1. 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 on CERAMIC BREEDER BLANKET INTERACTIONS Portland, USA, 11-16, September, 2011 1

2. Contents 1. Importance of Water Cooled Ceramic Breeder (WCCB) Test Blanket and assumed schedule of WCCB TBM R&D 2. Domestic cooperation and R&D flow chart 3. Module fabrication technology development - First Wall + Side Wall Assembly, Back Wall Partial Mockup 4. Advanced breeder and multiplier pebble development for DEMO - Increased chemical stability and soundness in high temperature 5. Advanced tritium recovery technology for blanket system 6. Neutronics engineering - Verification of tritium recovery rate of Li2TiO3 pebble bed with DT neutron in FNS 2

3. 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 Generator -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 600mm 500mm Water Cooled Ceramic Breeder (WCCB) TBM proposed by Japan Test Blanket Module （ TBM ） Structure of RAFM (F82H) Neutron Multiplier Pebble Bed (Be) Tritium Breeder Pebble Bed (Li 2 TiO 3 ) ITER Cross Section Provisional Port Allocation

4. 31 18 192021222324252627282930 1906 070809101112131415161718 ITER Construction ITER Operation PrSR FDR Tentative Milestones, under discussion JA PY AD 1/1 Breeder Box MU Testing 1/1 FW HHF Test TBMFabric. Tech. Eng R&D 1/1 TBM Box HHF Test 1/1 FW Mockuo Fabricat. 1/1 Breeder Box Mock- up Fabr. 1/1 SW Mockuo Fabrication Advanced Breeder, Multiplier Development Tritium Recovery Process Development Nuclear Performance Validation Tests Structural Material Validation/ Database PDRCDR TBM #1Safy Report Installat ion in ITER BW/Box Assembling 1/1 BW Fab. 1/1 BW Plate Fabrication 4 Material Proc., Parts Manufact. Ancillary System Manufact. Module Manufact., Assembling TBM #1 Fabrication TBMDelivery 1/1 BW Fabr. Tests Assembling of FW/SW Provisional WCCB TBM Delivery Schedule 20 21 TBM PA 1.66m.484m 0.6m WCCB TBM 32 33 TBM PPTS 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

5. Cooperation of Solid Breeder Blanket Development System Integration Out-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) Tritium Extraction System DEMO Reactor System 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 System Development, 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. - 6 -- 6 - 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 Properties Evaluation of Pebble Bed In-pile Irradiation Test Technol. Thermo-hydraulics of Coolant Network Partial Module Irradiation Tests, PIE Facility Integrated Functional Test by Large Scale Mockup Thermo-Mechanichs of Pebble and Container Including Compatibility Therm. Mech, Characteristics of Blanket Structure In-pile Irradiation Performance Evaluation Breeder / Multiplier Pebble Fabrication Tech. First Wall, Box Fabrication Technol. Neutronics of Blanket System Tritium Production Neutronics Tests Pebble Bed Thermo-Mechanichs Chemical Compatibility of Breed./ Mult. Pebble Beds TBM Fabrication & Installation in ITER Thermo-Hydro Dynamics of Coolant Fusion Neutronics Performance Evaluation Soundness, Safety, Chemical Compatibility Mass Fabrication Technol. for Breed. and Multipl. Pebble Fabrication Technology Blanket Safety Basic Tests, Coolant Corrosion and Permeation Safety Demonstration Tritium Permeation and Barrier Structure Corrosion by Coolant Flow Advanced Materials Fabrication Breeder Reprocessing Technol. Tritium Recovery Elementary Technology Development Tritium Recovery System Demonstration Tritium Recovery System Evaluation Blanket Tritium Recovery Technology Development of Integrated Simulation Code for Blanket Tritium Behavior Blanket System TPR Evaluation Blanket TPR Confirmation by Simulated Mocups Irradiation Mockup Design 2010 20052014 FW/ SW Asse mbli ng Tech. Simulation Code Development and Experimental Verification Corrosion Rate Evaluation FW/ SW Assembly Mockup Fabrication, BW Fabrication Tech. 2009 TBM Large Mockup Function Tests Development of High Temperature, Long Life Breeder / Multiplier Pebble Materials Fusion Neutron Tritium Recovery Experiment Advanced Tritium Recovery Process Developmentt

7. Progress of Fabrication Technology Development 2006 First Wall 2009 FW/SW Assembly 2007 Pebble Bed Container 2008 Side Wall 2012 Large Scale TBM Mockup 2010-2011 Back Wall 2021 Installation to ITER 2015 TBM Fabrication start 7

8. Real Scale FW Mockup and Heat Flux Test 176mm 25mm Cross-section of First Wall 8mm Coolant Inlet Coolant Outlet  The mockup was high heat flux tested with a heat load of 0.5 MW/m 2, 30sec for 80 cycles.  Neither hot spots nor thermal degradation were observed.  Expected heat removal performance was demonstrated. HHF tests in DATS facility Peak Heat Flux: 0.5 MW/m 2 Beam Duration:30 s Water Temperature: 300 o C Water Pressure: 15 MPa Flow Velocity:2 m/s 8

9. 10mm  x 1450mmL drilling 1.55m （ C. channel 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 1770 232 Welding test of 1/1 scale FW/SW using thick plates 9

