2021 Fiscal Year Final Research Report
Investigation of tritium decontamination method under the vacuum condition for fusion reactors
Project/Area Number |
18K04999
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Research Category |
Grant-in-Aid for Scientific Research (C)
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Allocation Type | Multi-year Fund |
Section | 一般 |
Review Section |
Basic Section 31010:Nuclear engineering-related
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Research Institution | National Institute for Fusion Science |
Principal Investigator |
Ashikawa Naoko 核融合科学研究所, ヘリカル研究部, 准教授 (00353441)
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Co-Investigator(Kenkyū-buntansha) |
鳥養 祐二 茨城大学, 理工学研究科(理学野), 教授 (80313592)
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Project Period (FY) |
2018-04-01 – 2022-03-31
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Keywords | トリチウム低減法 / トリチウム除染法 / 核融合原型炉 / 金属壁 / 等温脱離法 |
Outline of Final Research Achievements |
In DEMO, the tritium(T) decontamination scenario before the maintenance begins is a key issue. Currently, QST-DEMO team has not yet determined the allowable value of residual T in the plasma vacuum vessel, but it is necessary to indicate a candidate T decontamination technique. Furthermore, constructing a short-term maintenance scenario that includes T decontamination after stopped plasma operations is also important for DEMOs. The method using decay heat is being studied for T decontamination. The temperature of plasma-facing materials is 623 K, and an upper-temperature limit is 773-823 K due to structural issues of ferritic steel. Therefore, the available temperature range for tritium decontamination is 623-823 K. In this study, we investigated the possibility of tritium decontamination of tungsten materials in temperature control. The result shows T decontamination efficiency at 673K using the isothermal desorption.
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Free Research Field |
核融合炉工学
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Academic Significance and Societal Importance of the Research Achievements |
核融合エネルギー炉の実現に向け、定常運転後の残留トリチウム低減法を見出し、かつその運用条件を明らかにするための基礎研究を行った。本課題は核融合炉の定常運転後に必要なプラズマ真空容器の大気開放前の残留トリチウム低減、つまり真空下での除染法に関する研究である。このトリチウム除染の効率的な実施は、核融合装置が設置されているトーラスホール内でのトリチウム拡散防護や、除染によって回収した燃料トリチウムの再利用にもつながる。つまり、核融合プラントの安全性と性能向上につながる研究である。
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