Japanese Universities Perspective on LiV TBM T Muroga
Japanese Universities’ Perspective on Li/V TBM T. Muroga Fusion Engineering Research Center National Institute for Fusion Science
Outline of Presentation z. Position of Li/V Blanket and Li/V ITER-TBM for Japanese Universities z. Purpose of Li/V ITER-TBM z. Neutronics examination z. Consideration to the Russian design z. Present strategy z. Progress in Key Technology development for Li/V y. With emphasis on MHD coating development in Japan
Roadmap for Materials/Blanket Development Approximate calendar year 2015 2020 2030 Advanced Powerplant Power Generation Plant Design (Licencing) ITER 2040 Construction Operation (Blanket test) Blanket Module Test Materials and Blanket System Development Reference Material (RAFM) and System Advanced Materials (V-alloy, Si. C/Si. C --) and System IFMIF Irradiation Test, Materials Qualification and System Performance Test (Staged construction and operation) Fast realization (Mostly JAERI responsibility) Advanced option (Mostly NIFS/University responsibility)
Position of Li/V Blanket and Li/V ITER-TBM for Japanese Universities z. Li/V blanket is categorized into “advanced system” in contrast to RAFM/water blanket as “reference system” z. General plan for Li/V ITER-TBM is to start the test in the midst of the ITER operation phase z. However, first-day Li/V TBM will be explored in the following cases y If large technological progress is made, we will reconsider the schedule y If other parties propose first-day TBM, we will support it and make effort for Japanese idea to be incorporated into the proposal z. Now Russia is proposing first-day Li/V-TBM, we will evaluate the proposal and seek for possibility of cooperation
Purpose of Li/V ITER-TBM (current consensus in Japanese Universities) z. Feasibility of no-Be and natural Li blanket y. Use of 7 Li reaction for enhancing TBR in contrast to Russian Be+6 Li enriched TBM z. Validation of neutronics prediction z. Technology integration for V-alloy, Li and T
ITER with Li/V self-cooled blanket - MCNP calculation by T. Tanaka (NIFS) [ Inboard ] A Plasma SS, H 2 O Vacuum vessel 40 cm Blanket FW B SS (60%), V-4 Cr-4 Ti walls, Li coolant (40%) Natural Li Coil structure 40 cm Vacuum vessel Blanket Center + solenoid Filler 1 m B : ITER with V/Li full blanket Input geometry for MCNP calculation * (*Dimensions from ITER Nuclear Analysis Report) SS, HH 2 O 2 O A FW Blanket A : Standard ITEF-FEAT blanket [ Outboard ] Vacuum vessel B V-4 Cr-4 Ti walls, Natural Li SS (60%), Li coolan (40%)
ITER with Li/V self-cooled blanket - Local TBR (Full Coverage)* Inboard Outboard Total Contribution of 7 Li (%) Li/V blanket Coolant in filler 0. 30 0. 92 1. 22 33 0. 029 0. 15 0. 18 2. 6 Total 0. 33 1. 1 1. 4 --- (a) Inboard FW (b) Outboard FW Blanket (* JENDL 3. 2) Blanket Filler Distribution of tritium production rate ■ Tritium self-sufficiency is feasible
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
Tentative design of Li/V self-cooled TBM by NIFS/Universities 505 Verification of (1) Coolant circulation Plasma (2) MHD coating V-4 Cr-4 Ti Verification of (1) Neutron transport (2) Tritium production from 7 Li 1720 210 470 210 SS(60%), H 2 O(40%) Li layer SS 316 TBM frame Plasma Inlet/outlet pipes Li : ~0. 027 m 3 (Unit : mm) Tentative design of Li/V TBM SS(60%), H 2 O(40%) ■ Verification of TPR for 7 Li ■ Thick Li tanks for verification of neutron transport
