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Development of plant concept related to tritium handling in the water-cooling system for JA DEMO
https://repo.qst.go.jp/records/75612
https://repo.qst.go.jp/records/75612f699011f-6ab7-4dca-8b2d-ae866c062ae8
| Item type | 学術雑誌論文 / Journal Article(1) | |||||
|---|---|---|---|---|---|---|
| 公開日 | 2018-10-18 | |||||
| タイトル | ||||||
| タイトル | Development of plant concept related to tritium handling in the water-cooling system for JA DEMO | |||||
| 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_6501 | |||||
| 資源タイプ | journal article | |||||
| 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 |
Hiwatari, Ryoji
× Hiwatari, Ryoji× Katayama, Kazunari× Nakamura, Makoto× Miyoshi, Yuuya× Aoki, Akira× Asakura, Nobuyuki× Uto, Hiroyasu× Homma, Yuuki× Tokunaga, Shinsuke× Nakajima, Noriyasu× Someya, Yoji× Sakamoto, Yoshiteru× Tobita, Kenji× Special Design Team for Fusion DEMO, Joint× Hiwatari, Ryoji× Nakamura, Makoto× Miyoshi, Yuuya× Asakura, Nobuyuki× Uto, Hiroyasu× Homma, Yuuki× Tokunaga, Shinsuke× Someya, Yoji× Sakamoto, Yoshiteru× Tobita, Kenji |
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| 抄録 | ||||||
| 内容記述タイプ | Abstract | |||||
| 内容記述 | The conceptual design of Japan’s fusion demonstration plant (JA DEMO) is now being developed. In this paper, an overall plant system concept related to tritium handling in the water-cooling system is developed to give a concrete shape to the present JA DEMO concept as an electric power plant. The basic condition of tritium permeation from the in- vessel components to the primary cooling system is evaluated to be 5.7g-T/day. The tritium concentration of the primary coolant is assumed to be 1 TBq/kg similar to the heavy water reactor condition. The capacity of the water detritiation system (WDS) is assessed, and the bypass feed water from the primary cooling loop is evaluated to be 94kg/h under the tritium extraction efficiency of 0.96. Based on those specific parameters, the existing WDS in the heavy water reactor is found to be applicable to that of JA DEMO. Configuration of the primary heat transfer system (PHTS) is also discussed. Based on the heavy water reactor experience, tritium permeation through a steam generator (SG) to the secondary cooling system in PHTS is evaluated at 318 Ci/year/loop, which is found to be less than the restricted amount of tritium disposal for a pressurized water reactor in Japan. The key effect of the heavy water reactor experience is reduction of tritium permeation by oxide layer formed on SG pipes. Finally, confinement concept of tritium release from PHTS is discussed under the condition of an ex-vessel loss of coolant accident (LOCA). A pressure suppression system is installed to prevent the upper tokamak hall from pressurizing at the ex-vessel LOCA, and the tritium leakage from the upper tokamak hall is consequently restrained. The resultant early public dose at the plant site boundary can be reduced to 1.8 mSv, which is negligibly smaller than 100 mSv of the no-evacuation limit recommended by IAEA. | |||||
| 書誌情報 |
Fusion Engineering and Design 巻 143, 号 C, p. 259-266, 発行日 2019-04 |
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| 出版者 | ||||||
| 出版者 | Elsevier | |||||
| ISSN | ||||||
| 収録物識別子タイプ | ISSN | |||||
| 収録物識別子 | 0920-3796 | |||||
| DOI | ||||||
| 識別子タイプ | DOI | |||||
| 関連識別子 | 10.1016/j.fusengdes.2019.03.174 | |||||
| 関連サイト | ||||||
| 識別子タイプ | URI | |||||
| 関連識別子 | https://www.sciencedirect.com/science/article/pii/S0920379619305071 | |||||
