Regulatory perspective of seismic issues

Some of the current seismic issues facing the nuclear power industry, such as seismic design criteria (USI A-40), seismic qualification of equipment in operating nuclear power plants (USI A-46), eastern United States seismicity, operating basis earthquake (OBE) exceedance criteria, seismic instrumentation, post OBE inspection of nuclear power plants, anchor bolts too close to a free edge, seismic margins of plants, and the potential for external events to cause severe accidents, are presented and the Nuclear Regulatory Commission's perspective on the resolution of these issues are discussed.     

Seismic individual plant examination of external events of US nuclear power plants: insights and implications

As part of the implementation of the severe accident policy, nuclear power plants in the US are conducting the individual plant examination of external events (IPEEE). Seismic events are treated in these IPEEEs by either a seismic probabilistic risk assessment (PRA) or a seismic margin assessment. The major elements of a seismic PRA are the seismic hazard analysis, seismic fragility evaluation of structures and equipment and systems analysis using event tree and fault tree analysis techniques to develop accident sequences and calculate their frequencies of occurrence. The seismic margin assessment is a deterministic evaluation of the seismic margin of the plant beyond the design basis earthquake. A review level earthquake is selected and some of the components that are on the success paths are screened out as exceeding the review level earthquake; the remaining ones are evaluated for their seismic capacity using information from the original plant design criteria, test data and plant walkdown. The IPEEEs of over 100 operating nuclear power plants are nearing completion. This paper summarizes the lessons learned in conducting the IPEEEs and their applicability to nuclear power plants outside of the United States.     

Seismic reevaluation and upgrading of nuclear power facilities outside the US using US developed methodologies

Several seismic licensing and safety issues have emerged over the past fifteen years for commercial US Nuclear Power Plants and US Government research reactors, production reactors and process facilities. The methodologies for the resolution of these issues have been developed in numerous government and utility sponsored research programs. The resolution criteria have included conservative deterministic design criteria, deterministic seismic margins assessments criteria (SMA) and seismic probabilistic risk assessment criteria (SPRA). The criteria for SMAs and SPRAs have been based realistically on considering the inelastic energy absorption capability of ductile structures, equipment and piping and have incorporated the use of earthquake and testing experience to evaluate the operability of complex mechanical and electrical equipment. Most of the applications to date have been confined to the US, however there have been several applications to Asian, Western and Eastern Europe reactors. This paper summarizes the major issues addressed, the development of reevaluation criteria and selected applications to non US reactors including VVER reactors of Soviet origin.     

Nuclear power plant seismic design—A review of selected topics

This paper presents a background review of the basis for the current seismic design criteria employed in the United States with particular attention given to the so-called double earthquake approach to seismic design. This paper also provides details of approaches used in other countries, namely Canada and Japan, which at least in part do not use the two earthquake concept in design.The paper begins with a brief presentation of background material relative to the approach employed in the seismic design of nuclear plants in the 1960's along with comments on the development of the current procedures. The next section contains a brief discussion of the criteria contained in Appendix A of 10CFR100 which today largely governs the seismic design of nuclear power plants in the U.S. The last section discusses effective versus instrument acceleration in design and observations pertaining to other approaches that might be employed in terms of selecting and carrying out the seismic design.     

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.     

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.     

Philosophy and practice of the aseismic design of nuclear power plants — Summary of the guidelines in Japan

Aseismic design is considered to be one of the most important factors for the safety of the nuclear power plants built in zones of high seismicity such as Japan. All structures, equipment and piping are classified in accordance with the importance of their radioactive safety to the plant, and the dynamic analysis and/or factored seismic coefficient analysis are applied accordingly. Site and ground conditions, as well as seismicity, should be studied thoroughly in order to estimate the intensities of the design earthquake and the safety margin check earthquake. Dynamic analyses of buildings and structures are performed using the multi-lumped-mass-system supported by soil springs with time history analysis conceptions. This idea is also applied to the design of equipment and piping by coupled system to the major structure or by the floor response spectra criteria. Tolerances are applied to damping factors although some experiments show more realistic results. Allowable stresses of ferrous metals for equipment and piping during earthquakes are more scientifically precise.

This report summarizes a guideline for aseismic design of nuclear power plants. The guideline was prepared by the Japan Electric Association in April, 1970, after three years laborious work.

In sect. 1, the philosophy and criteria are described. All components of a plant should be classified into three classes in accordance with their contributions to reactor safeties. Design to earthquake loadings should be based on “design basis earthquake” which is decided in consideration of local seismicity.

