Thermohydraulic Analysis of Accident Scenarios of a Fusion DEMO Reactor Based on Water-Cooled Ceramic Breeder Blanket: Analysis of LOCAs and LOVA

M. Nakamura, K. Watanabe, K. Tobita, Y. Someya, H. Tanigawa, H. Utoh, Y. Sakamoto, T. Kunugi, T. Yokomine, W. Gulden

研究成果: Article査読

11 被引用数 (Scopus)

抄録

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.

本文言語English
論文番号7497584
ページ(範囲)1689-1699
ページ数11
ジャーナルIEEE Transactions on Plasma Science
44
9
DOI
出版ステータスPublished - 2016 9月
外部発表はい

ASJC Scopus subject areas

  • 核物理学および高エネルギー物理学
  • 凝縮系物理学

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