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Thermohydraulic Analysis of Accident Scenarios of a Fusion DEMO Reactor Based on Water-Cooled Ceramic Breeder Blanket: Analysis of LOCAs and LOVA
https://repo.qst.go.jp/records/47845
https://repo.qst.go.jp/records/47845929e5bc3-6172-4161-b275-0170240ea718
| Item type | 学術雑誌論文 / Journal Article(1) | |||||
|---|---|---|---|---|---|---|
| 公開日 | 2017-04-27 | |||||
| タイトル | ||||||
| タイトル | Thermohydraulic Analysis of Accident Scenarios of a Fusion DEMO Reactor Based on Water-Cooled Ceramic Breeder Blanket: Analysis of LOCAs and LOVA | |||||
| 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 |
中村, 誠
× 中村, 誠× 渡邊, 和仁× 飛田, 健次× 染谷, 洋二× 谷川, 尚× 宇藤, 裕康× 坂本, 宜照× 功刀, 資彰× 横峯, 健彦× 中村 誠× 飛田 健次× 染谷 洋二× 谷川 尚× 宇藤 裕康× 坂本 宜照 |
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| 抄録 | ||||||
| 内容記述タイプ | Abstract | |||||
| 内容記述 | Thermohydraulic analysis of postulated accidents will identify system responses to accident scenarios and aid in developing design of safety systems and strategies to prevent or mitigate accident propagation. This paper reports analyses of four accident scenarios of a fusion DEMO reactor based on water-cooled ceramic-pebble breeder blanket. The accidents analyzed, which were selected based on the previous logical accident analysis, are ex-vessel loss of coolant of the primary cooling system, in-vessel loss of coolant of the first wall (FW) cooling pipes, loss of coolant in blanket modules, and loss of vacuum. The analyses have identified thermohydraulic responses of the DEMO systems to these accidents, pressure loads to confinement barriers for radioactive materials. Effectiveness of the safety systems and the integrity of the primary and final (secondary) confinement barriers against the accidents are discussed. As for the final confinement barrier, we show for the first time that implementation of a pressure suppression system (PSS) to the cooling system vault and a vacuum breaker to the tokamak pit is effective in significantly keeping the integrity of the final confinement barrier against the ex-vessel loss-of-coolant and loss-of-vacuum accidents, respectively. As for the primary confinement barrier, we show for the first time that limitation of the number of blankets from which a helium purge gas line collects the bred tritium will be a key technical issue to prevent propagation of loss of coolant in a blanket box through the purge gas line and then suppress the pressurization of the vacuum vessel (VV). For the in-vessel loss of coolant of the FW cooling pipes, further optimization of the PSS or design solutions regarding in-vessel components and plasma control will be necessary to decrease the pressure load to the VV and ensure the integrity of the primary confinement barrier. | |||||
| 書誌情報 |
IEEE Transactions on Radiation and Plasma Medical Sciences 巻 44, 号 9, p. 1689-1699, 発行日 2016-09 |
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| 出版者 | ||||||
| 出版者 | IEEE Nuclear and Plasma Sciences Society | |||||
| DOI | ||||||
| 識別子タイプ | DOI | |||||
| 関連識別子 | 10.1109/TPS.2016.2578348 | |||||
