Seismic analysis of nuclear power plants by using three-dimensional finite element models: a review

ABSTRACT

Seismic design of nuclear power plants (NPPs) is important for ensuring their integrity during earthquakes. Seismic analysis has been conducted using lumped mass beam models (LMBMs) for the design of plants in Japan, whereas three-dimensional (3D) finite element models (FEMs) have been used for novel plants outside Japan. The purposes of this study are to organize issues related to the development and application of 3D FEMs for seismic analysis of Japanese NPPs and to indicate future study directions. To organize these issues, the authors systematically investigated: (1) international guides and standards related to seismic analysis and (2) 3D FEMs of novel NPPs outside Japan. By considering other studies on the issues, the authors suggest directions for future studies. Resolving the issues will contribute to application of 3D FEMs for seismic analysis in the design of Japanese NPPs.     

核设施抗震设计中的设计地震反应谱

对于除核电厂以外的其他核设施,国际原子能机构的技术文件中推荐使用一组适用于不同场地情况的标准设计反应谱。通过研究分析其特点,并将GB50011-2010规范中推荐的设计反应谱与其他核设施反应谱相对比,为核设施设计中适当选择设计反应谱提供参考。     

不同法规关于核电厂设计地震动合成的技术要求比较

详细介绍了我国核电厂地震安全评价及抗震分析与设计中用到的多部国内外的法规、标准和导则的技术规定,对其技术背景和要求进行了深入比较和分析,结合工程实践给出了相关的评述和应用建议.     

核电厂负刚度阻尼隔震结构的地震响应研究

隔震技术能有效减小核电厂上部结构的加速度响应,但强地震作用下隔震层位移过大会导致管道断裂。本文基于曲面运动原理及预压弹簧伸缩特性提出了一种负刚度阻尼系统,通过球铰在拱球面曲线运动实现负刚度特性,并在弹簧压缩方向提供黏滞阻尼性能。提出了负刚度系统的理论恢复力模型并进行了力学特性分析,设计了负刚度装置并完成了静力试验,结果显示理论恢复力模型与试验结果的一致性较理想。将核电厂负刚度阻尼隔震结构与核电厂隔震结构进行了地震响应对比分析,比较了不同地震波输入下的地震响应。结果表明负刚度阻尼系统可有效同时减小核电厂上部加速度响应和隔震层位移响应。     

运行核电厂抗震裕度评价研究

抗震裕度评价是对核电厂应对超过设计基准地震能力的评价,特别是日本福岛核事故发生后,评价核电厂应对超过设计基准外部事件时的安全裕量、优化和落实改进措施、提高改进措施的有效性就显得尤为重要。本文通过研究国际上广泛采用的抗震裕度评价方法,最终选定EPRI SMA方法对秦山第二核电厂进行抗震裕量分析。分析结果表明：秦山第二核电厂满足1.5倍SSE的抗震裕度要求,具有较强的抗震能力。     

大亚湾核电站停堆工况风险研究

基于传统ＰＳＡ方法学（适用于功率运行工况）及核电站停堆工况特征,提出了一套停堆ＰＳＡ特征方法,包括电站运行状态离散法,分阶段评价和主逻辑故障树评价。将该方法应用于大亚湾核电站（ＧＮＰＰ）停堆工况ＰＳＡ研究,得到了较真实反映ＧＮＰＰ实际情况的结果。研究结果对ＧＮＰＰ的停堆运行和管理有实际应用价值,以我国今后核电站设计、运行及管理也有现实意义。     

Seismic Isolation Design for JSFR

This paper describes the seismic design of Japan Sodium-Cooled Fast Reactor (JSFR), which includes the seismic condition, the seismic isolation system, and the seismic evaluation of the primary components. Since the design seismic loading is set out severely than ever since The Niigata-ken Chuetsu-oki Earthquake in 2007, an advanced seismic isolation system is aimed to reduce the seismic force loaded on the primary components of JSFR to be less than that of the previous seismic isolation system. The advanced seismic isolation system is developed by optimizing the performance based on the previous seismic isolation system considering the natural frequency of the primary components. The laminated rubber bearings thicker than the previous ones and oil dampers are adopted for the advanced seismic isolation system of SFR. The seismic evaluation of nuclear reactor components applying the advanced seismic isolation system is performed and its feasibility is confirmed.     

