Volume 13: Risk Assessments and Management最新文献_第2页

Reliability Analysis of Active Reactor Pit Flooding Function of EHR DFT System of a NPP Using an Enhanced Component-Connection Method 基于增强组件连接法的核电厂EHR DFT系统活性堆坑泛函可靠性分析

Volume 13: Risk Assessments and Management Pub Date : 2022-08-08 DOI: 10.1115/icone29-91669

D. Guo, Shaoxuan Wang, Daochuan Ge, L. Tong, Xuewu Cao

{"title":"Reliability Analysis of Active Reactor Pit Flooding Function of EHR DFT System of a NPP Using an Enhanced Component-Connection Method","authors":"D. Guo, Shaoxuan Wang, Daochuan Ge, L. Tong, Xuewu Cao","doi":"10.1115/icone29-91669","DOIUrl":"https://doi.org/10.1115/icone29-91669","url":null,"abstract":"\u0000 Redundancy design is often used to enhance the reliability of complex systems which often run as safety-critical systems of nuclear power plant (NPP). Due to the redundancy design and management strategies, safety-critical systems of a NPP often have complex sequential failure behaviors. For a safety-critical system of a NPP with sequential failure behaviors, it is very important to evaluate its reliability for a given mission time. In this paper, the active reactor pit flooding function of EHR (containment heat removal system) of a nuclear power plant (NPP) is modeled by dynamic fault tree (DFT) and analyzed by BDD. To build the BDD, an enhanced component-connection-based method is proposed. The failure probability of the active reactor pit flooding function is analyzed. Besides, the FV, RAW and RRW importance of the EHR equipment are also evaluated. The results show the cumulative failure probability at mission time 24 hour is 1.81E−05, which is very low. The cumulative failure probability of train A (train B) is 4.25E−03. It can be concluded that the active reactor pit flooding function of EHR has a very high reliability given a mission time 24h.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"16 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":"123100555","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

Comparative Study on Internal Flooding Prevention Design of Several Types of NPPs and Improvement Suggestions 几种核电厂内部防洪设计的比较研究及改进建议

Volume 13: Risk Assessments and Management Pub Date : 2022-08-08 DOI: 10.1115/icone29-91895

Yang Yinghao, Liang Zhaohua, Xu Xiaoyan

{"title":"Comparative Study on Internal Flooding Prevention Design of Several Types of NPPs and Improvement Suggestions","authors":"Yang Yinghao, Liang Zhaohua, Xu Xiaoyan","doi":"10.1115/icone29-91895","DOIUrl":"https://doi.org/10.1115/icone29-91895","url":null,"abstract":"\u0000 There are many tapes of nuclear power plants (NPPs) in operation and under construction in China, involving M310 reactor, VVER reactor, AP1000 reactor and HPR1000 reactor and so on. The design concepts of each type nuclear power plant are quite different, including active nuclear power plant and passive nuclear power plant, second-generation reactor and third-generation reactor. Their designs for preventing internal flooding are also very different, either through physical isolation or by adding flood prevention barrier, but each reactor type nuclear power plant has its own advantages in flood prevention. In order to further improve the internal flooding prevention ability of nuclear power plants in China, this paper makes a comparative study on the internal flooding prevention design of several mainstream tapes of nuclear power plants in China, and gives the advantages of each nuclear power plant in the internal flooding prevention design. By drawing on the design advantages of other reactor type nuclear power plants and using the analysis method of probabilistic safety assessment (PSA), this paper quantitatively evaluates the improvement of the ability of reactor type nuclear power plant that our company participated in the design or construction to resist internal flooding risk after considering these design optimization, and gives targeted design improvement suggestions.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"42 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":"122491696","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

Access Time and Safety Benefit Analysis of Additional Temporary Water Injection Measures on Primary After the Fukushima 福岛核事故后初级反应堆额外临时注水措施的获取时间和安全效益分析

Volume 13: Risk Assessments and Management Pub Date : 2022-08-08 DOI: 10.1115/icone29-91957

