Effects of different accident scenarios and sites on the reduction factor used for expressing sheltering effectiveness

Hirouchi, Jun; Watanabe, Masatoshi*; Hayashi, Naho; Nagakubo, Azusa; Takahara, Shogo

Journal of Radiological Protection, 45(1), p.011506_1 - 011506_11, 2025/03

https://doi.org/10.1088/1361-6498/adba6e

Public living in areas contaminated by nuclear accidents is exposed to radiation in the early phase and over the long term. Even under similar accident scenarios, radiation doses and sheltering effectiveness, which is one of the protective measures, depend on meteorological conditions and the surrounding environment. Radiation doses and sheltering effectiveness in the early phase of nuclear accidents are crucial information for the public as well as national and local governments planning a nuclear emergency preparedness. In this study, we assessed radiation doses and sheltering effectiveness at sites with nuclear facilities in Japan using the Off-Site Consequence Analysis code for Atmospheric Release accidents, which is one of the level-3 probabilistic risk assessment codes, for five accident scenarios: three scenarios from past severe accident studies, a scenario defined by the Nuclear Regulation Authority in Japan, and a scenario corresponding to the Fukushima-Daiichi Nuclear Power Station accident. The sheltering effectiveness differed by up to approximately 50% among the accident scenarios at the same sites and by approximately 20%50% among sites under the same accident scenario. Differences in the radionuclide composition among the accident scenarios and the differences in wind speeds among the sites primarily caused these differences in sheltering effectiveness.

Cutting-edge of application of AI technology to PRA, 1; Automatic fault tree creation

Futagami, Satoshi

Nihon Genshiryoku Gakkai-Shi ATOMO, 66(11), p.555 - 559, 2024/11

https://doi.org/10.3327/jaesjb.66.11_555

no abstracts in English

Evaluation study of fault activity on Shionohira and Kuruma Faults located at the border of Fukushima and Ibaraki Prefecture, Japan

Aoki, Kazuhiro; Imai, Hirotaro; Seshimo, Kazuyoshi; Kimura, Megumi; Kirita, Fumio; Nakanishi, Ryuji

JAEA-Research 2024-005, 177 Pages, 2024/10

JAEA-Research-2024-005.pdf:12.02MB

This study presents a method for evaluating displacements on active faults that lack clear markers of fault offset. The method uses geological surveys, core studies, and chemical analyses along with hydraulic and mechanical tests. We applied this method to three test sites along the Shionohira Fault (Shionohira and Betto sites) and the Kuruma Fault (Minakamikita site). Laboratory friction tests on the fault gouge using a variable-speed, rotating shear friction apparatus were conducted. The samples from the Shionohira and Betto sites showed velocity weakening or strengthening, but no velocity dependence was observed at the Minakamikita site. A small-scale test to induce fault slip was conducted using the SIMFIP method. At the Shionohira site, fault slip can be modeled as a Coulomb rupture and shows a frictional dependence on slip velocity. On the other hand, at the Minakamikita site, a complex response using multiple fractures and slip planes was observed. Based on the water pressure response, the hydraulic properties of the area between the faults were evaluated. The transmissivity and specific storage are larger at Shionohira than at Minakamikita. Fault slip data such as shear plane attitude or shear sense were obtained from core samples and stress inversion analysis was performed. We attempted to elucidate the history of the movement and stress that formed the fracture zone. The results reconstructed five activity stages at Shionohira site and two stages at Minakamikita site. As shown in this report, the frictional properties, fault rupture mode, hydraulic properties and the history of fault motion were found to be different between the Shionohira and Kuruma sites. However, the results are based on a few locational data, so case studies at other sites and more applications to other faults should be considered to improve the reliability of the evaluation.

