Volume 13: Risk Assessments and Management最新文献_第5页

A Scoping Study on the Use of Direct Quantification of Fault Tree Using Monte Carlo Simulation in Seismic Probabilistic Risk Assessments 蒙特卡罗模拟直接量化故障树在地震概率风险评估中的应用范围研究

Volume 13: Risk Assessments and Management Pub Date : 2022-08-08 DOI: 10.1115/icone29-88773

K. Kubo, Keita Fujiwara, Yoichi Tanaka, Yuto Hakuta, Daisuke Arake, T. Uchiyama, K. Muramatsu

{"title":"A Scoping Study on the Use of Direct Quantification of Fault Tree Using Monte Carlo Simulation in Seismic Probabilistic Risk Assessments","authors":"K. Kubo, Keita Fujiwara, Yoichi Tanaka, Yuto Hakuta, Daisuke Arake, T. Uchiyama, K. Muramatsu","doi":"10.1115/icone29-88773","DOIUrl":"https://doi.org/10.1115/icone29-88773","url":null,"abstract":"\u0000 After the Fukushima Daiichi Nuclear Power Plant accident, the importance of conducting probabilistic risk assessments (PRAs) of external events, especially seismic activities and tsunamis, was recognized. The Japan Atomic Energy Agency has been developing a computational methodology for seismic PRA, called the direct quantification of fault tree using Monte Carlo simulation (DQFM). When appropriate correlation matrices are available for seismic responses and capacities of components, the DQFM makes it possible to consider the effect of correlated failures of components connected through AND and/or OR gates in fault trees, which is practically difficult when methods using analytical solutions or multidimensional numerical integrations are used to obtain minimal cut set probabilities. The usefulness of DQFM has already been demonstrated. Nevertheless, a reduction of the computational time of DQFM would allow the large number of analyses required in PRAs conducted by regulators and/or operators. We therefore performed scoping calculations using three different approaches, namely quasi-Monte Carlo sampling, importance sampling, and parallel computing, to improve calculation efficiency. These were applied when calculating the conditional core damage probability of a simplified PRA model of a pressurized water reactor, using the DQFM method. The results indicated that the quasi-Monte Carlo sampling works well at assumed medium and high ground motion levels, the importance sampling is suitable for assumed low ground motion level, and that the parallel computing enables practical uncertainty and importance analyses. The combined implementation of these improvements in a PRA code is expected to provide a significant acceleration of computation and offers the prospect of practical use of DQFM in risk-informed decision-making.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"163 7 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129280795","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

PSA Application for UK HPR1000 ALARP Demonstration 英国HPR1000 ALARP示范的PSA应用

Volume 13: Risk Assessments and Management Pub Date : 2022-08-08 DOI: 10.1115/icone29-93109

Zhiwen Wang, Chun-Hao Yang, Bing Zhang, Jinkai Wang, Huina Sang, Lingmei Xiao, Fei Wang, Ming Wang