Weidong Liu, Zhao Wang, Leilei Liu, Kuan Zhang

{"title":"Access Time and Safety Benefit Analysis of Additional Temporary Water Injection Measures on Primary After the Fukushima","authors":"Weidong Liu, Zhao Wang, Leilei Liu, Kuan Zhang","doi":"10.1115/icone29-91957","DOIUrl":"https://doi.org/10.1115/icone29-91957","url":null,"abstract":"\u0000 After the Fukushima nuclear accident, National Nuclear Safety Administration carried out plant inspection nuclear safety on the nuclear plant in operation and under construction, and put forward improvement requirements for each nuclear power plant according to the inspection results. One of the improvement actions was to add temporary water injection measures on the primary side. At present, all power plants have plant requested to add mobile pumps and related water supply pipelines and operating procedures. By summarizing the calculation cases of plant, which selected representative accident sequences for deterministic analysis in the past, this paper determines the effect of implementing temporary water injection measures in primary circuit on accident mitigation at different times, and the results can provide time suggestions for the preparation of mobile equipment in plant. At the same time, the safety benefits after implementing temporary water injection measures in primary circuit in plant are analyzed by using probability analysis method. Based on the analysis of deterministic and probability theory, and considering the preparation time of accessing mobile equipments, it is considered that 4 hours is a reasonable access time for mobile equipments under SBO accident conditions. In the actual access process, this time can ensure that operators have a more suitable time to complete the preparation work of mobile equipments in operation, and at the same time contribute to the reduction of large release frequency. At the same time, it should be noted that at present, the contribution of implementing temporary water injection measures on the primary side under SBO accident is mainly analyzed. In fact, some accident situations with slow development except SBO accident also have great access possibility, so the actual safety benefits of temporary water injection measures on the primary circuit may be more.","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":"122932780","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

Research and Practice on Operational Risk Management of Nuclear Power Plants 核电站运行风险管理的研究与实践

Volume 13: Risk Assessments and Management Pub Date : 2022-08-08 DOI: 10.1115/icone29-91797

Haifeng Wei

{"title":"Research and Practice on Operational Risk Management of Nuclear Power Plants","authors":"Haifeng Wei","doi":"10.1115/icone29-91797","DOIUrl":"https://doi.org/10.1115/icone29-91797","url":null,"abstract":"\u0000 Risk management is a very important work for nuclear power plants, which is related to the sustainable, healthy and stable operation of nuclear power plants. Almost all the work of nuclear power plants requires risk identification and risk control. Based on the risk management theory, advanced concepts and good practices, this paper defines the operational risk management of nuclear power plants, puts forward five modes of operational risk management of nuclear power plants and corresponding control measures for the first time. Combined with the practical feedback of nuclear power plant operation, it clearly puts forward the principles and suggestions for formulating operational risk control and response measures of nuclear power plants. Referring to the analysis mode and corresponding control measures of nuclear power plant operational risk management proposed in this paper, we can accurately and comprehensively analyze the operational risk of nuclear power plants, formulate the risk control measures of validity, the effectiveness of operational risk management of nuclear power plants in guaranty, and prevent the occurrence of accidents/events. At the same time, the research results of operational risk management of nuclear power plants have important reference and guiding significance for the promotion and optimization of risk management of nuclear power plants.","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":"133440944","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 Qualitative and Quantitative Configuration Risk Assessment in AP1000 Unit Outage Scheduling Optimization 定性与定量配置风险评估在AP1000机组停运调度优化中的应用

Volume 13: Risk Assessments and Management Pub Date : 2022-08-08 DOI: 10.1115/icone29-90137

Zhen-qi Wang, Weibing Huang, Jian Zhou, H. Lin, Yinyong Tao, Jun Zhang

{"title":"Application of Qualitative and Quantitative Configuration Risk Assessment in AP1000 Unit Outage Scheduling Optimization","authors":"Zhen-qi Wang, Weibing Huang, Jian Zhou, H. Lin, Yinyong Tao, Jun Zhang","doi":"10.1115/icone29-90137","DOIUrl":"https://doi.org/10.1115/icone29-90137","url":null,"abstract":"\u0000 Configuration risk management refers to the use of probabilistic safety assessment technology to calculate risk indicators and perform risk management based on the actual operation configuration of nuclear power plant. Since its development in 1990, this method has been widely used in nuclear power plants. It can help power plant personnel optimize maintenance plan scheduling, control the nuclear risk level of units and improve the safety of power plants. During the shutdown and outage of the nuclear power plant, a lot of work has to be performed, resulting in the centralized shutdown of system equipment, and the nuclear safety during this period has attracted the attention of the power plant. Therefore, in order to ensure that the nuclear risk of the unit shutdown and outage can be controlled, Sanmen Nuclear Power, as the world’s first AP1000 unit, actively performs configuration risk management and control during the shutdown, It is the first time in China to optimize the outage plan scheduling through the combination of qualitative and quantitative configuration risk assessment method, so as to comprehensively investigate and eliminate the potential risk sources. This paper will mainly introduce the development background of configuration risk management, and take Sanmen Nuclear Power AP1000 unit as the object to introduce the application and achievements of qualitative and quantitative configuration risk assessment in outage scheduling optimization. In addition, it will also make a comparative analysis of qualitative and quantitative configuration risk assessment techniques.","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":"123085324","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

