| Project/Area Number |
09480107
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| Research Category |
Grant-in-Aid for Scientific Research (B).
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| Allocation Type | Single-year Grants |
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| Section | 一般 |
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| Research Field |
Nuclear fusion studies
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| Research Institution | The University of Tokyo |
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Principal Investigator |
TERAI Takayuki Graduate School of Engineering, The University of Tokyo, Prof., 大学院・工学系研究科, 教授 (90175472)
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| Co-Investigator(Kenkyū-buntansha) |
YONEOKA Toshiaki Graduate School of Engineering, The University of Tokyo, R. assoc., 大学院・工学系研究科, 助手 (40013221)
YAMAGUCHI Kenji Graduate School of Engineering, The University of Tokyo, A. Prof., 大学院・工学系研究科, 助教授 (50210357)
TANAKA Satoru Graduate School of Engineering, The University of Tokyo, Prof., 大学院・工学系研究科, 教授 (10114547)
小野 双葉 東京大学, 大学院・工学系研究科, 助手 (00011198)
沖津 康平 東京大学, 工学部附属総合試験所, 助手
小林 知洋 東京大学, 工学部附属総合試験所, 助手 (40282496)
小野 勝男 東京大学, 工学部・附属総合試験所, 助手 (20160905)
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| Project Period (FY) |
1997 – 1999
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| Project Status |
Completed (Fiscal Year 2001)
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| Budget Amount *help |
¥11,300,000 (Direct Cost: ¥11,300,000)
Fiscal Year 1999: ¥1,700,000 (Direct Cost: ¥1,700,000)
Fiscal Year 1998: ¥2,700,000 (Direct Cost: ¥2,700,000)
Fiscal Year 1997: ¥6,900,000 (Direct Cost: ¥6,900,000)
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| Keywords | fusion reactor / tritium breeding material / molten salt / tritium release / permeation / corrosion / structural material / blanket / 液体ブランケット / トリチウム / 増殖材料 / 溶融塩 / 放出 / LiF / BeF_2 / 熱力学 |
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| Research Abstract |
A molten salt mixture of LiF and BeF_2 (denoted as Flibe) has been proposed to be a candidate material for tritium breeding and cooling medium in fusion reactor blanket systems. In this study, its chemical properties such as tritium release property and compatibility with structural materials under blanket-simulated conditions were investigated.
In-situ tritium release experiments were carried out under neutron irradiation at the temperatures of 500-700 C using the fast neutron source reactor "YAYOI", University of Tokyo in order to measure the released chemical species of tritium, diffusion coefficient of tritium in molten Flibe and tritium release rate. Based on these parameters, a model on tritium release from Flibe was constructed. In addition, the tritium permeation rate through structural wall facing molten Flibe was measured. By using, all the date obtained in the experiments, an effective tritium recovery system with low tritium permeation was proposed.
A compatibility test of structural materials with molten Flibe was also carried out. Specimens of some candidate materials including Mo, W, V, Fe and their alloys as well as SiC/SiC_f composite were immersed in molten Flibe at 550 C for 1 - 30 days, and they were examined by XRD, RBS, SEM, XPS, etc. on the composition and the crystallographic change of the surface. The corrosion rate was evaluated and the mechanism was discussed in comparison with thermodynamic calculation.
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