Diffusion coefficient of tritium through molten salt Flibe and rate of tritium leak from fusion reactor system

Satoshi Fukada; Robert A. Anderl; Akio Sagara; Masabumi Nishikawa

doi:10.13182/FST05-A1013

Diffusion coefficient of tritium through molten salt Flibe and rate of tritium leak from fusion reactor system

Satoshi Fukada, Robert A. Anderl, Akio Sagara, Masabumi Nishikawa

Engineering Science for Advanced energy system

Research output: Contribution to journal › Article › peer-review

16 Citations (Scopus)

Abstract

Diffusion coefficients of hydrogen isotopes in Flibe were correlated with making reference to previous relating data of F ion self-diffusivity and Flibe viscosity and so on. Rates of tritium permeation through structural materials in a fusion reactor system with Flibe blanket were estimated comparatively under conditions with or without a Flibe permeation barrier. A way to lower the tritium leak rate below a level regulated by law was proposed, and its effectiveness was discussed.

Original language	English
Pages (from-to)	666-669
Number of pages	4
Journal	Fusion Science and Technology
Volume	48
Issue number	1
DOIs	https://doi.org/10.13182/FST05-A1013
Publication status	Published - 2005

All Science Journal Classification (ASJC) codes

Civil and Structural Engineering
Nuclear and High Energy Physics
Nuclear Energy and Engineering
Materials Science(all)
Mechanical Engineering

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10.13182/FST05-A1013

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AU - Nishikawa, Masabumi

PY - 2005

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AB - Diffusion coefficients of hydrogen isotopes in Flibe were correlated with making reference to previous relating data of F ion self-diffusivity and Flibe viscosity and so on. Rates of tritium permeation through structural materials in a fusion reactor system with Flibe blanket were estimated comparatively under conditions with or without a Flibe permeation barrier. A way to lower the tritium leak rate below a level regulated by law was proposed, and its effectiveness was discussed.

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Diffusion coefficient of tritium through molten salt Flibe and rate of tritium leak from fusion reactor system

Abstract

All Science Journal Classification (ASJC) codes

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