{"title":"PSA Application for UK HPR1000 ALARP Demonstration","authors":"Zhiwen Wang, Chun-Hao Yang, Bing Zhang, Jinkai Wang, Huina Sang, Lingmei Xiao, Fei Wang, Ming Wang","doi":"10.1115/icone29-93109","DOIUrl":"https://doi.org/10.1115/icone29-93109","url":null,"abstract":"\u0000 UK version of the Hua-long Pressurised Reactor (UK HPR1000), designed by China General Nuclear Power Corporation (CGN), has been successfully completed the Generic Design Assessment (GDA), with the issue of a Design Acceptance Confirmation (DAC) from Office for Nuclear Regulation (ONR) and a Statement of Design Acceptability (SoDA) from the Environment Agency on 7 February 2022. In the UK GDA process, one of the main fundamental requirements is that the nuclear safety risk should be reduced to ALARP (As Low As Reasonably Practicable). During the UK HPR1000 design process, Probabilistic Safety Assessment (PSA) has been widely used as a key tool to inform the design and evaluate risk levels, and to demonstrate that the nuclear safety risk has been reduced to ALARP.\u0000 The process of PSA ALARP demonstration is established in accordance with the methodology that reflects the UK context requirements and consistent with relevant good practice. The process is fully applied by using PSA to systematically review the risk insights and to support other topic areas. After the systematic review of PSA risk insights from internal events Level 1 PSA, internal fire Level 1 PSA, internal flooding Level 1 PSA, external hazards Level 1PSA, external flooding Level 1 PSA, seismic PSA, Spent fuel pool PSA and Level 2 PSA, the related conservatisms, potential design improvements have been identified and addressed.\u0000 Current UK HPR1000 PSA results and risk insights show that the risk level of UK HPR1000 design is relatively low and the risk of UK HPR1000 design has been reduced to ALARP in GDA stage.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133093070","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Prediction and Uncertainty Analysis of Stress Corrosion Fraction Failure Probability for Pressurized Water Reactor Materials 压水堆材料应力腐蚀分数失效概率预测与不确定性分析

Volume 13: Risk Assessments and Management Pub Date : 2022-08-08 DOI: 10.1115/icone29-91951

Deyan Kong, X. Liu, Jie Cheng, Jianjun Wang, Jinqiang Zhang, Dingyong Chen, Qingzhao Li, Yanting Liu

{"title":"Prediction and Uncertainty Analysis of Stress Corrosion Fraction Failure Probability for Pressurized Water Reactor Materials","authors":"Deyan Kong, X. Liu, Jie Cheng, Jianjun Wang, Jinqiang Zhang, Dingyong Chen, Qingzhao Li, Yanting Liu","doi":"10.1115/icone29-91951","DOIUrl":"https://doi.org/10.1115/icone29-91951","url":null,"abstract":"\u0000 The failure of stress corrosion fraction (SCF) will lead to the ejection of nuclear power plant related equipment, which will affect the safety and economy of nuclear power plant. In this paper, a SCF failure analysis method for primary coolant circuits materials was established based on Paris model, and the uncertainty of fracture toughness was transformed into an integral form to improve the calculation efficiency. Taking the weld of thermowell as an example, the probability of SCF failure was calculated, and the uncertainty was analyzed by Wilks’ method. The influences of simple random sampling (SRS), Latin hypercube sampling (LHS) and Halton low discrepancy sequence on the uncertainty quantification of the calculated results were studied. At the same time, a surrogate model was established based on polynomial chaos expansion (PCE) method to study whether this method was suitable for SCF probability calculation. The results show that Halton sequence with 1000 samples can make the mean and variance convergence of failure probability better than SRS and LHS. When calculating the upper limit of tolerance interval, the mean and median of results corresponding to LHS are similar to those of SRS, but the dispersion degree of LHS is lower than SRS, while the results corresponding to Halton sequence are smaller than those corresponding to SRS and LHS. The increase of Wilks’ order can reduce the conservatism. When the order is 4, both the computational efficiency and the computational accuracy are considered. The results of the surrogate model based on PCE are basically consistent with those of the original program, but the amount of calculation is greatly reduced. This method is suitable for SCF probability analysis.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"29 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126986024","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Discussion on the Application of Risk Matrix Method in Regional Nuclear and Radiation Risk Assessment 风险矩阵法在区域核与辐射风险评估中的应用探讨

Volume 13: Risk Assessments and Management Pub Date : 2022-08-08 DOI: 10.1115/icone29-93216

