Thermohydraulic responses of a water-cooled tokamak fusion DEMO to loss-of-coolant accidents

M. Nakamura, K. Tobita, Y. Someya, H. Utoh, Y. Sakamoto, W. Gulden

Research output: Contribution to journalArticle

7 Citations (Scopus)

Abstract

Major in- and ex-vessel loss-of-coolant accidents (LOCAs) of a water-cooled tokamak fusion DEMO reactor have been analysed. Analyses have identified responses of the DEMO systems to these accidents and pressure loads to confinement barriers for radioactive materials. As for the in-VV LOCA, we analysed the multiple double-ended break of the first wall cooling pipes around the outboard toroidal circumference. As for the ex-VV LOCA, we analysed the double-ended break of the primary cooling pipe. The thermohydraulic analysis results suggest that the in- and ex-vessel LOCAs crucially threaten integrity of the primary and final confinement barriers, respectively. Mitigations of the loads to the confinement barriers are also discussed.

Original languageEnglish
Article number123008
JournalNuclear Fusion
Volume55
Issue number12
DOIs
Publication statusPublished - 2015 Oct 30
Externally publishedYes

Keywords

  • fusion DEMO
  • loss-of-coolant accident
  • safety
  • thermohydraulics

ASJC Scopus subject areas

  • Nuclear and High Energy Physics
  • Condensed Matter Physics

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