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Futagami, Satoshi; Ichikawa, Kenta*; Mochida, Haruo*; Kurome, Kazuya*; Goto, Shoji; Enuma, Yasuhiro
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR26) (Internet), 6 Pages, 2026/05
Kam, D. H.*; Grabaskas, D.*; Okano, Yasushi; Uchibori, Akihiro; Starkus, T.*
Nuclear Technology, 212(2), p.347 - 364, 2026/02
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Onoda, Yuichi; Uchita, Masato*; Tokizaki, Minako*; Okazaki, Hitoshi*
Nuclear Technology, 211(11), p.2812 - 2831, 2025/11
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Kato, Shinya; Doda, Norihiro; Yokoyama, Kenji; Tanaka, Masaaki; Endo, Tomohiro*
Proceedings of 2025 International Congress on Advances in Nuclear Power Plants (ICAPP 2025) (Internet), 11 Pages, 2025/09
During a reactor power increase in ULOF and UTOP events in sodium-cooled fast reactors, core deformation due to thermal expansion of core elements is expected to cause a negative feed-back effect to suppress this power increase. An analytical evaluation method of core deformation reactivity for design is being developed in JAEA. However, the neutronics calculation module uses several approximations. This study aims to develop the detailed evaluation method as a reference neutron transport calculation code for confirmation of the validity of the calculated core deformation reactivity. Here, the two-dimensional finite volume method (FVM) code based on simplified P3 (SP3) approximation with unstructured mesh have been developed as the first step of the development. This paper describes the calculation theory of the FVM code, the procedure of introducing SP3 approximation into the code and the verification results of the functions developed.
Onoda, Yuichi; Kubota, Ryuzaburo*; Yamano, Hidemasa
Proceedings of 2025 International Congress on Advances in Nuclear Power Plants (ICAPP 2025) (Internet), 11 Pages, 2025/09
Ishida, Shinya; Tagami, Hirotaka*; Yamano, Hidemasa; Kubo, Shigenobu
Proceedings of 2025 International Congress on Advances in Nuclear Power Plants (ICAPP 2025) (Internet), 11 Pages, 2025/09
Wozniak, N.*; Ohgama, Kazuya; Doda, Norihiro; Ota, Hirokazu*; Shemon, E.*; Feng, B.*; Uwaba, Tomoyuki; Futagami, Satoshi; Tanaka, Masaaki; Yamano, Hidemasa; et al.
Progress in Nuclear Science and Technology (Internet), 8, p.165 - 170, 2025/09
To enhance the accuracy of the safety evaluations for sodium-cooled fast reactors, it is necessary to develop a method that can realistically evaluate the reactivity changes induced by core deformation. In this context, Japan and the United States jointly conducted a benchmark analysis of thermal bowing experiments using multiple ducts from a Joyo-type fuel assembly. The aim was to confirm the validity of the core bowing analysis codes. Comparisons of analysis and experimental results demonstrated that the codes used by both countries were able to reasonably predict the thermal bowing of a row of assemblies in multiple duct configuration.
Ohgama, Kazuya; Doda, Norihiro; Ota, Hirokazu*; Wozniak, N.*; Uwaba, Tomoyuki; Futagami, Satoshi; Tanaka, Masaaki; Yamano, Hidemasa; Ogata, Takanari*; Shemon, E.*; et al.
Progress in Nuclear Science and Technology (Internet), 8, p.160 - 164, 2025/09
To enhance the accuracy of the safety evaluations for sodium-cooled fast reactors, it is necessary to develop a method that can realistically evaluate the reactivity changes induced by core deformation. In this context, Japan and the United States jointly conducted a benchmark analysis of thermal bowing experiments using a single duct from a Joyo-type fuel assembly. The aim was to confirm the validity of the core bowing analysis codes. Comparisons of analysis and experimental results demonstrated that the codes used by both countries were able to reasonably predict the axial distribution of horizontal duct displacement caused by thermal bowing as well as the contact load on the duct pad.
Hayakawa, Satoshi*; Hagiwara, Hiroyuki*; Imamura, Akira*; Onoda, Yuichi; Tanaka, Masaaki; Nakamura, Hironori*
Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 8 Pages, 2024/11
In a sodium-cooled fast reactor, a cover gas region filled with argon gas is located above the sodium pool in the main vessel to prevent the hot sodium from contacting the structures. This region involves heat transportation by natural convection of the cover gas, radiation among liquid surface and structures, and sodium phase change between mist and vapor. In this study, the numerical evaluation method has been developed with a commercial CFD code, Fluent, incorporating the sodium mist transport and growth models, and the radiation scattering model. Simulations of a laboratory scale test with a cylindrical cover gas region was carried out for the validation of the method and showed that the temperature distribution and sodium mist concentration in the cover gas region are in good agreements with the test results. A simulation of a pool-type sodium cooled fast reactor has also conducted and the basic aspect of physical phenomena taking place in the cover gas region were evaluated.
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.
