PRESENT STATUS OF THE OMEGA PROGRAM IN JAPAN

PREPARED FORSECOND GENEWIL MEETING FOR INFORiVIATION EXCANGE MEETING

ON ACTINIDE AND FISSION PRODUCT SEPARATION AND TRANSMUTATION11-13 NOVEMBER 1992

Masato NAKAMURAScience and Technology Agency

Chiyoda-ku , Tokyo ,, Japan

Hiroyuki YOSHIIIAJapan Atomic Energy Research Institute

Tokai-mura , Ibaraki-ken , Japan

Satoshi TANIPower Reactor and Nuclear Fuel Development Corporation

Minato-ku , Tokyo , Japan

Tadashi INOUECentral Research Institute of Electric Power Industry

Komae-shi. , Tokyo , Japan

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PRESENT STATUS OF THE OMEGA PROGRAM IN JAPAN

1. INTRODUCTION

The management of high. level radioactive waste ( HLW )generated from the reprocessing of spent fuel is very importantas well as the safety assurance, to further develop nuclearelectricity generation.

National policy of managing this HLW in Japan is to solidi-fy in a stable form and to dispose of ultimately into deepgeological formations after 30 to 50 years of storage forcooling. In addition, the Japan’s Atomic Energy Commissionsubmitted in October 1988 a report entitled “Long-Term .Programfor Research .and Development on Nuclide Partitioning and Trans-mutation (P-.T)”, from the viewpoints of conversion of HLW intouseful resources and its disposal efficiency. The programplots a course for technological development up to the year2000 and is called “OMEGA” which is the acronym derived fromOptions for Making Extra Gains from Actinides and fissionp r o d u c t s .

The program was jointly stimulated by the collaborativeefforts of the Japan Atomic Energy Research Institute (JAERI)and the Power Reactor and Nuclear Fuel Development Corporation(PNC) . In the private sector, the Central Research Instituteof Electric Power Industry (CRIEPI) has been also carrying outR&D on this subjec~.

The Japanese government (represented by Science and Tech-nology Agency) propos.ed an international cooperation for infor-mation exchange relevant to partitioning and transmutationtechnology under the framework of the OECD Nuclear EnergyAgency, in January 1989. The first information exchange meet-ing on this subject was held in Japan in November 1990. ElevenOECD member countries and two international organizationsnamely IAEA and Joint Research Center of CEC participated i:the meeting. A workshop on partitioning was held in Japan inNovember 1991 , and a specialist meeting on accelerator–driventransmutation was held in Switzerland in March 1992.

-.2. OMEGA PROGRAM IN JAPAN

The program is to proceeded in two phases: the phase-I and-II. The phase-I covers a period up to about 1996 and thephase-II covers a period from about 1997 to about 2000. Ingeneral, the basic studies and testing are to be conducted inthe phase-I to evaluate various concepts and to develop re-quired technologies. In the phase-II, engineering tests oftechnologies and/or demonstration of concepts are planned.After the year 2000 , pilot facilities will be built to demon-strate the P–T Technology.

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2.1. PARTITIONING

The principal technologies to be developed consist of thefollowing three subjects and their R&D are to be conductedaccording to the R&D schedule shown in Fig.1.

Nuclide partitioning technology for HLW,Technology for recovering useful metals in insoluble resi-due fromreprocessing, andTechnology for utilizing the separated elements.The partitioning is to separate HLW into four groups:

i) TRU elements ii) Sr-Cs elements iii) Tc-Platinium elementsiv) Others

Harmonized and optimized system is to be studied to inte-grate the partitioning process with the PUREX process and toperform the best utilization and management of nuclear re-sources contained in spent fuel. Feasibility and applicabilityof dry process with molten salt, and sublimation/volatilizationprocesses are also to be studied.

The recovery technology is to be developed to obtainuseful elements, such as Ru, Rh and Pd, from insoluble residuesof dissolved solution from reprocessing. Purification andisolation techniques are also to be developed, for example, forremoving long–lived Pd-107 from the recovered Pd element.

The fabrication technology of the separated TRU and fis-sion products is to be developed for their further use ordisposal. Immobilization of the separated Sr and Cs is neededfor either utilization or disposal”

2.2. TRANSMUTATION ‘

The principal technologies to be developed consist of thefollowing two subjects and their R&D are to be conducted alongwith the R&D schedule shown in Fig.2 and Fig.3.

Application of nuclear reactor on transmutationApplication of accelerator on transmutation.Nuclear reactor provides an extremely rational means for

the possible earlier realization of transmutation technologyfor TRU nuclides. Sodium-cooled fast breeder reactor with MOXor metallic fuel can be applied for transmutation of many kindsof TRU nuclides due to its relatively large high–energy fis–sioning. TRU burner fast reactor is another candidate to bedeveloped due to its more efficient transmutation capabilitythan conventional fast reactor. Thermal neutron reactor is.also an alternative, when plutonium is used as nuclear fuel.

Owing to the recent remarkable development of acceleratortechnology, application of accelerators is becoming more at-tractive as a means of transmutation. Proton accelerator is tobe developed to transmute TRU and long-lived FPs” by spallationreaction and associated large number of emitted neutrons.Electron accelerator is also to be developed to transmute thenuclides having small neutron cross–section, such as Sr-90 andCS-137 , by photo-nuclear reaction. Hybrid system combining anaccelerator with a subcritical assembly will be studied to

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improve the overall energy balance of the transmutation system.