In sect 2, site selection and review for ground are described in the sense of seismic aspects.

In sect 3, deciding the earthquake motion for design is discussed. In Japan, semi-statistical approaches are used in normal practice.

In sect. 4, design philosophy and practice of building structures and containment vessels are described. They are designed under statical seismic forces, and the design of the class “A” structures should be checked by a dynamic response technique.

In sect. 5, design philosophy and practice of piping, vesels and equipment are described. Those which belong to class “A” items should be designed in a dynamic sense. Several programs for dynamic analyses of these items are prepared. Allowable stress under earthquake conditions is discussed in relation to other codes, for example, ASME Section III.

The greater part of the philosophy and design criteria have been adopted to all nuclear power plants which have been and are currently being built in Japan.     

关于核设施流出物监测和环境监测中存在问题的探讨

核设施流出物监测和环境监测体系是核设施安全体系的重要组成部分,随着我国核电建设的不断发展,监测技术和能力也得到了长足的发展,但仍然存在一些问题。通过对低水平监测中存在的问题的分析,并调研美国和欧盟对此问题的处理方法,给出我国解决监测中探测限问题的思路和建议。     

Seismic probabilistic risk assessment and its impact on margin studies

In recent years a number of seismic probabilistic risk assessments of nuclear power plants have been conducted. These studies have highlighted the significance of seismic events to the overall plant risk and have identified several dominant contributors to the seismic risk. It has been learnt from the seismic PRAs that the uncertainty in the seismic hazard results contribute to the large uncertainty in the core damage and severe release frequencies. A procedure is needed to assess the seismic safety of a plant which is somewhat removed from the influence of the uncertainties in seismic hazard estimates. In the last two years, seismic margin review methodologies have been developed based on the results and insights from the seismic probabilistic risk assessments. They focus on the question of how much larger an earthquake should be beyond the plant design basis before it compromises the safety of the plant. An indicator of the plant seismic capacity called the High Confidence Low Probability of Failure (HCLPF) capacity, is defined as the level of earthquake for which one could state with high confidence that the plant will have a low probability of severe core damage. The seismic margin review methodologies draw from the seismic PRAs, experience in seismic analyses, testing and actual earthquakes in order to minimize the review effort. The salient steps in the review consists of preliminary screening of components and systems, performance of detailed seismic walkdowns and evaluation of seismic margins for components, systems and plant.     

Seismic protection technology for nuclear power plants: a systematic review

Seismic protection systems (SPS) have been developed and used successfully in conventional structures, but their applications in nuclear power plants (NPPs) are scarce. However, valuable research has been conducted worldwide to include SPS in nuclear engineering design. This study aims to provide a state-of-the-art review of SPS in nuclear engineering and to answer four significant research questions: (1) why are SPS not adopted in the nuclear industry and what issues have prevented their deployment? (2) what types of SPS are being considered in nuclear engineering research? (3) what are the strategies for location of SPS within NPPs? and (4) how may SPS provide improved structural performance and safety of NPPs under seismic actions? This review is conducted following the procedures of systematic reviews, where possible.

The issues concerning the use of SPS in NPPs are identified: cost, safety, licensing and scarcity of applications. NPPs demand full structural integrity and reactor's safe shutdown during earthquake actions. Therefore, horizontal isolation may be insufficient in active seismic zones and isolation in the vertical direction may be required. Based on the results in this review, it is likely that next generation reactors in seismic zones will include state-of-the-art SPS to achieve full standardised design.     

Seismic IPEEE industry insights

The United States Nuclear Regulatory Commission initiated a formal review of the seismic margin of all operating nuclear power plants in the US with the issuance in 1991 of Generic Letter 88-20, Supplement 4 (‘Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities’). Virtually all of the US nuclear utilities have submitted their responses for seismic IPEEE and these submittals are in the process of being reviewed by the NRC. The objective of this paper is to provide an industry perspective on the results and the insights obtained from the utility seismic IPEEE submittals.     

