Development of the Water Cooled Ceramic Breeder Test Blanket Module in Japan M. Enoeda, Hisashi...
-
Upload
cory-lyons -
Category
Documents
-
view
221 -
download
0
Transcript of Development of the Water Cooled Ceramic Breeder Test Blanket Module in Japan M. Enoeda, Hisashi...
Development of the Water Cooled Ceramic Breeder Test Blanket Module in Japan
M. Enoeda, Hisashi Tanigawa, T. Hirose, S. Suzuki,
K. Ezato, Y. Seki, A. Yoshikawa, D. Tsuru,
K. Ochiai, C. Konno,
Y. Kawamura, T. Yamanishi,
T. Hoshino, M. Nakamichi,
Hiroyasu Tanigawa, M. Akiba
JAPAN ATOMIC ENERGY AGENCY
16th International Workshop onCERAMIC BREEDER BLANKET INTERACTIONS
Portland, USA, 11-16, September, 2011
1
Contents1. Importance of Water Cooled Ceramic Breeder (WCCB) Test
Blanket and assumed schedule of WCCB TBM R&D2. Domestic cooperation and R&D flow chart3. Module fabrication technology development - First Wall + Side Wall Assembly, Back Wall Partial Mockup4. Advanced breeder and multiplier pebble development for
DEMO - Increased chemical stability and soundness in high
temperature5. Advanced tritium recovery technology for blanket system6. Neutronics engineering - Verification of tritium recovery rate of Li2TiO3 pebble bed with
DT neutron in FNS
2
ITER Test Blanket Module Program- ITER Test Blanket Module (TBM) Program is to test essential functions of DEMO Blanket in
the real fusion environment with scalable module.- ITER TBM Program is one of the most important development steps.
VV
Plasma
Water Loop
Generat
or -Production of fusion fuel tritium
- Extraction of energy
Fusion Council of Japan stated that ITER TBM Test Program is one of the most important development step. (Aug. 2000)Japan has a position to - act as a Port Master and a TBM Leader to test the WCCB TBM. - participate as a Partner in HCCB/HCSB, LiPb-based TBMs and Li-based TBM.
3
1700mm
600mm500mm
Water Cooled Ceramic Breeder(WCCB) TBM proposed by Japan
Test Blanket Module ( TBM )
Structure of RAFM (F82H)
Neutron MultiplierPebble Bed (Be)
Tritium BreederPebble Bed (Li2TiO3)
ITER Cross SectionProvisional Port Allocation
QuickTimeý DzPhoto - JPEG êLí£ÉvÉçÉOÉâÉÄǙDZÇÃÉsÉNÉ̀ÉÉǾå©ÇÈÇ…ÇÕïKóvÇ ÇÅBQuickTimeý DzPhoto - JPEG êLí£ÉvÉçÉOÉâÉÄǙDZÇÃÉsÉNÉ̀ÉÉǾå©ÇÈÇ…ÇÕïKóvÇ ÇÅB1mm
QuickTimeý DzPhoto - JPEG êLí£ÉvÉçÉOÉâÉÄǙDZÇÃÉsÉNÉ̀ÉÉǾå©ÇÈÇ…ÇÕïKóvÇ ÇÅBQuickTimeý DzPhoto - JPEG êLí£ÉvÉçÉOÉâÉÄǙDZÇÃÉsÉNÉ̀ÉÉǾå©ÇÈÇ…ÇÕïKóvÇ ÇÅB1mm
3118 19 20 21 22 23 24 25 26 27 28 29 30
1906 07 08 09 10 11 12 13 14 15 16 17 18ITER Construction ITER Operation
PrS
R
FD
RTentative Milestones,
under discussion
JA PY
AD
1/1 BreederBox MUTesting
1/1 FWHHF Test
TBMFabric. Tech. Eng R&D1/1 TBM
Box HHF Test
1/1 FWMockuoFabricat.
1/1 BreederBox Mock-
up Fabr.
1/1 SWMockuo
Fabrication
Advanced Breeder, Multiplier Development
Tritium Recovery Process Development
Nuclear Performance Validation Tests
Structural Material Validation/ Database
PD
R
CD
R
TB
M #1 S
afy R
eport
Installation in ITER
BW/Box Assembling
1/1 BWFab.
