Study on severe accidents of oxide-fueled sodium-cooled fast reactors, 1; Project overview and progress until 2025

Yamano, Hidemasa; Matsuba, Kenichi; Sasa, Kyohei*; Kinoshita, Enki*; Morita, Koji*; Saito, Yasushi*; Ito, Kei*; Suzuki, Toru*

Dai-30-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2026/06

Demonstration SFR conceptual design project in Japan, 1; Plan and status

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

A Study for establishment of passive creep-fatigue test techniques using the difference of thermal expansion coefficients of materials

Wakai, Takashi; Ando, Masanori; Okajima, Satoshi; Toyota, Kodai; Ishigami, Katsuo*; Onuma, Terumitsu*; Takahashi, Ryoya*; Asayama, Tai

Nihon Kikai Gakkai Rombunshu (Internet), 12 Pages, 2026/00

https://doi.org/10.1299/transjsme.26-00004

This paper describes current efforts within the Japan Atomic Energy Agency to develop a new materials surveillance technology for sodium-cooled fast reactors that aims at raising the confidence levels in operating reactors' structural integrity to the ones which could otherwise be achieved only by a vast amount of long-term material test data that needs decades to accumulate. The technology would allow for monitoring on-going materials degradation processes in an actual reactor, namely fatigue and creep-fatigue, by inserting a newly developed specimen for this purpose into the reactor and establishing procedures to observe and evaluate the specimen to get desired information such as the extent of degradation and residual life. To establish the test procedure, a passive creep-fatigue test model was developed in the framework of the "Civil Nuclear Energy Research and Development Working Group (CNWG)", a collaboration agreement between the US and Japan, and a demonstration test was performed under cyclic thermal loads using an electric furnace. Temperature conditions had been determined based on finite element analyses. As a result, the test specimen failed by several hundred thermal cycles. Macroscopic inspection of the specimen after the test showed the test had been performed successfully without buckling. Fracture surface observations suggested that the failure caused probably by fatigue or creep-fatigue.

Application study of adaptive mesh refinement method on unsteady wake vortex analysis

Alzahrani, H.*; Matsushita, Kentaro; Sakai, Takaaki*; Ezure, Toshiki; Tanaka, Masaaki

Nuclear Technology, 211(10), p.2446 - 2458, 2025/10

https://doi.org/10.1080/00295450.2025.2472582

Development of evaluation method for cover gas entrainment by vortices generated at free surface in upper plenum of sodium-cooled fast reactor is required, and an evaluation method by predicting vortices from flow velocity distribution obtained by CFD analysis is developed. In this study, Adaptive Mesh Refinement (AMR) method is examined to improve efficiency of CFD analysis. Initial mesh was refined with two indexes: the first index (Index-1) is when the second invariant of velocity gradient tensor, Q, is negative and the second one (Index-2) is pressure gradient index added to Index-1. As a result of applying AMR method to unsteady vortices system with a flat plate and performing transient analyses with refined meshes, the result of pressure distribution and velocity around the flat plate in mesh using Index-2 was similar to the result of all refined mesh. It was also confirmed that vortices generation and growth was better simulated by refining meshes around separation area.

Study on the predictive evaluation method of nonlinear sloshing wave height and load of cylindrical tanks, 9; Setting method of sinusoidal wave excitation that generates sloshing loads equivalent to seismic wave excitation

Ikesue, Shunichi*; Yamamoto, Tomohiko; Yokoi, Shinobu*; Sago, Hiromi*; Iwasaki, Akihisa*

Nihon Kikai Gakkai 2025-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2025/09

The design basis ground motions have been revised to improve the seismic resistance of nuclear power plants. The reduction of seismic forces not only horizontally but also vertically has required more critical than in the past to ensure the seismic resistance of components. Notably, the design of a Sodium-Cooled Fast Reactor will require reducing the seismic forces applied to the components because of the components with thin wall thickness. To overcome this problem, the authors plan to introduce a seismic isolation system. However, the natural frequency of first order sloshing may be close to the response frequency of the Sodium-Cooled Fast Reactor with the seismic isolation system, and the sloshing wave height and the vertical load is expected to increase. The authors studied the setting method of sinusoidal wave excitation that generates sloshing loads equivalent to seismic wave excitation. This paper reports on an overview of this method.

Application of the GIF safety design criteria and safety design guidelines on passive reactor shutdown capability to next generation sodium-cooled fast reactor in Japan

Yamano, Hidemasa; Futagami, Satoshi; Sasa, Kyohei*; Nakamura, Hironori*; Tokizaki, Minako*; Kubota, Ryuzaburo*

Proceedings of 2025 International Congress on Advances in Nuclear Power Plants (ICAPP 2025) (Internet), 12 Pages, 2025/09

This study examined the application of safety design criteria (SDC) and safety design guideline (SDG) developed in the Generation-IV international forum on the passive reactor shutdown capability to sodium-cooled fast reactors (SFRs) recently designed in Japan.

Impact of fast reactor fuel type on backend processes in the nuclear fuel cycle

Takeshita, Kenji*; Okamura, Tomohiro*; Nakase, Masahiko*; Abe, Takumi; Nishihara, Kenji

Progress in Nuclear Science and Technology (Internet), 8, p.52 - 57, 2025/09

https://doi.org/10.15669/pnst.8.52

Development of a new materials surveillance technology for fast reactors

Okajima, Satoshi; Ando, Masanori; Toyota, Kodai; Ishigami, Katsuo*; Onuma, Terumitsu*; Takahashi, Ryoya*; Asayama, Tai; Wakai, Takashi

Proceedings of the ASME 2025 Pressure Vessels & Piping Conference (PVP2025) (Internet), 10 Pages, 2025/07

https://doi.org/10.1115/PVP2025-153438

This paper describes current efforts within the JAEA to develop a new materials surveillance technology for SFRs that aims at raising the confidence levels in operating reactors' structural integrity to the ones which could otherwise be achieved only by a vast amount of long-term material test data that needs decades to accumulate. The technology would allow for monitoring on-going materials degradation processes in an actual reactor, namely fatigue and creep-fatigue, by inserting a newly developed specimen for this purpose into the reactor and establishing procedures to observe and evaluate the specimen to get desired information such as the extent of degradation and residual life. To establish the test procedure, a passive creep-fatigue test model was developed in the framework of CNWG, a collaboration agreement between the US and Japan, and a demonstration test was performed under cyclic thermal loads using an electric furnace. As a result, the test specimen failed by several hundred thermal cycles. Macroscopic inspection of the specimen after the test showed the test had been performed successfully without buckling. Fracture surface observations suggested that the failure caused probably by fatigue or creep-fatigue.

Investigation on multi-dimensional short-term behaviour through benchmark analysis of a large-volume sodium combustion experiment

Sonehara, Masateru; Okano, Yasushi; Uchibori, Akihiro; Oki, Hiroshi*

Journal of Nuclear Science and Technology, 62(5), p.403 - 414, 2025/05

https://doi.org/10.1080/00223131.2024.2441427

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.

Difference in accumulation of plutonium and curium isotopes formed in americium targets irradiated in Joyo and JMTR

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

https://doi.org/10.1016/j.nucengdes.2024.113755