/

2.3. RELATED BASIC RESEARCH

Reliable data base of TRU and long-lived FPs is indispens-able to proceed the OMEGA program. Underlying studies onphysical and chemical properties of TRU and FPs will improveunderstanding of the science and technology for separation andrecovery of these nuclides from HLW, for fabrication of TRUfuel for recycling t.o reactors or accelerator-driven systemsfor transmutation, and for utilization of these nuclides.Nuclear data and thermodynamic data of these nuclides will bemeasured, compiled and evaluated for reactor Physics and mate-rial development.

he areas covered by the OMEGA program and R&D activitiesare illustrated in Fig.4.

3. RESEARCH AND DEVELOPMENT ACTIVITIES

Under the framework of the OMEGA program , the followingR&D activities are

3.1. PARTITIONING

Wet processesPNC. Dry processes

under way at JAERI , PNC AND CRIEPI.

of partitioning are development. at JAERI ANDare developed at CRIEPI AND PNC.

A. STUDIES AT JAERI

A partitioning process has been developed to separateelements in the HLW into four groups. This process consists ofthree major steps as shown in Fig.5. In the first step ., theHLW denigrated with formic acid is fed to the solvent extrac-tion system where the DIDPA (diisodechl phosphoric acid) isused as a solvent. The DIDPA is an excellent solvent to extractnot only tetra and. hexa valent ions but also penta valent ionwith addition of hydrogen peroxide. The TRU elements , such asresidual Pu In HLW and minor actinides (Np , Am and Cm) , arecoextracted ‘with rare earths (RE) into “the DIDPA , and thenseparation of the TRU from” the RE is carried out by using theDTPA (diethylene triamine pantaacetic acid) as a completingreagent. In the second step, raffinate is fed through activatedcarbon column for separation final step , the high-decay-heatgenerating elements , Sr and Cs , are captured by inorganic ionexchangers of titanic acid and zeorite respectively , leavingthe final solution virtually free of both ultra-long-livedradioactive nuclides and high-decay-heat generating nuclides.

In the mixer-settler experiments experiments with theactual HLW, it was demonstrated that more than 99.99% of Am andCm were extracted with the DIDPA. It was also Verified in aseries of the tests with the synthesized HLW that more than

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99.95% of Np was extracted when hydrogen peroxide was added asindicated in Fig.6. Regarding separation of the TRU from theRE, a preferential back extraction of the TRU from the DIDPAhas been studied by using the DIDPA as the completing reagentto hold the RE in the organic phase. Distribution ratios of Amand RE were measured in the batch shaker tests and compared inFig.7 by using a stripping solution composed of 0.05M DTPA, lMlactic acid and NH40H at PH value of 3.6. The separation fac-tors were observed to be larger than 10 between Am and RE.Mixer–settler experiments are in progress to verify the effectof PH variations due to extraction of NH4+.

The separation of Tc has been demonstrated by adsorptionwith the activated carbon column. More than 99% of Tc wasadsorbed from the synthesized HLW adjusted to 0.5M nitric acidconcentration and the absorbed Tc was eluted from the columnvery efficiently by using a eluant , 2M KSCN-4M NaOH ( alkalinethiocyanate solution). The adsorption and elution of platinum-group metals were also examined. And more than 99.9% of Sr andCs were captured by the inorganic ion exchangers describedabove.

The proposed partitioning process is planned to be testedin the sense’ of chemical engineering and in parallel , to betested with the actual HLW at the Nuclear Fuel Cycle SafetyEngineering Facility (NUCEF) of which cold operation will startin 1994.

B. STUDIES ATPNC [11[21 [31

PNC’S basic approa~h is to develop partitioning process asa part of advanced reprocessing system where the improved PUREXprocess is closely combined with TRUEX-like TRU separationprocess as indicated in Figure.8 with emphasis on TRU recyclein FBR system.

As an effort to improve PUREX process , studies are inprogress to control Np valence by electrochemical and photo-chemical methods so that it is completely recovered in Puproduct stream for recycling in FBR. Preliminary experimentsare also in progress to electrochemically recover platinum-group metals and Tc from dissolver solution.

A series of extraction tests have been conducted to recov–er TRU from the HLW by using CMPO(octyl phenyl N,N diisobutylcarbamoyl methyl phosphine oxide) mixed with TBR as an extract-ant. Fundamental batch tests were carried out to determine thedistribution ratios of TRU and FP elements as well as to clari-fy the fundamental behaviors of the CMPO-TBP solvent such asthe third phase formation and the precipitation. Followingthese fundamental studies counter-current experiments withminiature mixer-settlers wer’e conducted based on the simplifiedTRUEX flowsheet using the actual HLW generated from FBR fuelreprocessing tests. Concentration profiles for Am, Pu andtypical FP uclides are shown in Fi”g.9. It has been observedthat Am and Pu were completely extracted by the CMPO-TBP sol-vent whereas such FP elements as Cs and Sb were well retainedin the raffinate. However , the RE elements such as Ce were

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co-extracted as expected from the batch shaker measurements.Also extracted into the solvent was a significant fraction ofRu . Future studies are planned to improve the process for Puback-extraction, to improve extraction efficiency of Np, and todevelop a process to separate TRU from RE element.

As an effort to recover noble metals from the nuclearwaste, the Pb extraction tests have been conducted with theactual insoluble residue produced in the dissolution of irradi-ated mixed oxide fuel. The tests verified that noble metalssuch as Ru, Rh and Pd were efficiently recovered by this proc-ess.

In addition to the development of aqueous separationprocesses, experimental studies are in progress to explore thefeasibility of dry separation process. One of the processedmethods under investigation is the FP separation by super hightemperature treatment of the HLW as illustrated in Fig.10. Inthe preliminary experiments, the simulated HLW calcined at 700C was heated in inert atmosphere at 900 C at first and thenfurther heated up to about 1800 C. It was observed that morethan 90% of Cs was eliminated by sublimation in 2 hours at 900C, and most of noble and transition metals were separated asmolten alloy from the oxides of remaining FP elements (RE, Zr,alkali earths and TRU) at 1800 C.

