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Kamiyama, Kenji; Matsuba, Kenichi; Kato, Shinya; Imaizumi, Yuya; Mukhamedov, N.*; Akayev, A.*; Pakhnits, A.*; Vurim, A.*; Baklanov, V.*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 9 Pages, 2022/04
Li, C.-Y.; Uchibori, Akihiro; Takata, Takashi; Pellegrini, M.*; Erkan, N.*; Okamoto, Koji*
Dai-25-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2021/07
The capability of stable cooling and avoiding re-criticality on the debris bed are the main issues for achieving IVR (In-Vessel Retention). In the actual situation, the debris bed is composed of mixed-density debris particles. Hence, when these mixed-density debris particles were launched to re-distribute, the debris bed would possibly form a density-stratified distribution. For the proper evaluation of this scenario, the multi-physics model of CFD-DEM-Monte-Carlo based neutronics is established to investigate the coolability and re-criticality on the heterogeneous density-stratified debris bed with considering the particle relocation. The CFD-DEM model has been verified by utilizing water injection experiments on the mixed-density particle bed in the first portion of this research. In the second portion, the coupled system of the CFD-DEM-Monte-Carlo based neutronics model is applied to reactor cases. Afterward, the debris particles' movement, debris particles' and coolant's temperature, and the k-eff eigenvalue are successfully tracked. Ultimately, the relocation and stratification effects on debris bed's coolability and re-criticality had been quantitatively confirmed.
Takase, Kazuyuki; Ose, Yasuo*; Akimoto, Hajime
Nihon Kikai Gakkai Dai-12-Kai Keisan Rikigaku Koenkai Koen Rombunshu, p.395 - 396, 1999/00
no abstracts in English
Sogabe, Joji; Kamiyama, Kenji; Tobita, Yoshiharu; Okano, Yasushi
no journal, ,
During severe accidents by an anticipated transient without scram, it is important to evaluate multiphase multi-component flow behavior, when a part of the disrupted core material is discharged outside the disrupted core region through control rod guide tubes. In particular, the blockage behavior of the disrupted core material in a flow path is an important phenomenon that affects the amount of relocated fuels (the fuel discharged outside the disrupted core region and the fuel remaining in the disrupted core region). A fast reactor safety analysis code, SIMMER, is currently being developed for application to the post-accident material relocation (PAMR) phase. In the paper, aiming at actual reactor analyses for the PAMR phase of the SIMMER code, a model for the blockage in the flow path for possible phenomena in the PAMR phase. The model improves the applicability of the SIMMER code to the PAMR phase on the actual reactors.