Yu Gong, J. Zheng, Xing Xu, Hui Peng, Yun-huan Qu, Lili Yang

{"title":"Discussion on the Application of Risk Matrix Method in Regional Nuclear and Radiation Risk Assessment","authors":"Yu Gong, J. Zheng, Xing Xu, Hui Peng, Yun-huan Qu, Lili Yang","doi":"10.1115/icone29-93216","DOIUrl":"https://doi.org/10.1115/icone29-93216","url":null,"abstract":"\u0000 While nuclear technology benefits human beings, the ionizing radiation generated may also have adverse effects on human health and the environment. With the rapid development of social economy, nuclear energy and nuclear technology have been widely used, nuclear technology has become more and more close to people’s life, and nuclear and radiation safety has also attracted more and more attention from all walks of life. Therefore, while making effective use of nuclear technology, it is very necessary to do a good job of risk identification and safety management and control. Doing a good job in nuclear and radiation risk assessment is the premise of nuclear safety, nuclear security, and nuclear work. This paper is based on the risk matrix method proposed by the Australian and New Zealand standard “Risk Mangement, AS/NZS 4360:1999”, a qualitative risk assessment analysis method that can comprehensively assess the risk size of the possibility of hazard occurrence and the severity of injury. An attempt to carry out nuclear and radiation risk assessment in a city through the steps of risk source identification, risk possibility analysis, and risk consequence analysis. This paper briefly discusses the application of risk matrix method in regional nuclear and radiation risk assessment, and puts forward suggestions for the application of this method in regional nuclear and radiation risk assessment.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"10 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115409391","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Application of Probabilistic Safety Assessment in the Risk Analysis and Management of Interface Systems LOCA in a Nuclear Power Plant 概率安全评估在核电厂界面系统LOCA风险分析与管理中的应用

Volume 13: Risk Assessments and Management Pub Date : 2022-08-08 DOI: 10.1115/icone29-92253

Leilei Liu, Weidong Liu, Zhao Wang

{"title":"Application of Probabilistic Safety Assessment in the Risk Analysis and Management of Interface Systems LOCA in a Nuclear Power Plant","authors":"Leilei Liu, Weidong Liu, Zhao Wang","doi":"10.1115/icone29-92253","DOIUrl":"https://doi.org/10.1115/icone29-92253","url":null,"abstract":"\u0000 Interfacing systems LOCA is a kind of special loss of coolant accident in nuclear power plants, which has an important impact on the nuclear power plants safety. Because the break of this kind of accident is located outside the containment, the interfacing systems LOCA will cause a large amount of radioactivity to be directly released from the core to the atmospheric environment without the detention of the containment, so which is an important contribution to the large early release of nuclear power plants. The Probabilistic safety assessment have been applied to a nuclear power plant in Chinese mainland in this paper. The risk of interfacing systems LOCA in the nuclear power plant is comprehensively evaluated from the occurrence scenarios and frequency of initial events, accident mitigation strategy and event tree modeling, accident consequence and impact analysis. Through the application of probabilistic safety assessment, the key factors that may lead to a core damage and large release of the interface LOCA in the nuclear power plant has been find out, and the weakness existing in the power plant in dealing with such accidents has been discussed at the same time. According to the results of the analysis, how can the nuclear power plant deal with such accidents much more better is discussed, the improve suggestions are given.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128912054","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Probability Safety Analysis for Nuclear Power Plant Long-Term Temporary Shutdown State 核电厂长期临时停堆状态的概率安全分析

Volume 13: Risk Assessments and Management Pub Date : 2022-08-08 DOI: 10.1115/icone29-91782

