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Development of level-1 PSA method applicable to Japan sodium-cooled fast reactor, 1; Research plan and internal event evaluation related to reactor shutdown failure

Kurisaka, Kenichi; Sakai, Takaaki; Yamano, Hidemasa; Fujita, Satoshi*; Minagawa, Keisuke*; Yamaguchi, Akira*; Takata, Takashi*

This paper describes a study to develop the level-1 probabilistic safety assessment (PSA) method that is applicable to the Japan Sodium-cooled Fast Reactor (JSFR). This study has been started since August 2010 and aims to provide a new evaluation method of (1) passive safety architectures related to internal events and (2) an advanced seismic isolation system related to a seismic event as a representative external event in Japan. However, after the nuclear accident at Fukushima Dai-ichi Nuclear Power Plant on March 11 2011, the research-period was shortened from four to two years by political reason, and only a limited scope of the research subject will be performed in the two years. Regarding the internal events evaluation, a quantitative analysis on the frequency of the core damage caused by reactor shutdown failure was conducted. A failure in passive reactor shutdown was taken into account in the event tree model. The failure rate of sodium-cooled fast reactor (SFR) specific components was evaluated based on the operating experience in existing SFRs by applying the Hierarchical Bayesian Method, which can consider a plant-to-plant variability. By conducting an uncertainty analysis, it was found that the assumption about the correlation of the probability parameters between the main and backup reactor shutdown systems is sensitive to the mean value of the frequency of the core damage caused by reactor shutdown failure.

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