Development of hydrogen behavior integrated analysis system

Terada, Atsuhiko; Thwe Thwe, A.; Hino, Ryutaro*; Harai, Yasutaka*; Sasaki, Gaku*; Shingeya, Hideshi*; Yamashita, Toshiyuki*; Yoneda, Jiro*; Okabayashi, Kazuki*; Sakamoto, Hiroyuki*; et al.

JAEA-Data/Code 2025-012, 151 Pages, 2025/12

JAEA-Data-Code-2025-012.pdf:9.69MB

Based on the lessons learned from the Fukushima Daiichi Nuclear Power Station accident, we have highly paid attention to the advancement of the fundamental technologies which are indispensable in timely response to hydrogen safety measures and assessments especially in both nuclear reactors and decommissioning. Focusing on this attention, we developed an analysis system that predicts the behavior of hydrogen from generation to diffusion, combustion and explosion. The system utilizes the commercial computational fluid dynamics software (FLUENT, AUTODYN), and incorporates new modules and pre/post-processors in order to withstand the general practical use. We also developed a system by utilizing open-source code (OpenFOAM) that can be used in hydrogen disaster prevention plans for nuclear facilities. So far, we have expanded the system to deal with the phenomena that should be considered from the practical point of view for PWR (Pressurized Water Reactor) in nuclear power plants. This report summarizes the overview of the integrated analysis system for hydrogen behavior, the handling method, and real scale analysis examples.

Development of a theoretical scaling factor method for the inventory estimation of difficult-to-measure nuclide Cs-135 in fuel debris and radioactive wastes

Sakamoto, Masahiro; Okumura, Keisuke; Kanno, Ikuo; Matsumura, Taichi; Terashima, Kenichi; Riyana, E. S.; Kaneko, Junichi*; Mizokami, Masato*; Mizokami, Shinya*

Journal of Nuclear Science and Technology, 62(8), p.756 - 765, 2025/08

https://doi.org/10.1080/00223131.2025.2478926

Decommissioning report for Wastewater Treatment Facility (Part 2); Chapter on contamination inspection section

Yamamoto, Keisuke; Nakagawa, Takuya; Shimojo, Hiroto; Kijima, Jun; Miura, Daiya; Onose, Yoshihiko*; Namba, Koji*; Uchida, Hiroaki*; Sakamoto, Kazuhiko*; Ono, Chika*; et al.

JAEA-Technology 2024-019, 211 Pages, 2025/02

JAEA-Technology-2024-019.pdf:35.35MB

The uranium enrichment facilities at the Nuclear Fuel Cycle Engineering Laboratories of Japan Atomic Energy Agency (JAEA) were constructed sequentially to develop uranium enrichment technology with centrifugal separation method. The developed technologies were transferred to Japan Nuclear Fuel Limited until 2001. And the original purpose has been achieved. Wastewater Treatment Facility, one of the uranium enrichment facilities, was constructed in 1976 to treat radioactive liquid waste generated at the facilities, and it finished the role in 2008. In accordance with the Medium/Long-Term Management Plan of JAEA Facilities, interior equipment installed in this facility had been dismantled and removed since November 2021 to August 2023. This report summarizes the findings obtained through the work related to the contamination inspection methods cancellation the controlled area of Wastewater Treatment Facility from September 2023 to March 2024.

Update of conceptual design of Near Surface Disposal Facilities of Radioactive Wastes Generated from Research, Industrial and Medical Facilities

Iwamura, Toko; Nakata, Hisakazu; Maekawa, Keisuke; Sakai, Akihiro; Sakamoto, Yoshiaki

JAEA-Review 2024-032, 39 Pages, 2024/08

JAEA-Review-2024-032.pdf:6.5MB

Japan Atomic Energy Agency (JAEA) is responsible for the disposal of low-level radioactive waste generated by JAEA itself and research facilities under the revised JAEA Act of 2008 and subsequently developed a "Plan for the Implementation of Disposal Operations" (implementation plan) in 2009. Furthermore, based on the results of the survey on the amount of waste generated by research facilities, the quantity of wastes for the near surface disposal was set at 600,000 in terms of 200L drums, and the results of the consideration on the conceptual design of the disposal facility were summarized in 2012. In 2018 JAEA published its long-term outlook and policy regarding back-end measures in "Back-end Roadmap", and in this "Back-end Roadmap", the amount of waste generated by JAEA was also organized and published. Therefore, the amount of waste materials from waste generators outside JAEA was re-examined, and as a result, the size of the burial facility was changed from 600,000 to 750,000 in terms of 200L drums, and approval was obtained for a change in the implementation plan. In addition, the conceptual design of the disposal facility was revised to accommodate the increased size of the facility. This report summarizes the results of the updated assumptions and disposal facility design from the 2012 conceptual design.