利用HHT产生与多阻尼楼板谱匹配的地震波

美国原子能管理委员会（USNRC）规范规定了用于核电厂抗震分析和设计的地震波要求。在抗震分析和设计中，采用的地震波可与多阻尼目标反应谱匹配，也可与单阻尼目标反应谱匹配。然而，在对核电设备和部件进行动力时程分析时，则需要与多阻尼目标楼板谱匹配的地震波。基于此问题，提出利用希尔伯特-黄变换（HHT）方法，通过修改种子地震波的频率和振幅信息，使之与多阻尼目标楼板谱匹配，且完全符合USNRC规范的匹配标准，从而为核电设备和部件的地震安全评估提供合适的地震激励。      

The European pressurized water reactor Result of the French–German cooperation of experienced NPP suppliers and operators

In 1989 Framatome and Siemens, the two most experienced European nuclear power plant suppliers, decided to join the efforts for the development of a new reactor type for the next generation in their equally owned subsidiary Nuclear Power International (NPI). In 1992 Electricité de France and the major German utilities operating nuclear power plants merged their own development programs with that of Nuclear Power International and initiated the European Pressurized Water Reactor (EPR) project. In order to reach the two major targets of the project, the licensability in both countries, France and Germany, and the competitiveness of nuclear energy with other alternative energy sources, the design basis which had differently developed in the two countries needed to be harmonized. In parallel, the licensing authorities of both countries extended their existing cooperation in the field of a safety survey of existing nuclear power plants to the definition of safety criteria for the next generation of nuclear power plants. Through this cooperation the licensability of EPR in France and Germany will be assured. Continuously performed cost analysis show in addition that also the second target of the project, the competitiveness with alternative primary energy sources, can be achieved. Thanks to the fruitful cooperation between all parties involved, satisfactory results have been achieved not by a simple superposition of existing design features but through a careful evaluation and combination of the best available alternatives. At the end of 1997 the basic design results were compiled in a final report. Subsequently an optimization phase was launched that further improves the competitiveness of the power generation costs.     

Lifetime integrity requirements for containments in the United States

Assuring the lifetime integrity of containment structures for nuclear power plants is becoming increasingly important as existing design criteria are reexamined, as new requirements for containment inspection and testing are formulated, and as today's operating plants grow older. Regulatory requirements for containments in the United States are contained in the Code of Federal Regulations and in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. New requirements for the preservice examination and inservice inspection of Class MC (metal) containments have been published in Section XI, Subsection IWE, of the ASME Boiler and Pressure Vessel Code. Requirements for Class CC (concrete) containments have been published in Subsection IWL. Federal regulations that will require mandatory compliance with the ASME Code by nuclear plant owners in the United States are forthcoming. Parameters for extending the life of current United States nuclear plants beyond the 40 year design lifetime are presently being formulated. Two pilot plant life extension (PLEX) studies, one for a BWR and one for a PWR, serve as models for extending the life of today's aging plants.This paper presents an overview of the regulatory requirements for containments in the United States and the new preservice examination, inservice inspection and related requirements for Class MC and Class CC containments in the ASME Boiler and Pressure Vessel Code. Operational and life extension considerations for containment structures, including the findings from pilot studies of typical PWR and BWR containments, are also discussed. Together, the regulatory and Code requirements and recommendations from the plant life extension studies provide a basis for improved lifetime integrity for containments in the United States.     

基于经验-解析数据的某核电安全壳地震易损性分析

本文基于混合数据的地震易损性分析方法,对我国已运行核电厂地震易损性分析进行研究。首先基于地震危险性分析和分解结果,生成了我国华南地区某核电厂厂址条件谱;然后采用贪心优化算法,选取符合厂址危险性的地震动记录;基于增量动力分析方法,生成我国某核电厂安全壳地震易损性安全系数FS和FSA的解析数据;地震易损性其他参数采用经验数据,基于经验-解析数据,生成了我国某核电厂安全壳地震易损性曲线。建议将基于经验-解析数据的地震易损性分析方法应用于我国核电厂安全壳初步地震易损性分析中。     

Design of nuclear facilities other than NPPs in relation to external events

The safety of a nuclear installation requires in general that it is sited, designed, constructed and operated to protect individuals, society and the environment against an uncontrolled release of radioactivity. External events, both natural and human-induced, play a major role in challenging the plant defense. Therefore, appropriate design provisions are needed to assure an adequate safety margin in case of such events.In recent years, the development of design criteria, design methodologies and assessment approaches for external events received major emphasis for nuclear power plants. Other nuclear installations, however, received less attention even though their radioactive inventory may be quite significant (sometimes comparable with the NPPs, like in the case of some research reactors or fuel re-processing plants). Also the risks for radiological (and chemical) contamination is often rather high (as in the case of the fuel re-processing plants) and their location may be very close to densely populated areas.There is a lack of generally accepted international standards in this field and the direct application of general safety principles for nuclear installations is not straightforward.The IAEA addressed this issue in the past several years. Some of the results have been collected in an IAEA Technical Document (TECDOC).This paper is intended to address some of the main issues that have been identified and discussed in the referenced document.     

