1992 Fiscal Year Final Research Report Summary
Chemical Behavior of Tritium Breeding Material for Fusion Reactor
| Project/Area Number |
03453167
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| Research Category |
Grant-in-Aid for General Scientific Research (B)
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| Allocation Type | Single-year Grants |
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| Research Field |
Nuclear engineering
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| Research Institution | University of Tokyo |
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Principal Investigator |
YONEOKA Toshiaki The Univ. of Tokyo, Fac. of Eng., Research Associate, 工学部, 助手 (40013221)
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| Co-Investigator(Kenkyū-buntansha) |
KOKUBO Sanya The Univ. of Tokyo, Fac. of Eng., Research Associate, 工学部, 助手 (20107577)
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| Project Period (FY) |
1991 – 1992
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| Keywords | Fusion reactor / Tritium / Breeding material / Lithium-lead alloy / Diffusion / Release / Oxide layer / Permeation |
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| Research Abstract |
Molten lithium-lead (17Li-Pb) alloy is one of the most important candidate materials for tritium breeding in a fusion reactor blanket system. In this research, its chemical behavior, in particular, tritium behavior in the alloy was investigated by in-pile and out-of pile experiments. The results obtained here are summarized as follows. 1. Molten 17Li-Pb alloy samples were prepared in a glove box of an Ar atmosphere from reagents of the elements, and they were confirmed to have a good quality in crystal structure, thermodynamic phase, impurity concentration and melting point by XRD, SEM, chemical analysis and DSC. 2. The tritium generated in 17Li-Pb by neutron irradiation was released in the molecular forms of hydrogen such as HT and T_2. It is for the first time in the world that the diffusion coefficient of tritium in molten 17Li-Pb was measured under neutron irradiation at elevated temperatures. 3. The mass-transfer coefficient of tritium from molten 17Li-Pb to purge gas was measured by in-pile experiment, and it increased with the partial pressure of hyderogen in the purge gas. This results indicates that tritium release is controlled by the tritium diffusion in the liquid film of the molten alloy. 4. The permeation coefficient of tritium through structural material facing the molten alloy is a quite important parameter from the viewpoints of tritium safety and tritium economy. It was measured for iron and stainless steel type 304, and it was influenced strongly by the partial pressure of hydrogen in the purge gas facing the rear surface of the structural materials. This result sugggests that oxide layer formed on the surface of the structural materials can surpress tritium permeation, and that a layer of stable oxides such as Cr_2O_3 and FeCr_2O_4 formed on the surface of stainless steel can be utilized as a tritium permeation barrier.
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