10. 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 SW 1.5 m 0.4 m 1489.2, 1488 1489.5, 1489 417.6 417.7 417.6 417.8 418.3 417.7 417.6 1490 417.5 FW-SW Assembly Mockup EB Welding Support EB Welding 10 FW/SW Assembly Mockup after Assembling process on the welding support

11. Fabrication of Back Wall Partial Mockup by Normal Steel 232 mm 90 mm 60 mm EB 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

12. 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 o C 15MPa. - Flow parameter is estimated by comparison between Flow Visualization Experiments and numerical simulation. 12 Flow Visualization Experiments Numerical Simulation Pressure Cylinder Rotating Test Piece

13. 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 Velocity [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 direction Shear Stress on Wall Header Blanch Channel  w w 2 23 2 13 2 12  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 × 10 5 (4.4m/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

14. Development of Advanced Neutron Multiplier Pebble Beryllide 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

15. XRD profiles and EPMA analysis for clarification of sintering temp. [at 1273K] BeBe 2 Ti Be 17 Ti 2 Be 12 Ti (Beryllides: Be 12 Ti, Be 17 Ti 2 and Be 2 Ti) 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.

16. Blackened by reduction No change White sample (Li 2 TiO 3 ) Li 2 TiO 3 with added Li  The color of Li 2 TiO 3 changed from white to black in a hydrogen atmosphere at high temperatures. This color-change corresponds to reduction of Li 2 TiO 3. Li 2 TiO 3 without added Li After heating at 1273K for 10h in 1%H 2 -He Development of Advanced Tritium Breeder Durable to Reduction in Hydrogen Atomoshpere  Development of Li 2 TiO 3 with excess Li to improve the resistance of reduction at high temperatures  In the case of Li 2 TiO 3 with added Li, the color did not change, indicating that this sample was not reduced in the hydrogen atmosphere.  Chemical Stability

17. Gel Water Gelation LiOHH 2 O and H 2 TiO 3 Gel-spheres Li 2 TiO 3 with excess Li Diameter1.18mm Sphericity1.04 Density89%T.D Grain size2 - 10μm Development of Pebble Fabrication Technology of Li 2 TiO 3 with excess composition of Li  Trial fabrication of pebbles of Li 2 TiO 3 with excess Li composition by sol gel method Raw materialGranulationSintering Slurry Li 2 TiO 3 with excess Li  Pebbles of Li 2 TiO 3 with excess Li was granulated by sol gel method from slurry.  Li 2 TiO 3 with excess Li was synthesized from mixed LiOHH 2 O and H 2 TiO 3 Sol-gel method is applicable in pebble fabrication of Li 2 TiO 3 with excess amount of Li.

18. 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 gas Experimental conditions Sweep Gas In (H 2, HT, H 2 O, HTO/He Sweep Gas Out (He/O 2 Principle of Electrochemical Hydrogen Pump Voltage

19. 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.

20. PREVIOUS EXPERIMENT (Offline) Tritium Recovery Experiment from Li Ceramic Breeding Material Irradiated with DT Neutrons We 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 measurement Pebble dissolution method with a weak acid (HCl) JAEA/FNS DT neutron source Beryllium bulk Breeding material Container

21. Experimental Setup for On-line Measurement of DT Neutron Production Experiment The neutron intensity was about 1.5 x 10 11 neutron/sec. The sweep gas He + H 2 1.04% flow rate kept 100 standard cm 3 /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 cm 3 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 Li 2 TiO 3 Pebble Bed (6Li 7.5%) Heater Former Trap Bottles for HTO (H 2 O 100cm 3 /bottle) Exhaust Gas Latter Trap Bottles for HTO (H 2 O 100cm 3 /bottle) Trap bottle change by remote handling MFC 100cm 3 /min DT Neutron Source (1.5 x 10 11 /n/sec) Gas Cylinder (He+H 2 1.04%) CuO Bed (100g) For Oxidization of HTO Concrete Wall (2m thick) Be Block Assembly JAEA/FNS DT neutron source Beryllium bulk Breeding material Container DT neutron irradiation experiment Breeder capsule arrangement

22. Experimental Setup for On-line Measurement of DT Neutron Production Experiment The neutron intensity was about 1.5 x 10 11 neutron/sec. The sweep gas He + H 2 1.04% flow rate kept 100 standard cm 3 /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 cm 3 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 Li 2 TiO 3 Pebble Bed (6Li 7.5%) Heater Concrete Wall (2m thick) Be Block Assembly Former Trap Bottles for HTO (H 2 O 100cm 3 /bottle) Heater (773K) Exhaust Gas Latter Trap Bottles for HTO (H 2 O 100cm 3 /bottle) Trap bottle change by remote handling MFC 100cm 3 /min DT Neutron Source (1.5 x 10 11 /n/sec) Gas Cylinder (He+H 2 1.04%) CuO Bed (100g) For Oxidization of HTO DT neutron irradiation experiment On-line tritium measurement setup Water bottles for tritium recovery Breeder Capsule

23. 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 Li 2 TiO 3 pebbles easily reacts on water vapor rather than H 2 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 10 15. Thus the TRR was 7.11 x 10 -14 Bq/g/DT neutron (normalized in Li 2 TiO 3 weight and neutron flux) 012345678 0.0 0.1 0.2 0.3 Total T = 8.66 kBq Fraction (per total tritium Bq) Irradiation time (hour) HTO HT Temperature 573 K DT neutron irradiation TRR/TPR =0.96 HT HTO

24. 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