Yanzhu Chen, Zhi Yang

{"title":"Probability Safety Analysis for Nuclear Power Plant Long-Term Temporary Shutdown State","authors":"Yanzhu Chen, Zhi Yang","doi":"10.1115/icone29-91782","DOIUrl":"https://doi.org/10.1115/icone29-91782","url":null,"abstract":"\u0000 In recent years, due to the influence of power grid demand and other factors, some nuclear power plants have long-term temporary SHUTDOWN. At present, there is no mature quantitative analysis method to measure which Plant operational state is safer under long-term temporary shutdown. Taking a second-generation nuclear power plant as an example, this paper first briefly introduces the PSA analysis method; Secondly, the difference is analyzed qualitatively, which between standard long-term temporary shutdown condition and refueling shutdown condition. Thirdly, from the quantitative point of view, the model processing methods and analysis assumptions are introduced, the risk views are given, and the quantitative calculation and sensitivity analysis are carried out; Finally, the analysis conclusion is obtained.\u0000 Stay in the standard long-term temporary shutdown state for 15/30/45/60 days, after supplementing the calculation of the thermal shutdown condition,the risk per unit time of long-term temporary staying Maintenance Cold Shutdown mode is lower. The risk sequence is: Cooldown with the residual Heat Removal mode > Hot shutdown condition > Maintenance Cold Shutdown mode with closed primary loop. From the perspective of the risk of core damage, it is suggested that the Maintenance cold shut mode in which the primary loop is kept closed is recommended for long-term temporary shutdown of the power plant.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"89 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124588064","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Inadvertent Reactor Pit Flooding Risk Research on the Generation III Nuclear Power Plant 第三代核电站反应堆坑意外淹水风险研究

Volume 13: Risk Assessments and Management Pub Date : 2022-08-08 DOI: 10.1115/icone29-92450

Lingmei Xiao, Bing Zhang, Ming Wang, Zhiwen Wang

{"title":"Inadvertent Reactor Pit Flooding Risk Research on the Generation III Nuclear Power Plant","authors":"Lingmei Xiao, Bing Zhang, Ming Wang, Zhiwen Wang","doi":"10.1115/icone29-92450","DOIUrl":"https://doi.org/10.1115/icone29-92450","url":null,"abstract":"\u0000 In-vessel Melt Retention (IVR) system is a typical design feature of the the generation III nuclear power plant. In case of severe accident, enough water is injected into the reactor pit to cool the outside of the Reactor Pressure Vessel (RPV) to maintain the integrity of the reactor pressure vessel, so that the core melt is retained in RPV. The design concept of IVR is very meaningful for maintaining the integrity of RPV and reducing the risk of large radioactivity release, but on the other hand, it may also introduce the possibility of inadvertent reactor pit flooding during normal operation, which may threaten the integrity of RPV. At home and abroad, the research on IVR mainly focuses on whether its safety function can be realized, but less attention is paid to its potential negative impact. This paper takes the China generation III reactor with the IVR design feature for the study, carry out the possible inadvertent reactor pit flooding path analysis and frequency analysis. Based on the risk study, put forward the concerns to prevent inadvertent reactor pit flooding, which will provide guide for the subsequent IVR design and optimization.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"124 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124193820","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Construction and Application of Mitigation System Performance Index (MSPI) Platform for a China Nuclear Power Plant 中国某核电站缓解系统性能指标(MSPI)平台的构建与应用

Volume 13: Risk Assessments and Management Pub Date : 2022-08-08 DOI: 10.1115/icone29-91837

Jie Zhang, Jing Wang, Pei Zhang, Kun Tan, Minxue Yang

{"title":"Construction and Application of Mitigation System Performance Index (MSPI) Platform for a China Nuclear Power Plant","authors":"Jie Zhang, Jing Wang, Pei Zhang, Kun Tan, Minxue Yang","doi":"10.1115/icone29-91837","DOIUrl":"https://doi.org/10.1115/icone29-91837","url":null,"abstract":"\u0000 The mitigation system performance index is used to evaluate the ability of the safety-related system to be put into operation to mitigate the accident when an accident occurs in the nuclear power plant (NPP). This index can effectively monitor the status of the safety system and reasonably reflect the impact of equipment performance on the safety level of the nuclear power plant. Typical pressurized water reactor (PWR) nuclear power plant mitigation systems include emergency alternating current (AC) power system, high pressure safety injection system, auxiliary feedwater system, residual heat removal system and cooling water support system. Based on the probabilistic safety assessment (PSA) model, this paper develops a nuclear power plant MSPI computing platform to monitor and judge the risk level of the mitigation system, and guide the nuclear power plant to carry out production activities such as risk-informed preventive maintenance, corrective maintenance and production planning. This paper takes the Essential Service Water System (SEC) of a nuclear power plant in China as an example to verify the proposed method, and conducts the trend analysis of MSPI in the 8 quarters of 2020–2021. The results show that MSPI can timely reflect the influence of the unavailability and unreliability of system equipment on the overall safety status of the power plant from the perspective of risk.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"226 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133337774","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

