Volume 13: Risk Assessments and Management最新文献

{"title":"Probabilistic Analysis of Containment Overpressure Failure in Level 2 PSA","authors":"Y. Liu, Wenjing Li, Wen-kui Lu, Gaopeng Wang","doi":"10.1115/icone29-92038","DOIUrl":"https://doi.org/10.1115/icone29-92038","url":null,"abstract":"\u0000 Containment is the final barrier to prevent any leakage and release of radioactive materials. The containment internal pressure may exceed the design value during severe accident, which could threaten the containment integrity. Therefore, the study on fragility and failure probability of containment under accident conditions has become a crucial issue in nuclear safety. In this paper, firstly, the PWR containment failure modes are identified based on relevant researches and engineering practices. Secondly, the containment failure caused by overpressure is divided into the structural failure and functional failure, which are discussed in detail. Finally, combined with both deterministic and probabilistic analysis methodology, fragility curve of PWR containment was evaluated as a step on the evaluation of containment failure probability due to the overpressure in Level 2 PSA.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"25 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":"127198457","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}

{"title":"Research on Development of On-Line Monitoring of Fuel Damage in Nuclear Power Plant","authors":"Tian Yiduo, Huo Wenjun","doi":"10.1115/icone29-93128","DOIUrl":"https://doi.org/10.1115/icone29-93128","url":null,"abstract":"\u0000 The occurrence of fuel damage has serious impact on safety and economic efficiency of nuclear power plants. The on-line monitoring and diagnosis system of fuel damage in nuclear power plant can timely determine whether the fuel element cladding is damaged or not, and diagnose the relevant information such as the size of the rupture, the position of the damaged component and the location of the damage.\u0000 The paper investigates some fuel damage monitoring systems and analyses their basic structure and process. Based on the examples, the paper classifies the process into radioactive signal detection, decomposition and failure diagnosis part and discusses the theoretical and experimental basis. Different methods of fuel damage diagnosis along with factors resulting in errors and applicable conditions are also discussed.\u0000 Through the investigation, some results can be concluded that the reason that the online detection system of fuel damage is not widely used is mainly due to precision limitation. Future efforts should be made on decomposition and failure diagnosis part in order to reach a fast response and high accuracy.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"72 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":"127473881","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}

{"title":"Spent Fuel Pool Risk Analysis for a CPR1000 Unit","authors":"Liu Weidong, Liu Leilei","doi":"10.1115/icone29-91949","DOIUrl":"https://doi.org/10.1115/icone29-91949","url":null,"abstract":"\u0000 After the Fukushima nuclear accident, the safety level of the spent fuel pool has gradually attracted increasing attention. This paper develops a probabilistic safety assessment (PSA) model for the spent fuel pool of a CPR1000 unit, and quantitatively evaluates the frequency of fuel damage. Possible weaknesses of the spent fuel pool are identified, and suggestions are provided to optimize the operation and management of nuclear power plants, thereby further improving the accident prevention and mitigation capabilities of the spent fuel pool and enhancing the overall safety of nuclear power plants. Through the quantitative analysis of spent fuel pool PSA, SFP fuel damage frequency of 6.99E−09/reactor year is finally developed. Compared with the core damage risk, SFP fuel damage frequency is very low, less than 1% of the core damage frequency, and the risk level is far lower than the reactor core. From the quantitative results of the initiating events, PTR discharge piping breakup and loss of offsite power are the dominant contribution to fuel damage frequency, mainly due to 1) After PTR discharge piping breaks, the ruptures can’t be isolated, and the emergency make-up is not available, finally leading to fuel damage; 2) Under refueling state, the loss of offsite power, diesel generator failure, while the main transformer, RRI/SEC, LHP/LHQ is to carry out planned maintenance, and also the emergency water supply failure jointly lead to fuel damage.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"18 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":"122515216","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}

{"title":"Case Study on Significance Determination Process for NPPs Abnormal Events at Power","authors":"Shijun Chen, Xiang Zou, Lihui Chen, Wen-Bo Luo","doi":"10.1115/icone29-93590","DOIUrl":"https://doi.org/10.1115/icone29-93590","url":null,"abstract":"With the research and application of Significance Determination Process (SDP) in China, it is urgent to carry out the detailed analysis of the significance of abnormal events at power, especially the NRC reactor regulation concept is being accepted to regulate the nuclear power plant in china. Based on Significance Determination Process (SDP) and the actual case analysis, this paper analyzes the significance of license operation event in a nuclear power plant with the current developed PSA model. The analysis process includes the initial significance determination, detailed analysis and external events/LERF analysis, finally determine the significance of license operation event, and provide a reference for the subsequent analysis of abnormal events.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"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":"125413273","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}

