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Hirouchi, Jun; Watanabe, Masatoshi*; Hayashi, Naho; Nagakubo, Azusa; Takahara, Shogo
Journal of Radiological Protection, 45(1), p.011506_1 - 011506_11, 2025/03
Times Cited Count:0 Percentile:0.00(Environmental Sciences)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.
Shigyo, Nobuhiro*; Furuta, Takuya; Iwamoto, Yosuke
JAEA-Conf 2024-002, 216 Pages, 2024/11
The 2023 Symposium on Nuclear Data was held at Tokai Industry and Information Plaza "iVil" on November 15-17, 2023. The symposium was organized by the Nuclear Data Division of the Atomic Energy Society of Japan (AESJ) in cooperation with Radiation Engineering Division of AESJ, North Kanto Branch of AESJ, Investigation Committee on Nuclear Data in AESJ, Nuclear Science and Engineering Center of Japan Atomic Energy Agency, and High Energy Accelerator Research Organization. In the symposium, tutorials "Overview of the nuclear data processing code, FRENDY version 2" was proposed and held. Two sessions of lectures and discussions were held: "Recent Topics on Nuclear Data and Particle and Heavy Ion Transport code System (PHITS)". In addition, recent research progress on experiments, nuclear theory, evaluation, benchmark, and applications were presented in the poster session. The total number of participants was 108 participants. Each oral and poster presentation was followed by an active question and answer session. This report consists of a total of 36 papers including 17 oral and 19 poster presentations.
Futagami, Satoshi
Nihon Genshiryoku Gakkai-Shi ATOMO, 66(11), p.555 - 559, 2024/11
no abstracts in English
Kowatari, Munehiko*; Yoshitomi, Hiroshi; Tani, Kotaro*; Tanimura, Yoshihiko; Kurihara, Osamu*
Radiation Protection Dosimetry, 200(16-18), p.1574 - 1579, 2024/11
Times Cited Count:0 Percentile:0.00(Environmental Sciences)Aoki, Kazuhiro; Imai, Hirotaro; Seshimo, Kazuyoshi; Kimura, Megumi; Kirita, Fumio; Nakanishi, Ryuji
JAEA-Research 2024-005, 177 Pages, 2024/10
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.
Sato, Kaoru; Furuta, Takuya; Satoh, Daiki; Tsuda, Shuichi
PLOS ONE (Internet), 19(10), p.e0309753_1 - e0309753_26, 2024/10
Times Cited Count:0 Percentile:0.00(Multidisciplinary Sciences)The authors previously developed the adult male (JM-103) and female (JF-103) voxel phantoms with standard Japanese body sizes for dose assessment of radiation accidents and medical exposures. However, JM-103 and JF-103 were not applicable to dose assessment considering posture at the time of exposure due to limitations in description format and resolution. In this study, we developed the polygon mesh-type adult Japanese phantoms (male: JPM, female: JPF) based on JM-103 and JF-103. The detailed models of skin and lens with radiosensitive regions less than 1 mm thick were incorporated into JPM and JPF. The effective doses, and skin and lens (entire and radiosensitive regions) doses were calculated for external irradiation with photons or electrons in anterior-posterior geometry. It was confirmed that dose analysis results by JPM and JPF were consistent with the previous reports. In the future, we will develop a detailed dose assessment method for individuals, taking into account their postures at the time of exposure, by applying the posture deformation technique currently under development to the JPM and JPF.
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
Wang, Y. W.*; Xu, P. G.; Su, Y. H.; Ma, Y. L.*; Wang, H. H.*
Physics Examination and Testing, 42(4), p.32 - 41, 2024/08
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
Times Cited Count:2 Percentile:63.37(Nanoscience & Nanotechnology)Kikuchi, Hirohito*; Uda, Toshiaki*; Hayashi, Daisuke*; Emori, Minoru*; Kimura, Shun
Genshiryoku Bakkuendo Kenkyu (CD-ROM), 31(1), p.11 - 20, 2024/06
no abstracts in English
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.
