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Japanese Universities’ Perspective on Li/V TBM

Japanese Universities’ Perspective on Li/V TBM. T. Muroga Fusion Engineering Research Center National Institute for Fusion Science. Position of Li/V Blanket and Li/V ITER-TBM for Japanese Universities Purpose of Li/V ITER-TBM Neutronics examination Consideration to the Russian design

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Japanese Universities’ Perspective on Li/V TBM

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  1. Japanese Universities’ Perspective on Li/V TBM T. Muroga Fusion Engineering Research Center National Institute for Fusion Science

  2. Position of Li/V Blanket and Li/V ITER-TBM for Japanese Universities Purpose of Li/V ITER-TBM Neutronics examination Consideration to the Russian design Present strategy Progress in Key Technology development for Li/V With emphasis on MHD coating development in Japan Outline of Presentation

  3. Approximate calendar year 2015 2020 2030 2040 Advanced Powerplant Design Power Generation Plant Design Construction Operation (Licencing) (Blanket test) ITER Blanket Module Test Materials and Blanket System Development Reference Material (RAFM) and System Advanced Materials (V-alloy, SiC/SiC --) and System IFMIF Irradiation Test, Materials Qualification and System Performance Test (Staged construction and operation) Roadmap for Materials/Blanket Development Fast realization (Mostly JAERI responsibility) Advanced option (Mostly NIFS/University responsibility)

  4. Li/V blanket is categorized into “advanced system” in contrast to RAFM/water blanket as “reference system” General plan for Li/V ITER-TBM is to start the test in the midst of the ITER operation phase However, first-day Li/V TBM will be explored in the following cases If large technological progress is made, we will reconsider the schedule If other parties propose first-day TBM, we will support it and make effort for Japanese idea to be incorporated into the proposal Now Russia is proposing ｆirst-day Li/V-TBM, we will evaluate the proposal and seek for possibility of cooperation Position of Li/V Blanket and Li/V ITER-TBM for Japanese Universities

  5. Feasibility of no-Be and natural Li blanket Use of 7Li reaction for enhancing TBR in contrast to Russian Be+6Li enriched TBM Validation of neutronics prediction Technology integration for V-alloy, Li and T Purpose of Li/V ITER-TBM(current consensus in Japanese Universities)

  6. ITER with Li/V self-cooled blanket - MCNP calculation by T. Tanaka (NIFS) - [ Inboard ] 40 cm SS, H2O A Vacuum vessel Blanket FW Plasma B SS (60%), Li coolant (40%) V-4Cr-4Ti walls, Natural Li [ Outboard ] Coil structure 40 cm 1 m Vacuum vessel + Filler SS, H2O SS, H2O Center solenoid A Blanket Vacuum vessel FW Blanket A : Standard ITEF-FEAT blanket B B : ITER with V/Li full blanket Input geometry for MCNP calculation * V-4Cr-4Ti walls, Natural Li SS (60%), Li coolan (40%) (*Dimensions from ITER Nuclear Analysis Report)

  7. ITER with Li/V self-cooled blanket - Local TBR - Local TBR (Full Coverage)* (* JENDL 3.2) (a) Inboard (b) Outboard FW FW Blanket Blanket Filler Filler Distribution of tritium production rate ■ Tritium self-sufficiency is feasible

  8. Neutron spectrum at first wall of Standard and V/Li Blanket Comparison of Neutron Flux at Outboard First Wall Cross Section for Tritium Production (JENDL 3.2) ■ Significant difference between thermal neutron component in ITER-FEAT and ITER-Li/V ■ Thermal neutron should be shielded in the TBM area of ITER-FEAT for the purpose of simulating V/Li blanket condition

  9. Tentative design of Li/V self-cooled TBM by NIFS/Universities 505 Verification of (1) Coolant circulation (2) MHD coating SS(60%), H2O(40%) Plasma Li layer V-4Cr-4Ti Verification of (1) Neutron transport (2) Tritium production from 7Li SS316 TBM frame 210 1720 210 470 Plasma Inlet/outlet pipes SS(60%), H2O(40%) Li : ~0.027 m3 Li/V TBM (Unit : mm) Tentative design of Li/V TBM ■ Verification of TPR for 7Li ■ Thick Li tanks for verification of neutron transport