日本柏崎·刈羽核电厂新潟地震震害经验

通过对日本柏崎·刈羽核电厂在2007年7月16日新潟地震中的主要震害情况的描述,分析说明该核电厂的抗震设计基本情况,最后提出核电厂抗震设计与评价中应引起重视的问题。     

A probabilistic seismic risk assessment procedure for nuclear power plants: (I) Methodology

A new procedure for probabilistic seismic risk assessment of nuclear power plants (NPPs) is proposed. This procedure modifies the current procedures using tools developed recently for performance-based earthquake engineering of buildings. The proposed procedure uses (a) response-based fragility curves to represent the capacity of structural and nonstructural components of NPPs, (b) nonlinear response-history analysis to characterize the demands on those components, and (c) Monte Carlo simulations to determine the damage state of the components. The use of response-rather than ground-motion-based fragility curves enables the curves to be independent of seismic hazard and closely related to component capacity. The use of Monte Carlo procedure enables the correlation in the responses of components to be directly included in the risk assessment. An example of the methodology is presented in a companion paper to demonstrate its use and provide the technical basis for aspects of the methodology.     

先进压水堆核电厂安全壳地坑滤网设计

核电厂的地坑滤网堵塞问题是目前世界核能领域比较关注的重要安全问题之一,先进压水堆的设计也充分体现了改善地坑性能的最新研究成果。本文介绍了在我国已经开工建设的先进压水堆核电厂（AP1000以及EPR）的地坑滤网设计,可以为我们解决地坑滤网堵塞问题提供有益的借鉴。     

内陆核电厂发展状况与安全监管探讨

对国际上内陆核电厂的发展状况进行了分析,详细介绍了我国内陆核电厂厂址的特点和评价情况,进而对国内首批内陆核电项目的审查情况和监管要求进行了全面的论述。在总结沿海核电厂建设实践和审评经验的基础上,结合我国发展内陆核电厂的实际情况,研究并提出了建设内陆核电厂的审评重点和监管要求。     

Seismic design of nuclear power plants — an assessment

This paper presents a review and evaluation of the design standards and the analytical and experimental methods used in the seismic design of nuclear power plants with emphasis on United States practice. Three major areas were investigated: (a) soils, siting, and seismic ground motion specification; (b) soil-structure interaction; and (c) the response of major nuclear power plant structures and components. The purpose of this review and evaluation program was to prepare an independent assessment of the state-of-the-art of the seismic design of nuclear power plants and to identify seismic analysis and design research areas meriting support by the various organizations comprising the ‘nuclear power industry’. Criteria used for evaluating the relative importance of alternative research areas included the potential research impact on nuclear power plant siting, design, construction, cost, safety, licensing, and regulation.Three methods were used in the study herein. The first involved the review of current literature, focusing primarily on publications dated later than 1970. This review included the results of numerous studies, of which those of Japanese origin and those presented in recent international conferences were predominant. The second method entailed a review of international experience in the dynamic testing of nuclear power plant structures and components, and related experience with scaled and model tests. Included in this experience, in addition to the questions of analysis, design, and measurement of dynamic parameters, are related efforts involving a review of responses obtained during measured earthquake response and investigations into appropriate methods for backfitting or upgrading older nuclear power plants to meet new seismic criteria.The third approach was to obtain the opinions and recommendations of technically knowledgeable individuals in the US ‘nuclear industry’; the survey results are shown in the Appendix.     

