Analytical study on stress behavior of core graphite components using simplified viscoelastic evaluation model

Saijo, Tomoaki; Shimazaki, Yosuke; Ishihara, Masahiro

JAEA-Technology 2025-010, 126 Pages, 2025/12

JAEA-Technology-2025-010.pdf:12.52MB

During the operation of the High Temperature Engineering Test Reactor (HTTR), thermal stress is generated in the graphite components. In addition, graphite exhibits dimensional shrinkage and creep deformation under neutron irradiation. As a result, residual stress remains in the graphite components during reactor shutdown. Therefore, in the design of the HTTR core graphite structures, stress analyses of the graphite components have previously been performed using the finite element analysis code VIENUS. In the HTTR, the graphite components are exposed to a wide range of temperature, from approximately 400C to 1200C, depending on their location. Consequently, irradiation-induced behaviors such as material property changes and irradiation shrinkage vary among the graphite components. On the other hand, since VIENUS code evaluates stress based on thermal fluid and heat conduction analysis results, it is not suitable for parametric studies. In this study, the influence of irradiation behavior on the stress behavior of graphite components in the wide temperature range (400C to 1200C) was analyzed using simplified viscoelastic evaluation model, consisting of two beam elements, to conduct efficient parametric studies. Operational stress exhibits two distinct patterns depending on whether the irradiation temperature is below or above 800C, due to irradiation shrinkage. Residual stress approaches the thermal stress, preventing excessive increase even when irradiation shrinkage is large. Moreover good agreement in stress behavior trends was observed between the stress analysis results by the simplified viscoelastic evaluation model and VIENUS code. These results indicate that the simplified viscoelastic evaluation model is beneficial in simulating stress behavior.

Fukushima Daiichi Nuclear Power Plant Unit 2 Accident analysis considering the thermal stratification and containment leakage

Nakamura, Yuki*; Kojima, Yoshihiro*; Yamashita, Takuya; Shimomura, Kenta; Mizokami, Shinya

Journal of Nuclear Science and Technology, 62(12), p.1226 - 1230, 2025/12

https://doi.org/10.1080/00223131.2025.2570371

Re-evaluation of nuclear criticality characteristics for infinite and finite heterogeneous lattice systems composed of uranium-zirconium hydride fuel rods used in the TRIGA annular core pulse reactor NSRR

Yanagisawa, Hiroshi; Motome, Yuiko

JAEA-Research 2025-010, 197 Pages, 2025/11

JAEA-Research-2025-010.pdf:3.5MB

For understandings of nuclear criticality risks of TRIGA fuel rods and review of safety measures for handling them, nuclear criticality characteristics for infinite and finite heterogeneous lattice systems composed of the NSRR fuel rods were re-evaluated with the use of a detailed computational model for the fuel rod. The MVP version 3 code was used with the JENDL libraries including the latest version, JENDL-5, for the re-evaluation. As the criticality characteristics, variations of neutron multiplication factors of the infinite and water-reflected finite systems were examined in detail with parameters of the lattice pitch and density of moderator water. From the results of the re-evaluated criticality characteristics, the minimum critical number of fuel rods for the water-reflected hexagonal shaped lattice system was obtained to be 46.8 0.2 using the JENDL-5 library. Moreover, the attainability of criticality without the water as moderator and reflector was examined because the zirconium hydride moderator and graphite reflector are equipped with the TRIGA fuel rod. It was found that the criticality is possible to be attained by 115.7 0.6 of the number of fuel rods, which is the smaller number of fuel rods than loaded in the NSRR standard core, even though no water exists.

Corrosion behavior of extra-high-purity Type 316 austenitic stainless steel in a liquid lead-bismuth eutectic with oxygen saturation or low oxygen concentrations

Irisawa, Eriko; Kato, Chiaki

Corrosion Science, 256, p.113173_1 - 113173_16, 2025/11

https://doi.org/10.1016/j.corsci.2025.113173

Uncertainty reduction of the FPs transport mechanism and debris degradation behavior and evaluation of the reactor contamination of debris state on the basis of the accident progression scenario of Fukushima Daiichi Nuclear Power Station Unit 2 and 3 (Contract research); FY2023 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; Institute of Science Tokyo*

JAEA-Review 2025-012, 96 Pages, 2025/10

JAEA-Review-2025-012.pdf:3.99MB

JAEA/CLADS had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. (TEPCO). Among the adopted proposals in FY2021, this report summarizes the research results of the "Uncertainty reduction of the FPs transport mechanism and debris degradation behavior and evaluation of the reactor contamination of debris state on the basis of the accident progression scenario of Fukushima Daiichi Nuclear Power Station unit 2 and 3" conducted from FY2021 to FY2023. The present study aims to elucidate the cause of the high dosage under shield plug by clarification of to the cesium behavior of migration, adhesion to structure and deposition as well as evaluate the properties of metal-rich debris predeceasing melted through the materials science approach based on the most probable scenario of accident progression of Unit 2 and 3. Through three years investigation, chemi-absorption configuration of Cs has been elucidated to change with acidity of steel surface during oxidation in humid atmosphere and CsO trapping compound as well as penetration depth have found to be importantly considered. For metallic debris, solid oxidation was found to be controlled by FeO formation and molten state was found to tend to preferentially yield ZrO resulting in formation of slate structure during solidification. Present findings obtained are contributing to better improve the accuracy of accident progression scenario in FDNPP in viewpoint of backward analysis.

