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Development of physics and engineering designs for japan’s demo concept
https://repo.qst.go.jp/records/54923
https://repo.qst.go.jp/records/54923ace353ca-c72e-4e39-a89c-700d4d744623
| Item type | 会議発表論文 / Conference Paper(1) | |||||
|---|---|---|---|---|---|---|
| 公開日 | 2018-10-03 | |||||
| タイトル | ||||||
| タイトル | Development of physics and engineering designs for japan’s demo concept | |||||
| 言語 | ||||||
| 言語 | eng | |||||
| 資源タイプ | ||||||
| 資源タイプ識別子 | http://purl.org/coar/resource_type/c_5794 | |||||
| 資源タイプ | conference paper | |||||
| アクセス権 | ||||||
| アクセス権 | metadata only access | |||||
| アクセス権URI | http://purl.org/coar/access_right/c_14cb | |||||
| 著者 |
坂本, 宜照
× 坂本, 宜照× 飛田, 健次× 朝倉, 伸幸× 日渡, 良爾× 染谷, 洋二× 宇藤, 裕康× 徳永, 晋介× 本間, 裕貴× 三善, 悠矢× 相羽, 信行× 松山, 顕之× Sakamoto, Yoshiteru× Tobita, Kenji× Asakura, Nobuyuki× Hiwatari, Ryoji× Someya, Yoji× Uto, Hiroyasu× Tokunaga, Shinsuke× Homma, Yuuki× Miyoshi, Yuuya× Aiba, Nobuyuki× Matsuyama, Akinobu |
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| 抄録 | ||||||
| 内容記述タイプ | Abstract | |||||
| 内容記述 | Recent progress of Japan's DEMO design is presented. The key concept is a steady-state DEMO with a major radius of 8 m class and fusion power of 1.5 GW level, which is proposed based on ITER physics and technology bases, and characterized by operational flexibility from pulse to steady-state operations. Even in a steady-state DEMO, the pulse operation is required for the commissioning of plant systems and also suitable for early demonstration of fusion electricity by moderate plasma performance. Regarding the physics design, divertor plasma simulation clarifies that the lower density to be compatible with detached plasma, which is consistent with the operational density of JA DEMO. Vertical stability evaluation by 3D eddy current and plasma control model shows that plasma elongation of 1.75 is sustainable by applying the double-loop type shells. The development of plasma operation scenario indicates the importance of off-axis ECCD for controlling the internal transport barriers. In addition to physics design, engineering designs are performed in wide area. The divertor cassette design is developed for reducing the fast neutron flux to protect the vacuum vessel and for replacement of the power exhaust units. The breeding blanket concept based on JA ITER-TBM strategy is developed to increase the pressure-tightness of the modules by considering safety assessment of in-box LOCA. On the TF coil design, assessment of the error field indicates that the fabrication tolerance can be mitigated by ~2.5 times as large as ITER's with correction coil current of several 100 kAT/coil. The concept of remote maintenance for the blanket segments is developed such as the stable transfer mechanism in the vertical, radial and toroidal directions. The rad-wastes generated by the maintenance can be disposed of in shallow land burial after 10-year storage. The concept of primary cooling water system is developed for effective use of thermal power removed from not only blanket but also divertor. | |||||
| 書誌情報 |
Proceedings of the 27th IAEA Fusion Energy Conference (FEC 2018) 発行日 2018-10 |
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| 出版者 | ||||||
| 出版者 | International Atomic Energy Agency, Vienna | |||||
