Analysis on ex-vessel loss of coolant accident for a water-cooled fusion DEMO reactor

Kazuhito Watanabe, Makoto Nakamura, Kenji Tobita, Youji Someya, Hisashi Tanigawa, Hiroyasu Utoh, Yoshiteru Sakamoto, Takao Araki, Shiro Asano, Kazuhito Asano

研究成果: Conference contribution

抄録

Safety studies of a water-cooled fusion DEMO reactor have been performed. In the DEMO design, the blanket primary cooling system involves a large amount of energy due to pressurized water coolant (290-325 °C, 15.5 MPa). Moreover, it contains radioactive materials such as tritium and activated corrosion products. Therefore, in the event of the blanket cooling pipe break outside the vacuum vessel, i.e. ex-vacuum vessel loss of coolant accident (ex-VV LOCA), the pressurized steam and air may lead to damage reactor building walls which have confinement function, and to release the radioactive materials to the environment. In response to this accident, we proposed three options of confinement strategies. In each option, the pressure and thermal loads to the confinement boundaries and total mass of tritium released to the environment were analyzed by accident analysis code MELCOR modified for fusion reactor. These analyses developed design parameters to maintain the integrity of the confinement boundaries.

本文言語English
ホスト出版物のタイトル2015 IEEE 26th Symposium on Fusion Engineering, SOFE 2015
出版社Institute of Electrical and Electronics Engineers Inc.
ISBN(電子版)9781479982646
DOI
出版ステータスPublished - 2016 5 31
外部発表はい
イベント26th IEEE Symposium on Fusion Engineering, SOFE 2015 - Austin, United States
継続期間: 2015 5 312015 6 4

出版物シリーズ

名前Proceedings - Symposium on Fusion Engineering
2016-May

Conference

Conference26th IEEE Symposium on Fusion Engineering, SOFE 2015
CountryUnited States
CityAustin
Period15/5/3115/6/4

ASJC Scopus subject areas

  • Nuclear and High Energy Physics
  • Nuclear Energy and Engineering

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