Tentative design of Li/V self-cooled TBM Covering by B 4 C Plasma SS(60%), H 2 O(40%) Li layer (1) - Tritium production - ■ For verification of tritium production from 7 Li (n, na)T reaction - Reduction of thermal neutrons by B 4 C shielding (2) (3) (4)(5) Li layer (1) (2) (3) (4) (5) Tritium production rate in Li layers (3) (4) (5) Contribution of 7 Li to tritium production
Experimental parameter for Li/V TBM - Adjustment by B 4 C 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 7 Li by B 4 C covering ■ Contribution of 7 Li to TPR can be adjusted by thickness of B 4 C shield
Russian Li/V self-cooled test blanket module - Structure 505 Be multiplier 1720 Plasma V-5 Cr-5 Ti Li layer (6 Li : 90%) WC Shield (Reflector) SS(60%) + H 2 O(40%) Structure of Russian Li/V TBM (Unit : mm) ■ 6 Li enriched coolant (7. 5 % ==> 90%) ■ Li layer x 2, Be multiplier ==> 6 Li (n, a) T ■ Maximize the 6 Li reaction to demonstrate DEMO reactor breeding tritium by 6 Li
Russian Li/V self-cooled test blanket module - Tritium production - Total : 0. 09 (g/FPD) Plasma SUS + H 2 O Li layer (1) TBM surface Li/V TBM SS 316 TBM frame Li layer (2) Tritium production rate in Li layers and contribution of 6 Li and 7 Li Be Plasma WC SS+H 2 O Li layer (1) Li layer (2) [6 Li : 90%]
Is Russian Li-Be-V an Attractive Option? z “No-beryllium” is an attractive potentiality of Li/V system z Does Li-Be-V have alternative merits compensating the demerit of using Be? y. High TBR? – Excess TBR probably not necessary y. More space for shielding? – Requirement is system dependent
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 ■ First Calculation Using FFHR Torus [MCNP 4 C+JENDL 3. 2] ■ Side View 2 m
Li/Be/V Blanket into FFHR 1 MW/m2 FW V-4 Cr-4 Ti [2 cm] JLF 1 Li-1 V alloy-wall Carbon (70%) JLF 1 [10 cm] [20 cm] [5 cm] + [1 cm x 2] B 4 C Be (30%) [5 cm] [33 cm] V/Li blanket: 36 cm Li-2 [15 cm] 6 Li: 30% Filler: 58 cm V-4 Cr-4 Ti => Local TBR (Full coverage): 1. 41 [2 cm] Total: 92 cm
Parametric Survey * *: Standard parameter shown in the first page 6 Li enrichment Thickness of Be * (Changing carbon reflector to JLF-1, TBR: 1. 41 =>1. 41 (same). Usage of JLF-1 is more attractive for shielding ability) Thickness of 2 nd Li layer *
Results V Layer (cm) Li Layer (cm) Be Layer (cm) RAFM shield layer (cm) TBR (full coverage) Neutron flux at SCM (1010 n/m 2 s) 6 25 6 20 6 15 4 12 5 5 10 10 10 58 63 63 68 73 1. 41 1. 54 1. 50 6. 3 3. 1 2. 2 1. 3 0. 78 4. 3 (FFHR-II) ■ Neutron flux at SCM could be reduced to the target level 0. 71 (Target)
Present Position to Russian TBM Design z Li-Be-V system may have a merit of enhancing shielding for SCM z Could be considered as attractive if this degree of enhancement is crucial for protecting SCM z There may be other methods to protect SCM z Probably, this potential merit will not deserve abandoning the merit of no-Be z Present philosophy is to cooperate with Russia and seek for opportunity of testing no-Be TBM
Progress in Key Technology development for Li/V system in Japan z V-alloy development y. Production of purified large ingots – feasibility of recycling y. Manufacturing technology (tube, welding--) y. Radiation effects of weld joints z Technology development in relation to IFMIFKEP y. Li loop technology y. Basic study for tritium recovery with Y z MHD coating development (JUPITER-II and domestic activity)
