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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

JSFRのレベル1PSA手法の開発,1; 研究計画と原子炉停止失敗に関する内的事象評価

栗坂 健一; 堺 公明; 山野 秀将; 藤田 聡*; 皆川 佳祐*; 山口 彰*; 高田 孝*

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

本論文は、ナトリウム冷却型高速炉JSFRに適用可能なレベル1PSA手法の開発について記述する。本研究は2010年8月に開始し、(1)内的事象について受動的安全特性,(2)外的事象(地震)について高速炉用免震システムの新評価手法の開発を目標としている。2011年3月11日の福島第一原子力発電所事故後、研究計画が当初の4年から2年へ短縮され、一部の研究課題のみがこの2年間に遂行されることとなった。内的事象評価に関して、原子炉停止失敗による炉心損傷の頻度の評価を実施し、受動的原子炉停止の失敗をイベントツリーモデルにて考慮した。高速炉特有の機器故障率を既存の高速炉の運転経験をもとにプラント間のばらつきを考慮できる階層ベイズ法を適用して評価した。不確実さ解析の結果、主炉停止系と後備炉停止系間での確率パラメータの相関性についての想定が炉心損傷頻度の平均値へ感度を有することがわかった。

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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