  10. Tentative design of Li/V self-cooled TBM - Tritium production - Covering by B4C ■ For verification of tritium production from 7Li (n, na)T reaction Plasma SS(60%), H2O(40%) - Reduction of thermal neutrons by B4C shielding Li layer (1) (2) (3) (4) (5) Li layer (1) Li layer (1) (2) (2) (3) (4) (5) (3) (4) (5) Contribution of 7Li to tritium production Tritium production rate in Li layers

  11. Experimental parameter for Li/V TBM - Adjustment by B4C shield - 10 cm in front side Li/V TBM 10 cm in rear side 10 cm in front side Li/V blanket 10 cm in rear side Russian TBM Changes in contribution of 7Li by B4C covering ■ Contribution of 7Li to TPR can be adjusted by thickness of B4C shield

  12. Russian Li/V self-cooled test blanket module- Structure - 505 WC Shield (Reflector) Be multiplier SS(60%) + H2O(40%) Plasma 1720 V-5Cr-5Ti Li layer (6Li : 90%) Structure of Russian Li/V TBM (Unit : mm) ■6Li enriched coolant(7.5 % ==> 90%) ■ Li layer x 2, Be multiplier ==> 6Li (n, a) T ■ Maximize the 6Li reaction to demonstrate DEMO reactor breeding tritium by 6Li

  13. Be WC Plasma SS+H2O Li layer (1) Li layer (2) [6Li : 90%] Russian Li/V self-cooled test blanket module- Tritium production - Total : 0.09 (g/FPD) SUS + H2O Plasma Li layer (1) Li layer (2) TBM surface Tritium production rate in Li layers and contribution of 6Li and 7Li SS316 TBM frame Li/V TBM

  14. Is Russian Li-Be-V an Attractive Option? • “No-beryllium” is an attractive potentiality of Li/V system • Does Li-Be-V have alternative merits compensating the demerit of using Be? • High TBR? – Excess TBR probably not necessary • More space for shielding? – Requirement is system dependent

  15. Consideration on FFHR ■ FFHR-II Original Design ■ Flibe Blanket →V/Be/Li Blanket ■ Top View 10 m ■ Shielding for SCM is one of the critical issues for FFHR ■ Side View ■ First Calculation Using FFHR Torus [MCNP4C+JENDL3.2] 2 m

  16. Li/Be/V Blanket into FFHR FW V-4Cr-4Ti [2cm] １MW/ｍ２ Li-1 [10cm] JLF1 (70%) + B4C (30%) [33cm] Carbon [20cm] V alloy-wall [1cm x 2] JLF1 [5cm] Be [5cm] V/Li blanket: 36cm Li-2 [15cm] Filler: 58cm 6Li: 30% => Local TBR (Full coverage): 1.41 V-4Cr-4Ti [2cm] Total: 92cm

  17. Parametric Survey * *: Standard parameter shown in the first page 6Li enrichment (Changing carbon reflector to JLF-1, TBR: 1.41 =>1.41 (same). Usage of JLF-1 is more attractive for shielding ability) Thickness of Be Thickness of 2nd Li layer * *

  18. Results ■ Neutron flux at SCM could be reduced to the target level

  19. Present Position to Russian TBM Design • Li-Be-V system may have a merit of enhancing shielding for SCM • Could be considered as attractive if this degree of enhancement is crucial for protecting SCM • There may be other methods to protect SCM • Probably, this potential merit will not deserve abandoning the merit of no-Be • Present philosophy is to cooperate with Russia and seek for opportunity of testing no-Be TBM

  20. Progress in Key Technology development for Li/V system in Japan • V-alloy development • Production of purified large ingots – feasibility of recycling • Manufacturing technology (tube, welding--) • Radiation effects of weld joints • Technology development in relation to IFMIF-KEP • Li loop technology • Basic study for tritium recovery with Y • MHD coating development (JUPITER-II and domestic activity)