C. STUDIES AT CRIEPI [41

CRIEPI has proposed a concept to separate TRUS from HLW bydry process with pyrometallurgical methods and then to trans-form to short-lived nuclides in a metallic fuel FBR under de-veloped .[4]. This dry.process can be applied on the removal ofTRUS from undissolved residue and cladding after the leachingby nitric acid without much efforts for the modification.According to this concept, TRUS produced in LWR fuel cycle areconfined in the FBR cycle. Fig.11 shows the flow diagram of thedry process for partitioning of TRUS. This process consists of(1) denigration to obtain oxides, (2) chlorination to oxides tochlorides, (3) reductive extraction to reduce TRUS from moltenchlorides in liquid cadmium by using of lithium as reductant,and (4) electrorefining to increase the purity of TRUS recove-red. Both the denigration and chlorination steps are pre-treatment process prior to the application of pyrometallurgicalprocess. The first target on the development of dry process isto recover more than 99% of TRUS containing in the HLW byallowing the introduction of the same amount of rare earths asTRUS .

Microwave heating device was selected as the preferredmethod for denigration of HLW because of the expectation of theproduction of small amount secondary wastes and operatingsafety in a hot cell. The technical feasibility of this methodwas examined experimentally, and a design study was carried outfor the assumed scale of the denigration step based on theexperiments. The specification of the facility with suitablecondition is presented in Table.1.

Chlorination using chlorine gas together with carbon asreductant was selected as the preferred method because of

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recycling of chlorine gas by processing of salt wastes. Thefeasibility examined experimentally, by using the apparatusshown in Fig.12, sh.owed that nearly 100% of chlorination can beattained under suitable condition with evaporation of someamount of specified elements with chloride form, which can berecovered for the following processing of” reductive extraction.The assumed scale of chlorination step was designed based onthe experimental study. The specification of the facility isalso presented in Table.1.

The TRUS, accompanied with some amount of rare earths,contained in molten salts are reduced into liquid cadmium addedlithium as reductant. The principle of the reductive extractionwith the subsequent step of electrorefining is schematicallyshown in Fig.13. In order to eveluate the purity of TRUS recov-ered in liquid cadmium, it is necessary to obtain the distribu-tion coefficients between salt and cadmium phases, and therelated thermodynamic data, especially activity coefficientsand standard free energy of formation of chlorides forTRUs andrare earths. The distribution coefficients measured for acti-nides of U and Np and for rare earths of La, Ce, Pr, Nd and ypredict the separation factor between actinides and rare earths151[61.

The electrorefining technology can be applied as prefera-ble method in order to get the TRUS with rare earth as lowConcentration as possible. The electrorefining with liquidcadmium anode or inert anode and solid cathode covered withmolten salt electrolyte, i.e. KCI1-LiCl eutectic salt, has beenapplied. The separation study using rare earths has been car-ried out in CRIEPI, and, moreover the anode dissolution andcathode deposition study using acti~ides has started within theframework of the TRUMP-S Program organized by CRIEPI, KawasakiHeavy Industry (KHI), Rockwell International corp. (RI) andUniv.of Missouri. The development of process technology includ-ing pre-treatment process and the reductive extraction andelectrorefining steps will start from the beginning of 1993 inCRIEPI .

The activity coefficients and free energy of formation ofchlorides in the eutectic salt and activity coefficients inliquid cadmium were obtained by EMF measurements for rareearths, of La, Ce, Pr, Nd, Sm, Eu, Gd and Y, and for actinidesof U, NP, Pu and Am [7]. The data gives the prediction ofseparation efficiency of actinides from rare earths, as shownin Fig.14. The figure suggests that recovery more than 99% ofTRUS can be a“ttained with the accompaniment of rare earth lessthan 10% to the initial amount in HLW.

Cadmium used in the pyrometallurgical process is recycledto the reductive extraction step after recovery of TRUS andanother elements, such as noble metals, by the distillation.The waste chloride salts containing fission products mainly ofalkali and/or disposal. It was clarified that the alkali andalkaline earth elements were reduced to metals by using thegraphite anode and liquid cadmium or zinc cathode. The chlorinegas produced in the process can be”’ recycled to the chlorinationprocess.

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3.3. TRANSMUTATION WITH FISSION REACTORS

Optimization of minor actinide recycling into MOX-FBR isstudied at PNC. Studies of minor actinide recycling intometallic fuel FBR are conducted at CRIEPI. Actinide burnerreactor design study and integral experiments to improve crosssection data of minor actinidesare carried out at JAERI.

3:3.1. MOX-FBR AND METALLIC FUEL FBR

A. STUDIES ON TRU TRANSMUTATION IN MOX-FBR POWER PLANT AT PNC[8][9]

Feasibility studies of TRU transmutation in LMFBRs havebeen’ performed to establish TRU. (exactly minor actinides)transmutation technology. Systematic parameter survey calcula-tions were performed for a conventional 1000MWe LMFBR core toinvestigate basic characteristics of TRU transmutation in anLMFBR core and also to establish TRU loading method which hasno serious influence on core design.

A heterogeneous TRU-loading method, where a few number ofsubassemblies with concentrated TRU fuel (target S/As ) areloaded in the core, can have an advantage in manufacturing andmanaging the TRU-Loaded fuel since the number of the TRU-loadedfuel subassemblies is smaller than that of the homogeneousmethod. An effort has been made so as to make the heterogeneousmethod feasible. The fuel pin design and the loading pattern ofthe target S/A were studied. It was found that the heterogene-ous TRU-loading”method with target S/As is feasible by reducingthe fuel pin diameter loaded with TRU and optimizing the load-’ing pattern of target S/As as shown in Fig.15.

