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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Project Status |
Completed (Fiscal Year 2021)
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Budget Amount *help |
¥4,420,000 (Direct Cost: ¥3,400,000、Indirect Cost: ¥1,020,000)
Fiscal Year 2020: ¥780,000 (Direct Cost: ¥600,000、Indirect Cost: ¥180,000)
Fiscal Year 2019: ¥1,040,000 (Direct Cost: ¥800,000、Indirect Cost: ¥240,000)
Fiscal Year 2018: ¥2,600,000 (Direct Cost: ¥2,000,000、Indirect Cost: ¥600,000)
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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 / トリチウム除染 / 等温脱離 / タングステン / 原型炉 / 真空下 / 水素同位体分析 / グロー放電 / GD-OES / 定温脱離法 |
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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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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Report
(5 results)
Research Products
(21 results)
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[Presentation] Observations of Tritium Inventory in JET ILW Dust Particles and Applications to Metal Wall Fusion Devices2019
Author(s)
N.Ashikawa, T.Otsuka, Y.Torikai, N.Asakura, A.Widdowson, M.Rubel, H.Furuta, M.Hara, S.Masuzaki, Y.Hatano, H.Nakamura, S.Jachmich, T.Hayashi, JET Contributors
Organizer
The 12th International Conference on Tritium Science and Technology
Related Report
Int'l Joint Research
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[Presentation] Determination of retained tritium from ILW dust particles in JET2018
Author(s)
N. Ashikawa, Y. Torikai, N. Asakura, T. Otsuka, A. Widdowson, M. Rubel, M. Oyaizu, M. Hara, S. Masuzaki, K. Isobe, Y. Hatano, K. Heinola, A. Baron-Wiechec, S. Jachmich, T. Hayashi and JET Contributors
Organizer
International Conference on Plasma Surface Interactions in Controlled Fusion Devices
Related Report
Int'l Joint Research
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[Presentation] Dust Particle Generation from the Controlled thermonuclear Fusion Device Jet with Metallic Plasma-Facing Components2018
Author(s)
N. Ashikawa, N. Asakura, A. Widdowson, M. Rubel, Y. Torikai, T. Otsuka, M. Hara, S. Masuzaki, M. Oyaidzu, K. Isobe,Y. Hatano, D. Hamaguchi, H. Kurotaki, S. Nakano, J.H. Kim, M. Tokitani, R. Sakamoto, J. Grzonka , K. Heinola11), A. Baron-Wiechec, M. Miyamoto, H. Tanigawa, M. Nakamichi, T. Hayashi, and JET Contributors
Organizer
2018 Asia-Pacific Conference on Plasma and Terahertz Science
Related Report
Int'l Joint Research / Invited