多机组核电厂总体风险的一级PSA方法研究

福岛核事故发生后,多机组核电厂的总体风险受到越来越多的关注,但国内外缺乏评价多机组核电厂总体风险的方法或导则。本文结合有关法规对核电厂的总体安全要求,探索将单机组的一级概率安全评价（PSA）方法拓展为多机组的风险评价方法。以双机组核电厂为例,讨论了多机组厂址PSA定量化的一些问题,提出了机组间相关性的一些见解,并阐明了数学原理。本文讨论的方法对研究多机组厂址PSA方法具有重要价值。     

Seismic safety of nuclear power plants in eastern Europe

This paper summarizes the work performed by the International Atomic Energy Agency in the areas of safety review and applied research in support of programmes for the assessment and enhancement of seismic safety in Eastern Europe and in particular, WWER type nuclear power plants during the past seven years. Three major topics are discussed; engineering safety review services in relation to external events, technical guidelines for the assessment and upgrading of WWER type nuclear power plants, and the Coordinated Research Programme on "Benchmark study for the seismic analysis and testing of WWER type nuclear power plants". These topics are summarized in a way to provide an overview of the past and present safety situation in selected WWER type plants which are all located in Eastern European countries. The main conclusion of this paper is that even though there is now a thorough understanding of the seismic safety issues in these operating nuclear power plants, the implementation of seismic upgrades to structures, systems and components are lagging behind, particularly for those cases in which re-evaluation indicated the necessity to strengthen the safety related structures or install new safety systems.     

内陆核电厂液态流出物排放口设计及环境影响评价初探

对美国内陆核电厂液态流出物的扩散器工程实践进行了概括和总结,根据《核动力厂环境辐射防护规定》(GB 6249—2011)的要求,在分析我国内陆核电厂液态氚排放面临较高的排放要求的基础上,对我国内陆核电厂液态流出物的排放设计及环境影响模拟进行了归纳和总结,并提出建议。成果可为内陆核电厂的扩散器研究提供参考。     

Safety assessment of nuclear power plant under volcanic phenomena – Background and technical basis of the revision of JEAG4625

Japan is one of the countries with abundant active volcanoes and has a long history of developing countermeasures to mitigate volcanic disasters. In the field of nuclear energy, it is also necessary to assess safety against volcanic hazards, and in 2009, a voluntary guideline was published as JEAG4625 in order to set up requirements of site assessments and basic designs of nuclear power plants (NPPs). This guideline has been revised to satisfy the requirements for examining the necessity of considering volcanic phenomena and concrete countermeasures in detailed designs of NPPs. This paper focuses on the background and technical basis of this voluntary guideline and shows the basic policy to ensure safety of NPPs and the requirements to prevent nuclear hazards due to volcanic phenomena based on the defense in depth concept.     

核电厂功率工况异常重要性判定模型的开发

核电厂异常重要性判定程序（SDP）是由美国核管会（NRC）首先使用的一种基于风险指引型的安全事项重要性判定工具。运用此方法,核安全监管人员对核安全相关事项进行筛选和评估,进而给出其风险重要性程度。本文对已有的PSA模型进行简化,开发了功率工况SDP第2阶段判定模型,对该模型进行了验证。验证结果表明,该模型满足核安全监管人员的使用要求,能够对核电厂事项进行快速有效的风险重要性判定。     

分散控制系统在核电厂的应用

文章简要介绍了分散控制系统(DCS)的特点,通过近几年DCS在我国核电厂的应用情况,并基于我国核电建设的发展现状,详细地论述了几种已经在核电厂中应用的DCS产品特点,并结合DCS在核电厂的应用,在DCS系统设计、安装、调试、维护等阶段分别提出了需要关注的要点。     

基于网络的核电站PSA数据库管理系统设计研究

核电站PSA数据库管理系统是进行核电站安全评估的手段之一.按照数据库规范设计法,设计了基于网络的核电站PSA数据库管理系统NDMS,NDMS能为PSA提供方便而充分的数据.根据核电站实际情况,NDMS采用了网络化设计,并设计了用户管理机制及智能编码规则管理机制.目前NDMS已作为PSA分析软件RiskA的数据库平台得到了较好的应用.