Study on an innovative core concept for TRY transmutationwas also carried out taking into account the results of system–atic parameter survey calculations. Feasibility from neutronicpoint of view was studied on 1000MWe and 300MWe Super Long-LifeCores with no need of fuel exchange during plant life. It wasfound that remarkable reduction of both reactivity change andpower variation during burnup is possible by optimizing theamount and zoning of TRU loaded. The reactivity change and andpower swing during 30 years of “the optimized core are less thanhalf of those of the Super Long-Life Cores with no TRU”loaded.The Super Long-Life Cores loaded with TRU fuel were foundfeasible from the neutronic point of view.

R&D program are being pursued to establish the transmuta-tion technology of TRU nuclides in LMFBRs. Irradiation of MOXfuel pin including small amount of 241Am was started in JOYOfrom August 1992.

B.STUDIES ON TRANSMUTATION IN METALLIC FUEL FBR AT CRIEPI

TRUS recovered by pyrometallurgical process will be trans-

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formed. to short-lived nuclides by fissioning in metallic fuelwith the matrix of U-Pu-Zr in FBR. The research is directed to(1) the analysis of transmutation rare by simulation code anddesign study of fuel element and core, and the characterizationstudy of U–Pu-Zr alloy contained minor actinides (Np, Am andCm) and irradiation study. The latter activity has been pro-ceeding by the cooperation with the European Institute forTransuranium Elements (ITU), of the Commission of the EuropeanCommunities in Karlsruhe, Germany.

The CITATION-TRU code, developed by modifying the burn–upcalculation code of the CITATION, pointed out that a metallicfuel FBR is superior in the transmutation TRUS because of thehard neutron spectrum. The transmutation rate of 13 to 14 wt%per year of the loaded TRUS is expected in the metallic. fuelFBR. If the metallic fuel allows 5 wt% of minor actinides inthe matrix, a metallic fuel FBR with the capacity of 1000MWecan accept the minor actinides produced from 5 LWR plants withthe same capacity per year at the equilibrium recycle condition[10]. Moreover, the CITE-BURN code, which can make the burn-upanalysis at the fuel rod unit, was developed for the analysisof irradiation study. The analysis of sodium void effect gaveno large effect’ on reactor operation so far as the content of

minor actinides is up to 5% [11]. In addition, using the ana-lytical data for oxide samples with the bench mark calculationsfor transmutation was carried out in the cooperation with KfK,ITU and CRIEPI. This study showed that the prediction by calcu-lation of transmutation of minor actinides is accordance withthe experimental results.

As the first step, the alloys with binary alloys of acti-nides, and U-Zr with rare earths, and U-Pu-Zr with minor acti-nides and rare earths were fabricated for the miscibility and-characterization studies in the Institute for TransuraniumElements. Through the study, the volubility of minor actinidesand rare earths in the U-Pu–Zr alloy as well as the miscibilityof actinides were clarified. It was found that the rare earthinclusion with americium distributed uniformly along the grainboundaries in the matrix, when the content of minor actinideswas up to 5 wt% [12]. In addition, the important parameters ofU-Pu-Zr alloy with minor actinides and rare earths, such asmetallography and phase composition, melting point,Young’smodulus thermal conductivity, dilatometry, and redistributionof alloy components in a temperature gradient, were measured.The chemical interaction between the alloy with minor actinidesand stainless steel, and compatibility of the alloy with sodiumwere also examined prior to the irradiation study. Base on theresults, feasibility study was carried out for the irradiationin CEA. Fig.16 shows the composition of the irradiation fueland the figure of the irradiation capsule in the PHENIX fastreactor.

3.3.2. ACTINIDE BURNER REACTOR STUDIES AT JAERI

Conceptual design study has been carried out on ActinideBurner Reactor (ABR) which is a sort of fast reactor with very

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hard neutron spectrum, specially designed for efficientlyburning of minor actinides (MA). Two types of reactor designwith about 1000MW thermal power have been evolved, which areNa-cooled MA alloy fuelled ABR (M-ABR) and He-cooled MA nitrideparticle fuelled ABR (P-ABR).

Low thermal conductivity and melting point of MA alloyfuel are the limiting factors for the high MA burnup in the M-ABR. Therefore, the particle bed reactor concept was selectedas the reference, which has the high efficiency in heat trans-fer since small particle size produces a large heat transfersurface per volume. The bed of coated particle fuel containedin double concentric porous frits iS directly cooled by helium.The fuel is a microsphere of MA nitride which is coated with arefractory material such as TiN. The fuel concept of P-ABR iscompared with that of M-ABR in Fig.17.

In a cold fuel concept as adopted to the P-ABR, fueltemperature is to be kept lower than one third of its meltingpoint to reduce mass transport. Since reduced mass transportresults in smaller swelling and gas release, thickness ofcoating layer can be minimised to give large heavy–metal densi-ty in order to obtain hard neutron spectrum in a core. -

The reactor core design parameters of M-ABR and P-ABR attheir equilibrium state are shown in Table.2. In the P-ABR, theneutron flux is very high and the MA burnup per year is conse-quently very high (17%), owing to the efficient heat removalcharacteristics of particle fuel. A fuel cycle based on the P-ABR is being studied.

Reliabe database of MA is indispensable to increase theaccuracy of the ABR design. The integral experiments have beenconducted in the fast critical facility (FCA) of JAERI toevaluate and refine neutron cross-sections for MA, and accuracyof nuclear data of MA compiled in JENDL-3 nuclear data libraryhas also been examined. The phase diagram and thermodynamicstudies on MA compounds are being carried out under the cooper-ation with the Oak Ridge National Laboratory (ORNL).

