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Onoda, Yuichi; Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa
Proceedings of Probabilistic Safety Assessment and Management & Asian Symposium on Risk Assessment and Management (PSAM17 & ASRAM2024) (Internet), 10 Pages, 2024/10
We developed the measures for improving resilience of the sodium-cooled fast reactor structure using the failure mitigation technology and evaluated the effectiveness of the measures. To prevent core damage in the event of an accident progressing to an ultra-high temperature state, both measures to prevent overpressure in the reactor vessel and measures to cool the reactor core are required. As a core cooling measure, we developed a core cooling concept that promotes radiant heat transfer from the reactor vessel and cools the containment vessel outer surface by natural convection named Containment Vessel Auxiliary Cooling System (CVACS). We developed a method to use the reduction rate of core damage frequency as an indicator for effectiveness of the measures for improving resilience. The core damage frequency was evaluated by calculating the core cooling performance using CVACS, reflecting the results of structural analysis and human reliability analysis. By implementing measures for improving resilience in addition to existing measures, the core damage frequency of Japan loop-type sodium-cooled fast reactor caused by LOHRS has been reduced to about one-hundredth of the previous level.
Onoda, Yuichi; Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa
no journal, ,
Focusing on the accident sequence leading to ultra-high temperatures due to loss of heat removal systems, the effectiveness of the measures for improving reactor structural resilience was evaluated for next-generation sodium-cooled fast reactors. Focusing on countermeasures to promote heat dissipation from the reactor vessel, the cooling performance of these measures was evaluated. In addition, we examined the uncertainty of accident progression and the success or failure of the measures for improving resilience, and evaluated the branch probability of the event tree. As a result, the reduction rate of the core damage frequency due to the assumed countermeasures was quantified.
Onoda, Yuichi; Kurisaka, Kenichi; Yamano, Hidemasa
no journal, ,
Focusing on the accident sequence leading to ultra-high temperatures due to loss of heat removal systems, the effectiveness of the measures for improving reactor structural resilience was preliminarily evaluated for next-generation sodium-cooled fast reactors. Focusing on countermeasures to promote heat dissipation from the reactor vessel, the cooling performance of these measures was preliminarily evaluated. In addition, we examined the uncertainty of accident progression and the success or failure of the measures for improving resilience, and preliminarily evaluated the branch probability of the event tree. As a result, it was confirmed that the reduction rate of the core damage frequency due to the assumed countermeasures can be quantified.