Recent applications of PSA for managing nuclear power plant safety

The safety design and regulation of nuclear power plants has traditionally been based upon deterministic approaches that consider a set of challenges to safety, e.g. design basis accidents, and determine how those challenges should be handled. The approach has been very successful since no plant designed or regulated to United States standards has ever harmed a member of the public. The arbitrary nature of these safety criteria, the potential inconsistencies in the judgments on relative probabilities, and the lack of definition for ‘safety’ became increasingly evident during the 1960s. Probabilistic approaches to reactor safety were proposed ^1,2,3 but did not take off in the United States until publication of the Reactor Safety Study ⁴ in 1975. Even as the methodology matured, there remained a challenge to integrate it into the regulatory process. This article will describe this integration process. A probabilistic approach to regulation enhances and extends the traditional deterministic approach by introducing the concept of safety (risk) significance that allows the designer/operator to focus on important issues. Emphasis was initially placed on relative risk but now regulatory decision-making is employing both relative and absolute risk. Measures of importance will be defined. Risk information can be used to prioritize the allocation of resources and three examples will be described. Equipment configuration control systems are being installed and used at nuclear power plants to enhance safety and to reduce Operating and Maintenance costs; they will be described. Finally, the US Nuclear Regulatory Commission's introduction of risk-informed decision-making into the regulatory process will be discussed.     

Aging management of containment structures in nuclear power plants

Research is being conducted by Oak Ridge National Laboratory under US Nuclear Regulatory commission (USNRC) sponsorship to address aging management of nuclear power plant containment and other safety-related structures. Documentation is being prepared to provide the USNRC with potential structural safety issues and acceptance criteria for use in continued service evaluations of nuclear power plants. Accomplishments include development of a Structural Materials Information Center containing data and information on the time variation of 144 material properties under the influence of pertinent environmental stressors or aging factors, evaluation of models for potential concrete containment degradation factors, development of a procedure to identify critical structures and degradation factors important to aging management, evaluations of non-destructive evaluation techniques, assessments of European and North American repair practices for concrete, review of parameters affecting corrosion of metals embedded in concrete, and development of methodologies for making current condition assessments and service life predictions of new or existing reinforced concrete structures in nuclear power plants.     

福清核电厂1、2号机组抗震裕量评价

地震是核电厂主要外部灾害之一,地震风险评估对于核电厂的安全评价具有重要的价值。抗震裕量评价(SMA)是开展核电厂地震灾害风险分析的重要方法之一,其目的是为了判断核电厂的抗震设计能力相对于设计基准地震的抗震裕量,找出核电厂的抗震薄弱环节,提高核电厂的抗震能力。本文针对福建福清核电厂1、2号机组进行抗震裕量评价,分析表明电力支持系统和一回路辅助管道的抗震能力相对薄弱,是导致核电厂抗震能力薄弱的主要原因,电力支持系统和一回路辅助管道需进一步提高其抗震能力,且核电厂需考虑编制地震应急规程。     

The evolution of the structure and application of U.S. NRC regulations and standards

NRC regulations and standards and their implementation have evolved from early adaptations of conventional engineering practices to a mature, cohesive set of regulations that govern NRC regulation of nuclear power plant safety in the United States.From a simple set of rules and design criteria and from the standards of the professional engineering societies, a hierarchy of practices, standards, guides, rules and goals has developed. Resting on a foundation of industrial practices, this hierarchy rises through levels of national standards, regulatory guides and standard review plans, policy statements and NRC regulations.The licensing process is evolving today toward one that permits both site approval and standard design certification before the plant is constructed. At the present time, NRC is reviewing five standard designs for certification for a period of 15 years. NRC focuses its regulation of operating nuclear plants on inspections conducted from five regional offices. Resident inspectors, specialist inspectors, and multi-disciplinary inspection teams examine specific plant situations. The results of all these inspections are used to develop a complete understanding of a plant's physical condition, its operation, maintenance and management.To improve safe operation of nuclear plants in the U.S., a most important program, the Systematic Assessment of Licensee Performance, measures operational performance, using a broad spectrum of functional areas.