1/1 BWPlate
Fabrication
4
Material Proc., Parts Manufact.
Ancillary System Manufact.
Module Manufact., Assembling
TBM #1 Fabrication
TB
M
Delivery
1/1 BWFabr.Tests
Assemblingof FW/SW
Provisional WCCB TBM Delivery Schedule
20 21
TB
M P
A1.66m
.484m
0.6m
WCCB TBM
32 33
TBMPPTS
Testing
- Large Size Mockup Fabr.- LSMU Internal Pressure Test- Detailed Fabr. Validation- RH Demonstration
- Water Ingress Safety Validation Tests
Prototype #2 TBM Fabrication- Structure Endurance Tests (Errosion-corrosion, Be pebble bed)- Ancil. System Testing- HHF Tests/Heating Tests, function tests of prototype
Large Size Mockupo Testing
Cooperation of Solid Breeder Blanket DevelopmentSystem IntegrationOut-pile R&D, Module fabrication technology, Thermal hydraulic research, - Blanket Engineering Lab. (JAEA)
Material Development, Fabrication Tech.- Fusion Mater. Development Gr., Radiation Effects and Analyses Gr. (JAEA)- Profs Kohyama, Kimura (Kyoto Univ.)- Prof. Serizawa (Osaka Univ)
TritiumExtraction System
DEMO ReactorSystem
Blanket Module
DEMO Reactor Design- Reactor System Gr. (JAEA)- Prof. Ogawa (the Univ. Tokyo)- Dr. Okano (CREIPI)
In-pile R&D, Breeder/multiplier development- Blanket Irradiation Tec. Gr. (JAEA)
Neutronics / Tritium Production Tests with 14MeV neutrons- Fusion Neutronics Gr. (JAEA) Cooling System
5
Tritium Recovery SystemDevelopment, Tritium Control- Tritium Tech. Gr. (JAEA)- Profs Fukada, Nishikawa (Kyushu Univ.)- Profs Tanaka, Terai (the Univ. Tokyo) - Prof. Hino (Hokkaido Univ.)- Profs Okuno, Oya (Shizuoka Univ.)
Plasma Facing Materials- Prof. Hino (Hokkaido Univ.)- Prof. Tanabe (Kyushu Univ.)- Prof. Ueda (Osaka Univ)
- 6 -
QuickTime? Ç?Photo - JPEG êLí£ÉvÉçÉOÉâÉÄǙDZÇÃÉsÉNÉ̀ÉÉÇ?å©ÇÈÇ…ÇÕïKóvÇ-Ç?ÅBQuickTime? Ç?Photo - JPEG êLí£ÉvÉçÉOÉâÉÄǙDZÇÃÉsÉNÉ̀ÉÉÇ?å©ÇÈÇ…ÇÕïKóvÇ-Ç?ÅB1mm
QuickTime? Ç?Photo - JPEG êLí£ÉvÉçÉOÉâÉÄǙDZÇÃÉsÉNÉ̀ÉÉÇ?å©ÇÈÇ…ÇÕïKóvÇ-Ç?ÅBQuickTime? Ç?Photo - JPEG êLí£ÉvÉçÉOÉâÉÄǙDZÇÃÉsÉNÉ̀ÉÉÇ?å©ÇÈÇ…ÇÕïKóvÇ-Ç?ÅB1mm
R&D Items and Development Flow of Blanket
Element Tech.Development
Engineering Scale Evaluation
Pebble Bed Fabrication Technol.
Thermo-Hydraulics of High Heat Flux Cooling
Endurance and PropertiesEvaluation of Pebble Bed
In-pile Irradiation Test Technol.
Thermo-hydraulicsof Coolant Network
Partial Module Irradiation Tests, PIE Facility
Integrated Functional Testby Large Scale Mockup
Thermo-Mechanichs ofPebble and ContainerIncluding Compatibility
Therm. Mech, Characteristicsof Blanket Structure
In-pile IrradiationPerformance Evaluation
Breeder / Multiplier Pebble Fabrication Tech.
