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Futagami, Satoshi; Ando, Masanori; Yamano, Hidemasa
Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 10 Pages, 2024/03
Onoda, Yuichi; Kurisaka, Kenichi; Yamano, Hidemasa
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 7 Pages, 2023/05
The objective of this study is to develop an effectiveness evaluation methodology of the measures for improving resilience of nuclear structures at ultra-high temperature by using the failure mitigation technology. At the beginning, to identify the accident sequences having the potential to improve resilience, the characteristics of a next-generation loop-type sodium-cooled fast reactor (SFR) in Japan has been investigated by analyzing the event tree of level-1 and level-2 probabilistic risk assessment. As a result, event sequences of loss of heat removal systems (LOHRS) are identified. The effectiveness of the measures for improving resilience is evaluated by quantifying the reduction rate of core damage frequency before and after the introduction of the measures for improving resilience for all the accident sequences leading to LOHRS. To examine applicability of the developed methodology, a trial evaluation has conducted for a next-generation loop-type SFR in Japan. Through the applicability examined, the method for the effectiveness evaluation was developed successfully. The refinement of the conditional success probability of the measures for improving resilience is the future work.
Onoda, Yuichi; Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa
Proceedings of Asian Symposium on Risk Assessment and Management 2021 (ASRAM 2021) (Internet), 11 Pages, 2021/10
The effectiveness evaluations technology of the measures for improving resilience by applying a fracture control concept under ultra-high temperature conditions has developed for prototype sodium-cooled fast reactor Monju as a model plant, and the trial evaluation has conducted using this technology in this paper. The important accident sequences to which the fracture control concept is expected to be applied under ultra-high temperature condition are identified by investigating the results of the existing researches of level-2 probabilistic risk assessment for Monju. Accident sequences categorized in protected loss of heat sink and loss of reactor level are both identified as such important accident sequences which has the potential to prevent core damage. This study has developed the technology to evaluate the effectiveness of improving resilience, where the headings which stand for success or failure of the measures to improve resilience are introduced into the event tree, the branch probability of them is set, and the effectiveness of improving resilience is expressed as the reduction of core damage frequency. As a result of the trial evaluation of the effectiveness for the measures to improve resilience, it is confirmed that core damage frequency can be reduced by applying fracture control concept. The branch probability of the measures to improve resilience proposed in this study is tentatively assigned based on the assumption. This value is expected to be quantified by the forthcoming analyses of the integrity for the reactor vessel structure at ultra-high temperature. The technology developed in this study will be applied for the evaluation of improving resilience of the next generation sodium-cooled fast reactor.
Futagami, Satoshi; Ando, Masanori; Yamano, Hidemasa; Kasahara, Naoto*
no journal, ,
To improve resilience of next-generation nuclear power plants, the authors have performed a structural analysis assuming a protected loss of heat sink event, which may cause all decay heat removal systems to lose their functions immediately after reactor shutdown. The results show how the reactor vessel and guard vessel deform under extremely high temperatures.
Futagami, Satoshi; Ando, Masanori; Yamano, Hidemasa; Kasahara, Naoto*
no journal, ,
To improve resilience of next-generation nuclear power plants, the authors have performed structural analyses assuming a protected loss of heat sink event, which may cause all decay heat removal systems to lose their functions immediately after reactor shutdown. From the results of these analysis, the authors developed a passive safety structure under extremely high temperatures.