3.4. TWINSMUTATION WITH ACCELERATORS

A transmutation system driven by a proton accelerator isstudied and development of an intense proton accelerator isunder way at JAERI. At PNC, accelerator assisted or fusiondriven ’systems for Sr-90 and Cs-137” transmutation are evaluatedand an intense electron accelerator is under development.

A. STUDIES OF A TRANSMUTATION SYSTEM DRIVEN BY AN INTENSEPROTON ACCELERATOR AT JAERI

A conceptual design study on proton accelerator–driventransmutation plant has been made at JAERI, which consist of ahigh intensity proton accelerator, a sodium-cooled subcriticalcore, with MA alloy fuels and a tungsten target. As shown inFig.18, the central tungsten target is surrounded by the annu-lar MA fuel region, which in turn is surrounded by radial and

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axial reflectors of stainless steel. A proton beam window islocated right above the target region and is designed to bereplaceable. High energy protons are injected vertically down-ward through the window into the target. The tungsten targetacts as a spallation n e u t r o n s o u r c e . T h e p r i n c i p a l d e s i g nparameters of the plant is shown in Table.3. The table showst h a t t h e Plant t r a n s m u t e s about 250kg of MA by f i s s i o n a n dproduces 820MW thermal power by usinga 1.5GeV proton beam with a current of 39mA. Electr i c power o f2 4 6 M W i s g e n e r a t e d w i t h a c o n v e n t i o n a l s t e a m t u r b i n e , a n ds u p p l i e s s u f f i c i e n t e l e c t r i c i t y t o Power t h e a c c e l e r a t o r .

Study is a lso in progress on the system with a chlor idemolten salt target as shown in Fig.19. The system has a poten-t ia l to establ ish a cont inuous MA transmutat ion concept witho n – l i n e s e p a r a t i o n f a c i l i t i e s f o r t h e spallation a n d f i s s i o nproducts , and to avoid the complicated processes being fol lowedby batch reprocessing for solid spent fue ls and/or targets . Thepreliminary system design parameters and the schematic image ofi n e s e p a r a t i o n system are given in Table.4 and Fig.20, respec–tively. I n a d d i t i o n , s tudy on transmutat ion o f Tc and I byu s i n g a c c e l e r a t o r i s a l s o u n d e r w a y , i n w h i c h B i - P b f l u i dtarget surrounded by a heavy water b l a n k e t i s p r e l i m i n a r i l yused.

I n o r d e r t o v e r i f y t h e r e l i a b i l i t y o f t h e h i g h - e n e r g ynuclear data and the computer codes deve loped at JAERI, theintegral experiment with a lead bulky target and 500MeV protonshas been carried out since 1990. The experiment to use tungstenand TRUS will-also be conducted in near future.

I n o r d e r t o c o n d u c t v a r i o u s e n g i n e e r i n g t e s t s f o r t h eaccelerator-driven transmutation plant, a plan to construct theE n g i n e e r i n g T e s t A c c e l e r a t o r ( E T A ) w i t h a p r o t o n e n e r g y o f1.5GEv and a current of 10mA shown in Fig.21 has been proposed.As the f irst step of developing the ETA, the Basic TechnologyA c c e l e r a t o r (BTA) w i t h t h e p r o t o n e n e r g y o f 10MeV a n d t h es u r r e n t o f 10m.~ i s r e a d y t o b e c o n s t r u c t e d t o study a l o w e renergy portion of the ETA.

B. STUDIES OF TRANSMUTATION BY USE OF ACCELERATOR AT PNC[13] [14] [15]

The transmutation methods of CS-137 using acceleratorshave been studied. The transmutation rate and the transmutationenergy have berm evaluated by Monte Carlo calculation codes toexamine the feasibilities. The new transmutation methods havebeen also proposed and studied to achieve both high transmuta-tion rate and low transmutation energy.

A high current electron linear accelerator with an elec-tron energy of 10MeV and a maximum current of 100mA (average20mA) is under development. A schematic drawing of the acceler-ator technology development facility is shown in Fig.22.

Fabrication of prototype components is in progress. Proto-types of main components such as an accelerating tube and aklyptron were assembled and their RF power test is being con-ducted. Reflecting the test results, the design will be modi-

34

fied.The construction of the whole linac system is scheduled to

start in 1993, and the first operation is scheduled in 1996.

4. CONCLUSIONS

Various R&D activities under the framework of the OMEGAprogram in Japan were presented here. These activities will bereviewed by Atomic Energy Commission, and the partitioning andtransmutation technology will become more and more worthy toreduce the long-term burdens of nuclear waste disposal and toenhance the effective utilization of resources. However, itshould be recognized that the OMEGA program is not to seek ashort-term alternatives for established or planned fuel cycleback-end policies, but is conceived as a research effort topursue benefits for future generations through long-term basicR&D .

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REFERENCES

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[2] Y.Wada et al.; “Recovery and Utilization of Valuable Metalsfrom Spent Nuclear Fuel”, Proc. 1989 Joint InternationalWaste Management Conf., Kyoto(1989).