Tada, Kenichi; Yamamoto, Akio*; Kunieda, Satoshi; Konno, Chikara; Kondo, Ryoichi; Endo, Tomohiro*; Chiba, Go*; Ono, Michitaka*; Tojo, Masayuki*
Journal of Nuclear Science and Technology, 61(6), p.830 - 839, 2024/06
Times Cited Count:8 Percentile:89.79(Nuclear Science & Technology)Nuclear data processing code is important to connect evaluated nuclear data libraries and radiation transport codes. The nuclear data processing code FRENDY version 1 was released in 2019 to generate ACE formatted cross section files with simple input data. After we released FRENDY version 1, many functions were developed, e.g., neutron multi-group cross section generation, explicit consideration of the resonance interference effect among different nuclides in a material, consideration of the resonance upscattering, ACE file perturbation, and modification of ENDF-6 formatted file. FRENDY version 2 was released including these new functions. It generates GENDF and MATXS formatted neutron multi-group cross section files from an ACE formatted cross section file or an evaluated nuclear data file. This paper explains the features of the new functions implemented in FRENDY version 2 and the verification of the neutron multigroup cross section generation function of this code.
Nakayama, Shinsuke
JAEA-Review 2024-009, 16 Pages, 2024/05
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.
Tanaka, Masaaki; Enuma, Yasuhiro; Okano, Yasushi; Uchibori, Akihiro; Yokoyama, Kenji; Seki, Akiyuki; Wakai, Takashi; Asayama, Tai
Mechanical Engineering Journal (Internet), 11(2), p.23-00424_1 - 23-00424_13, 2024/04
The outline and development status of element functions and design optimization process in ARKADIA to transform advanced nuclear reactor design to meet expectations of a safe, economic, and sustainable carbon-free energy source are introduced. It is also briefly explained that ARKADIA will realize Artificial Intelligence (AI)-aided integrated numerical analysis to offer the best possible solutions for the design and operation of a nuclear plant including optimization of safety equipment, and merge state-of-the-art numerical simulation technologies and a knowledge base that stores data and insights from past nuclear reactor development projects and R&Ds with AI technologies.
Yamaguchi, Masaaki; Suzuki, Yuji*; Kabasawa, Satsuki; Kato, Tomoko
JAEA-Data/Code 2024-001, 21 Pages, 2024/03
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.
Shimada, Kazumasa; Sakurahara, Tatsuya*; Farshadmanesh, P.*; Reihani, S.*; Mohagehgh, Z.*
Annals of Nuclear Energy, 197, p.110243_1 - 110243_12, 2024/03
Times Cited Count:1 Percentile:25.62(Nuclear Science & Technology)This research improves the realism of Level 3 probabilistic risk assessment (PRA) for nuclear power plants (NPP) to avoid subjective expert judgment when setting evacuation behavior for residents. Therefore, the evacuation speed output by the traffic simulation code MATSim was input to the level 3 PRA code MACCS. Furthermore, to set the priority of the places where road closure is to be considered, a method to evaluate the road closure risk due to the earthquake using the natural disaster risk assessment code HAZUS was developed. Then, the relationship between the evacuation routes and the radiation dose was evaluated for the case study of the Sequoyah NPP adopted in the SOARCA study conducted by the US NRC. As a result, the present study found an evacuation route with low closure risk but causing high radiation dose of residents when it is closed. This showed effectiveness of the proposed Level 3 PRA methodology for supporting decision-makers to enhance evacuation routes.
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
Times Cited Count:12 Percentile:96.41(Nuclear Science & Technology)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.
Tachi, Yukio
Kagaku To Kyoiku, 71(10), p.420 - 423, 2023/10
no abstracts in English
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.
Tanaka, Masaaki; Uchibori, Akihiro; Okano, Yasushi; Yokoyama, Kenji; Uwaba, Tomoyuki; Enuma, Yasuhiro; Wakai, Takashi; Asayama, Tai
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09
The book, JSME Series in Thermal and Nuclear Power Generation Vol.3 Sodium-cooled Fast Reactor, was published as a 30th anniversary memorial project of Power & Energy Systems Division. This paper describes an introduction of the book on a part of key technologies regarding safety assessment, thermal-hydraulics, neutronics, and fuel and material development. This introductory paper also provides an overview of an integrated evaluation system named ARKADIA to offer the best possible solutions for challenges arising during the design process, safety assessment, and operation of a nuclear plant over its life cycle, in active use of the R&D efforts and knowledges on thermal-hydraulics and safety assessment with state-of-the-art numerical analysis technologies.