First Wall, BoxFabrication Technol.
Neutronics ofBlanket SystemTritium Production
Neutronics Tests
Pebble Bed Thermo-Mechanichs
Chemical Compatibilityof Breed./ Mult. Pebble Beds
TB
M F
abrication & Installa
tion in
ITE
R
Thermo-Hydro Dynamics of Coolant
Fusion Neutronics Performance Evaluation
Soundness, Safety, Chemical Compatibility
Mass Fabrication Technol.for Breed. and Multipl.
Pebble FabricationTechnology
Blanket Safety Basic Tests, Coolant Corrosion and Permeation
SafetyDemonstration
Tritium Permeation and Barrier
Structure Corrosionby Coolant Flow
Advanced Materials FabricationBreeder Reprocessing Technol.
Tritium Recovery ElementaryTechnology Development
Tritium RecoverySystem Demonstration
Tritium Recovery System EvaluationBlanket Tritium
Recovery Technology
Development of Integrated Simulation Code for Blanket Tritium Behavior
Blanket System TPR Evaluation
Blanket TPR Confirmationby Simulated Mocups
Irradiation Mockup Design
20102005 2014
FW/SW Assembling Tech.
Simulation Code Development and Experimental Verification
Corrosion Rate Evaluation
FW/ SW Assembly Mockup Fabrication, BW Fabrication Tech.
2009TBM Large Mockup
Function Tests
Development of High Temperature, Long Life Breeder / Multiplier Pebble Materials
Fusion Neutron Tritium Recovery Experiment
Advanced Tritium Recovery Process Developmentt
QuickTimeý DzPhoto - JPEG êLí£ÉvÉçÉOÉâÉÄǙDZÇÃÉsÉNÉ̀ÉÉǾå©ÇÈÇ…ÇÕïKóvÇ ÇÅBQuickTimeý DzPhoto - JPEG êLí£ÉvÉçÉOÉâÉÄǙDZÇÃÉsÉNÉ̀ÉÉǾå©ÇÈÇ…ÇÕïKóvÇ ÇÅB1mm
QuickTimeý DzPhoto - JPEG êLí£ÉvÉçÉOÉâÉÄǙDZÇÃÉsÉNÉ̀ÉÉǾå©ÇÈÇ…ÇÕïKóvÇ ÇÅBQuickTimeý DzPhoto - JPEG êLí£ÉvÉçÉOÉâÉÄǙDZÇÃÉsÉNÉ̀ÉÉǾå©ÇÈÇ…ÇÕïKóvÇ ÇÅB1mm
Progress of Fabrication Technology Development
2006First Wall
2009FW/SW
Assembly
2007Pebble BedContainer
2008Side Wall
2012Large Scale
TBM Mockup
2010-2011Back Wall
2021Installation
to ITER
2015TBM
Fabrication start
7
Real Scale FW Mockup and Heat Flux Test
176mm
25mm
Cross-section of First Wall
8mm
Coolant Inlet
CoolantOutlet
The mockup was high heat flux tested with a heat load of 0.5 MW/m2, 30sec for 80 cycles.
Neither hot spots nor thermal degradation were observed.
Expected heat removal performance was demonstrated.
HHF tests in DATS facilityPeak Heat Flux: 0.5 MW/m2
Beam Duration: 30 sWater Temperature: 300 oCWater Pressure: 15 MPaFlow Velocity: 2 m/s
8
10mm x 1450mmL drilling
1.55
m(
C. c
hann
el 1
.45m
)
0.4m
1 mm accuracy was achieved at the end of 1.45 m depth drilled holes.
Fabrication of Real Scale Side Wall
WCSB TBM
- Real scale Side Wall was fabricated. Cooling channels were machined by drilling.
-10 mm x 1450 mm L cooling channels were formed within 1 mm accuracy at the end of the drilled holes. 1700 mm L is available.
Fabricated Side Wall
17701770
232
232
Welding test of 1/1 scale FW/SW using thick plates
9
Fabrication of FW and SW Assembly Mockup
1/1 FW mockup (with cooling channels) and 1/1 SW mockups (with cooling channels) were assembled by EB welding.
