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Journal Articles

Effectiveness evaluation of the measures for improving resilience at ultra-high temperatures

Onoda, Yuichi; Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa

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

We developed the measures for improving resilience of the sodium-cooled fast reactor structure using the failure mitigation technology and evaluated the effectiveness of the measures. To prevent core damage in the event of an accident progressing to an ultra-high temperature state, both measures to prevent overpressure in the reactor vessel and measures to cool the reactor core are required. As a core cooling measure, we developed a core cooling concept that promotes radiant heat transfer from the reactor vessel and cools the containment vessel outer surface by natural convection named Containment Vessel Auxiliary Cooling System (CVACS). We developed a method to use the reduction rate of core damage frequency as an indicator for effectiveness of the measures for improving resilience. The core damage frequency was evaluated by calculating the core cooling performance using CVACS, reflecting the results of structural analysis and human reliability analysis. By implementing measures for improving resilience in addition to existing measures, the core damage frequency of Japan loop-type sodium-cooled fast reactor caused by LOHRS has been reduced to about one-hundredth of the previous level.

Journal Articles

Study on public exposure risk assessment for dismantling of radioactive components in decommissioning phase of nuclear reactor facilities

Shimada, Taro; Kabata, Kazuhiko*; Takai, Shizuka; Takeda, Seiji

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

Nuclear regulatory inspections during the decommissioning phase of nuclear power plants need to be conducted based on risk information, but a method for quantitatively evaluating this risk has not been developed. Therefore, in this study, an event tree of accident events that may occur in the decommissioning phase has been developed, and a code DecAssess-R has been developed to evaluate the exposure risk, which is expressed as the product of the exposure dose and probability of occurrence according to the accident sequence for each equipment to be dismantled. In particular, we have taken into account that the amount of mobile radioactivity that may accumulate in HEPA filters and be released all at once during an accident varies temporally and spatially with the progress of dismantling work. The event tree was constructed based on the results of the survey of domestic and international trouble information in the decommissioning phase and similar dismantling and replacement operations. The event frequencies are based on information from general industries, and the event progression probabilities are based on the equipment failure probabilities in the operation phase. The safety functions to be reduced with the progress of decommissioning were taken into account according to the dismantling work schedule. As a result of the exposure risk assessment for dismantling operations of BWRs and PWRs in Japan, the exposure risk for fire events was the largest. In particular, the exposure risk was greater for the dismantling of components in the reactor building by airborne cutting than for the dismantling of reactor internals, which has the greatest radioactivity in underwater dismantling.

Journal Articles

Effectiveness evaluation of the measures for improving resilience of nuclear structures against excessive earthquake, 1; Fragility evaluation of reactor vessel based on structural analysis

Nishino, Hiroyuki; Kurisaka, Kenichi; Futagami, Satoshi; Watakabe, Tomoyoshi; Yamano, Hidemasa

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

The reactor vessel (RV) buckling was a dominant contributor to core damage. However, even if the RV is buckled due to seismic shaking, it is expected that the RV maintains stable state without unstable failure such as rupture, collapse. Realistic consideration of the post-buckling behavior is regarded as a measure for improving the resilience in this study. The purpose of this study is to understand the post-buckling deformation behavior of the RV and to evaluate the RV fragility based on fatigue failure. This study performed structural analysis using a finite element method to quantify time histories of displacement, strain, etc. As the result of the analysis, wrinkles of the buckling appeared at the elevation higher than the liquid level in the RV. The largest strain value was also indicated around this elevation. The cumulative fatigue damage fraction was evaluated in this analysis to evaluate the fragility of fatigue failure in addition to the buckling fragility. The result showed that the seismic intensity for the median fragility of the fatigue failure was about six times larger than the design-basis ground motion. This is 1.2 times larger than the buckling-based result, which suggests that realistic evaluation of the post-buckling behavior could contribute to improving the resilience of the nuclear structure.

