轻水反应堆RBIPart 1.docx

1、轻水反应堆RBIPart 1CRTD-Vol. 20-2基于风险的检测的导则制订轻水反应堆(LWR)核电站部件PREPARED BYThe Research Task Force on Risk-Based Inspection GuidelinesAPPOTNTED BYThe Codes and Standards Research Planning Committee of the ASME Center for Research and Technology DevelopmentForThe ASME Council on Codes and StandardsThe United

2、States Nuclear Regulatory CommissionThe National Board of Boiler and Pressure Vessel InspectorsThe Pressure Vessel Research Committee - Welding Research CouncilAmerican Nuclear InsurersThe Hartford Steam Boiler Inspection and Insurance CompanyIndustrial Risk InsurersThe American Petroleum InstituteT

3、he National Rural Electric Cooperative AssociationThe United States Department of EnergyOil Insurance LimitedEdison Electric InstituteREVIEWED AND EDITED BYSteering Commit tee on Risk-Based Inspection Guidelines and an Independent Peer Review CommitteeNRC GRANT NO. NRC-04-89- 102THE AMERICAN SOCIETY

4、 OF MECHANICAL ENGINEERSUnited Engineering Center 345 East 47th Street Now York, N.Y. 100 17DISCLAIMERThis report was prepared as an account of work sponsored through the American Society of Mechanical Engineers (the Society) Center for Research and Technology Development by the United States Nuclea

5、r Regulatory Commission, the National Board of Boiler and Pressure Vessel Inspectors, the Pressure Vessel Research Committee - Welding Research Council, the American Nuclear Insurers, The Hartford Steam and Boiler- Inspection and Insurance Company, the Industrial Risk Insurers, the American Petroleu

6、m Institute, the National Rural Electric Cooperative association, The United States Department of Energy, Oil Insurance: Limited, and Edison Electric Institute (collectively referred to herein as the Sponsors).Neither the Society. nor the Sponsors, nor Westinghouse Electric Corporation, the Universi

7、ty of Maryland, Rolls Royce and Associates Ltd., Battelle Pacific Northwest Laboratories, Failure Analysis Associates, Inc., Factory Mutual Research Corporation. Idaho National Engineering Laboratory, and the McDonnell Aircraft Company (collectively referred to herein as the Sponsorees), nor any fin

8、ancial contributors or others involved in the preparation or review of this report. nor any of their respective employees, members, or persons acting on their behalf, makes my warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness

9、 of any information, apparatus, product, or process disclosed, or represents that its use would not infringe upon privately owned rights.Reference herein to any specific commercial product, process or service by trade name. trademark, manufacturer, or otherwise does not necessarily constitute or imp

10、ly its endorsement, recommendation, or favoring by the Society, the Sponsors, the Sponsorees, or any financial contributors or others involved in the preparation or review of this report, or any agency thereof. The views and opinions of the authors, contributors, and reviewers of the report expresse

11、d herein do not necessarily reflect those of the Society, the Sponsors, the Sponsorees, or any financial contributors or others involved in the preparation or review of this report, or any agency thereof, Statement from By-Laws: The Society shall not be responsible for statement or opinions advanced

12、 in papers or printed in its publications ( 7.1.8)ISBN Nn. 0.791 8-0658-8Library of CongressCatalog Number 92-5.1327Copyright: I- 1992 byTHE AMERICAN SOCIETY OF MECHANTCAL ENGINEERSAll Rights ReservedPrinted in U.S.A.ACKNOWLEDGMENTSAlthough this document represents the work of the research task forc

13、c mcmbers, this study would not bo possible without the contributioris of a large number of leaders in their respective fields from academia, government, and indust1y.The steering committee members have carefully guided the project. and the independent peer review members teamed with the steering co

14、mmittee to diligently review and edit this document. The valuable and gencrous cnntrihution of these members, who are identified in this document, is most appreciated. The rosuachta sk force acknowledges with appreciation thc contributions of Truong Vo of Battelle Pacific Northwest Laboratories, who

15、 attended most of the meetings and provided results from scvcral of the pilot studies cited herein. He has been invited to be an honorary member of the task force because of his significant contribution. Comments by Dr. Lee Abramson of the U.S+ Nuclear Regulatory Commission regarding the elicitation

16、 of expert opinion and his provision of an equation format for the risk-based inspection methodology were much appreciated.Charles H. Boyd, Barney L. Silverblatt, Richard E. Schwirian, and Barry D. Sloane of Westinghouse Electric Corporations Nuclear and Advanced Technology Division are noted for th

