Numerical analysis of a potential Reactor Pressure Vessel (RPV) boundary failure mechanism in Fukushima Daiichi Nuclear Power Station Unit-2

Li, X.; Yamaji, Akifumi*; Sato, Ikken*; Yamashita, Takuya

Annals of Nuclear Energy, 214, p.111217_1 - 111217_13, 2025/05

https://doi.org/10.1016/j.anucene.2025.111217

Calculation method improvement of the original one-dimensional code GENGTC

Fuyushima, Takumi; Sayato, Natsuki; Otsuka, Kaoru; Endo, Yasuichi; Tobita, Masahiro*; Takemoto, Noriyuki

JAEA-Testing 2024-008, 38 Pages, 2025/03

JAEA-Testing-2024-008.pdf:2.37MB

In Japan Materials Testing Reactor (JMTR), irradiation tests had been conducted by loading specimens into capsules for irradiating fuels and materials. The thermal design calculation of capsules is significant to irradiate various types of specimens at the target temperature. The decommissioning plan of JMTR was approved in March 2021, and the Department of Waste Management and Decommissioning Technology Development is currently working on irradiation plans by foreign testing reactors as an alternative for JMTR. A one-dimensional thermal calculation code "GENGTC", which was developed at the Oak Ridge National Laboratory in U.S., is used for capsule design and irradiation tests. GENGTC has been repeatedly improved as improvements of computer performance, but there were some defects in calculation function. Therefore, we investigated the cause of the problem and changed the program from the currently used FORTRAN77 language program to a Visual Basic language program that uses the macro calculation function of Excel. In addition, the program was improved to make it easier to use the calculation code.

Assessment report of research and development activities in FY 2023; Activity "Implementation and technological development of decommissioning of nuclear facilities" and "Implementation and technological development of radioactive waste processing" (In-advance evaluation)

Nuclear Backend Technology Development

JAEA-Evaluation 2024-002, 86 Pages, 2025/03

JAEA-Evaluation-2024-002.pdf:8.29MB

Japan Atomic Energy Agency (hereinafter referred to as "JAEA") consulted the "Evaluation Committee on Research and Development Activities for Decommissioning" (hereinafter referred to as "Committee") about the in-advance evaluation of research themes which had started after FY2021 of R&D plan for the 4th medium- and long-term objectives period concerning "Decommissioning of nuclear facilities and development of technologies for the treatment of related radioactive waste" project in accordance with the "Guideline for evaluation of government R&D activities". In response, the Committee evaluated to the evaluation points of view made by JAEA.

Measurements of neutron capture cross-section for nuclides of interest in decommissioning (II); Fe(n,)Fe

Nakamura, Shoji; Shibahara, Yuji*; Endo, Shunsuke; Rovira Leveroni, G.; Kimura, Atsushi

Journal of Nuclear Science and Technology, 62(3), p.300 - 307, 2025/03

https://doi.org/10.1080/00223131.2024.2427244

Verification of direct coupling code system using FRENDY version 2 and GENESIS for light water reactor lattices

Fujita, Tatsuya; Yamamoto, Akio*

Journal of Nuclear Science and Technology, 62(2), p.179 - 196, 2025/02

https://doi.org/10.1080/00223131.2024.2410918

This study newly established a direct coupling code system consisting of the nuclear data processing code FRENDY version 2, and the three-dimensional heterogeneous transport code GENESIS (FRENDY-V2/GENESIS) for easy implementation of the random-sampling-based uncertainty quantification considering the implicit effect due to nuclear cross-section (XS) perturbations. The multi-group macroscopic XSs prepared for GENESIS were generated by FRENDY version 2, where the Dancoff factor was calculated by the neutron current method. Then the background XSs were evaluated based on the Carlvik two-term rational approximation. The infinite multiplication factor (k-infinity) and the fission reaction rate distribution in UO and MOX lattice geometries were compared with MVP3 to verify the calculation accuracy of FRENDY-V2/GENESIS. The sensitivity analyses on the discretization conditions such as the ray tracing of the method of characteristics were also carried out. Through several comparisons between FRENDY-V2/GENESIS and MVP3, FRENDY-V2/GENESIS with the SHEM 361-group structure calculates the k-infinity within approximately 50 pcm and the fission reaction rate distribution within approximately 0.1% by the root mean square, respectively. Consequently, the applicability of FRENDY-V2/GENESIS was verified, and FRENDY-V2/GENESIS can be used to discuss the implicit effect due to multi-group XS perturbations.

Measurements of neutron capture cross-section for nuclides of interest in decommissioning (III); Er(n,)Er and Hf(n,)Hf reactions

Nakamura, Shoji; Shibahara, Yuji*; Endo, Shunsuke; Rovira Leveroni, G.; Kimura, Atsushi

Journal of Nuclear Science and Technology, 14 Pages, 2025/00

https://doi.org/10.1080/00223131.2025.2457589

New insight on the thermal impact on cementitious materials due to high-temperature with water supply; Continuous expansive spalling in water

Miura, Taito*; Miyamoto, Shintoro*; Maruyama, Ippei*; Aili, A.*; Sato, Takumi; Nagae, Yuji; Igarashi, Go*

Case Studies in Construction Materials, 21, p.e03571_1 - e03571_14, 2024/12

https://doi.org/10.1016/j.cscm.2024.e03571

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, 12 Pages, 2024/12

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.

In-situ measurement of radiation driven back-conversion from para to ortho liquid hydrogen state in cold moderators at J-PARC

Teshigawara, Makoto; Lee, Y.*; Tatsumoto, Hideki*; Hartl, M.*; Aso, Tomokazu; Iverson, E. B.*; Ariyoshi, Gen; Ikeda, Yujiro*; Hasegawa, Takumi*

Nuclear Instruments and Methods in Physics Research B, 557, p.165534_1 - 165534_10, 2024/12

https://doi.org/10.1016/j.nimb.2024.165534

At Japanese Spallation Neutron Source in J-PARC, the para-hydrogen fraction was measured by using Raman spectroscopy in-situ for an integrated beam power of 9.4 MWh at 1 MW operation, to evaluate the functionality of the ferric oxyhydroxide catalyst. This result showed that full functionality of the catalyst was retained up to the 1 MW operation. We attempted to study the effect of neutron scattering driven para to ortho-hydrogen back-conversion rate in the absence of the catalyst effect with a bypass line without catalyst. The measured increase of ortho-hydrogen fraction was 0.44% for an integrated beam power of 2.4 MWh at 500 kW operation, however, which was considered to be due to not only to neutron collisions in cold moderators but also to the high ortho-hydrogen fraction of initially static liquid hydrogen in the bypass line and passive exudation of quasi-static hydrogen in the catalyst vessel to the main loop.

Development of a nondestructive assay technique using a laser-driven neutron source

Koizumi, Mitsuo; Ito, Fumiaki*; Lee, J.; Hironaka, Kota; Takahashi, Tone; Suzuki, Satoshi*; Arikawa, Yasunobu*; Abe, Yuki*; Wei, T.*; Yogo, Akifumi*; et al.

Dai-45-Kai Nihon Kaku Busshitsu Kanri Gakkai Nenji Taikai Kaigi Rombunshu (Internet), 4 Pages, 2024/11