[3] M.Horie “Super-High-Temperature of High-Level LiquidWaste’’,Trans. Am. Nucl. SOC. ,62, 111 (1990)

[4] T.Inoue et al.;” Development of partitioning and transmutat-ion technology for long-lived nuclides”, Nucl. Technol. 93206 (1991)

[5] M.Sakata et al.;” Equilibrium distribution of rare earthelements between molten KC1-LiCl eutecric salt and liquidcadmium”, J.Nuc1. Mater. 185 56 (1991)

[6] M.Sakata et al.;” Development of pyrometallurgical parti-tioning of transuranium elements prior to transmutation inmetallic fuel FBR”, Int. Conf. on Fast Reactor and RelatedFuel Cycles, Ott.28to Nov.1, 1991, Kyoto, Japan (1991)

[7] J.J.Roy et al.;” Standard potential of lanthanide and acti-nide trichlorides in molten eutecric LiC1-KCl electrolyte”,Mater. Sci. Forum 83 547 (1991)

[8] M.Yamaoka et al.;” Feasibility Study of TRU TransmutatiotnLMFBRs”, Int. Conference on Fast Reactors and Related FuelCycles, VO1.111, Kyoto, Oct. 28 - Nov.1 (1991)

[9] M.Yamaoka et al.;” Design Study of a Super Long Life CoreLoaded with TRU Fuel”, Int, Conference on Design and Safetyof Advanced Nuclear Power Plants, P3-3, Tokyo, Oct. 25-29(1992)

[lO]A.Sasahara et al.;” Transmutation of transuranium nuclideswith FBR”, Proc. Int. Conf. Nuclear Data for Science andTechnology, Mito, Japan, May 30 - June 3, 1988, p .955(1988)

[lllA.Sasahara et al.;” The status of minor actinide transmuta-tion study in CRIEPI”, Int. Working Group ON Fast ReactorsSpecialist Meeting on Use of Fast Breeder Reactors forActinide Transmutation, Obninsk, IPPE, Sep.22-24, 1992

[121T.Inoue et al.;” Characterization of alloy fuels with minoractinides”, Trans. Am. Nucl. Sot. , 64 552 (1991)

[13]T.Kase et al.; “Transmutation of CS-137 Using Muon Cata-lyzed Fusion Reaction”, Muon Catalysed Fusion, 6 (1991) 521

[14]K.Konashi et al.;” New Transmutation Method by Using MovingTarget”, Fusion Technology, 20 (1991) 664

[15]K.Haga et al.;” Progress of Super-Power CW Electron LinacDevelopment in PNC”, to be submitted to the SecondOECD/NEA Information Exchange Meeting on Separation andTransmutation of Actinides and Fission Products, Chicago(1992)

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Fig. 1. R&D SCHEDU,LE OF NUCLIDES P A R T I T I O N I N G

~ phase I @ Phase IIV: Check & Review

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F3.g. 2 R&D SCHEDULE OF TRANSMUTATION BY REACTOR

[. Reactor Physics/Phys i=l Property

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L App[i&tion of FBR(1) Design Study

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’83

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of ktal Fuel of Hetal. Fuel0 I

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.+. ,

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[

-

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u “

s ,.L

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Il. I - I

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CREPI ‘PNCJAERI

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Utilization

X&&l =’: ‘Ccc’emtor ‘Nc.Utilization‘o.

-@9

II I I

LE!s%El+eInsoluble Residue

Fig .4 P-T R/D ACTIVITIES UNDER OMEGA PROGRANI

.

t

Filtration

tDTPA

DIDPA Solvcn~ SolventExtracllon Extracllon l--,=.

1 I

AcWalcdCarbon

‘.RIzz!l

Zeorile & tTl~anlc Acid I \

Ion Exchange Sr-Cs Group

‘1 [(Sr, Cs)

‘: LzzJ’

.Fig.5 FLOW CHART OF ADVANCED PARTITIONINGPROCESS DEVELOPED BY J,AERI

41

103

.—-

z= 10’0.—+

.“ 2.—~ lo”(n

.—a

i(?

I (i2

I I I I I I I I I

,.

\9

\0’

\ @

-@-

Org :0.5 M DIDPA -0.1 M TBP in n-DodecaneAq” :0.5 M OTPA -1 M Laclic Acid - NH40H,

pH 3.6 after Back-Extraction

Am LCI Ce P( If’d Pm Sm. Eu Gd

Fig.6 COMPARISON OF DISTRIBUTION RATIOSBETWEEN Am AND RARE”EARTHS

42

10-9

I I I I I I I I I I I 1 I I

45 cL

\

1 I I I 1 I I I I I I I I.-0 2 4 6 8 {0 (2 [4

Siege Number from Feed Poird .

Np Recovery Frac[lon = 99.0 % for Run (14) (7 stages)> 99.96’% for Run (15) (14 slagos)

Fig. 7 CONCENTRATION PROFILE OF NP IN AQUEOUSPHASE AT THE EXTRACTION SECTION

43

.“.“

.“.“

PUREX

I

‘.“.“•*

● .*‘.

““’”4

Recycle

J’

Present Reprocessing Concept

Recycle

mRecycle

.

- Utilization w

@=@-#

.●

●✎

✎

?: Salt - Free method

( Electrochemical!, Photochemicaf )

Recycle

ouRecycle

Advanced Reprocessing Concept

Fig. 8 ADVANCED REPROCESSING SYSTEM

44

@wI

-108

107

106

105

-if-

104

7A--A---~T’0T’0~

+

~ [Am241]org: --v”””-Q””*””*”’----v-=”-4””-4. , ” . :. .’

[Sb125Jaq /_ / A A A ~144ce]aqm A A

:, ----. ---- 4.....

“-”-”-”””@----@-u”” [Sbl 25, CslsY]org

1 3 5 7 9 1 1 1 3 1 5 1 7 1 9Stage

(43TRU.

