 90Sr, 106Ru, 125I, 131I, 132I, 133I, 134I, 135I, 134Cs, 137Cs, 233U, 234U, 235U, 238U, 238Pu, 239Pu, and 241Am. The methods of Revision 1 to NUREG/CR‐5631 are considered interim as efforts continue to further develop the bases and calculational methods for estimating prenatal radiation doses. Revision 1 to NUREG/CR‐5631 provides details of the data and bases for the dosimetric features that were used for the radionuclides listed above.

It is expected that the embryo/fetus dose assessment methods will evolve over the next several years as more research is conducted in this area. As additional research is conducted, better estimates of actual embryo/fetus doses resulting from the exposure of the declared pregnant woman will be possible. For internal doses, research that categorizes the degree of placental transfer, the resulting embryo/fetus/placenta concentrations, and the potential radiation exposures of the embryo/fetus from radionuclides in their more usual chemical forms should simplify assessment of the dose to the embryo/fetus based on the maternal exposure. The ICRP is considering the formulation of dose assessment methods specific for the embryo/fetus.

This regulatory guide provides acceptable methods that may be used in determining the dose to the embryo/fetus. For internal exposure, a simplified approach and a more detailed methodology are presented for conducting dose evaluations. The regulatory position specified in Section 1 provides guidance on the threshold criteria for use in determining when the dose to the embryo/fetus needs to be evaluated. The regulatory position specified in Section 2 presents a simplified approach for estimating the dose to the embryo/fetus from intakes by the declared pregnant woman. The regulatory position specified in Section 3 provides an alternative, more detailed methodology for a limited number of radionuclides, using the gestation‐time dependent dosimetric data from Revision 1 to NUREG/CR‐5631 (Ref. 2).

A graded approach for determining when to evaluate, with both a simple and more detailed dose assessment methodology, is provided. Both methods are acceptable for evaluating the dose to the embryo/fetus. It is recognized that some licensees will only need to demonstrate that the dose to the embryo/fetus is not likely to exceed the 0.05 rem (0.5 mSv) monitoring threshold of 10 CFR 20.1502, while other licensees may need to determine an embryo/fetus dose for demonstrating compliance with the dose limit of 10 CFR 20.1208 and the recordkeeping requirements of 10 CFR 20.2106(e).

Appendix A provides information on and a table of dose equivalent factors for use in approximating the embryo/fetus dose from radionuclides in maternal blood. Appendix B is a table of blood uptake fractions for ingested activity. Appendix C contains tables of gestation‐time dependent doses to the embryo/fetus following introduction of specified radionuclides and chemical forms into maternal blood. Examples of the use of dose assessment methods are provided in Appendix D.

The total radiation dose to the embryo/fetus is the sum of the deep‐dose equivalent to the declared pregnant worker and the dose to the embryo/fetus from intakes of the declared pregnant worker. If multiple dosimetric devices are used to measure the deep‐dose equivalent to the declared pregnant worker, the results of monitoring that are most representative of the deep dose to the embryo/fetus may be used. The licensee need not use the deep dose to the maximally exposed portion of the whole body of the mother as the deep dose to the embryo/fetus. The licensee may employ temporary or permanent shielding to reduce the deep dose to the embryo/fetus. Alternatively, deep dose to the embryo/fetus may be limited by placing more stringent restrictions on the exposure of the declared pregnant woman than on other members of the occupational work force.

As specified in 10 CFR 20.1208(a), the dose to the embryo/fetus from occupational exposure of the declared pregnant woman during the entire gestation period is not to exceed 0.5 rem (5 mSv). In addition, the licensee is required to make efforts to avoid substantial variation in the monthly exposure throughout the period of gestation. If the dose to the embryo/fetus is found to have exceeded 0.5 rem (5 mSv) or is within 0.05 rem (0.5 mSv) of this dose by the time the woman declares the pregnancy to the licensee, the licensee is required to limit the additional dose to the embryo/fetus to 0.05 rem (0.5 mSv) during the remainder of the pregnancy.

