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Sonehara, Masateru; Okano, Yasushi; Uchibori, Akihiro; Oki, Hiroshi*
Journal of Nuclear Science and Technology, 62(5), p.403 - 414, 2025/05
Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)For sodium-cooled fast reactors, understanding sodium combustion behaviour is crucial for managing sodium leakage accidents. In this study, we perform benchmark analyses of the Sandia National Laboratories (SNL) T3 experiment using the multi-dimensional thermal hydraulic code AQUA-SF. Conducted in an enclosed space with a large vessel volume of 100 m and a sodium mass flow rate of 1 kg/s, the experiment highlighted the multi-dimensional effects of local temperature increase shortly after sodium injection. This study aims to extend the capabilities of AQUA-SF by focusing on the simulation of these multi-dimensional temperature variations, in particular the formation of high temperature regions at the bottom of the vessel. The proposed models include the temporary stopping of sodium droplet ignition and spray combustion of sodium splash on the floor. Furthermore, it has been shown that additional heat source near the floor is essential to enhance the reproduction of the high temperature region at the bottom. Therefore, case studies including sensitivity analyses of spray cone angle and prolonged combustion of droplets on the floor are conducted. This comprehensive approach provides valuable insights into the dynamics of sodium combustion and safety measures in sodium-cooled fast reactors.
Matsushita, Kentaro; Ezure, Toshiki; Tanaka, Masaaki; Imai, Yasutomo*; Fujisaki, Tatsuya*; Sakai, Takaaki*
Nuclear Engineering and Design, 432, p.113785_1 - 113785_16, 2025/02
Times Cited Count:1 Percentile:0.00(Nuclear Science & Technology)Establishing an evaluation method for the gas entrainment (GE) of argon cover gas due to surface vortices is required in terms of safety design of sodium-cooled fast reactors. To modify the evaluation model in an in-house evaluation tool for GE, StreamViewer, a modified evaluation model on the pressure distribution along the vortex center line (PVL model) was proposed to identify the vortex center lines by connecting continuous vortex center points from the suction port to the surface and evaluate gas core length based on the balance between the hydrostatic pressure and the pressure decrease distribution along the vortex center line. PVL model was applied the three-dimensional numerical analysis results for the experiments where a plate induced unsteady traveling vortices in the open channel flow. Consequently, the GE evaluation using StreamViewer with PVL model could reproduce the relation between the inlet flow velocity and the gas core length in the unsteady vortex flow experiments.
Onishi, Takashi; Koyama, Shinichi*; Yokoyama, Keisuke; Morishita, Kazuki; Watanabe, Masashi; Maeda, Shigetaka; Yano, Yasuhide; Oki, Shigeo
Nuclear Engineering and Design, 432, p.113755_1 - 113755_17, 2025/02
Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)Doi, Daisuke
International Journal of Hydrogen Energy, 91, p.1245 - 1252, 2024/11
Times Cited Count:1 Percentile:25.55(Chemistry, Physical)Emura, Yuki; Matsuba, Kenichi; Kikuchi, Shin; Yamano, Hidemasa
Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 8 Pages, 2024/11
Miyazawa, Takeshi; Uwaba, Tomoyuki; Yano, Yasuhide; Tanno, Takashi; Otsuka, Satoshi; Onizawa, Takashi; Ando, Masanori; Kaito, Takeji
JAEA-Technology 2024-009, 140 Pages, 2024/10
For the purpose of enhancing the reliability of fast reactor fuel designing using modified type 316 steel, the out-of-pile and in-pile mechanical data of modified type 316 steel cladding tubes and wrapper tubes were statistically analyzed with the knowledge on material science and engineering; the high-temperature strength equations of modified type 316 steel, which can be applied to high-dose neutron irradiation environment, were derived. The out-of-pile high-temperature tensile and creep data of modified type 316 steel cladding tubes and wrapper tubes were derived up to 900C, which is higher than the upper limit temperature of anticipated transient event of fast reactor. Using the extended database, the best-fit equation and the lower limit equation were derived for out-of-pile 0.2% proof strength, ultimate tensile strength and creep rupture strength while the best-fit equation and the upper and lower limit equations for creep strain. Furthermore, the degradation factors for tensile and creep strength, which will be produced by in-reactor environment (i.e., neutron irradiation in liquid sodium), were evaluated using the existing neutron irradiation data of modified type 316 steel, which were derived using the experimental fast reactor Joyo, the French proto-type fast reactor Phenix, the American experimental fast reactor FFTF. The derived equations were validated by the comparison with the experimental data.
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.
Yamano, Hidemasa; Futagami, Satoshi; Higurashi, Koichi*
Proceedings of Advanced Reactor Safety (ARS 2024), p.121 - 130, 2024/08
This paper describes the application of safety design criteria (SDC) and safety design guideline (SDG) developed in the Generation-IV international forum on decay heat removal system (DHRS) enhancing reliability to sodium-cooled fast reactors (SFRs) recently designed in Japan.
Yamano, Hidemasa; Futagami, Satoshi; Shibata, Akihiro*
Proceedings of Advanced Reactor Safety (ARS 2024), p.151 - 160, 2024/08
This study examined the application of safety design criteria (SDC) and safety design guideline (SDG) developed in the Generation-IV international forum on the active reactor shutdown system (RSS) to sodium-cooled fast reactors (SFRs) recently designed in Japan.