Practical development of accident tolerant FeCrAl-ODS fuel claddings for BWRs in Japan

Sakamoto, Kan*; Sakaguchi, Chisato*; Miura, Yusuke*; Yokoyama, Hironori*; Matsunaga, Junji*; Kasahara, Hideyuki*; Miyata, Hajime*; Ioka, Ikuo; Yamashita, Shinichiro; Osaka, Masahiko

Proceedings of 2023 Water Reactor Fuel Performance Meeting (WRFPM 2023), p.20 - 28, 2024/00

An oxide-dispersion-strengthened FeCrAl (FeCrAl-ODS) has been continuously developed in Japan as a promising candidate alloy for the accident tolerant fuel cladding of BWRs (boiling water reactors). This paper will introduce the progress in practical development of accident tolerant FeCrAl-ODS fuel claddings for BWRs in the program fully or partially supported and organized by the Ministry of Economy, Trade and Industry (METI) of Japan. The experimental studies have been conducted to obtain and accumulate key material properties of FeCrAl-ODS fuel claddings to support the evaluations in the analytical studies. For the evaluation at normal operation condition, fatigue test of unirradiated fuel cladding and tensile test of irradiated sheet specimen were conducted. In the fatigue test, a tensile-compressive bending strain was loaded on the C-shaped specimens by cyclic movement of a push-pull rod. Test temperature was 623 K, frequency was 1 Hz, and strain amplitude were 0.27, 0.34 and 0.55 %. The results of fatigue tests demonstrated that cycles to failure of the FeCrAl-ODS cladding were higher than that of the O'Donnell and Langer fatigue curve of Zr-based alloy. The tensile test was conducted in a hot cell using the SS-J2 type specimens at ambient temperature, 573 K and 623 K at a strain rate of 10-3 s-1. The specimens were irradiated up to 7.8 and 13 dpa at 573 K in the High Flux Isotope Reactor at ORNL. The irradiation hardening and ductility loss obtained at 7.8 and 13 dpa were comparable to those at 3.9 dpa.

Alloy design and characterization of a recrystallized FeCrAl-ODS cladding for accident-tolerant BWR fuels; An Overview of research activity in Japan

Ukai, Shigeharu; Sakamoto, Kan*; Otsuka, Satoshi; Yamashita, Shinichiro; Kimura, Akihiko*

Journal of Nuclear Materials, 583, p.154508_1 - 154508_24, 2023/09

https://doi.org/10.1016/j.jnucmat.2023.154508

Modelling heterogeneous hydration behaviour of bentonite by a FracMan-Thames coupling method for the Bentonite Rock Interaction Experiment (BRIE) at sp HRL

Sawada, Atsushi; Sakamoto, Kazuhiko*; Watahiki, Takanori*; Imai, Hisashi*

SKB P-17-06, 154 Pages, 2023/08

Cr doping-induced ferromagnetism in the spin-glass CdMnTe studied by X-ray magnetic circular dichroism

Verma, V.*; Sakamoto, Shoya*; Ishikawa, Koichiro*; Singh, V. R.*; Ishigami, Keisuke*; Shibata, Goro; Kadono, Toshiharu*; Koide, Tsuneharu*; Kuroda, Shinji*; Fujimori, Atsushi*

Physica B; Condensed Matter, 642, p.414129_1 - 414129_5, 2022/10

https://doi.org/10.1016/j.physb.2022.414129

Development of accident tolerant FeCrAl-ODS fuel cladding for BWRs in Japan

Sakamoto, Kan*; Miura, Yusuke*; Ukai, Shigeharu; Ono, Naoko*; Kimura, Akihiko*; Yamaji, Akifumi*; Kusagaya, Kazuyuki*; Takano, Sho*; Kondo, Takao*; Ikegawa, Tomohiko*; et al.

Journal of Nuclear Materials, 557, p.153276_1 - 153276_11, 2021/12

https://doi.org/10.1016/j.jnucmat.2021.153276

A FeCrAl-oxide dispersion strengthened (ODS) alloy is a promising candidate alloy for the accident tolerant fuel (ATF) cladding of light water reactors (LWRs) and being developed in Japan recently. This paper will introduce the progress of development of accident tolerant FeCrAl-ODS fuel claddings for boiling water reactors (BWRs) in Japan. Both the experimental and the analytical studies have been performed to evaluate the influence of implementation of the FeCrAl-ODS fuel claddings to the current BWRs. The experimental studies have been conducted to obtain and accumulate key material properties of FeCrAl-ODS fuel claddings by using bar, sheet and tube-shaped materials to support the evaluations in the analytical studies. At the end of paper, the challenges and prospects found in the program are highlighted to enhance international collaborations to accelerate the development of FeCrAl-ODS fuel cladding.

Study on the evaluation method to determine the radioactivity concentration in radioactive waste on Oarai Research and Development Institute (FY2020)

Asakura, Kazuki; Shimomura, Yusuke; Donomae, Yasushi; Abe, Kazuyuki; Kitamura, Ryoichi; Miyakoshi, Hiroyuki; Takamatsu, Misao; Sakamoto, Naoki; Isozaki, Ryosuke; Onishi, Takashi; et al.

JAEA-Review 2021-020, 42 Pages, 2021/10

JAEA-Review-2021-020.pdf:2.95MB

The disposal of radioactive waste from the research facility need to calculate from the radioactivity concentration that based on variously nuclear fuels and materials. In Japan Atomic Energy Agency Oarai Research and Development Institute, the study on considering disposal is being advanced among the facilities which generate radioactive waste as well as the facilities which process radioactive waste. This report summarizes a study result in FY2020 about the evaluation method to determine the radioactivity concentration in radioactive waste on Oarai Research and Development Institute.