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The tables in the appendices to this guide were prepared directly from the computer outputs, which led to the values generally being expressed to three significant figures. This indicates greater accuracy than is warranted by the dosimetry model, but the results are presented in this form to avoid roundoff errors in calculations. In general, final results should be rounded to the nearest thousandth of a rem.

f) C. REGULATORY POSITION

i) 1. CRITERIA FOR DETERMINING DOSE TO THE EMBRYO/FETUS 1.1 Monitoring

The dose equivalent to the embryo/fetus should be determined based on the monitoring of the declared pregnant woman as required by 10 CFR 20.1502. Specifically, 10 CFR 20.1502(a)(2) requires monitoring the exposure of a declared pregnant woman when the dose to the embryo/fetus is likely to exceed, in 1 year, a dose from external sources in excess of 10% of the limit of 10 CFR 20.1208 (i.e., 0.05 rem). According to 10 CFR 20.1502(b)(2), the licensee must monitor the occupational intakes of radioactive material for the declared pregnant woman if her intake is likely to exceed, in 1 year, a committed effective dose equivalent in excess of 0.05 rem (0.5 mSv). Based on this 0.05 rem (0.5 mSv) threshold, the dose to the embryo/fetus should be determined if the intake is likely to exceed 1% of ALI (stochastic) during the entire period of gestation.

These monitoring thresholds will ensure that any potentially significant exposures to the embryo/fetus are evaluated and, as appropriate, doses are determined. The conditions specified in 10 CFR 20.1502(a) and (b) are based on a 1‐year period. Prior to declaration of pregnancy, the woman may not have been subject to monitoring based on conditions specified in 10 CFR 20.1502(a)(1) and 10 CFR 20.1502(b)(1). In this case, the licensee should estimate the exposure during the period monitoring was not provided, using any combination of surveys or other available data (for example, air monitoring, area monitoring, bioassay).

The monitoring criteria contained in 10 CFR 20.1502 do not establish required levels of detection sensitivity. For some radionuclides it may not be feasible to actually confirm by bioassay measurements an intake of 1% of their stochastic ALI. Workplace monitoring, occupancy factors, and access control should be considered as appropriate in evaluating potential exposures and monitoring requirements.

1.2 Evaluation of Dose to the Embryo/Fetus

The appropriate dose to be evaluated for the embryo/fetus is the dose equivalent for the duration of the pregnancy. An assessment of the 50‐year committed dose is not appropriate. Also, it is not appropriate to use effective dose equivalent or committed effective dose equivalent. (Note: the committed dose equivalent to the uterus may be applied to the embryo/fetus under certain conditions as a simplified approach as described in the regulatory position specified in Section 2.)

1.3 External Dose to the Embryo/Fetus

According to 10 CFR 20.1208(c)(1), the deep‐dose equivalent to the declared pregnant woman will be taken as the external dose component to the embryo/fetus. The determination of external dose should consider all occupational exposures of the declared pregnant woman since the estimated date of conception. The deep‐dose equivalent that should be assigned is that dose that would be most representative of the exposure of the embryo/fetus (i.e., in the motherʹs lower torso region). If multiple measurements have been made, assignment of the highest deep‐dose equivalent for the declared pregnant woman to the embryo/fetus is not required unless that dose is also the most representative deep‐dose equivalent for the region of the embryo/fetus.

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1.4 Internal Dose to the Embryo/Fetus

The internal dose to the embryo/fetus should consider the exposure to the embryo/fetus from radionuclides in the declared pregnant woman and in the embryo/fetus. The dose to the embryo/fetus should include the contribution from any radionuclides in the declared pregnant woman (body burden) from occupational intakes occurring prior to conception. The intake for the declared pregnant woman should be determined using air sample data, bioassay data, or a combination of the two. Guidance on bioassay measurements used to quantify intake is being developed and has been issued for public comment as Draft Regulatory Guide DG‐8009, ʺInterpretation of Bioassay Measurements.ʺ Specific guidance on workplace air sampling is in Revision 1 to Regulatory Guide 8.25, ʺAir Sampling in the Workplace.ʺ

1.5 Evaluating Continuous Exposure

For continuous or near‐continuous exposure to radioactive material that may be inhaled or ingested, the cumulative intake should be quantified and the dose determined at least every 30 days. If significant variation in the exposure levels may have occurred, the time interval for quantifying the intake should be reduced. More frequent evaluations should be considered as the potential dose to the embryo/fetus approaches the limit.