The Development and Application of Real-Time Risk Monitor System Based on Level 2 PSA for NPP 基于二级PSA的核电厂风险实时监测系统的开发与应用

Volume 13: Risk Assessments and Management Pub Date : 2022-08-08 DOI: 10.1115/icone29-91785

Wenjing Li, Yu Liu, Wei Wei, Xinli Yu

{"title":"The Development and Application of Real-Time Risk Monitor System Based on Level 2 PSA for NPP","authors":"Wenjing Li, Yu Liu, Wei Wei, Xinli Yu","doi":"10.1115/icone29-91785","DOIUrl":"https://doi.org/10.1115/icone29-91785","url":null,"abstract":"\u0000 The method of real-time risk monitor system (RM) development based on Level 2 probability safety analysis (PSA) is established and a real-time risk model for a nuclear power plant (NPP) in China is developed in this paper. The risk monitor system can calculate not only the core damage frequency (CDF) but also the large early release frequency (LERF) with different plant configurations. In this paper, the application of risk monitor system will be discussed in two aspects: Configuration risk management and risk-informed technical specification (TS) optimization. Configuration risk management method based on risk monitor system is studied and case analysis is carried out for the NPP, which will improve the plant risk management ability and safety. Risk-informed technical specification optimization approach based on RM is studied and case analysis is performed for the TS, which can prolong the completion time for a single equipment/system failure in TS to improve the operation economy of NPP. In addition, the evaluation of completion time for multiple equipment failures is the supplement to the traditional TS. The Level 2 PSA risk monitor system development is a good engineering practice of Level 2 PSA in NPP, and is of great significance for improving both the safety and economy of the plant.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125514084","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Dependence Analysis of Nuclear Power Plant Operators Based on HRA 基于HRA的核电站运营商相关性分析

Volume 13: Risk Assessments and Management Pub Date : 2022-08-08 DOI: 10.1115/icone29-89671

Jie Xu, Jianzhang Zhou, Deyi Liu, Jian Wu, Zilong Wang, Mingsha Zhao, Yang Luo, Honghao Chen

{"title":"Dependence Analysis of Nuclear Power Plant Operators Based on HRA","authors":"Jie Xu, Jianzhang Zhou, Deyi Liu, Jian Wu, Zilong Wang, Mingsha Zhao, Yang Luo, Honghao Chen","doi":"10.1115/icone29-89671","DOIUrl":"https://doi.org/10.1115/icone29-89671","url":null,"abstract":"\u0000 As an important part of probabilistic safety assessment (PSA), human reliability analysis (HRA) provides human error probabilities for assessing the risk of nuclear power plants (NPPs). Human reliability analysis (HRA) uses a systematic method to qualitatively and quantitatively assess the mechanism and probability of human error in order to assess the human contribution to the risks of NPPs. Dependence anslysis is an important part of HRA. Without dependency analysis, the result of human error probabilities may be over optimistic, which will lead to the underestimation of the safety risk level of the NPPs. Based on the feedback of human factor dependence from the experts review of level 1 PSA in NPPs. This paper summarized the technical insights of HRA dependence analysis, and used two dependence analysis methods to re-analyze the human factor dependence in the PSA model of nuclear power plant.\u0000 In this paper, starting from the problem of human dependence raised by experts in the peer assessment of first-level PSA in power operation, some technical insights on human dependence were summarized in the analysis and discussion of the existing HRA dependence method. Meanwhile, the human dependence of power PSA was analyzed again by SPAR-H method and NUREG-1921 dependence method.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"116131731","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0