17、eir contribution to the inspection prioritization example for reactor internals subcomponents. Dr. Robert Perduc and Bruce A. Bishop of Westinghouse Electric Corporations Science and Technology Center and Nuclear and Advanced Technology Division. respectively, are cited for their valuable assistance

18、 in preparing the decision analysis example for choosing an inspection strategy.Finally, the Westinghouse Energy Center Word Processing staff and the ASME Technical Publishing Department members are acknowledged for their dedicated and diligent efforts in compiling, editing, and publishing this docu

19、ment.基于风险的检测导则研究指导委员会Raymond J. Art, Assistant Director, ASME Center for Research and TechnologyDcvclopment , Washington, D.C.Robert J. Bosnak, ASME Council on Codes and Standards, ASME Codes andStandards Research Planning Committee, ASME Board on Research andTechnology Development, Deputy Director

20、- Division of Engineering, Officeof Nuclear Regulatory Research, U.S. Nuclear Regulatory Cornmission.Washington, D.C.Dr. Spencer J. Bush, Consultant, Past Chairman - ASME Section XI, Richland,WashingtonJohn Blackburn, American Petroleum Institute, Washington, D.C.Ray Davies, Dc t Norse Veritas Indus

21、trial Services, Inc.Thdore A. Meyer, Manager, Westinghouse Electric Corporation, Pittsburgh,PennsylvaniaEvangelos Michalopoulos, P.E., Senior Engineer. The Hartford Steam BoilerInspection and Insurance Company, Hartford, ConnecticutDr. Joseph Muscara, Senior Metallurgical Engineer, U. S + Nuclear Re

22、gulatoryCommission. Washington, D.C.Michael E. G. Schmidt, P.E., Research Consultant, Industrial Risk Insurers,Hartford, ConnecticutErnest W. Thrmkmorton, Virginia Power, Glen Allen, VirginiaWilliam G. Wendland, P.E., Manager - Engineering Projects, American NuclearInsuruers, Farmington, Connecticut

23、viii独立同业互查委员会Dr. Vicki Bier, Professor, University of Wisconsin, Madison, WisconsinJohn D. Boardman, Inservice Inspection Engineer, Southern CaliforniaEdison, San Clemente. CaliforniaMichael Beford, Robin L. Dyle, and Dennis M. Swann, Engineers, Inspectionand Testing Services, Southern Nuclear Opera

24、ting Company,Birmingham, AlabamaT. N. (Bud) Epps, Chairman ASME BPVC Section XI Long Range PlanningCommittee, T.K.S. International Inc., Birmingham, AlabamaCRTD-Vol. 20-2基于风险的检测的导则制订轻水反应堆(LWR)核电站部件执行摘要本文件是题为“基于风险的检测的导则制订”系列的第二部分，正由美国机械工程师学会(ASME)的多学科特别小组制订。第一刊第一卷一般性文件(ASME 1991)说明了一个总体的基于风险的程序，该程序可被

25、用来制订任何工业设施或结构系统的检验指南。该系列的后续各卷给出了阐明特定工业的结构完整性问题的一般方法论的具体应用。本文件(第二卷第一部分)即是首个这种应用。文件是针对轻水反应堆(LWR)核电站部件的检验。第一卷中推荐的以基于风险的检测的一般性程序提供了一个以节省成本的方式配置检验资源的总体框架，并有利于将检查运用到最需要的地方。该一般性方法论已对在核部件中的应用作了进一步地规定和扩充。 该程序包括以下五个部分：(a) 系统的定义；(b) 定性风险评估；(c) 定量风险分析，包括失效模式、影响与危害性分析(FMECA)以选择待检验的部件并对其进行评级；(d) 采用分摊法选择单个部件的目标失效概

26、率以使因所有部件的失效而产生的影响保持在总风险目标值以下；(e) 确定一个使用具有结构可靠性/风险评估(SRRA)方法的风险决策分析法将部件失效概率保持在目标值以下的费用低的检验计划。对于程序的头两部分，已经为LWR核电站作了充分说明，定性风险方法也已暗含在现行的检验计划之中。说明制订核部件检验计划的定量风险分析的使用是本第二卷所报告的研究工作的主焦点。尤其是：现在已经为许多核电站而产生的风险概率评估信息的使用已并入该方法论中，从而可以提高与部件压力边界失效相关的风险的量化程度。 已根据这些定量风险估计为规定通过检验活动保持的目标部件失效概率值推荐了一个程序。已经阐述了确定一个检验计划必须具有