In-situ Coating z. The in-situ coating method has advantages as, ypossibility of coating on the complex surface after fabrication of component ypotentiality to heal the cracks without disassembling the component y. Ca. O coating was explored in the US
Problems of the Ca. O Coating and New Effort on Er 2 O 3 z. It was found that the Ca. O coating, after formation, dissolved at high temperature (600, 700 C) y Ca. O bulk is inherently unstable in pure Li at high temperature, continuous supply of oxygen is necessary to maintain the coating y Once the oxygen is exhausted in Valloy, the coating start to dissolve z. Er 2 O 3 is much more stable at high temperature y It is expected Er 2 O 3, once formed, be stable in Li for a long time y Er 2 O 3 is stable in air, combination of dry-coating and in-situ coating is more feasible y Solubility of Er (<<1%) is much lower Ca. O Er 2 O 3
In-situ Er 2 O 3 Coating on V-4 Cr-4 Ti z. Er 2 O 3 layer was formed on V-4 Cr-4 Ti by oxidation, anneal and exposure to Li (0. 15 wt% Er) at 600 C z. The coating was stable to 750 hrs, also 700 C 100 hrs Oxidation and anneal at 700 C for 16 hr Oxidation only 5 5 Er 2 O 3 -layer-0030_1. PRO 2 O 1 s Er 4 d 1. 5 Intensity Oxidation at 700 C x 10 6 hr V 2 p 3 1 Er 2 O 3 -layer-0035_1. PRO Er 4 d V 2 p 3 1 0. 5 0 5 10 15 20 25 0 30 0 5 Sputter Time (min) 5 x 10 5 Er 2 O 3 -layer-0067_1. PRO 2 O 1 s Er 4 d V 2 p 3 Intensity 1. 5 1 hr 1 0. 5 0 Yao. 2003 10 15 20 25 30 Sputter Time (min) x 10 Er 2 O 3 -layer-0062_1. PRO O 1 s Er 4 d 1. 5 Intensity 2 O 1 s ~100 nm 0. 5 0 x 10 1. 5 Intensity 2 V 2 p 3 1 0. 5 0 5 10 15 20 Sputter Time (min) 25 30 0 0 5 10 15 20 25 30 Sputter Time (min) XPS depth profile after exposure to Li (Er) at 600 C for 100 hr
Oxygen Supply Mechanism z In the oxidized annealed condition, oxygen is stored as Ti-O precipitates oriented to <100> directions y The stability of the precipitates depends on the oxygen level y During Li exposure, the precipitates dissolve, supplying oxygen into matrix NIFS-HEAT-2 (V-4 Cr-4 Ti) As-received (annealing 1272 K, 2 h) Oxidation (973 K, 1 h) + Annealing (973 K, 16 h) In-situ coating condition
Coating Resistivity z. Resistivity increases by ~12 order of magnitude by formation of Er 2 O 3 layer z. Analysis of crack allowance limit (by Sze) suggested 10(4)~10(-6) lower crack limit than the practical coating defect density. z. 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
MHD Coating in JUPITER-II New Candidates Found by Bulk Exposure Tests (Eu 2 O 3, Y 2 O 3, Al. N) Coating Development and Characterization (Eu 2 O 3, Y 2 O 3, Al. N, in progress) Crack allowance estimate Two layer coating development In-situ Coating with Er 2 O 3 (feasibility demonstrated) Compatibility of the first layer material Resistivity measurements in Li Proposal of the coating system TBM design
Summary z. Japanese Universities have an interest in participating in Li/V ITER-TBM z. General philosophy is to plan to start the test in the later stage of ITER operation z. However, collaboration with Russia (and other potential countries) for first-day Li/V TBM will also be explored z. Key technology development is being enhanced by domestic program, IFMIF program and JUPITER-II z. The MHD coating is making significant progress. The achievements in JUPITER-II will be applied to designing TBM
Neutron Irradiation Effects on SCM (at RTNS-II. LLNL. 1988)
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