Applicability Study on Human Reliability Analysis for Severe Accidents in Nuclear Power Plants 核电厂重大事故人的可靠性分析的适用性研究

Volume 13: Risk Assessments and Management Pub Date : 2022-08-08 DOI: 10.1115/icone29-91929

Mingsha Zhao, Shuai Chen, Jie Xu, Jianzhang Zhou, Tao Qing

{"title":"Applicability Study on Human Reliability Analysis for Severe Accidents in Nuclear Power Plants","authors":"Mingsha Zhao, Shuai Chen, Jie Xu, Jianzhang Zhou, Tao Qing","doi":"10.1115/icone29-91929","DOIUrl":"https://doi.org/10.1115/icone29-91929","url":null,"abstract":"\u0000 Three severe nuclear power accidents (Fukushima accident, Chernobyl accident and Three Mile Island accident) have proved that human errors in the mitigation of severe accidents will exacerbate the deterioration of accidents and cause immeasurable consequences. Therefore, there is an urgent need for finding human reliability analysis methods to effectively analyze and predict possible human errors in severe accident mitigation.In order to compare the applicability of the Human Reliability Analysis (HRA) method under the severe accident in the nuclear power plant, this paper put forward four criteria, qualitative analysis, quantitative analysis, traceability and availability, based on the characteristics of the personnel in the process of severe accident mitigation. Several commonly used HRA methods were compared and analyzed. The results show that the common HRA methods which applied to the analysis have shortcomings, so it is necessary to establish a new HRA method, or to improve the original methods to solve the issues under severe accident conditions and the IDAC model and Phoenix HRA methodology are considered better choice. Combined with computer programs,the IDAC model can simulate the process of accident and the dynamic response of personnel, which helps to explain the dynamic interaction characteristics of human behavior and environment, but the model still has some limitations in the aspects of uncertainty analysis and simulation data availability. While Phoenix method provides some Performance influencing factors and crew failure modes that can reflect the response characteristics of emergency personnel. This applicability study can provide guidance for the establishment of HRA method in severe accidents.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"19 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":"125784729","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 Risk-Informed Technology in Emergency Operating Procedure Development 风险信息技术在应急操作程序开发中的应用

Volume 13: Risk Assessments and Management Pub Date : 2022-08-08 DOI: 10.1115/icone29-93180

L. Liu, Zhiyun Liu, Zongqiang Yang

{"title":"Application of Risk-Informed Technology in Emergency Operating Procedure Development","authors":"L. Liu, Zhiyun Liu, Zongqiang Yang","doi":"10.1115/icone29-93180","DOIUrl":"https://doi.org/10.1115/icone29-93180","url":null,"abstract":"\u0000 Emergency operating procedures (EOPs) are used to instruct the operators to manage accidents in a nuclear power plant. Many plants are presently in the process of improving their EOP. It is important to recognize that the operators will have to deal with very unusual situations, use systems they are less familiar with, and possibly face unexpected plant behavior they have not experienced. In these situations the operators need reliable operating procedures to adequately respond to the complex and stressful situations and identify and take the appropriate actions. While typical operating procedures for many nuclear power plants are based on the deterministic analysis and the system design where important human actions and weakness are not being sufficiently identified. This report specifies the process by which plant operating procedures are developed and maintained using risk-informed technology so as to ensure that the operating procedures developed are technically accurate, comprehensive, explicit and user-friendly. Moreover, methods and benefits of risk-informed technology for developing procedures and verifying that the important human actions are addressed. Necessary enhancements and applications stressing the important human actions and weakness in developing operating procedures are also proposed which are of great significance in improving the safety operation of the plant.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"84 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":"121142244","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

Missile Risk Assessment for Nuclear Power Plant Under Super Typhoon 超强台风下核电站导弹风险评估

Volume 13: Risk Assessments and Management Pub Date : 2022-08-08 DOI: 10.1115/icone29-93391

