- Количество слайдов: 24
16 th International Workshop on CERAMIC BREEDER BLANKET INTERACTIONS Portland, USA, 11 -16, September, 2011 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 1
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 Li 2 Ti. O 3 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 Test Blanket Module（TBM） Water Loop ITER Cross Section Generator Plasma Structure of RAFM (F 82 H) 1700 mm -Production of fusion fuel tritium - Extraction of energy Neutron Multiplier Pebble Bed (Be) Tritium Breeder Pebble Bed (Li 2 Ti. O 3) 500 mm 600 mm Water Cooled Ceramic Breeder Provisional Port Allocation (WCCB) TBM proposed by Japan 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, Li. Pb-based TBMs and Li-based TBM. 3
Provisional WCCB TBM Delivery Schedule JA 18 PY AD 06 19 07 20 08 21 09 23 11 24 25 26 27 12 13 14 15 ITER Construction 28 16 29 17 30 18 32 33 31 20 21 19 ITER Operation TBM #1 Fabrication TBM #1 Safy Report TBM Delivery FDR PDR CDR TBM PA Pr. SR Tentative Milestones, under discussion 22 10 Material Proc. , Parts Manufact. TBM Installat PPTS ion in Module Manufact. , Assembling Testing ITER Ancillary System Manufact. 0. 6 m TBMFabric. Tech. Eng R&D Large Size Mockupo Testing 1/1 TBM 1/1 BW 1/1 Breeder Box 1/1 FW Fabr. Box MU HHF Tests HHF Testing 1/1 BW Fab. - Large Size Mockup Fabr. - LSMU Internal Pressure Test - Detailed Fabr. Validation - RH Demonstration 1/1 BW 1/1 FW 1/1 Breeder 1/1 SW Assembling Plate BW/Box Prototype #2 TBM Fabrication Mockuo Box Mock- Mockuo Assembling - Structure Endurance Tests (Errosion-corrosion, Be pebble bed) of FW/SW Fabrication - Water Ingress Safety Fabricat. up Fabrication - Ancil. System Testing Validation Tests - HHF Tests/Heating Tests, function tests of prototype 1. 66 m Advanced Breeder, Multiplier Development Tritium Recovery Process Development Nuclear Performance Validation Tests Structural Material Validation/ Database . 484 m WCCB TBM 4
Cooperation of Solid Breeder Blanket Development System Integration Out-pile R&D, Module fabrication technology, Thermal hydraulic research, - Blanket Engineering Lab. (JAEA) DEMO Reactor System Plasma Facing Materials - Prof. Hino (Hokkaido Univ. ) - Prof. Tanabe (Kyushu Univ. ) - Prof. Ueda (Osaka Univ) Neutronics / Tritium Production Tests with 14 Me. V neutrons - Fusion Neutronics Gr. (JAEA) Cooling 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) 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 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. ) 5
R&D Items and Development Flow of Blanket Element Tech. Development First Wall, Box Fabrication Technol. Thermo-Hydro Dynamics of Coolant Thermo-Hydraulics of High Heat Flux Cooling Pebble Fabrication Technology Chemical Compatibility of Breed. / Mult. Pebble Beds Soundness, Safety, Chemical Compatibility In-pile Irradiation Performance Evaluation Structure Corrosion by Coolant Flow Thermo-hydraulics of Coolant Network Endurance and Properties Evaluation of Pebble Bed FW/ SW Assembly Mockup Fabrication, BW Fabrication Tech. Advanced Materials Fabrication Breeder Reprocessing Technol. Mass Fabrication Technol. for Breed. and Multipl. Development of High Temperature, Long Life Breeder / Multiplier Pebble Materials Blanket Safety Basic Tests, Coolant Corrosion and Permeation Safety Demonstration Corrosion Rate Evaluation Development of Integrated Simulation Code for Blanket Tritium Behavior Simulation Code Development and Experimental Verification Irradiation Mockup Design Partial Module Irradiation Tests, PIE Facility Neutronics of Blanket System TPR Evaluation Blanket TPR Confirmation by Simulated Mocups In-pile Irradiation Test Technol. Tritium Production Neutronics Tests FW/ SW Asse mblin g Tech. Integrated Functional Test by Large Scale Mockup Tritium Permeation and Barrier Fusion Neutronics Performance Evaluation Blanket Tritium Recovery Technology Thermo-Mechanichs of Pebble and Container Including Compatibility 2014 TBM Large Mockup Function Tests Fusion Neutron Tritium Recovery Experiment Tritium Recovery System Evaluation Tritium Recovery Elementary Technology Development Tritium Recovery System Demonstration Advanced Tritium Recovery Process Developmentt - 6 - TBM Fabrication & Installation in ITER Pebble Bed Thermo-Mechanichs 2009 2010 Engineering Scale Evaluation Pebble Bed Fabrication Technol. Therm. Mech, Characteristics of Blanket Structure Breeder / Multiplier Pebble Fabrication Tech. 2005