Emura, Yuki; Takai, Toshihide; Kikuchi, Shin; Kamiyama, Kenji; Yamano, Hidemasa; Yokoyama, Hiroki*; Sakamoto, Kan*
Journal of Nuclear Science and Technology, 61(7), p.911 - 920, 2024/07
Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)Ishida, Shinya; Uchibori, Akihiro; Okano, Yasushi
Dai-28-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2024/06
no abstracts in English
Miyazawa, Takeshi; Tanno, Takashi; Imagawa, Yuya; Hashidate, Ryuta; Yano, Yasuhide; Kaito, Takeji; Otsuka, Satoshi; Mitsuhara, Masatoshi*; Toyama, Takeshi*; Onuma, Masato*; et al.
Journal of Nuclear Materials, 593, p.155008_1 - 155008_16, 2024/05
Times Cited Count:1 Percentile:51.66(Materials Science, Multidisciplinary)Yamamoto, Tomohiko; Kato, Atsushi; Hayakawa, Masato; Shimoyama, Kazuhito; Ara, Kuniaki; Hatakeyama, Nozomu*; Yamauchi, Kanau*; Eda, Yuhei*; Yui, Masahiro*
Nuclear Engineering and Technology, 56(3), p.893 - 899, 2024/03
Times Cited Count:1 Percentile:51.66(Nuclear Science & Technology)Kamide, Hideki; Kawasaki, Nobuchika; Hayafune, Hiroki; Kubo, Shigenobu; Chikazawa, Yoshitaka; Maeda, Seiichiro; Sagayama, Yutaka; Nishihara, Tetsuo; Sumita, Junya; Shibata, Taiju; et al.
Jisedai Genshiro Ga Hiraku Atarashii Shijo; NSA/Commentaries, No.28, p.14 - 36, 2023/10
Developments of next generation nuclear reactors, e.g., Fast Reactor, and High Temperature Gas cooled Reactor, are in progress. They can contribute to markets of electricity and industrial heat utilization in the world including Japan. Here, current status of reactor developments in Japan and also situation in the world are summarized, especially for activities of Generation IV International Forum (GIF), developments of Fast Reactor and High Temperature Gas cooled Reactor in Japan, and SMR movements in the world.
Ohno, Shuji; Maeda, Seiichiro
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 3 Pages, 2023/09
Kosaka, Wataru; Uchibori, Akihiro; Okano, Yasushi; Yanagisawa, Hideki*
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.1150 - 1163, 2023/08
The leakage of pressurized water from a steam generator (SG) and the progress after that are a key issue in the safety assessment or design of a SG in sodium-cooled fast reactor. The analysis code LEAP-III can evaluate a rate of water leakage during the long-term event progress, i.e., from the self-wastage initiated by an occurrence of a microscopic crack in a tube wall to the water leak detection and water/water-vapor blowdown. Since LEAP-III consists of semi-empirical formulae and one-dimensional equations of conservation, it has an advantage in short computation time. Thus, LEAP-III can facilitate the exploration of various new SG designs in the development of innovative reactors. However, there are several problems, such as an excessive conservative result in some case and the need for numerous experiments or preliminary analyses to determine tuning parameters of models in LEAP-III. Hence, we have developed a Lagrangian particle method code, which is characterized by a simpler computational principle and faster calculation. In this study, we have improved the existing particle pair search method for interparticle interaction in this code and developed an alternative model without the pair search. Through the trial analysis simulating in a tube bundle system, it was confirmed that new models reduced the computation time. In addition, it was shown that representative temperatures of the heat-transfer tubes evaluated by this particle method code, which is used to predict the tube failure in LEAP-III, were good agreement with that by SERAPHIM, which is a detailed mechanistic analysis method code.
Li, C.-Y.; Wang, K.*; Uchibori, Akihiro; Okano, Yasushi; Pellegrini, M.*; Erkan, N.*; Takata, Takashi*; Okamoto, Koji*
Applied Sciences (Internet), 13(13), p.7705_1 - 7705_29, 2023/07
Times Cited Count:2 Percentile:36.16(Chemistry, Multidisciplinary)Tsai, T.-H.; Sasaki, Shinji; Maeda, Koji
Journal of Nuclear Science and Technology, 60(6), p.715 - 723, 2023/06
Times Cited Count:1 Percentile:14.04(Nuclear Science & Technology)Takino, Kazuo; Oki, Shigeo
JAEA-Data/Code 2023-003, 26 Pages, 2023/05
Since next-generation fast reactors aim to achieve a higher core discharge burn-up than conventional reactors do, core neutronics design methods must be refined. Therefore, a suitable analysis condition is required for the analysis of burn-up nuclear characteristics to accomplish sufficient estimation accuracy while maintaining a low computational cost. We investigated the effect of the analysis conditions on the accuracy of estimation of the burn-up nuclear characteristics of next-generation fast reactors in terms of neutron energy groups, neutron transport theory, and spatial mesh. This study treated the following burn-up nuclear characteristics: criticality, burn-up reactivity, control rod worth, breeding ratio, assembly-wise power distribution, maximum linear heat rate, sodium void reactivity, and Doppler coefficient for the equilibrium operation cycle. As a result, it was found that the following conditions were the most suitable: 18-energy-group structure, 6 spatial meshes per assembly with diffusion approximation. Additionally, these conditions should apply to correction factors for energy group structure, spatial mesh and transport effects.
Yoshida, Hiroyuki; Horiguchi, Naoki; Furuichi, Hajime*; Katono, Kenichi*
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 7 Pages, 2023/05