Bingchen Feng, Jie Zhou

{"title":"Missile Risk Assessment for Nuclear Power Plant Under Super Typhoon","authors":"Bingchen Feng, Jie Zhou","doi":"10.1115/icone29-93391","DOIUrl":"https://doi.org/10.1115/icone29-93391","url":null,"abstract":"\u0000 Structures, systems and components (SSCs) of nuclear power plants are susceptible to wind-generated missiles under super typhoon, especially for water tanks and electrical equipment arranged in the open air. The tank of the conventional island demineralized water distribution (SER) system is the largest open-air tank of CPR1000 nuclear power plant, which has a high probability of being affected by missiles. To evaluate the missile risk for nuclear power plant under super typhoon, taking SER tank for example, an analysis using probabilistic safety assessment (PSA) method focused on typical accident sequences is carried out. Based on qualitative analysis of initiating events and mitigation functions required, the missile PSA model is established considered the failure probabilities of SER tank under different typhoon conditions. The quantitative evaluation is carried out from the aspect of accident mitigation and overall risk of the unit. The results show that the SER tank is one of the main weaknesses in accident mitigation and has a certain contribution to the overall risk of the unit. Therefore, the risk of missile for nuclear power plant under super typhoon should be concerned.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"47 3 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":"121192611","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

Selection of Complex Accident Sequences for Sodium-Cooled Fast Reactor Design Extension Conditions 钠冷快堆设计扩展条件下复杂事故序列的选择

Volume 13: Risk Assessments and Management Pub Date : 2022-08-08 DOI: 10.1115/icone29-93065

Jingke Jiang, Nanfei Zhao, D. Yang

{"title":"Selection of Complex Accident Sequences for Sodium-Cooled Fast Reactor Design Extension Conditions","authors":"Jingke Jiang, Nanfei Zhao, D. Yang","doi":"10.1115/icone29-93065","DOIUrl":"https://doi.org/10.1115/icone29-93065","url":null,"abstract":"\u0000 Design Extension Conditions (DEC) are accident conditions beyond design basis for nuclear power plants, including complex accident sequences (DEC-A) and severe accidents (DEC-B). As required by the nuclear safety regulation HAF102-2016, DEC must be considered by identifying additional accident scenarios and develop practical prevention and mitigation measures for such accidents in design process for nuclear power plants. The aim is to enhance nuclear power plant’s response to beyond design basis accidents or multi-fault accident. It’s to contribute to increase the design requirements and expand the design consideration for nuclear power plants to consider DEC in design process, which may increase the design difficulty and construction and operation costs for nuclear power plants. The selection of DEC is great significance to balance the safety and economy for nuclear power plants.\u0000 Compared with DEC-B, the DEC-A is a newer concept, and the mature cognition of DEC-A has not yet been reached. The selection method of DEC-A for sodium-cooled fast reactor is analyzed systematically, which includes how to define, identify and screen DEC-A. Taking a sodium-cooled fast reactor as an example, DEC-A are identified and screened in the method of the combination of probabilistic safety assessment, determinism and engineering experience, and the entire selection process of DEC-A is performed. Besides, It is analyzed and discussed about the application of DEC-A in safety analysis and safety design of sodium-cooled fast reactors. Existing study results will provide theoretical reference for determining DEC-A lists for similar reactors.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"98 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":"126094434","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

Study on Interfacing System LOCA of VVER-1200 VVER-1200接口系统LOCA的研究

Volume 13: Risk Assessments and Management Pub Date : 2022-08-08 DOI: 10.1115/icone29-89803

Jinyan Du, Fangyu Dong, Chao Ma

{"title":"Study on Interfacing System LOCA of VVER-1200","authors":"Jinyan Du, Fangyu Dong, Chao Ma","doi":"10.1115/icone29-89803","DOIUrl":"https://doi.org/10.1115/icone29-89803","url":null,"abstract":"\u0000 The evaluation of public risk at nuclear power plants has generally found the risk to be very low. Part of the reason for the low assessed risk can be attributed to the “defense-in-depth” philosophy of nuclear power plants. This defense-in-depth includes multiple barriers to prevent the release of radionuclides to the environment. Nevertheless, accident scenarios have been postulated which might compromise the defense-in-depth by bypassing the reactor coolant system boundary and the containment. One such accident type is referred to as the interfacing system LOCA (ISLOCA). This type of sequence involves the loss of isolation between the high pressure RCS and a low pressure system outside containment. ISLOCA could lead to the loss of coolant outside containment while simultaneously disabling ECCS injection. Such a scenario could result in core damage and might have substantially higher consequences than some of other postulated severe accidents because the containment is also bypassed. Thus, it is necessary to calculate the ISLOCA risk for a specific nuclear power plant. Based on summarizing the ISLOCA analysis method and process, this article carried out the ISLOCA analysis for the VVER-1200 nuclear power plant, identified its typical ISLOCA path, and quantified the impact on the safety of nuclear power plants.","PeriodicalId":407628,"journal":{"name":"Volume 13: Risk Assessments and Management","volume":"39 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":"126769514","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