{"title":"The Research and Application of Configuration Risk Management in Daya Bay NPP","authors":"Kuan Zhang, Changsheng Cheng, Dongyuan Guo, Qiongzhe Li, Heng Zhang","doi":"10.1115/icone29-93537","DOIUrl":"https://doi.org/10.1115/icone29-93537","url":null,"abstract":"\u0000 National Nuclear Safety Administration issuance “The Technical Policy of Nuclear Power Plant Configuration Risk Management” in December 2019 to support the risk management of multiple system/equipment failures and the flexible operation of nuclear power plant (NPP). Subsequently, Suzhou Nuclear Power Research Institute PSA teams and Daya Bay NPP nuclear safety advisory/Operation operator/planning Engineer established a joint project team to study the implementation of configuration risk management. The configuration risk management system framework and operation process have been determined through the in-depth discussion of multiple disciplines and departments, and the trial operation of risk management system has been started in January 2021 to optimize iterative software and implementation processes. This paper introduces the composition of Daya Bay configuration risk management system, the responsibilities of each discipline in the implementation of configuration risk management, the establishment method and results of risk threshold and risk management matrix, the functions of operation and maintenance risk assessment tools, the characteristics of risk calculation engine, the implementation process of operation configuration risk management and maintenance configuration risk management, the problems found during the trial operation, the management measures for subsequent upgrades, so as to provide reference for other NPPs to carry out the configuration risk management.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"26 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":"125811311","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}

{"title":"Research on Key Problems of Digital Instrument and Control System Reliability Analysis for Probabilistic Safety Assessment","authors":"S. Lin","doi":"10.1115/icone29-93001","DOIUrl":"https://doi.org/10.1115/icone29-93001","url":null,"abstract":"\u0000 Under accident conditions, the support of digital instrument control system especially the normal operation of protection and safety monitoring system has an important impact on the safety of nuclear power plants. As the basic core components of digital I&C system, the card modules integrate and cooperate to perform nuclear power plant operation, data display, control, protection and safety monitoring functions. The existing card reliability analysis methods appeared to be too conservative, and lack enough operation experience and field test as sample data. In this paper, analysis tools such as RiskSpectrum PSA professional software are used, through the methods of static fault tree model refinement, card reliability test, Bayesian/frequentist statistical analysis, a quantification method of card reliability parameter are put forward, which is more suitable for PSA. The PSA model of digital I&C PMS is refined and optimized according to the results, truly reflects the reliability level of the digital I&C system. While verifying the advancement of digital I&C system, this work provides technical basis for design and operation revision or optimization decision, thus improving the safety and economy of nuclear power plants.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"32 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":"133931846","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}

{"title":"Level 3 PSA Application in a UK Generic Design Assessment Project","authors":"Jinkai Wang, Qi Wang","doi":"10.1115/icone29-89218","DOIUrl":"https://doi.org/10.1115/icone29-89218","url":null,"abstract":"\u0000 Level 3 Probabilistic Safety Assessment (L3 PSA) is required in UK Generic Design Assessment (GDA) to demonstrate that a new nuclear power plant is suitable to be built in UK. L3 PSA is used to assess the individual and societal risk and compare the results against the offsite Radiation Protection Targets (RPTs) for fault and accident conditions. There is little relevant good practice for L3 PSAs that have recently been performed worldwide and there is lack of mature standard to implement L3 PSA. In this study, a pilot L3 PSA is performed for UK HPR1000 to reflect the UK context and relevant good practices. It introduces the methodology and the processes to be followed to perform conditional consequences calculations for the faults and accident scenarios. All radiation sources are considered and analyzed. The radiological risks to a potential UK site are analyzed and compared against RPTs. A widely used code - PC COSYMA, is selected as a primary tool for the accident consequence calculations. The strengths and limitations of the code are identified based on the project situation, and either qualitative arguments or supplementary analysis is subsequently proposed to overcome the limitations. The final L3 PSA results has supported the demonstration that the offsite radiological risks for UK HPR1000 have been achieved as low as reasonably practicable (ALARP) and meet the UK regulatory expectation.","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":"131134758","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}