1.6 Existing Maternal Body Burdens

Maternal body burdens resulting from internal occupational exposures prior to conception should be included in determining the embryo/fetus dose. The contribution to the embryo/fetus dose from a maternal burden existing at the time of conception should be evaluated if the maternal burden at the time of pregnancy exceeds 1% of the radionuclideʹs stochastic ALI value for the appropriate mode of intake and class (for inhalation intakes). For multiple radionuclide burdens, the dose should be evaluated if the sum of the quotients of each burden divided by its stochastic ALI exceeds 0.01. Only body burdens existing at the time of conception need to be considered in evaluating this threshold; radioactive material already eliminated from the body should not be included.

This threshold of 1% ALI provides a simplified approach for determining when pre‐existing body burdens should be evaluated. At this threshold, it is unlikely that any resultant dose to the embryo/fetus would be significant (i.e., greater than 10% of the 0.5 rem (5 mSv) limit). As an alternative, the dose assessment methods presented in the regulatory position specified in Section 3 of this guide may be used for determining whether a pre‐existing body burden represents a potentially significant dose (i.e., greater than 0.05 rem (0.5 mSv)).

ii) 2. SIMPLIFIED METHOD FOR DETERMINING EMBRYO/FETUS DOSE FROM MATERNAL INTAKES

The determination of the dose to the embryo/fetus from the intake of radioactive material by the pregnant woman should be based on the best available scientific data. At present, the NRC staff considers Revision 1 to NUREG/CR‐5631 (Ref. 2) to provide such data. For most radionuclides, the dose to the embryo/fetus will be similar to or less than the dose to the maternal uterus (Ref. 1). However, the data in Revision 1 to NUREG/CR‐5631 indicate that for some radionuclides the embryo/fetus dose may be significantly different, either greater than or less than the dose to the uterus.

Based on these premises (uterus dose similar to fetal dose and the data in Revision 1 to NUREG/CR‐5631 (Ref. 2)), a set of dose factors has been developed for use in calculating an embryo/fetus dose. Except for those radionuclides addressed in Revision 1 to NUREG/CR‐5631 (Ref. 2), the dose factors presented in Appendix A to this guide represent the committed dose equivalent to the uterus per introduction of unit activity into the first transfer compartment (i.e., blood) of the woman.1 For the radionuclides in Revision 1 to NUREG/CR‐5631, the dose factors in Appendix A represent the maximum dose equivalent to the embryo/fetus for the gestation period from the

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introduction of unit activity into the first transfer compartment of the woman at any time during the gestation period.

The dose limit for the embryo/fetus is expressed as a 9‐month gestation dose equivalent. Particularly for certain radionuclides with both long radiological half‐lives and long‐term biological retention, the committed dose equivalent to the uterus may be significantly different from a 9‐month gestation dose equivalent to the embryo/fetus. Several radionuclides of this type have been evaluated in Revision 1 to NUREG/CR‐5631 (Ref. 2), and data have been developed for calculating an embryo/fetus gestation dose instead of using the committed dose equivalent to the uterus.

For demonstrating compliance with the dose limits of 10 CFR 20.1208, the dose factors in Appendix A may be used for approximating the embryo/fetus dose equivalent for the entire gestation period.

The steps for determining the embryo/fetus dose, using the simplified method, are as follows:

2.1 Include all the intakes by the declared pregnant woman at any time during the gestation period in the calculation of the embryo/fetus dose.

2.2 For ingested radionuclides, determine the activity uptake by the first transfer compartment (blood) by multiplying the intake (I) by the appropriate uptake factor (f1) from Appendix B (adapted from Federal Guidance Report No. 11, Table 3 (Ref. 4)). The uptake factor, f1, is the fraction of an ingested compound of a radionuclide that is transferred into the first transfer compartment (i.e., blood uptake fraction).

2.3 For inhaled radionuclides, determining the fraction of initial intake that is transferred to the blood involves an evaluation of the deposition in the three compartments of the lung and the subsequent time‐dependent transfer to the body fluids and to the GI tract. Unless it is known otherwise, it should be assumed that the transfer from the lung to body fluids and from lung to GI tract to body fluids follows the ICRP 30 (Ref. 3) modeling (which is the basis for this guide).