Progress of Fabrication Technology Development 2006 First Wall 2009 FW/SW Assembly 2021 Installation to ITER 2010 -2011 Back Wall 2007 Pebble Bed Container 2012 Large Scale TBM Mockup 2008 Side Wall 2015 TBM Fabrication start 7
Real Scale FW Mockup and Heat Flux Test Coolant Outlet Coolant Inlet 8 mm 25 mm 176 mm Cross-section of First Wall 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 q The mockup was high heat flux tested with a heat load of 0. 5 MW/m 2, 30 sec for 80 cycles. q Neither hot spots nor thermal degradation were observed. q Expected heat removal performance was demonstrated. 8
Fabrication of Real Scale Side Wall - Real scale Side Wall was fabricated. Cooling channels were machined by drilling. -10 mmf x 1450 mm L cooling channels were formed within 1 mm accuracy at the end of the drilled holes. 1700 mm L is available. 1 mm accuracy was achieved at the end of 1. 45 m depth drilled holes. 232 （C. c 1. 55 hann m el 1. 45 m ） 17 70 0. 4 m WCSB TBM 10 mmf x 1450 mm. L drilling Fabricated Side Wall 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 1 mm, and distortion in hight is less than 3 mm. Welding soundness was inspected by UT. Welding technique and procedure, welding support were confirmed. EB Welding FW-SW Assembly Mockup 1. 5 m 0. 4 m SW FW EB Welding Support FW/SW Assembly Mockup after Assembling process on the welding support 417. 7 417. 6 1489. 2, 1488 1489. 5, 1489 1490 417. 8 418. 3 417. 5 417. 6 417. 7 10
Fabrication of Back Wall Partial Mockup by Normal Steel 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. 60 mm EB Welding 232 mm Coolant Header Cooling Channel 90 mm 11
Experimental Apparatus for Flow Assistred Corrosion of Structural Material by High Temperature and Pressure Water Flow - A disc of a test material is rotated in an autoclave of high pressure and Rotating temperature water. Test Piece - Test specimen of 100 mm 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 Pressure Cylinder 15 MPa. Flow Assisted Corrosion Experimental Apparatus- Flow parameter is estimated by comparison between Flow Visualization by High Pressure and Temperature Water Experiments and numerical simulation. Numerical Simulation Flow Visualization Experiments 12
Evaluation of Corrosion of Structural Material by High Temperature and Pressure Water Flow 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 He Re Flow where coolant split into blanch channel from the header. ad =9. 4 × 1 05 Shear Stress on Wall dir ect er ion (5. 0 m /s ) 2 2 t w = ( 12 + t 13 + t 23 ) t 2 w (Pa) 8. 7 4. 3 0. 7 Flow Line 平均流速流線 (m/s) Blanch Channel Re=3. 6 × 105 (4. 4 m/s) 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. Experiments of Flow Assisted Corrosion by High Pressure and Temperature Water 3 m/s (300℃, 15 MPa) Corrosion layer was peeled off. 5. 2 m/s Velocity [m/s] Observed Flow Velocity Distribution (7 mm above the disk) Consistency with the simulation was confirmed. Visualization Experiment 3 m/s Trace Particle Hydraulic Simulation Distance from the center [mm] 1 m/s No Flow 13
Development of Advanced Neutron Multiplier Pebble Beryllide synthesis process -Plasma sintering Plasma Sintering Conditions Additions of : 1) Pressure 2) Current (for activation and heating) Raw material Powder purity : >99 wt% Raw material powder Punch and Die unit : Be & Ti powder Powder size : <50µm Sintering time : 20 min Pressure : 50 MPa Temperature : 1273 K The plasma sintering direct sintering from material powder - Enhancing powder particle activeness for sintering - Reducing high temperature exposure
XRD profiles and EPMA analysis for clarification of sintering temp. Be 2 Ti Be Be : 2% Beryllides: 98% [at 1273 K] Be 12 Ti Be 17 Ti 2 (Beryllides: Be 12 Ti, Be 17 Ti 2 and Be 2 Ti) (1) It was shown that spark plasma sintering is applicable for synthesis of Be 12 Ti. (2) By the experiments of sintering temperature effect on Be 12 Ti synthesis, It was clarified that sintering in 1273 K achieved largest fraction of Be 12 Ti.