Distortion on FW side is less than 1mm, and distortion in hight is less than 3mm. Welding soundness was inspected by UT.
Welding technique and procedure, welding support were confirmed.
FW
SW1.5 m
0.4 m
1489.2, 1488
1489.5, 1489
417.6417.7
417.7417.6
417.8418.3
417.7417.6
1490
417
.5
FW-SW Assembly Mockup
EB Welding Support
EB Welding
10
FW/SW Assembly Mockup after Assembling process on the welding support
Fabrication of Back Wall Partial Mockup by Normal Steel
232 mm
90 mm
60 mmEB Welding
Cooling Channel
Coolant Header
A partial mockup of the back wall of the Test Blanket Module, which has major structure feature of coolant channels, header and a shear key, was fabricated by using conventional steel, by EB welding of the shear key.
The fabrication technique and procedure for the back wall were confirmed.
11
Experimental Apparatus for Flow Assistred Corrosion of Structural Material by High Temperature and Pressure Water Flow
Flow Assisted Corrosion Experimental Apparatus by High Pressure and Temperature Water
- A disc of a test material is rotated in an autoclave of high pressure and temperature water.
- Test specimen of 100mm diameter disk is rotated up to 2000 rpm. Equivalent superficial velosity at the edge of the disk is 10 m/s
- Water condition is available up to 340 oC 15MPa.
- Flow parameter is estimated by comparison between Flow Visualization Experiments and numerical simulation.
12Flow Visualization Experiments
Numerical Simulation
Pressure Cylinder
Rotating Test Piece
Hydraulic Analysis of the Rotation Disc Test Section
Hydraulic Analysis of the Rotation Disc Test Section
It was found that the shear stress is higher than 740 Pa where corrosion layer was peeled off.
3m/s (300℃,15MPa) Corrosion layer was peeled off.
0.7
8.7
4.3
( m / s )
Hydraulic Simulation
Experiments of Flow Assisted Corrosion by High Pressure and Temperature Water
Experiments of Flow Assisted Corrosion by High Pressure and Temperature Water
Vel
ocity
[m/s
]
Distance from the center [mm]
Observed Flow Velocity Distribution (7mm above the disk)Consistency with the simulation was confirmed.
Visualization Experiment
1 m/s No Flow3 m/s
5.2 m/s
Trace Particle
Evaluation of Corrosion of Structural Material by High Temperature and Pressure Water Flow
Flow directionShear Stress on Wall
Header
Blanch Channel
ww
223
213
212
Re=9.4× 10 5 (5.0 m/s)
(Pa)
(m/s)
平均流速流線(m/s)
Flow Line
Re=3.6×105
(4.4 m/s)
Hydraulic analysis of coolant flow in Side Wall headers and channels.
By Hydraulic analysis, it was clarified that shear stress of more than 740 Pa appeared near the part where coolant split into blanch channel from the header.
13
Development of Advanced Neutron Multiplier PebbleBeryllide synthesis process -Plasma sintering -
Raw material powder
Punch and Die unit
Additions of : 1) Pressure 2) Current (for activation and heating)
The plasma sintering direct sintering from material powder - Enhancing powder particle activeness for sintering - Reducing high temperature exposure
Plasma Sintering Conditions
Raw material : Be & Ti powder
Powder purity : >99wt%
Powder size : <50µm
Sintering time : 20min
Pressure : 50MPa
Temperature : 1273K
XRD profiles and EPMA analysis for clarification of sintering temp.
[at 1273K]
Be Be2Ti
Be17Ti2Be12Ti(Beryllides: Be12Ti, Be17Ti2 and Be2Ti)
Be : 2%Beryllides: 98%
(1) It was shown that spark plasma sintering is applicable for synthesis of Be12Ti.
(2) By the experiments of sintering temperature effect on Be12Ti synthesis, It was clarified that sintering in 1273 K achieved largest fraction of Be12Ti.
Blackened by reduction
No change
White sample(Li2TiO3)
Li2TiO3
with added Li
The color of Li2TiO3 changed from white to black in a hydrogen atmosphere at high temperatures. This color-change corresponds to reduction of Li2TiO3.