2.4 For simplicity and conservatism in the modeling, the total uptake into the blood from the maternal intake is assumed to be instantaneous. However, for radionuclides with lung clearance class of W (10‐ to 100‐day half‐life clearance) or Y (greater than 100‐day half‐life clearance), the actual translocation from the lung and uptake in the blood may occur over a time period that exceeds the gestation period. Clearance from the lung may take up to several years. All the initially deposited material is not immediately available for uptake by the first transfer compartment (blood). However, an incremental transfer from the lung to the blood may be assessed based on the lung model as described in ICRP Publications 30 and 19 (Refs. 3 and 5).2

Table 1, adapted from the data in Figure 5.2 of ICRP 30 (Ref. 3), may be used for determining the total transfer from the lung to the first transfer compartment (i.e., blood), where f1 is the blood uptake fraction from Appendix B.3 The lung clearance class (D, W, or Y) for a particular chemical form of a particular radionuclide may be obtained from Appendix B to 10 CFR 20.1001‐20.2401.

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Table 1

Transfer Fraction of Inhaled Activity to First Transfer Compartment

Class Transfer Fraction (TF)

D 0.46 + 0.15 f1

X 0.12 + 0.51 f1

Y 0.05 + 0.58 f1

2.5 Based on the determination of the maternal intake, the dose to the embryo/fetus for the entire gestation period should be calculated using the following equations:

For ingestion intakes:

DE = Σ Ii x f 1,i x DF1 (Equation 1)

For inhalation intakes:

DE = Σ Ii x TFi x DF1 (Equation 2)

where:

DE = dose equivalent to the embryo/fetus for the entire gestation period from the acute intakes of all radionuclides during the gestation period (rem)

Ii = intake of radionuclide i by the declared pregnant woman at any time during the gestation period (μCi)

DFi = dose factor for use in approximating the dose equivalent to the embryo/fetus for the entire gestation period from the introduction of unit activity (1 μCi) into the maternal blood at any time during the gestation period, from tabular data presented in Appendix A to this guide (rem/μCi in maternal blood)

f1,i = the fraction of radionuclide i reaching the body fluids following ingestion (i.e., the fraction of ingested activity of radionuclide i that enters the blood), from data presented in Appendix B to this guide

TFi = transfer fraction of inhaled activity to the first transfer compartment (i.e., the fraction of inhaled activity of radionuclide i that enters the blood, see Table 1 of this guide)

2.6 For pre‐existing body burdens, the total burden determined to exist at time of pregnancy should be assumed to be available for uptake in the blood of the woman. The dose should be assigned to the embryo/fetus as if the maternal blood uptake occurs within the first month of pregnancy. The embryo/fetus dose is calculated by multiplying the maternal burden of the radionuclide by its dose factor from Appendix A using the equation:

DE = Σ Ai x DF1 (Equation 3)

where:

DE = dose equivalent to the embryo/fetus

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Ai = maternal burden existing at time of pregnancy (μCi)

DFi = dose conversion factor (Appendix A)

This method provides a simplified and conservative approach for evaluating the significance of pre‐existing conditions. If the embryo/fetus is likely to receive a dose in excess of 25% of the limit from pre‐existing burdens (i.e., greater than 0.125 rem (1.25 mSv)), more detailed modeling should be considered.4

2.7 Doses from multiple nuclides or multiple intakes should be evaluated on a frequency corresponding to the determination of the intake. Multiple dose determinations should be added to determine the total dose. Doses may need to be reevaluated if better estimates of intakes are provided by followup bioassay measurements.

iii) 3. DETERMINING GESTATION‐TIME DEPENDENT DOSE TO THE EMBRYO/FETUS USING REVISION 1 TO NUREG/CR‐5631 METHODS

As an alternative to the simplified methods presented above, a gestation‐time dependent dose to the embryo/fetus may be calculated for the radionuclides addressed in Revision 1 to NUREG/CR‐5631 (Ref. 2). Revision 1 to NUREG/CR‐5631 presents dosimetric methods for calculating the dose to the embryo/fetus following the instantaneous introduction of unit activity into the first transfer compartment (blood) of the pregnant woman at successive stages of gestation. These methods include the contribution to the embryo/fetus dose from the resultant body burdens of the declared pregnant woman and from activity in the embryo/fetus resulting from transfer across the placenta. Refer to Revision 1 to NUREG/CR‐5631 (Ref. 2) for a detailed description of the modeling.