I?ig. 9 CONCENTRATION PROFILES OF Cs, Sb, Ce,Am AND Pu IN MD(ER-SEITLER EXPERIMENTWITH ACTUAL tll-W

45

Calcination’-700 ‘c .*

I SublimationVoiatilizatlon

800-1000 “C +‘EI1. .,:5::..s

Nldten A1ioy

I“El

Noble andMeltinp Transition

- 1 8 0 0 C’ MetaisUtilization

‘ Stable C)xideSolidification

Fig.10

SUPER HIGH TEMPERATURE TREATMENT OF HLW

46

c12

chloride (TRU+FP+CP+SC)Cd ..,. ” Li ‘

~~= ‘3.,.. .:f%,.’..’ : “@$” y.”. <:. .;”:; : : :6< :~&. “:”

1Liauid Metal Molten Salt

,. (TRU+FP) (FP+CP+SC).

‘*

I I I ~(Anode) (Cathode) ICd+FP

,~el

>

1,

.....●✎

bF B R

Fig. 11 Flow diagram of pyrometahrgicalprocess for partitioning of TRUS fkomHLW in CRIEPI

47

Table 1. Theand

Specification of Facilitiesthe Operating Conditions.

(1) for DenigrationConditions for Design

HLW 2,0001 / Day

Temperature 700 “C -800 “CCondition

Absorption Efficiency of Microwave 0.’7Microwave Power per Unit Volume of Oven

().()s ~ 11 (Tentative Value)

J

Specifications

Major Apparatus O v e n 2 m W X 2 m L X 2 m HMicrowave Oscillator 100 KW X 2

Number of Batch 4 Batches/Day

Hot Cell 5 m W X 8 m L X 4 m I I

(2) for Chlorination

Conditions for Disign

Denigrated Material 300 kg/ Day

Conditions Temperature 780 “C -Carbon 36 kgCMorine Gas 1100 kg

Specifications

Major Apparatus Chlorination Furnace 1.2 mDX 2.3 mHCondenser 0.9 mDx 1.9 mHPowder Blender

1.8 mW X 0.5 mL X 0.45 mH

Number of Batch 3 Batches/Day

Hot Cell 5 . 5 m W X 6 m L X 8 . 5 m H

Quartz Tube .

Manometer \ h

FlowMeter

HIII Trap forReco.eryofVolatilizedSpecies

Fr-”ff’as\

Air Pump

Thermocouple’

----

v Blowpipe of ,Chlorine Gas

’11I I Reaction

Separation Plate

I?i g. 1 z Schematic Illustration of Chlorination

49

Test Appratus

DChloridesRU + FP)I I

1Reductive Extraction I “ I

( Liquid Cd 1

=?-”*Y

Fig. 13 schematic [illustration of Pyror?wtallurgica!Partitioning Process

50

CIJC-J

Fig.14

‘0.01 . 0 . 1 1 10

The amount of rare earth in actinides recovered(Wt%)

Prediction of recovery rate of actinides as function ofthe amount of rare earth contained in actinides reco-vered by electrorefining using Cd anode

51

a)&s

o

m

c1

la

.&a

.

IIm

cd

-@cd

P-1

M$4

‘%Cd0

Al

52

Actinide Fuel ;U 19Pu 10Zr 0.6Am 1.2Np 0.2Cm 2REU 19Pu 10Zr 1.6Am 3.ONp 0.4Cm 5RE

Reference Fuel ;U 19Pu 10Zr

Operating conditionsTmetals 750 “C <<1150 “C (fuel melting)Tc~ad <520 *C <<725 “C (eutectic

U-Pu-clad)

o :3 metal pins

:16 oxide pins

Irradiation capsule

F@. 16 Figure of Irradiation Study

53

...

.... .. .. .

-. . . ..-.- “&-Ksluz -

‘ . .4%+. .,45.22”.

$Kj)+:J ~. . . . . . !. . . . . . . . . . . . .. . . . . . . . . . . . .. . . . . . . . . . . .. . . . . . . . . .. . ......... .

.

0

La 03’ :w .;. .. .. .

“. . “.- ;X\:-Rod-------- . . .

I%5q. . “Cookim Flow

/r TIN(O.1O5’)

- ’ *

pOrOUS TiNIO.03’)

2

,... . . . . ....- -..,, ..:.:.:.:,:.:,.,.j. .*#?*.$$, . . . . . . . . ........ .......... . ... .. ----

- - - - - ~~<~g ,:::::,::,,. . . . . . . . . . ... --” .-4 . . . . . . . . . . . .., .......................?.+ : .$WV.: . . . . . . . . . . . .. . . . . . . . . . . . . . .. . . . . . . . . . . . .. ... w$j, ........................... . . . . . . . . . . . . ..“ .........................2:.. . . . . . . . . . . . .. . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . ...=---- -.. ... ----- ..................... . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . .

I m.u Nloidc\ ---- “ >

“$I..... -------...... ““~------....... .........-. .-.~:. ...~.-... ......... .. ‘------)7. . . . . . . . . . . . . . . . . .“ - -. - - . . . . . . . . . . . . . . . ........

Inner Frit . . . . . . . . . ....................... ... Pardcle Bcd 1.47 ‘- -:.:.:.:-..:. ~ Oukr Frit Cored Fuel Pmicle—. —.. — .