Li2TiO3
without added Li
After heating at 1273K for 10h in 1%H2-He
Development of Advanced Tritium Breeder Durable to Reduction in Hydrogen Atomoshpere
Development of Li2TiO3 with excess Li to improve the resistance of reduction at high temperatures
In the case of Li2TiO3 with added Li, the color did not change, indicating that this sample was not reduced in the hydrogen atmosphere. Chemical Stability
Gel
Water
GelationLiOH•H2O and H2TiO3
Gel-spheres
Li2TiO3 with excess Li
Diameter 1.18mm
Sphericity 1.04
Density 89%T.D
Grain size 2 - 10μm
Development of Pebble Fabrication Technology of Li2TiO3 with excess composition of Li
Trial fabrication of pebbles of Li2TiO3 with excess Li composition by sol gel method
Raw material Granulation SinteringSlurryLi2TiO3 with excess Li
Pebbles of Li2TiO3 with excess Li was granulated by sol gel method from slurry.
Li2TiO3 with excess Li was synthesized from mixed LiOH•H2O and H2TiO3
Sol-gel method is applicable in pebble fabrication of Li2TiO3 with excess amount of Li.
Advanced Tritium Recovery Technology Development - Principle -
Study on Application of Electrochemical Hydrogen Pump (Ceramic Proton Conductor) as a continuously operating HT and HTO recovery process
Driving force of tritium extraction•Pressure difference•Electric potential difference
Experimental validation of Tritium transport property to evaluate applicability, using tritium gasExperimental conditions
Sweep Gas In (H2, HT, H2O, HTO/He
Sweep Gas Out (He/O2
Principle of Electrochemical Hydrogen Pump
Voltage
Advanced Tritium Recovery Technology Development - Result of transport property measurement -
Tritium was extracted by applying voltage. DF and recovery ratio were 1.5 and 0.4.
- Principle was demonstrated.- One-through Decontamination Factor and T recovery rate
were 1.5 and 0.4 by a single tube with 0.2 l/min He + HT gas flow.
- Scale up is the further issue for adopting this principle.
PREVIOUS EXPERIMENT (Offline) Tritium Recovery Experiment from Li Ceramic Breeding Material
Irradiated with DT NeutronsWe have conducted a tritium recovery experiment for solid breeding blanket with DT neutrons for the first time in the world.
Tritium recovery measurement
DT neutron irradiation arrangement
The experiment shows the tritium recovery ratio for the mock-up is 1.05 0.08 at 873 K, which indicates that the design of Japanese solid breeder blanket promises a good prospect of tritium recovery up 873 K.
Tritium production measurementPebble dissolution method with a weak acid (HCl)
JAEA/FNSDT neutron source
Beryllium bulk
Breeding material Container
Gas cylinderHe gas (H2 1.04%)
MFC
100sccmTC
CoolantAir in
Purge Air InLi2TiO3 pebble67g 6-Li: 7.5%)
Heater
Up to 873 K
Heater
773K
CuO Bed100.0g
CoolantAir out
Compressor
Bubbler 1 Bubbler 2
Silica Gel
(124.3g)(water : 100cc/bottle)
MFC
100sccm
Pump(for purge)
Purge Airout
Experimental Setup for On-line Measurement of DT Neutron Production Experiment
• The neutron intensity was about 1.5 x 1011 neutron/sec.
• The sweep gas He + H2 1.04% flow rate kept 100 standard cm3/min
• After the irradiation, water vapor fraction in the sweep gas line was measured with a dew-point meter. It was an order of 1000 ppm.
• After the run, 1 cm3 water in each trap bottle was mixed into a liquid scintillator and measured with a liquid scintillation counter (LSC), which was calibrated with a standard HTO (50 Bq/cc) sample within 2 % accuracy.