The methods and data of Revision 1 to NUREG/CR‐5631 (Ref. 2) may be used for determining the dose to the embryo/fetus from maternal intakes at successive stages of gestation for the radionuclides 3H, 14C, 57Co, 58Co, 60Co, 89Sr, 90Sr, 106Ru, 125I, 131I, 132I, 133I, 134I, 135I, 134Cs, 137Cs, 233U, 234U, 235U, 238U, 238Pu, 239Pu, and 241Am.

The steps for determining the embryo/fetus dose using the Revision 1 to NUREG/CR‐5631 (Ref. 2) methods are as follows:

3.1 The methods presented in the regulatory position in Sections 2.1 through 2.4 should be used for determining the uptake in the first transfer compartment (blood) of the declared pregnant woman.

3.2 Equations 1 and 2 of the regulatory position specified in Section 2.5 may be used for determining the embryo/fetus dose with the following clarifications:

3.2.1 For Equations 1 and 2, in place of the dose factor parameter, DF, the dose values should be taken from Appendix C to this guide for the time period representing the time of intake relative to stage of gestation. The data in Appendix C to this guide are for an absorbed dose (in rads) from the introduction of 1 μCi of the radionuclide into the first transfer compartment (blood) of the woman at the beginning of the specified month of gestation. To convert from an absorbed dose (rad) to a dose equivalent (rem), the data in Appendix C should be multiplied by the appropriate quality factor from Table 1004(b).1 of 10 CFR Part 20. For 3H, 14C, 57Co, 58Co, 60Co, 89Sr, 90Sr, 106Ru, 125I, 131I, 132I, 133I, 134I, 135I, 134Cs, a quality factor of 1 should be applied. For 233U, 234U, 235U, 238U, 238Pu, 239Pu, and 241Am, a quality factor of 20 should be applied, recognizing that most of the embryo/fetus dose results from alpha decay.

For some radionuclides (e.g., 235U), a blood uptake at the beginning of the gestation period results in a negligible dose contribution to the embryo/fetus. These radionuclides are identified in the tables in Appendix C to this guide by an ʺNʺ entry in the row for the 0‐day of gestation at radionuclide introduction (i.e., the first row of dose factor data). For an intake of these radionuclides within the first month of gestation, a time‐weighted dose factor using the second month data (31‐day row) should be used. The 31‐day dose factor should be multiplied by the quotient of the days‐to‐

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date in the first gestation month at time of intake divided by 30 days. For example, assuming a maternal intake of 14C resulting in a 1‐μCi blood uptake on the 20th day of the pregnancy, the embryo/fetus dose should be determined by multiplying the cumulated dose from an intake at day 31 (i.e., Table C3, Cumulated Dose column, 1.89E‐04 rads) by the ratio of 20 days to 30 days (i.e., 20 divided by 30).

3.2.2 For using the tabular dose data in calculating the embryo/fetus dose, it may be assumed that all intakes occurring within any of the 30‐day periods of gestation occur at the beginning of that period.5 The cumulated dose column should be used in order to determine the total dose for the remainder of the gestation period.

3.2.3 For pre‐existing body burdens from occupational exposure, the total burden determined to exist at time of pregnancy should be assumed to be available for uptake in the blood of the woman. The dose should be assigned to the embryo/fetus as if the maternal blood uptake occurs within the first month of pregnancy. The embryo/fetus dose is calculated by multiplying the maternal burden of the radionuclide by its dose factor (Equation 3). The dose factor to be used from the Appendix C tables is that factor corresponding to the cumulated dose for a 0‐day of gestation at radionuclide introduction (i.e., right‐most column, first data entry). However, for those radionuclides with an ʺNʺ for this 0‐day entry, the entry for the second gestation month should be used (i.e., the right‐most column, second data entry). Alternatively, time‐dependent release kinetics may be used for calculating that fraction of the body burden that is translocated to the blood through the duration of the pregnancy. The time‐dependent release is described in ICRP Publications 30 and 54 (Refs. 3 and 6). This approach is complex, involving interlinking differential equations, and is considered outside the scope of a routine health physics program.

3.3 Doses from multiple nuclides and multiple intakes should be evaluated with a frequency corresponding to the intake (i.e., at least once every 30 days). Multiple dose determinations should be added to determine the total dose. Doses may need to be reevaluated if better estimates of intakes are provided by followup bioassay measurements.

g) D. IMPLEMENTATION

The purpose of this section is to provide information to applicants and licensees regarding the NRC staffʹs plans for using this regulatory guide.