Journal Articles

Development of a probabilistic risk assessment method for combined hazards; A Classification and modeling framework for multihazard events

Choi, B.; Nishida, Akemi; Tsutsumi, Hideaki*; Takada, Tsuyoshi

Proceedings of Probabilistic Safety Assessment and Management & Asian Symposium on Risk Assessment and Management (PSAM17 & ASRAM2024) (Internet), 8 Pages, 2024/10

The Fukushima Daiichi Nuclear Power Plant accident highlighted the importance of developing safety assessment methods that consider multihazard events involving numerous simultaneously occurring events such as earthquakes (shaking) and tsunamis (submersion). When addressing such multihazard events, traditional methods often focus on assessing the load combinations of general structures in their structural designs and adopt simple selection criteria. However, these methods fall short when evaluating, countering, and screening external events, such as earthquakes, tsunamis, strong winds, and concentrated heavy rainfall, that occur simultaneously or in a chain. To address this, we reviewed existing literature on multihazard assessment methods, focusing particularly on scenarios involving earthquake and tsunami events. Based on concepts and basic theories, we examined various methods for addressing multihazard scenarios and classified their characteristics. Specifically, several multihazard scenarios were surveyed, and the relationships between multiple hazards were organized. In addition, common causes leading to combined events, their mutual influences, and potential cascading effects were analyzed.

Journal Articles

Uncertainty analysis of dynamic PRA using nested Monte Carlo simulations and multi-fidelity models

Zheng, X.; Tamaki, Hitoshi; Takahara, Shogo; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of Probabilistic Safety Assessment and Management (PSAM16) (Internet), 10 Pages, 2022/09

Journal Articles

Simulation-based Level 2 multi-unit PRA using RAVEN and a simplified thermal-hydraulic code

Zheng, X.; Mandelli, D.*; Alfonsi, A.*; Smith, C.*; Sugiyama, Tomoyuki

Proceedings of 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference (ESREL 2020 and PSAM-15) (Internet), p.2176 - 2183, 2020/11

Journal Articles

Enhancement of the treatment of system interactions in a dynamic PRA tool

Tanaka, Yoichi; Tamaki, Hitoshi; Zheng, X.; Sugiyama, Tomoyuki

Proceedings of 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference (ESREL 2020 and PSAM-15) (Internet), p.2195 - 2201, 2020/11

Journal Articles

Dynamic PRA of flooding-initiated accident scenarios using THALES2-RAPID

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

Proceedings of 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference (ESREL 2020 and PSAM-15) (Internet), p.2279 - 2286, 2020/11

Probabilistic risk assessment (PRA) is one of the methods used to assess the risks associated with large and complex systems. When the risk of an external event is evaluated using conventional PRA, a particular limitation is the difficulty in considering the timing at which nuclear power plant structures, systems, and components fail. To overcome this limitation, we coupled thermal-hydraulic and external-event simulations using Risk Assessment with Plant Interactive Dynamics (RAPID). Internal flooding was chosen as the representative external event, and a pressurized water reactor plant model was used. Equations based on Bernoulli's theorem were applied to flooding propagation in the turbine building. In the analysis, uncertainties were taken into account, including the flow rate of the flood water source and the failure criteria for the mitigation systems. In terms of recovery action, isolation of the flood water source by the operator and drainage using a pump were modeled based on several assumptions. The results indicate that the isolation action became more effective when combined with drainage.

Journal Articles

Severe accident scenario uncertainty analysis using the dynamic event tree method

Zheng, X.; Tamaki, Hitoshi; Ishikawa, Jun; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of 14th International Conference on Probabilistic Safety Assessment and Management (PSAM-14) (USB Flash Drive), 10 Pages, 2018/09

Journal Articles

Application of Bayesian optimal experimental design to reduce parameter uncertainty in the fracture boundary of a fuel cladding tube under LOCA conditions

Narukawa, Takafumi; Yamaguchi, Akira*; Jang, S.*; Amaya, Masaki

Proceedings of 14th International Conference on Probabilistic Safety Assessment and Management (PSAM-14) (USB Flash Drive), 10 Pages, 2018/09

Journal Articles

PRA on mixed foreign substances into core of Japanese prototype FBR

Nishimura, Masahiro; Fukano, Yoshitaka; Kurisaka, Kenichi; Naruto, Kenichi*

Proceedings of 13th Probabilistic Safety Assessment and Management Conference (PSAM-13) (USB Flash Drive), 12 Pages, 2016/10

Fuel subassemblies of fast breeder reactors (FBRs) are densely arranged and have high power densities. Therefore, the local fault (LF) has been considered as one of the possible initiating events of severe accidents. According to the LF evaluation under the condition of total flow blockage of one sub-channel in the analyses of design basis accident (DBA) for Monju, it was confirmed that the pin failures were limited locally without severe core damage. In addition, local flow blockage (LB) of 66% central planar in the subassembly was investigated as one of the beyond-DBA. However, it became clear that these deterministic analyses were not based on a realistic assumption by experimental studies. Therefore, PRA on LF which was initiated from LB was performed reflecting the state-of-the-art knowledge in this study. As the result, damage propagation from LF caused by LB in Monju can be included in CDF of ATWS or PLOHS in the viewpoint of both probability and consequence.