Schematic view of the DT Neutron Tritium Recovery Online Experiment
Li2TiO3 Pebble Bed (6Li 7.5%)
Heater
Former Trap Bottles for HTO (H2O
100cm3/bottle)
Exhaust Gas
Latter Trap Bottles forHTO (H2O 100cm3/bottle)
Trap bottle change by remote handling
MFC 100cm3/min
DT Neutron Source(1.5 x 1011/n/sec)
Gas Cylinder(He+H2 1.04%)
CuO Bed (100g)For Oxidization of HTO
Concrete Wall (2m thick)
Be Block Assembly JAEA/FNSDT neutron source
Beryllium bulk
Breeding material Container
DT neutron irradiation experimentBreeder capsule arrangement
Experimental Setup for On-line Measurement of DT Neutron Production Experiment
• The neutron intensity was about 1.5 x 1011 neutron/sec. • The sweep gas He + H2 1.04% flow rate kept 100 standard cm3/min
• After the irradiation, water vapor fraction in the sweep gas line was measured with a dew-point meter. It was an order of 1000 ppm.
• After the run, 1 cm3 water in each trap bottle was mixed into a liquid scintillator and measured with a liquid scintillation counter (LSC), which was calibrated with a standard HTO (50 Bq/cc) sample within 2 % accuracy.
Schematic view of the DT Neutron Tritium Recovery Online Experiment
Li2TiO3 Pebble Bed (6Li 7.5%)
Heater Concrete Wall (2m
thick)
Be Block AssemblyFormer Trap Bottles forHTO (H2O
100cm3/bottle)
Heater (773K)
Exhaust Gas
Latter Trap Bottles forHTO (H2O
100cm3/bottle)
Trap bottle change by remote handling
MFC 100cm3/min
DT Neutron Source(1.5 x 1011/n/sec)
Gas Cylinder(He+H2 1.04%)
CuO Bed (100g)For Oxidization of HTO
DT neutron irradiation experimentOn-line tritium measurement setup
Water bottles for tritium recovery
Breeder Capsule
Result
In order to deduce the tritium recovery ratio, we adopted our previously measured TPR data, 7.46 x 10-14 Bq/g/DT neutron with experiment error of 8 %. As a result, the present tritium recovery ratio was 0.96. It is indicated that the tritium recovery of Japanese TBM has a good prospective at 573 K.
It is also shown from the result that the total recovered HTO was significantly larger than the total recovered HT and its ratio is about 0.9. It was considered that larger HTO recovery was due to larger water vapor (1000 ppm) in the sweep gas line. It seems that the tritium produced in the Li2TiO3 pebbles easily reacts on water vapor rather than H2 in the sweep gas at such low temperature. In future, we will conduct an additional experiment with a cold trap system (e.g. dry ice and/or molecular sieve) in the sweep gas line.
From the measurement, HT release showed delay compared with HTO release.
The horizontal axis is elapsed time of tritium recovery and the vertical one is fraction of recovered tritium radioactivity
The total tritium radioactivity was about 8.66 kBq. The number of DT neutron irradiation was 1.74 x 1015. Thus the TRR was 7.11 x 10-14 Bq/g/DT neutron (normalized in Li2TiO3 weight and neutron flux)
-1 0 1 2 3 4 5 6 7 80.0
0.1
0.2
0.3
Total T = 8.66 kBq
Fra
ctio
n (
pe
r to
tal t
ritiu
m B
q)
Irradiation time (hour)
HTO HT
Temperature 573 K
DT neutronirradiation TRR/TPR =0.96
HT
HTO
Conclusions
1. In the fabrication technology development of WCCB TBM, real scale F82H First Wall and Side Walls were successfully assembled with enough small distortion. Also, partial mockup of the Back Wall was fabricated to confirm the fabrication route.
2. In the advanced multiplier and breeder pebble development for DEMO blanket, Be12Ti rod , pebble of Li2TiO3 with excess Li composition, which have increased chemical stability in high temperature, were clarified.
3. In Advanced Tritium Recovery Technology Development, the principle of Electrochemical Hydrogen Pump was demonstrated and basic tritium recovery property was clarified. It was observed that scale up is a further issue.
4. Tritium Production and On-line Recovery Experiment by DT neutron irradiation at JAEA-FNS showed that the tritium recovery ratio was 0.96 0.08 compared to the evaluation by neutronics experiments. It was expected that the tritium recovery data is used for verification of Tritium production performance of TBM.
24