Except in those cases in which an applicant proposes an acceptable alternative method of complying with specified portions of the Commissionʹs regulations, the methods described in this guide will be used

in the evaluation of applications for new licenses, license renewals, and license amendments and for evaluating compliance with 10 CFR 20.1001‐ 20.2401.

Appendix A ‐‐ Dose Equivalent Factors for Use in Approximating The Embryo/fetus Dose from Radionuclides in Maternal Blood

Appendix B ‐‐ Blood Uptake Fractions for Ingested Activity

Appendix C ‐‐ Radiation Absorbed Dose to the Embryo/fetus Following Introduction of Specified Radionuclides and Chemical Forms into the Maternal Transfer Compartment (Blood)

Appendix D ‐‐ Examples of Embryo/Fetus Dose Calculations

h) REFERENCES

1. International Commission on Radiological Protection, ʺAge‐Dependent Doses to Members of the Public from Intake of Radionuclides: Part 1,ʺ ICRP No. 56, Pergamon Press Inc., 1989.

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2. M. R. Sikov et al., ʺContribution of Maternal Radionuclide Burdens to Prenatal Radiation Doses‐‐Interim Recommendations,ʺ NUREG/CR‐5631, Revision 1 (PNL‐7445), U.S. Nuclear Regulatory Commission, March 1992.

3. International Commission on Radiological Protection, ʺLimits for Intakes of Radionuclides by Workers,ʺ ICRP No. 30, Parts 1 through 4, including supplements, Annals of the ICRP, Volume 2, No. 3/4, Pergamon Press Inc., 1979.

4. K. F. Eckerman, A. B. Wolbarst, and A. C. B. Richardson, ʺLimiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion,ʺ Environmental Protection Agency, Federal Guidance Report No. 11 (EPA‐ 520/1‐ 88‐020), September 1988.

5. International Commission on Radiological Protection, ʺThe Metabolism of Compounds of Plutonium and Other Actinides,ʺ ICRP No. 19, Pergamon Press Inc., May 1972.

6. International Commission on Radiological Protection, ʺIndividual Monitoring for Intake of Radionuclides by Workers: Design and Interpretation,ʺ ICRP No. 54, Annals of the ICRP, Volume 19, No. 1‐3, Pergamon Press Inc., 1988.

Footnotes

1 The committed dose equivalent factors for the uterus presented in Appendix A were calculated based on the modeling employed during the development of the ICRP 30 (Ref. 3) data. It is recognized that the metabolism of the pregnant woman may not be adequately represented by the standard metabolic model. However, partly because of the lack of more definitive data, this modeling has been used for determining the dose commitment factors for the uterus that may be used for evaluating compliance with the embryo/fetus dose limit.

2 As modeled in ICRP Publications 19 and 30, the clearance from the different lung compartments is assumed to follow first‐order kinetics. This approach is complex, involving interlinking differential equations, and is considered outside the scope of a routine operational health physics program.

3 The coefficients for the transfer fraction equations in Table 1 are applicable to particles with a 1‐micrometer activity median aerodynamic diameter (AMAD). As a default, these equations may be used for all particle sizes. However, if the actual particle size distribution is known, transfer fractions for other AMAD particle sizes may be derived from data in Figure 5.2 of ICRP 30 (Ref. 3).

4 This approach for evaluating pre‐existing body burdens does not specifically address time‐dependent releases as could occur for certain radionuclides with both a long biological retention and radiological half‐life. However, the assumption of blood uptake of the total burden in the first month of the gestation period provides a simple method with reasonable assurance that any actual dose to the embryo/fetus will not be significantly underestimated. More detailed evaluations may be needed for unusual circumstances in which a pre‐existing body burden could present a significant source of exposure to the embryo/fetus. An evaluation of this nature should be conducted by individuals knowledgeable in the area of internal dosimetry. Such a detailed evaluation could consider the element retention functions as presented in ICRP Publications 30 and 54 (Refs. 3 and 6). Also, the modeling presented in Revision 1 to NUREG/CR‐5631 (Ref. 2) could be applied. The details of this type of an evaluation are beyond the types of analyses that are considered routinely required and, as such, are outside the scope of this guide.

5 The correlation of intake to actual stage of gestation can only be roughly estimated. For this reason, it is believed that the correlation should be limited to the best estimate of the month of gestation.

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