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

Research output: Chapter in Book/Report/Conference proceedingConference contribution

Abstract

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.

Original languageEnglish
Title of host publication2015 IEEE 26th Symposium on Fusion Engineering, SOFE 2015
PublisherInstitute of Electrical and Electronics Engineers Inc.
ISBN (Electronic)9781479982646
DOIs
Publication statusPublished - 2016 May 31
Externally publishedYes
Event26th IEEE Symposium on Fusion Engineering, SOFE 2015 - Austin, United States
Duration: 2015 May 312015 Jun 4

Publication series

NameProceedings - Symposium on Fusion Engineering
Volume2016-May

Conference

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

Keywords

  • accident scenario analysis
  • safety study
  • safety system
  • water-cooled fusion DEMO

ASJC Scopus subject areas

  • Nuclear and High Energy Physics
  • Nuclear Energy and Engineering

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