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Terada, Atsuhiko; Thwe Thwe, A.; Hino, Ryutaro*
JAEA-Review 2024-049, 400 Pages, 2025/03
JAEA-Review-2024-049.pdf:13.94MB
In the aftermath of the Fukushima Daiichi Nuclear Power Station accident, safety measures against hydrogen in severe accident has been recognized as a serious technical problem in Japan. As one of efforts to form a common knowledge base between nuclear engineers and experts on combustion and explosion, we issued the "Handbook of Advanced Nuclear Hydrogen Safety (1st edition)" in 2017. For improvement of the rational advancement of hydrogen safety measures and further reliability of hydrogen safety evaluation, a CFD analysis is highly expected to produce more precisely and quantitative results. We have been developing an integrated CFD analysis code system which can analyze hydrogen diffusion, explosion-combustion and structural integrity at the severe accident especially for pressurized water reactors (PWRs). We organized the role of LP and the CFD analyses and their utilization examples of hydrogen safety validation. Based on these results, we made the "Handbook of Advanced Nuclear Hydrogen Safety (2nd volume)". The analysis results of real scale PWR described in 2nd volume are confirmed by cross-analysis models and existing data obtained through representative small, medium and large-scale tests.
Brumm, S.*; Gabrielli, F.*; Sanchez Espinoza, V.*; Stakhanova, A.*; Groudev, P.*; Petrova, P.*; Vryashkova, P.*; Ou, P.*; Zhang, W.*; Malkhasyan, A.*; et al.
Annals of Nuclear Energy, 211, p.110962_1 - 110962_16, 2025/02
Times Cited Count：5 Percentile：90.97(Nuclear Science & Technology)Tsukimori, Kazuyuki; Yada, Hiroki
Journal of Pressure Vessel Technology, 147, p.031901_1 - 031901_9, 2025/00
After the accident at the Fukushima Daiichi Nuclear Power Plant, very strict safety measures were implemented for nuclear power plants in Japan. It thus becomes a crucial issue if the safety of a plant is maintained or not at beyond design basis events. In this study, head plates and bellows were examined as components that compose the parts of the boundary of vessels that contain the primary coolant of a prototype fast breeder reactor. The behaviors of buckling, post-buckling deformation, and penetration failure, that is, loss of boundary function of these components with increasing pressure were investigated. The series of this research program started in FY2013 and the research proceeded step by step. The new result in this paper is the application of the proposed criteria to head plates and bellows, and a conservative estimation of penetration failure of these components is obtained.
Yamada, Takeshi*; Li, X.; Yamashita, Takuya; Yamaji, Akifumi*
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 10 Pages, 2024/11
In this study, a new crust model is being developed to analyze MCCI, which involves continuous concrete ablation with presence of the crust layer between the corium and the concrete walls, which may gradually move with the slow concrete wall ablation process over long time. The new crust model must enable accumulation of physical displacement of the crust particle over long time (i.e., enable physical creeping) while preventing accumulation of numerical displacement of the crust particles over long time (i.e., preventing numerical creeping), Hence, in the new crust model, the PS has been effectively disabled for the crust particles. Qualitative validity of such numerical modeling was confirmed through some trial analyses of VULCANO-VBS test using a set of tentative calculation conditions and parameters, which should be carefully revised for future quantitative discussions including validation of the analysis results with experimental results.
Sogabe, Joji; Ishida, Shinya; Tagami, Hirotaka; Okano, Yasushi; Kamiyama, Kenji; Onoda, Yuichi; Matsuba, Kenichi; Yamano, Hidemasa; Kubo, Shigenobu; Kubota, Ryuzaburo*; et al.
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
In the frame of France-Japan collaboration, the calculational methodologies were defined and assessed, and the phenomenology and the severe accident consequences were investigated in a pool-type sodium-cooled fast reactor.
Onoda, Yuichi; Ishida, Shinya; Fukano, Yoshitaka; Kamiyama, Kenji; Yamano, Hidemasa; Kubo, Shigenobu; Shibata, Akihiro*; Bertrand, F.*; Seiler, N.*
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
at temperature range 170 - 290
CLuu, V. N.; Nakajima, Kunihisa
Nuclear Engineering and Design, 426, p.113402_1 - 113402_7, 2024/09
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Matsumoto, Toshinori; Hibiki, Takashi*; Maruyama, Yu
International Journal of Energy Research, 2024(1), p.9748588_1 - 9748588_18, 2024/08
Times Cited Count：0 Percentile：0.00(Energy & Fuels)To evaluate the effectiveness of the wet cavity strategy, the authors developed a stochastic evaluation method that considers the uncertainties of the molten material conditions ejected from reactor pressure vessels. The first step was uncertainty analysis using the MELCOR code to obtain the melt condition. Five uncertain parameters related to the core degradation process were chosen. The input parameter sets were generated using Latin hypercube sampling. The second step was analyzing of the melt-behavior using the JASMINE code. The probabilistic distribution of parameters for the JASMINE analyses was determined from the MELCOR analysis results. The initial water depth was set to 0.5, 1.0, and 2.0 m. The debris height was compared with the criterion to judge its coolability. Consequently, the success probability of debris cooling was obtained through a sequence of calculations. The feasibility and technical difficulties in the MELCOR-JASMINE combined analysis were also discussed.