27、的特性以便在考虑经济因素的同时符合目标失效概率的方法。已经进行了基于风险的检测方法论的初步试验，包括萨里-1(Surry-1)发电厂的一个主要研究。在整个第二卷第一部分介绍该程序的各个部分的结果。先导性研究证明以下：整个程序中可计算并使用定量风险值，说明该方法论切实可行；使用风险决策分析和SRRA方法可以结合安全和经济因素，从而选择最佳的部件检验策略；还需要合理的资源投入；结果合理，并符合常识性质量评定。尽管基于风险的检测的方法可用于整个轻水堆核电站,今后努力的主要目标将是全面推荐采用ASME锅炉压力容器规范(BPVC)第XI卷。为了BPVC第XI卷相关组的考虑，第二卷第二部分计划为核部件推荐

28、一个基于风险的检测计划，包括一个坚实的技术基础。今后的工作应：完成萨里-1核电站各部件失效风险；进一步确定整个电站上结果能够普及化的程度；论证用于所选部件的检验策略的制订将以一种节约成本的方式使由于元件失效引起的风险保持在目标风险值以下。尽管完成此项工作需要相当长的时间，但基于风险的检测的一些关键好处如下：运用其程序和方法时获得的真知卓见；在参与保持商用核电站的安全可靠运行的 众多专业和组织之间的交流得到增强。目录致谢执行摘要基于风险的检验导则特别研究小组基于风险的检验导则研究指导委员会独立同业互查委员会基于风险的检测导则制订：出版物清单出版物清单ASME研究和技术开发中心1 引言1.1 与一

29、般的基于风险的检测方法论的关系 1.2 目标与范围 2 基于风险的检测的全过程 2.1 综述 2.2 系统定义 2.3 定性风险评估 2.4 定量风险分析 2.5 故障模式影响及危害性分析(FMECA)方法论 2.6 目标失效概率的选择 2.7 考虑安全与经济因素的检验计划的制订 3 定义 4 摘要与建议 4.1 方法论概述 4.2 进一步制订状况与计划 4.3 结论 5 参考文献 5.1 正文参考文献 5.2 附录参考文献 插图 1-1 一个加压水冷反应堆系统的基本元件 1-2 直接循环沸水反应堆系统 2-1 轻水堆核电站部件的基于风险的检测程序 2-2 核电站安全系统的典型分类 2-3 瑞

30、典在确定检验时间间隔时采用的方法 2-4 技术信息与核电站部件的基于风险的检测的FMECA的整合 2-5 使用专家判断估计失效概率的程序 2-6辅助给水(AFW)系统部件的失效频率估计 2-7 反应堆压力容器的失效频率估计 2-8 轻水堆核电站系统与部件的基于风险的评级技术方法与信息 2-9 六个代表性加压水冷却反应堆(PWR)电站的基于风险的评估 2-10 萨里-1电站所选系统的基于风险的评估 2-1 1 两个代表性PWR电站上所选系统的基于风险的评估 2-1 2 大海湾-1(Grand Gulf-1)核电站所选系统的基于风险的评估 2- 13 奥科尼-3(Oconee-3)的应急给水(EF

31、W)管段 2- 14 萨里-1部件堆芯损坏频率的单个影响 2- 15 萨里-1部件的累积风险影响 2- 16 基于堆芯损坏风险的反应堆内部子部件的风险评估 2-17 基于经济风险的反应堆内部子部件的风险评估 2- 18 基于堆芯损坏频率的萨里-1部件的风险评估 2- 19 考虑安全与经济因素后对检查方案的改进 2-20 用于选择备用检验策略的决策树构架 2-21 核电站系统和部件使用寿命期间的在役检验(ISI)的作用 2-22 用于选择备用检验策略的决策树构架 2-23 用于选择备用检验策略的决策树构架 2-24 超声检验时一个裂纹探测不到的概率与其深度的关系 2-25 使用ASME锅炉和压力

32、容器规范要求进行超声波检验时不能探测到容器缺陷的概率 2-26 使用特殊的程序进行超声波检验时不能探测到容器缺陷的概率 2-27 用于选择检验策略的决策树架构 2-28 对于一个疲劳裂纹扩展损伤机理评价管道可靠性的 SRRA过程步骤示意图 2-29 一个范例中对于各种裂纹深度分布和检验计划一根双头管破裂(DEPB)的累积条件概率 2-30 适用于管道应力腐蚀裂纹SRRA评估的一个扩展的“PRAISE”模型的各种部件的示意图 2-31 加压水冷反应堆容器环带区的辐射脆化SRRA评估的各种部件的示意图 2-32 利用循环试验检测10英寸不锈钢管中晶粒间应力腐蚀裂纹(IGSCC) 2-33 沸水反应堆(BWR)管道中的晶间应力腐蚀裂纹 2-34 根据电力研究院(