Development of Advanced Tritium Breeder Durable to Reduction in Hydrogen Atomoshpere q Development of Li 2 Ti. O 3 with excess Li to improve the resistance of reduction at high temperatures Li 2 Ti. O 3 without added Li White sample (Li 2 Ti. O 3) Blackened by reduction After heating at 1273 K for 10 h in 1%H 2 -He Li 2 Ti. O 3 with added Li No change u The color of Li 2 Ti. O 3 changed from white to black in a hydrogen atmosphere at high temperatures. This color-change corresponds to reduction of Li 2 Ti. O 3. u In the case of Li 2 Ti. O 3 with added Li, the color did not change, indicating that this sample was not reduced in the hydrogen atmosphere. Chemical Stability
Development of Pebble Fabrication Technology of Li 2 Ti. O 3 with excess composition of Li q Trial fabrication of pebbles of Li 2 Ti. O 3 with excess Li composition by sol gel method Raw material Li. OH • H 2 O and H 2 Ti. O 3 Gelation Granulation Slurry Li 2 Ti. O 3 with excess Li Diameter Water Gel Li 2 Ti. O 3 with excess Li Sintering 1. 18 mm Sphericity 1. 04 Gel-spheres Density 89%T. D Grain size 2 - 10μm u Li 2 Ti. O 3 with excess Li was synthesized from mixed Li. OH • H 2 O and H 2 Ti. O 3 u Pebbles of Li 2 Ti. O 3 with excess Li was granulated by sol gel method from slurry. Sol-gel method is applicable in pebble fabrication of Li 2 Ti. O 3 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 Sweep Gas In (H 2, HT, H 2 O, HTO/He Voltage Sweep Gas Out (He/O 2 Principle of Electrochemical Hydrogen Pump Driving force of tritium extraction • Pressure difference • Electric potential difference n Experimental validation of Tritium transport property to evaluate applicability, using tritium gas Experimental conditions
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 Neutrons We have conducted a tritium recovery experiment for solid breeding blanket with DT neutrons for the first time in the world. Tritium production measurement Pebble dissolution method with a weak acid (HCl) Tritium recovery measurement Li 2 Ti. O 3 pebble JAEA/FNS DT neutron source Coolant Air out Beryllium bulk Purge Air In 67 g 6 -Li: 7. 5%) TC Heater Coolant Air in K Up Cu. O Bed to 873 Breeding material Container Heater 100. 0 g 773 K 100 sccm MFC Gas cylinder He gas (H 2 1. 04%) Compressor 100 sccm Purge Air out MFC DT neutron irradiation arrangement Silica Gel Bubbler 1 Bubbler 2 (124. 3 g) (water : 100 cc/bottle) Pump (for purge) The experiment shows the tritium recovery ratio for the mockup 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.
Experimental Setup for On-line Measurement of DT Neutron Production Experiment Li 2 Ti. O 3 Pebble Bed (6 Li 7. 5%) Heater Concrete Wall (2 m thick) MFC 100 cm 3/min Former Trap Be Block Assembly Bottles for HTO (H 2 O DT Neutron Source 100 cm 3/bottle) (1. 5 x 1011/n/sec) Gas Cylinder (He+H 2 1. 04%) Cu. O Bed (100 g) For Oxidization of HTO Trap bottle change by remote handling Exhaust Gas Latter Trap Bottles for HTO (H 2 O 100 cm 3/bottle) JAEA/FNS DT neutron source Beryllium bulk Breeding material Container DT neutron irradiation experiment Breeder capsule arrangement Schematic view of the DT Neutron Tritium Recovery Online Experiment • • The neutron intensity was about 1. 5 x 1011 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.
Experimental Setup for On-line Measurement of DT Neutron Production Experiment Li 2 Ti. O 3 Pebble Bed (6 Li 7. 5%) Heater Concrete Wall (2 m thick) MFC 100 cm 3/min Former Trap Be Block Assembly Bottles for HTO (H 2 O DT Neutron Source 100 cm 3/bottle) (1. 5 x 1011/n/sec) Trap bottle change by remote handling Gas Cylinder (He+H 2 1. 04%) Cu. O Bed (100 g) For Oxidization of HTO Heater (773 K) Exhaust Gas Latter Trap Bottles for HTO (H 2 O 100 cm 3/bottle) Breeder Capsule Water bottles for tritium recovery DT neutron irradiation experiment On-line tritium measurement setup Schematic view of the DT Neutron Tritium Recovery Online Experiment • • The neutron intensity was about 1. 5 x 1011 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.
Result Fraction (per total tritium Bq) The total tritium radioactivity was about 8. 66 k. Bq. 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 Li 2 Ti. O 3 weight and neutron flux) Temperature 573 K 0. 3 0. 2 HTO HT DT neutron irradiation TRR/TPR =0. 96 Total T = 8. 66 k. Bq HTO 0. 1 HT 0. 0 -1 0 1 2 3 4 5 6 Irradiation time (hour) 7 The horizontal axis is elapsed time of tritium recovery and the vertical one is fraction of recovered tritium radioactivity 8 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 Ti. O 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.
Conclusions 1. In the fabrication technology development of WCCB TBM, real scale F 82 H 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, Be 12 Ti rod , pebble of Li 2 Ti. O 3 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