Nakamura, Yuki*; Kojima, Yoshihiro*; Yamashita, Takuya; Shimomura, Kenta; Mizokami, Shinya
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NTHOS-14) (Internet), 12 Pages, 2024/08
Ishida, Shinya; Uchibori, Akihiro; Okano, Yasushi
Dai-28-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2024/06
no abstracts in English
Ishida, Shinya; Fukano, Yoshitaka; Tobita, Yoshiharu; Okano, Yasushi
Journal of Nuclear Science and Technology, 61(5), p.582 - 594, 2024/05
Times Cited Count：1 Percentile：27.70(Nuclear Science & Technology)Li, X.; Yamaji, Akifumi*; Sato, Ikken*; Yamashita, Takuya; Nagae, Yuji
Proceedings of 11th European Review Meeting on Severe Accident Research Conference (ERMSAR 2024) (Internet), 12 Pages, 2024/05
Li, N.*; Sun, Y.*; Nakajima, Kunihisa; Kurosaki, Ken*
Journal of Nuclear Science and Technology, 61(3), p.343 - 353, 2024/03
Times Cited Count：1 Percentile：27.70(Nuclear Science & Technology)During the Fukushima Daiichi nuclear power plant (1F) accident, an overwhelming amount of the cesium remaining in the pressure vessel could have been deposited onto 304 stainless steel (SS304) steam separators and dryers, both with large surface areas. During 1F's decommissioning, the deposited cesium is a safety hazard as it can generate radioactive dust. However, the cohesive and adhesive strengths of CsOH-chemisorbed oxide scales are yet to be defined. In this study, we investigated how CsOH-chemisorption affects the cohesive and adhesive strengths between oxide scales and SS304 substrates with a scratch tester. The scratch test results revealed that the cohesive strengths of the oxide scales decreased after CsOH-chemisorption, while adhesive failure could not be reached.
Sato, Ikken; Yoshikawa, Shinji; Yamashita, Takuya; Shimomura, Kenta; Cibula, M.*; Mizokami, Shinya*
Nuclear Engineering and Design, 414, p.112574_1 - 112574_20, 2023/12
Katsumura, Kosuke*; Takagi, Junichi*; Hosomi, Kenji*; Miyahara, Naoya*; Koma, Yoshikazu; Imoto, Jumpei; Karasawa, Hidetoshi; Miwa, Shuhei; Shiotsu, Hiroyuki; Hidaka, Akihide*; et al.
Nihon Genshiryoku Gakkai-Shi ATOMO
, 65(11), p.674 - 679, 2023/11
no abstracts in English
Maruyama, Yu; Sugiyama, Tomoyuki*; Shimada, Asako; Lind, T.*; Bentaib, A.*; Sogalla, M.*; Pellegrini, M.*; Albright, L.*; Clayton, D.*
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.4782 - 4795, 2023/08
Nanjo, Kotaro; Shiotsu, Hiroyuki; Maruyama, Yu; Sugiyama, Tomoyuki; Okamoto, Koji*
Journal of Nuclear Science and Technology, 60(7), p.816 - 823, 2023/07
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Yamashita, Takuya; Honda, Takeshi*; Mizokami, Masato*; Nozaki, Kenichiro*; Suzuki, Hiroyuki*; Pellegrini, M.*; Sakai, Takeshi*; Sato, Ikken; Mizokami, Shinya*
Nuclear Technology, 209(6), p.902 - 927, 2023/06
Times Cited Count：5 Percentile：82.11(Nuclear Science & Technology)Li, C.-Y.; Watanabe, Akira*; Uchibori, Akihiro; Okano, Yasushi
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05
pellet in molten Zr claddingIto, Ayumi*; Yamashita, Susumu; Tasaki, Yudai; Kakiuchi, Kazuo; Kobayashi, Yoshinao*
Journal of Nuclear Science and Technology, 60(4), p.450 - 459, 2023/04
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)