Journal Articles

Dynamic and interactive approach to level 2 PRA using continuous Markov process with Monte Carlo Method

Jang, S.*; Yamaguchi, Akira*; Takata, Takashi

Proceedings of 13th Probabilistic Safety Assessment and Management Conference (PSAM-13) (USB Flash Drive), 11 Pages, 2016/10

The current approach to Level 2 probabilistic risk assessment (PRA) using the conventional event-tree (ET)/fault-tree (FT) methodology requires pre-specifications of event order occurrence and component failure probabilities which may vary significantly in the presence of uncertainties. In the present study, a new methodology is proposed to quantify the level 2 PRA in which the accident progression scenarios are dynamic and interactive with the instantaneous plant state and related phenomena. The accident progression is treated as a continuous Markov process and the transition probabilities are evaluated based on the computation of plant system thermal-hydraulic dynamics. A Monte Carlo method is used to obtain the resultant probability of the radioactive material release scenarios. The methodology is applied to the protected loss of heat sink accident scenario of the level 2 PRA of a generation IV fast reactor.

Journal Articles

Event sequence assessment of tornado and strong wind in sodium cooled fast reactor based on continuous Markov chain Monte Carlo method with plant dynamics analysis

Takata, Takashi; Azuma, Emiko*

Proceedings of 13th Probabilistic Safety Assessment and Management Conference (PSAM-13) (USB Flash Drive), 10 Pages, 2016/10

A new approach has been developed to assess event sequences under external hazard considering a plant status quantitatively and stochastically so as to take various scenarios into account automatically by applying a Continuous Markov Chain Monte Carlo (CMMC) method coupled with a plant dynamics analysis. In the paper, a tornado and a strong wind are selected as the external hazard to assess the plant safety in a loop type sodium cooled fast reactor (SFR). As a result, it is demonstrated that the various scenarios where the order of the occurrence event and its occurrence time differs from each other can be assessed simultaneously as well as the statistical characteristics of plant parameter such as the coolant temperature. Furthermore, a weight factor is introduced so as to investigate the low failure probability events with a comparative small number of the sampling.

Journal Articles

Reliability enhancement of seismic risk assessment of NPP as risk management fundamentals; Quantifying epistemic uncertainty in fragility assessment using expert opinions and sensitivity analysis

Choi, B.; Nishida, Akemi; Itoi, Tatsuya*; Takada, Tsuyoshi*; Furuya, Osamu*; Muta, Hitoshi*; Muramatsu, Ken

Proceedings of 13th Probabilistic Safety Assessment and Management Conference (PSAM-13) (USB Flash Drive), 8 Pages, 2016/10

In this study, we address epistemic uncertainty in structure fragility estimation of nuclear power plants (NPPs). In order to identify and quantify dominant factors in fragility assessment, sensitivity analyses of seismic analysis results are conducted for a target NPP building using a three-dimensional finite element model and a conventional lumped mass model (embedded sway rocking model), and the uncertainty caused by the major factors is then evaluated. The results are used to classify epistemic uncertainty levels in a fragility estimation workflow for NPPs in several stages, and a graded knowledge tree technique, which can be used for future fragility estimations, is proposed.