+Coolm-lt i-kc

Ductless Fuel Assembly of M-ABR Piirticle Fuel of P-A13R

Fig. 17 Fuel Cocept of ABR (unit mm)

54

Tab 1 e. 2 Rc=ior design parameters of Actinide Bumcr Reac[ors

M.ABRII p.ABR2JFud Conccpi pin-bundle

Mat:ridcoaled ~“cle

IcJJ:Np-22Pu-202r f66NpAmCm-34pu),xN,4.OCAmCm-35Pu-5 Y

MA inirial loading?) kg 666 “ 2065NplAm.c* 255f199j212 765159W02

Rcac[or power, MWrh 1 7 0 1 2 0 0

Coolant mz!aid S o d i u m Heliumvelocity, rds 8 total flow. kg/s 1088

inlet pressure, MPi 10pressure drop. kpa 13

irk tempcrarurt.C 300 127Ourkt Wrrpmnsrc(corc Snax).c 1C484 0C:440 340

Fuel lempcraiurc,C ms# IC:834 0(2809 722.Clad t&&GNISiC.C nd) . 1C517 0C484 Fait tcmpcrarrarc. m= 5 6 0

Neutron flux, 10’ +kn?.s= .‘ IC4.1 oc23.4 8.4

Ncurron flucncc (EXilMcV)..10%km2 IC2.2 OC: 1.7 2.2Core avcrdgd mcm ncurron &rcrgy, kcV IC766 0C:785 743

RCXlhity (% Ak/k)Na-void r:zc[ivity/core 2.52bpptcr reactivi&/corc(A t=s~~) -0.01 -0.01

Kinetic pafzrrwers

~ err 135X103 1.72X1O’Lp, SC.C 6.84 X 10; 10.8X10 S

Cycle Icng[h, fell-power days7) 730 3CQ”MA oarumrmriori. %/cycle 26.0 25.3MA bumup. %/cycle -

17.8 17.31) M-ABiCMA metallic ficl bumcr reactor2)

- 3)4)

““ 5)

6)7)

P-ABR :MA panicle frscl bumcr rcaaorKM-ma Core, OCOuter coreAftcrls( cycle.only Np.Am.Cm arc added.Predicted mc!ting point of fuel 9(2OC for M-ABRMax. allowable temp. of fuel 727C (1/3 of M.P. 3000K) for P-A BRML<. aliowablc temp. of claddin~rit (HT-9) 650cFuel inadiarion (imc

55

146 tvfWe

“ r ’

P r o t o n ‘LINAC +

1 . 5 GeV 39 mA

Protons ji.

820 Miit

‘r\_- ,----. . . . . . . . . . .. . . . . . . . . ..::::::: :::. - .:,:.:..< . . . . . . . . . . . . Na. . . . . . . . .- .. . . . . . . . .:.:.:.: . . . .“.. .-.: x.:.:.:.. . . . . - :.....:.:.:.:. -. ‘::- -.-.:.:.:.:. . . . .“... -. . . ..-. :.:.:.:.:.: . . . . . . . . . . .. . . . . . . . . . . . .:.:.:.:.: . . . . . . ...4.:.:.:.:.:.:. :.::.:.:.: . . . . . . . . . . . .

:- ;? .[

f. . . . . . .:.:-:-:.:. . . . . . .. . . . . .. . . . . . . . . . . .:::::::::::, . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . .. . . . . ,. . . . . . . . ...< ..:.:.:.:.: . . . . . . . . . . . ...:.:.: .:.:. :.:.x.:. . . . . . . . . . . . .. . . . . . . . . . . . . . . . . .I

—d 1+–-~-l ‘---- “;Core

. . . . . . . . . . . . - . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . .0 . . . . . . .:.:.:.: . . . . . It-!xnq

100 &IWe—+

246 MWe

‘L1’li{mtern.tor-J

Primary ,, S e c o n d a r y Feed PumpPump rump

Fig.18, TRANSMUTATION PLANT CONCEPT DRIVENBY AN INTENSE PROTON ACCELERATOR

56

Table. 3 ‘D~@n parameter of minor actinide transmutationsystem asskted by an intense proton L1lfAc.

Proton beam current 39 mAAc[hidc inventory 3160kgEffec[ive neumon multiplication fac[or 0.89No. of ncu[rons per incident photon 40 nlpNo. of fissions” (>15McV) 0.45 f/p

{<15MCV) loo f/p .Neutron flux 4X 1015&m2 . sIvfeari neutron Energy 690 keVBumup 250 kg/y

MA ~hennal output fuel 800 MWlungst& 20 Mw

,3

total 820 MW ‘

Power densily . maximum 930 MW/m3average @O MW/m3

Linear power ratio maximum 61 kW/maverage 27 kW/m .

Coolant temperatureOu[lcl maximum 473 z

average .430C /Maximum, temperature

fuel center 890 Tsurface 548 ~

clad inside 528 Toutside 484 ~

57

INTERNAL P&FLECTOR

Fig. 19 concept of the molten salt target system

58

.

Table. 4 Des[gn parat-neters of target system and results of thecalculation for MA transmutation

T

Target Salt -

Composition : 63NaC1-36(Pu, MA)CIO(Pu : MA= 15: 85)(MA : Np, Am, Cm)

Volume: 2.6 maWeight “:9 tonTRU inventory :5 ton

Effective Multification Factork = 0492

Be% Power :1.5 GeV, 25 mAThermal Output :800 MWTransmutation Rate of MA :250 kg/y

59

Salt Pump Proton 13eaiIII

n IMolten Salt Flow I

=Q+-7-)Molten

[JsaltHe bjector Gas Separator Taxget

~550=500”C

?;:)!-T-J--Cd I

L—

‘ Cold Trap ;OO .C

Fig. 20 Concept of the on-line separation system

60

. .

.

. .

t)evelopment for Basic. . —Accelerolor Technology

oO T L

q’!jj’-’yy;k;k 8!/_ 150 MeV “ 400MeV “ L 5 GeV

~1=]WrfunSdeInteryi&p. FToIqw Exp, Irrodio[ion Test

Fig. 21’ Concept of Engineerhg Test Accelc.rater (ETA)

61

L4?0 I

UJn

n “za

-J

z