{"title":"Applying the PIRT Approach to the Meteorological Hazards Wind and Rainfall in Relation to Industrial Needs: Benefits and Lessons Learned","authors":"Tiphaine Le Morvan, Arièle Défossez, Mickaël Kreitz","doi":"10.1115/icone29-90907","DOIUrl":"https://doi.org/10.1115/icone29-90907","url":null,"abstract":"\u0000 As an integrated nuclear engineer and operator, EDF has a strong background in studying extreme natural hazards for its Nuclear Power Plants (NPPs) fleet safety demonstration. The state-of-the art evolution and the development of natural hazards Probabilistic Safety Analysis provide an opportunity to go more in depth in our studies. To reinforce or, if need be, to adjust our hypotheses regarding the detailed characterization of natural hazards, we decided to investigate the detailed phenomenology associated with strong winds and rainfall in France.\u0000 The PIRT process (Phenomena Identification and Ranking Table) allows a physics-oriented approach to complex phenomena by systematically analyzing relevant parameters and their influence on a scenario of interest. This approach appears like a legitimate way to go more in depth in the phenomenology associated with strong winds and rainfall.\u0000 This approach, used in the nuclear energy field, for example, as support for complex transients numerical modelling, is here tested on complex natural phenomena.\u0000 Strong winds, rainfall and other associated phenomena are very complex if one wishes to consider a high detail level. It was decided during the exercise to limit the technical complexity to a level relevant to industrial concerns.\u0000 Even with this simplification, the lessons learned will allow EDF to be more precise in future studies, such as those regarding strong winds directions and some phenomena specific to certain geographical locations.\u0000 This article describes the PIRT approach and modifications that were made to the classical PIRT steps both for strong winds and rainfall. We also outline the main lessons we learned, from the PIRT application and regarding the analysis we were able to gain regarding specific needs for our studies.\u0000 We found that the PIRT approach can be used to study meteorological phenomenology with the following adaptations:\u0000 - An adapted detail level — some very complex aspects, coming under Research & Development (R&D) or advanced meteorological expertise have not been taken into account in the present exercise\u0000 - Use the PIRT as a way to present and organize advanced knowledge on a subject. The aim is to shed more light on an end user issue, like to better comprehend the underlying physical processes for strong winds safety demonstration studies: speed, direction, duration ...\u0000 The exercise was possible thanks to the strong collaboration with Météo France, the national French Meteorological Agency, who was the referent expert bringing the technical and scientific knowledge on meteorology.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"11 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":"115196586","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}

{"title":"Study on Different Ways for Selecting NPP Design Extension Conditions","authors":"Yuan Ma, C. Feng, Chao Ma","doi":"10.1115/icone29-93071","DOIUrl":"https://doi.org/10.1115/icone29-93071","url":null,"abstract":"\u0000 With the further improvement of the safety requirements, the Design Extension Conditions need to be fully considered as an important design requirement in the design of advanced nuclear power plants. For Chinese new nuclear power plants and future new reactor types, in principle, a set of Design Extension Conditions must be obtained on the basis of engineering judgment, determinism and probability theory evaluation to further improve the safety of nuclear power plants. In current practice, PSA methods and models are often used to identify and determine extremely unlikely events and multiple failure events in the selection of the Design Extension Condition sequence list, while considering the design extended conditions given by determinism and engineering judgment. However, even with the same frequency screening values, the way in which the PSA model is used can be different, so it is necessary to study the effect of different ways on the results of DEC selection. In this paper, combined with the determination process of the DEC-A list of a third-generation reactor type, two different ways of identifying the DEC-A event sequences based on the PSA event sequence and based on the cutset-importance, are described, and the results are compared and discussed.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"22 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":"132446173","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}

{"title":"Review of Fire Common Mode Analysis and Probabilistic Safety Assessment of Fire Common Modes for an Operating Nuclear Power Plant","authors":"Lihua Huang, Bin Long","doi":"10.1115/icone29-93474","DOIUrl":"https://doi.org/10.1115/icone29-93474","url":null,"abstract":"\u0000 Fire common mode analysis (a.k.a. Fire Vulnerability Analysis, FVA) is usually performed after the completion of the design of a nuclear power plant. Any discrepancies in equipment or cable layout during the construction or operation phases could result in new potential fire common modes, which pose new safety threats to the plant. Consequently, it is necessary to perform a review of the fire common mode analysis for an operating nuclear power plant. Any newly identified fire common modes that need to be addressed might require modification either by protection or by physical separation. In prioritizing the modifications regarding these fire common modes, probabilistic safety assessment (PSA) provides quantitative insights into the safety benefits of eliminating any newly identified fire common modes. This article introduces the general process of FVA and discusses the tasks of cable data collecting and plant walkdown, which are two fundamental steps of the process of FVA for an operating nuclear power plant. In the second part, an example is also provided of performing probabilistic safety assessment on fire common modes, based on which the analyst can prioritize the common mode list in terms of safety benefits if they are to be eliminated.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"126 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":"123082513","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}