Journal Articles

Bayesian optimization analysis of containment venting operation in a BWR severe accident

Zheng, X.; Ishikawa, Jun; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of 13th Probabilistic Safety Assessment and Management Conference (PSAM-13) (USB Flash Drive), 10 Pages, 2016/10

Journal Articles

Study on next generation seismic PRA methodology, 2; Quantifying effects of epistemic uncertainty on fragility assessment

Nishida, Akemi; Takada, Tsuyoshi*; Itoi, Tatsuya*; Furuya, Osamu*; Muramatsu, Ken*

Proceedings of 12th Probabilistic Safety Assessment and Management Conference (PSAM-12) (USB Flash Drive), 12 Pages, 2014/06

This study focused on uncertainty-assessment frameworks and utilization of expertise develops methodology for quantification of uncertainty associated with final results from SPRA in the framework of risk management of Nuclear Power Plant (NPP) facilities. This research aimed to contribute to the development of probabilistic models for uncertainty quantification- and software (1); to the aggregation of expert opinions on structure/equipment fragility estimation and development of implementation guidance on epistemic uncertainty (2); and to the study of applicability of newly proposed SPRA models to plant models (3). In particular, we focused on the second goal. There were two different groups of experts used: those in the field of civil engineering, and those in the fields of mechanical engineering. With these groups, we conducted a pilot study on the use of expert-opinion elicitation for identification and quantification of parameters of fragility assessment.

Journal Articles

Development of margin assessment methodology of decay heat removal function against external hazards; Project overview and preliminary risk assessment against snow

Yamano, Hidemasa; Nishino, Hiroyuki; Kurisaka, Kenichi; Sakai, Takaaki; Yamamoto, Takahiro*; Ishizuka, Yoshihiro*; Geshi, Nobuo*; Furukawa, Ryuta*; Nanayama, Futoshi*; Takata, Takashi*

Proceedings of 12th Probabilistic Safety Assessment and Management Conference (PSAM-12) (USB Flash Drive), 11 Pages, 2014/06

This paper describes mainly preliminary risk assessment against snow in addition to the project overview. The snow hazard indexes are the annual maximum snow depth and the annual maximum daily snowfall depth. Snow hazard curves for the two indexes were developed using 50 year weather data at the typical sodium-cooled fast reactor site in Japan. In this paper, the snow risk assessment showed less than 10$$^{-6}$$/reactor-year of core damage frequency. The dominant snow hazard category was the combination of 1-2m/day of snowfall velocity and 0.75-1.0 day of snowfall duration. Sensitivity analyses indicated important human actions, which were the improvement of snow removal velocity and the awareness of snow removal necessity.

Journal Articles

Probability of adventitious fuel pin failures in fast breeder reactors and event tree analysis on damage propagation up to severe accident in Monju

Fukano, Yoshitaka; Naruto, Kenichi*; Kurisaka, Kenichi; Nishimura, Masahiro

Proceedings of 12th Probabilistic Safety Assessment and Management Conference (PSAM-12) (USB Flash Drive), 12 Pages, 2014/06

Experimental studies, deterministic and probabilistic and risk assessments (PRAs) on local fault (LF) propagation in sodium cooled fast reactors (SFRs) have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Adventitious fuel pin failures were considered to be the most dominant initiators of LFs in these PRAs because of high frequency of occurrence during reactor operation and possibility of subsequent pin-to-pin failure propagation. Therefore event tree analysis (ETA) on fuel element failure propagation initiated from adventitious fuel pin failure (FEFPA) in Monju was performed in this study based on state-of-the-art knowledge on experimental and analytical studies on FEFPA and reflecting latest operation procedure at emergency in Monju. Probability of adventitious fuel pin failures in SFRs which is the initiating event of this ETA was also updated in this study. It was clarified that FEFPA in Monju was negligible and could be included in core damage fraction of the anticipated transient without scram and protected loss of heat sink in the viewpoint of both probability and consequence.

Journal Articles

Development of accident consequence assessment scheme using accident cost and consideration of decontamination model

Silva, K.*; Okamoto, Koji*; Ishiwatari, Yuki*; Takahara, Shogo; Promping, J.*

Proceedings of 12th Probabilistic Safety Assessment and Management Conference (PSAM-12) (USB Flash Drive), 12 Pages, 2014/06

no abstracts in English

Journal Articles

Research & development of safety approach and safety assessment for the next generation SFR

Okano, Yasushi; Kurisaka, Kenichi; Yamano, Hidemasa; Fujita, Satoshi; Nishino, Hiroyuki; Sakai, Takaaki

Proceedings of Probabilistic Safety Assessment and Management Topical Conference; In light of the Fukushima Dai-ichi Accident (PSAM 2013) (USB Flash Drive), 6 Pages, 2013/04

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