OSCAAR version 2.0 User's Manual

Risk Analysis Research Group, Nuclear Safety Research Center

JAEA-Testing 2025-007, 110 Pages, 2026/03

JAEA-Testing-2025-007.pdf:2.57MB

The Japan Atomic Energy Agency's Nuclear Safety Research Center is developing the Level 3 PRA code OSCAAR as part of its research on probabilistic risk assessment (PRA) for nuclear power plant accidents. OSCAAR is a computational code that evaluates the advection, diffusion, and deposition of radioactive materials released into the environment under various meteorological conditions, based on source terms obtained from Level 2 PRA. It can probabilistically assess the radiation doses and health effects to the public caused by these radioactive materials. OSCAAR can account for the dose reduction effects of protective measures implemented during an actual nuclear power plant accident, thereby contributing to the pre-planning of countermeasures and plans to reduce the exposure of residents near nuclear power plants during an accident. This report is a manual for users to create input files and execute the OSCAAR program.

Engineering decision making and seismic PRA

Takada, Tsuyoshi

Hozengaku, 24(4), p.24 - 27, 2026/01

This article discusses how engineering decision-making should be approached, assuming the use of seismic PRA based on risk theory, in response to the case of non-approval caused by the regulation of Tsuruga Unit 2. It also attempts to identify future challenges regarding the approach to regulatory review and the conception of review standards.

Report of summer holiday practical training on 2024

Hasegawa, Toshinari; Nagasumi, Satoru; Ishitsuka, Etsuo; Egashira, Keiichiro*; Furuya, Aoi*; Ando, Ryota*; Sakaguchi, Akira*; Sakurai, Yosuke; Nakano, Yumi*; Iigaki, Kazuhiko

JAEA-Technology 2025-004, 20 Pages, 2025/07

JAEA-Technology-2025-004.pdf:1.67MB

Four people from three universities participated in the 2024 summer holiday practical training with the theme of "Technical development on HTTR". The participants practiced the analysis of the HTTR core, the analysis of Cs deposition behavior in the primary cooling system, and the feasibility study of nuclear rockets using HTGR. In the questionnaire after this training, there were comments from participants that it was beneficial as a work experience and that it was meaningful because of many opportunities to communicate with staff. These impressions suggest that this training was generally evaluated as good.

Evaluation of exposure doses and reduction factor for sheltering for each nuclear site under each accident scenario (Contract research)

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

JAEA-Research 2024-015, 114 Pages, 2025/03

JAEA-Research-2024-015.pdf:10.03MB

The public living in areas contaminated by nuclear accidents is exposed to radiation in the early phase and over the long term. Even under the same accident scenario, the exposure doses and the effectiveness of sheltering, which is one of the protective measures, vary depending on the meteorological condition and the surrounding environment. The exposure doses and sheltering effectiveness in the early phase are important information for the public and the national and local governments planning a nuclear emergency preparedness. In this report, we evaluate the exposure doses and sheltering effectiveness at sites with nuclear facilities in Japan using OSCAAR, one of the probabilistic risk assessment codes, for five accident scenarios: three scenarios from past severe accident studies; a scenario defined by the Nuclear Regulatory Authority; and a scenario assuming the Fukushima Daiichi Nuclear Power Station accident. The sheltering effectiveness differed by approximately 20% among the sites. This was due to the differences in wind speed among the sites.

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

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

Report of summer holiday practical training on 2023

Ishitsuka, Etsuo; Nagasumi, Satoru; Hasegawa, Toshinari; Kawai, Hiromi*; Wakisaka, Shinji*; Nagase, Sota*; Nakamura, Kento*; Yaguchi, Hiroki*; Ishii, Toshiaki; Nakano, Yumi*; et al.