Takeda, Takeshi; Shibata, Taiju
JAEA-Review 2024-040, 29 Pages, 2024/09
JAEA-Review-2024-040.pdf:1.33MB
An important theme of Japan's 6th strategic energy plan is to indicate the energy policy path towards carbon neutrality by 2050. Policy responses for Japan's nuclear energy research and development (R&D) towards 2030 contain the demonstrations of technologies for small modular reactors (SMRs) through international cooperation by 2030. In light of this energy plan, basic policy initiatives over the next 10 years have been compiled to realize Green Transformation (GX), which simultaneously achieves decarbonization and economic growth. Looking overseas, activities of SMR R&D are active internationally, mainly in the US, Canada, Europe, China, and Russia. These activities are not only by heavy industry manufactures and R&D institutes, but also by venture companies. Under these circumstances, the NEA CSNI has gathered an Expert Group on SMRs (EGSMR) to help estimate the safety effects of SMRs. The EGSMR efforts required the submission of responses to several questionnaires whose main purpose was to collect the latest information on the efforts of SMR deployment and research. The first author of this report responded to this based on information from Hitachi-GE Nuclear Energy, Ltd. and Mitsubishi Heavy Industries, Ltd. as well as JAEA. Most of the responses from Japan to the questionnaires are the information that serves as the basis of CSNI Technical Opinion Paper No. 21 (TOP-21). In this report, the Japan's publicly available responses to the questionnaires arranged and additional information are explained, which complements some of the content of the TOP-21. In this manner, the investigation results of R&D related to SMR in Japan, focusing on the EGSMR activities (2022-2023), are summarized. The target of this report is to provide useful information for future discussions on international cooperation concerning SMR as well as nuclear power field human resources development internationally and domestically.
Ishida, Shinya; Tagami, Hirotaka; Okano, Yasushi; Yamano, Hidemasa; Kubo, Shigenobu; Tobita, Yoshiharu
Proceedings of 11th European Review Meeting on Severe Accident Research Conference (ERMSAR 2024) (Internet), 10 Pages, 2024/05
Yamamoto, Tomohiko; Watakabe, Tomoyoshi; Miyazaki, Masashi; Okamura, Shigeki; Miyagawa, Takayuki; Yokoi, Shinobu*; Fukasawa, Tsuyoshi*; Fujita, Satoshi*
Mechanical Engineering Journal (Internet), 11(2), p.23-00393_1 - 23-00393_21, 2024/04
Yamamoto, Tomohiko; Watakabe, Tomoyoshi; Miyazaki, Masashi; Miyagawa, Takayuki*; Yokoi, Shinobu*; Okamura, Shigeki*; Fukasawa, Tsuyoshi*; Fujita, Satoshi*
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 7 Pages, 2023/05
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.
Ishida, Shinya; Fukano, Yoshitaka; Tobita, Yoshiharu; Okano, Yasushi
Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 8 Pages, 2023/04
Yamamoto, Tomohiko; Kato, Atsushi; Hayakawa, Masato; Shimoyama, Kazuhito; Ara, Kuniaki; Hatakeyama, Nozomu*; Yamauchi, Kanau*; Eda, Yuhei*; Yui, Masahiro*
Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 6 Pages, 2023/04
Onoda, Yuichi; Uchita, Masato*; Tokizaki, Minako*; Okazaki, Hitoshi*
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 6 Pages, 2022/08
The safety analyses were carried out to confirm the sufficiency of the function of the plant protection system against the pump/diagrid link rupture. The target plant is a pool-type SFR of about 600 MWe class equipped with an axially homogeneous core currently under development in Japan. In the pool-type SFR, the primary system piping connects primary pump and the high-pressure sodium plenum located at the inlet of fuel sub-assemblies and called "pump/diagrid link". Because this piping is submerged in the reactor vessel, it is difficult to detect small scale sodium leakage in this piping, and thus a certain large pipe break like guillotine should be assumed and evaluated as a design basis event. In order to confirm the detectability of pump/diagrid link rupture by safety protection system signals, a series of analyses of the guillotine break for a pump/diagrid link were carried out. Sensitivity study had also been performed to consider the uncertainty of the reactivity coefficient in the analyses. The sufficiency of the function of the plant protection system against the pump/diagrid link rupture was confirmed by the analysis results that at least two signals are transmitted for the detection of the event, which is the development target of the plant protection system in pool-type SFR.
Li, C.-Y.; Watanabe, Akira*; Uchibori, Akihiro; Okano, Yasushi
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2022/07
Identifying accident scenarios that could lead to severe accidents and evaluating their frequency of occurrence are essential issues. This study aims to establish the methodology of the dynamic Probabilistic Risk Assessment (PRA) for sodium-cooled fast reactors that can consider the time dependency and the interdependence of each event. Specifically, the Continuous Markov chain Monte Carlo (CMMC) method is newly applied to the SPECTRA code, which analyzes the severe accident conditions of nuclear reactors, to develop an evaluation methodology for typical external hazards. Currently, a fault-tree model of air coolers of decay heat removal system is implemented as the CMMC method, and a series of preliminary analysis of the plant's transient characteristics under the scenario of volcanic ashfall has been conducted.
Quaini, A.*; Goss
, S.*; Payot, F.*; Suteau, C.*; Delacroix, J.*; Saas, L.*; Gubernatis, P.*; Martin-Lopez, E.*; Yamano, Hidemasa; Takai, Toshihide; et al.
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 10 Pages, 2022/04
CEA and JAEA defined new sub-tasks under the current implementing arrangement: Kinetics of interaction in core material mixtures- Physical properties of core material mixtures, High temperature thermodynamic data for the UO
-Fe-B
C system, Experimental studies on BC-SS kinetics and BC-SS eutectic material relocation (freezing), BC/SS eutectic and kinetics models for SIMMER code systems, Methodology for the modelling of mixtures liquefaction kinetics. The paper describes major R&D results obtained in the France-Japan collaboration under the previous implementing arrangement as well as experimental and analytical roadmaps under the current arrangement.