  21. The in-situ coating method has advantages as, possibility of coating on the complex surfaceafter fabrication of component potentiality to heal the cracks without disassembling the component CaO coating was explored in the US In-situ Coating

  22. It was found that the CaO coating, after formation, dissolved at high temperature (600, 700C) CaO bulk is inherently unstable in pure Li at high temperature, continuous supply of oxygen is necessary to maintain the coating Once the oxygen is exhausted in V-alloy, the coating start to dissolve Er2O3 is much more stable at high temperature It is expected Er2O3, once formed, be stable in Li for a long time Er2O3 is stable in air, combination of dry-coating and in-situ coating is more feasible Solubility of Er (<<1%) is much lower than CaO CaO Er2O3 Problems of the CaO Coating and New Effort on Er2O3

  23. Er2O3 layer was formed on V-4Cr-4Ti by oxidation, anneal and exposure to Li (0.15 wt% Er) at 600C The coating was stable to 750 hrs, also 700C 100hrs 5 5 x 10 Er2O3 - layer - 0030_1.PRO x 10 Er2O3 - layer - 0035_1.PRO 2 2 O1s O1s Er4d 1.5 1.5 Er4d V2p3 V2p3 Intensity Intensity 1 1 0.5 0.5 0 0 0 5 10 15 20 25 30 0 5 10 15 20 25 30 Sputter Time (min) Sputter Time (min) 5 5 x 10 Er2O3 - layer - 0067_1.PRO x 10 Er2O3 - layer - 0062_1.PRO 2 2 O1s O1s Er4d Er4d 1.5 1.5 V2p3 V2p3 Intensity Intensity 1 1 0.5 0.5 0 0 0 5 10 15 20 25 30 0 5 10 15 20 25 30 Sputter Time (min) Sputter Time (min) In-situ Er2O3 Coating on V-4Cr-4Ti Oxidation and anneal at 700C for 16 hr Oxidation only Oxidation at 700C ~100 nm 6 hr 1 hr Yao. 2003 XPS depth profile after exposure to Li (Er) at 600C for 100 hr

  24. In the oxidized and annealed condition, oxygen is stored as Ti-O precipitates oriented to <100> directions The stability of the precipitates depends on the oxygen level During Li exposure, the precipitates dissolve, supplying oxygen into matrix Oxygen Supply Mechanism NIFS-HEAT-2 (V-4Cr-4Ti) Oxidation (973K, 1h) + Annealing (973K, 16h) In-situ coating condition As-received (annealing 1272K, 2h) Oxidation (973K, 1h)

  25. Resistivity increasesby ~12 order of magnitude by formation of Er2O3 layer Analysis of crack allowance limit (by Sze) suggested 10(-4)~10(-6) lower crack limit than the practical coating defect density. Goal of the in-situ healing may be set to increase the resistivity of cracked area from complete conduction by 4~6 orders of magnitude – Seems to be feasible –Need further research Coating Resistivity

  26. MHD Coating in JUPITER-II New Candidates Found by Bulk Exposure Tests (Eu2O3, Y2O3, AlN) Coating Development and Characterization (Eu2O3, Y2O3, AlN, in progress) Crack allowance estimate Two layer coating development In-situ Coating with Er2O3 (feasibility demonstrated) Compatibility of the first layer material Resistivity measurements in Li TBM design Proposal of the coating system

  27. Japanese Universities have an interest in participating in Li/V ITER-TBM General philosophy is to plan to start the test in the later stage of ITER operation However, collaboration with Russia (and other potential countries) for first-day Li/V TBM will also be explored Key technology development is being enhanced by domestic program, IFMIF program and JUPITER-II The MHD coating is making significant progress. The achievements in JUPITER-II will be applied to designing TBM Summary

  28. Neutron Irradiation Effects on SCM (at RTNS-II. LLNL. 1988) (at RTNS-II. LLNL. 1988)

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