JAEA-Technology 2024-008, 23 Pages, 2024/07

JAEA-Technology-2024-008.pdf:1.69MB

Five people from three universities participated in the 2023 summer holiday practical training with the theme of "Technical development on HTTR". The participants practiced the analysis of HTTR core, the analysis of behavior on loss of forced cooling test, the analysis of Iodine deposition behavior in primary cooling system and the feasibility study of energy storage system for HTGRs. In the questionnaire after this training, there were impressions such as that it was useful as a work experience and some students found it useful for their own research. These impressions suggest that this training was generally evaluated as good.

Simulation-based dynamic probabilistic risk assessment of an internal flooding-initiated accident in nuclear power plant using THALES2 and RAPID

Kubo, Kotaro; Zheng, X.; Tanaka, Yoichi; Tamaki, Hitoshi; Sugiyama, Tomoyuki; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*

Proceedings of the Institution of Mechanical Engineers, Part O; Journal of Risk and Reliability, 237(5), p.947 - 957, 2023/10

https://doi.org/10.1177/1748006X221091604

Probabilistic risk assessment (PRA) is a method used to assess the risks associated with large and complex systems. However, the timing at which nuclear power plant structures, systems, and components are damaged is difficult to estimate if the risk of an external event is evaluated using conventional PRA based on event trees and fault trees. A methodology coupling thermal-hydraulic analysis with external event simulations using Risk Assessment with Plant Interactive Dynamics (RAPID) is therefore proposed to overcome this limitation. A flood propagation model based on Bernoulli's theorem was applied to represent internal flooding in the turbine building of the pressurized water reactor. Uncertainties were also taken into account, including the flow rate of the floodwater source and the failure criteria for the mitigation systems. The simulated recovery actions included the operator isolating the floodwater source and using a drainage pump; these actions were modeled using several simplifications. Overall, the results indicate that combining isolation and drainage can reduce the conditional core damage probability upon the occurrence of flooding by approximately 90%.

Application of quasi-Monte Carlo and importance sampling to Monte Carlo-based fault tree quantification for seismic probabilistic risk assessment of nuclear power plants

Kubo, Kotaro; Tanaka, Yoichi; Hakuta, Yuto*; Arake, Daisuke*; Uchiyama, Tomoaki*; Muramatsu, Ken

Mechanical Engineering Journal (Internet), 10(4), p.23-00051_1 - 23-00051_17, 2023/08

https://doi.org/10.1299/mej.23-00051

The significance of probabilistic risk assessments (PRAs) of nuclear power plants against external events was re-recognized after the Fukushima Daiichi Nuclear Power Plant accident. Regarding the seismic PRA, handling correlated failures of systems, components, and structures (SSCs) is very important because this type of failure negatively affects the redundancy of accident mitigation systems. The Japan Atomic Energy Research Institute initially developed a fault tree quantification methodology named the direct quantification of fault tree using Monte Carlo simulation (DQFM) to handle SSCs' correlated failures in detail and realistically. This methodology allows quantifying the top event occurrence probability by considering correlated uncertainties related to seismic responses and capacities with Monte Carlo sampling. The usefulness of DQFM has already been demonstrated. However, improving its computational efficiency would allow risk analysts to perform several analyses. Therefore, we applied quasi-Monte Carlo and importance sampling to the DQFM calculation of simplified seismic PRA and examined their effects. Specifically, the conditional core damage probability of a hypothetical pressurized water reactor was analyzed with some assumptions. Applying the quasi-Monte Carlo sampling accelerates the convergence of results at intermediate and high ground motion levels by an order of magnitude over Monte Carlo sampling. The application of importance sampling allows us to obtain a statistically significant result at a low ground motion level, which cannot be obtained through Monte Carlo and quasi-Monte Carlo sampling. These results indicate that these applications provide a notable acceleration of computation and raise the potential for the practical use of DQFM in risk-informed decision-making.

Accident sequence precursor analysis of an incident in a Japanese nuclear power plant based on dynamic probabilistic risk assessment

Kubo, Kotaro

Science and Technology of Nuclear Installations, 2023, p.7402217_1 - 7402217_12, 2023/06

https://doi.org/10.1155/2023/7402217