Development of probabilistic risk assessment methodology using artificial intelligence technology, 3; Automatic fault tree creation tools for failure mode level fault tree

Futagami, Satoshi; Kondo, Yuki; Yamano, Hidemasa; Kurisaka, Kenichi

Proceedings of Probabilistic Safety Assessment and Management & Asian Symposium on Risk Assessment and Management (PSAM17 & ASRAM2024) (Internet), 9 Pages, 2024/10

Influence of interstitial carbon on bulk texture evolution of carbide-free high-entropy alloys during cold rolling using neutron diffraction

Fang, W.*; Liu, C.*; Zhang, J.*; Xu, P. G.; Peng, T.*; Liu, B.*; Morooka, Satoshi; Yin, F.*

Scripta Materialia, 249, p.116046_1 - 116046_6, 2024/08

https://doi.org/10.1016/j.scriptamat.2024.116046

Development of new simulation method; Design support of next generation advanced reactors

Uchibori, Akihiro; Okano, Yasushi

Isotope News, (793), p.32 - 35, 2024/06

The design of a containment vessel in a sodium-cooled fast reactor was optimized from simulation on the hypothetical severe accident including sodium leakage and combustion. The simulation method is one of the base technologies of the design optimization system, ARKADIA. The simulation was performed on the different design conditions including volume of the containment vessel and the safety equipment as optimization parameters. The iterative simulation successfully found that the safety under this accident was kept even in the downsized containment vessel by selecting an effective safety equipment. This study demonstrated that the developed method has basic capability for design optimization in ARKADIA.

Report of nuclear data roadmap 2023

Nakayama, Shinsuke

JAEA-Review 2024-009, 16 Pages, 2024/05

JAEA-Review-2024-009.pdf:1.17MB

Nuclear data is fundamental data for nuclear energy research and development, and its importance has been widely recognized. On the other hand, for future nuclear data research, it is necessary to sort out which types of data (target nuclides, energies, physical quantities, etc.) should be prioritized. Therefore, a "Task Force (TF) for Nuclear Data Roadmap" was established within Investigation Committee on Nuclear Data in the Atomic Energy Society of Japan to discuss a roadmap for future nuclear data research and development. This document reports the results of the discussion in the TF.

Model catchments for testing biosphere assessment models of geological disposal

Yamaguchi, Masaaki; Suzuki, Yuji*; Kabasawa, Satsuki; Kato, Tomoko

JAEA-Data/Code 2024-001, 21 Pages, 2024/03

JAEA-Data-Code-2024-001.pdf:3.45MB
JAEA-Data-Code-2024-001-appendix(CD-ROM).zip:8.0MB

Model catchments have developed for use in testing various assessment models that can consider specific surface environmental conditions such as topography, riverine systems, and land use in the biosphere assessment of HLW geological disposal. The model catchments consist of the topography and riverine system of the catchment area created using existing tools, as well as land use and population distribution, river discharge, sediment flux data set by algorithms from topographical data. Datasets of three types of model watersheds (Types 1 to 3, watershed area: 730 to 770 km) with different topographical characteristics have released as raster data that can be handled by geographic information systems (GIS). Since the model catchments were created virtually reflecting as much as possible the main characteristics of Japan's surface environment, they can be used as a test bed for conducting hydraulic/mass transport analysis to set the GBI and compartment model.

JENDL-5 benchmarking for fission reactor applications

Tada, Kenichi; Nagaya, Yasunobu; Taninaka, Hiroshi; Yokoyama, Kenji; Okita, Shoichiro; Oizumi, Akito; Fukushima, Masahiro; Nakayama, Shinsuke

Journal of Nuclear Science and Technology, 61(1), p.2 - 22, 2024/01

https://doi.org/10.1080/00223131.2023.2197439

The new version of the Japanese evaluated nuclear data library, JENDL-5, was released in December 2021. This paper demonstrates the validation of JENDL-5 for fission reactor applications. Benchmark calculations are performed with the continuous-energy Monte Carlo codes MVP and MCNP and the deterministic code system MARBLE. The benchmark calculation results indicate that the performance of JENDL-5 for fission reactor applications is better than that of the former library JENDL-4.0.

Development of probabilistic risk assessment methodology using artificial intelligence technology, 1; Automatic fault tree creation

Futagami, Satoshi; Yamano, Hidemasa; Kurisaka, Kenichi; Ujita, Hiroshi*

Proceedings of PSAM 2023 Topical Conference AI & Risk Analysis for Probabilistic Safety/Security Assessment & Management, 8 Pages, 2023/10

To create an innovation for efficient and effective social implementation of nuclear power plant PRA, automatic construction tool for fault tree architecture and automatic failure judgment tool to construct reliability database are developed by using AI and digitization technology. This paper describes overall development plan of PRA methodology using the AI technology and the progress of automatic FT creation tools development.