Neutron emission during fission and its impact on fission-fragment mass distribution studied by a Langevin model

Takagi, Shinya*; Harada, Sota*; Aritomo, Yoshihiro*; Hirose, Kentaro; Nishio, Katsuhisa

Physical Review C, 112(4), p.044607_1 - 044607_9, 2025/10

https://doi.org/10.1103/ml4c-3k8t

Research on radioactive aerosol control and decontamination at Fukushima Daiichi Nuclear Power Station Decommissioning (Contract research); FY2023 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; The University of Tokyo*

JAEA-Review 2025-015, 73 Pages, 2025/09

JAEA-Review-2025-015.pdf:5.9MB

The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2023. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2021, this report summarizes the research results of the "Research on radioactive aerosol control and decontamination at Fukushima Daiichi Nuclear Power Station decommissioning" conducted from FY2021 to FY2023. The present study aims to develop a safe laser decontamination system that integrates advanced particle detection and characterization systems with aerosol dispersion control, in collaboration with UK researchers. For aerosol dispersion control, new containment methods ranging from simple mechanical containment hoods to optical laser shields are jointly investigated. The Japanese team will develop a radioactive dispersion control method based on the use of water mist and water spray to reduce radiation risks during laser cutting and decontamination applications. Based on data on aerosol particles provided by the UK team, the potential to enhance aerosol scrubbing efficiency will be explored. Additionally, the effect of charge enhancement will be evaluated. The UK team will develop laser-based aerosol containment methods and conducting experiments to improve the condensation of aerosol particles and mist in spray scrubbing. The development of aerosol removal technologies and strategies will be carried out through comprehensive experimental and computational studies. Experiments will be conducted at UTARTS facility to verify simultaneous operations such as laser decontamination, cutting, and spray operations. Aerosol measurements will also be performed to obtain high spatial resolution data that can better validate CFD models. In the final year, mock-up tests will be conducted to verify the system's operation and decontamination performance at full scale, evaluating the decontamination system developed through this research.

Assessment report on research and development activities in FY2024 Activity; Research and development related to the accident at TEPCO's Fukushima Daiichi Nuclear Power Station (Achievements in environmental behavior research) (Ex-post evaluations)

Fukushima Research and Engineering Institute*

JAEA-Evaluation 2025-001, 23 Pages, 2025/09

JAEA-Evaluation-2025-001.pdf:0.72MB
JAEA-Evaluation-2025-001-appendix(CD-ROM).zip:185.38MB

Japan Atomic Energy Agency (hereinafter referred to as "JAEA") consulted an assessment committee, "Evaluation Committee of Research and Development related to the Accident at TEPCO's Fukushima Daiichi Nuclear Power Station" (hereinafter referred to as "Committee") for ex-ante evaluations of "Research and Development Related to the Accident at TEPCO's Fukushima Daiichi Nuclear Power Station" in accordance with "General Guideline for the Evaluation of Government Research and Development (R&D) Activities" by Cabinet Office, Government of Japan, "Guideline for Evaluation of R&D in Ministry of Education, Culture, Sports, Science and Technology" and "Regulation on Conduct for Evaluation of R&D Activities" by JAEA. Based on a request from JAEA, the committee conducted a post-evaluation of research and development activities in the fourth phase (April 2022 to March 2029) of the medium-term research plan. This report summarizes the results of the assessment by the Committee.

Applicability of equivalent linear three-dimensional FEM analysis of reactor buildings to the seismic response of a soil-structure interaction system

Ichihara, Yoshitaka*; Nakamura, Naohiro*; Nabeshima, Kunihiko*; Choi, B.; Nishida, Akemi

Nuclear Engineering and Design, 441, p.114160_1 - 114160_10, 2025/09

https://doi.org/10.1016/j.nucengdes.2025.114160

This paper evaluates the applicability of equivalent linear analysis of reinforced concrete model, which uses frequency-independent complex damping with a small computational load, to the seismic design of nuclear power plant reactor buildings. To this end, a three-dimensional finite element method analysis of the soil-structure interaction focusing on nonlinear and equivalent linear seismic behavior of the building embedded in an ideally uniform soil condition was performed for the Kashiwazaki-Kariwa Nuclear Power Plant Unit 7 reactor building. The equivalent linear analysis results correlated well with the nonlinear analysis results of the shear strain, acceleration, displacement, and acceleration response spectrum, demonstrating the effectiveness of the equivalent linear analysis method. Moreover, the equivalent linear analysis results were more conservative than those of nonlinear analysis using the material constitutive law in evaluating the shear strain of the external wall of the reactor building. From this result, equivalent linear analysis method tended to obtain a lower building stiffness than nonlinear analysis under the analysis conditions used in this paper.

Three-dimensional numerical simulation of shaft wall collapse and lining crack initiation in deep shafts

Maeda, Soshu*; Nago, Makito*; Aoyagi, Kazuhei; Sugawara, Kentaro*; Kodama, Junichi*

Shigen, Sozai Koenshu (Internet), 12(2), 5 Pages, 2025/09

no abstracts in English