Recent extensions to seismic margin review methodology

Two seismic margin review methodologies — one by USNRC and the other by EPRI — have been developed in the last four years. The focus is on assessing the capability of existing nuclear power plants to withstand earthquakes larger than the design basis earthquakes. The methods restrict the analysis to a selected few systems and components using the insights from past seismic PRAs, seismic analysis and qualification results, and earthquake experience data. The objective of this paper is to describe recent and on-going studies in extending the NRC seismic margin review methodology. Specifically, three topics are discussed: (1) extension of the HCLPF capacity to analyse radiological releases and importance of human factors and non-seismic failures; (2) importance of BWR plant systems and functions to seismic margins; and (3) extensions of seismic margin review results to obtain seismic risk estimates.     

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.     

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.     

On estimated modes of failure of nuclear power plants by potential earthquakes

This paper deals with the estimated ‘modes of failure’ of nuclear power plants during future violent earthquakes. The authors have been surveying the damage to industrial plants caused by several violent earthquakes since 1960. Some of them have already been reported in English, but here the authors try to rearrange them from the viewpoint of ‘modes of failures’ of nuclear power plant buildings, equipment, vessels and piping. The authors categorize the mechanisms of failure as follows: (i) damaged by the dynamic effect of acceleration waves, (ii) by resonance in displacement waves, (iii) by the static effect of seismic force, (iv) by external force from attached piping and others, or forced deformation, and (v) by liquefaction of soil.The authors try to determine the modes of failure of the following items in a matrix form of the mechanisms: (i) the reactor building, (ii) steel containment vessel, (iii) auxiliary building, (iv) reactor vessel, (v) core internals, (vi) primary and secondary coolant system, (vii) emergency power supply system, (viii) emergency gas treatment system and stack, (ix) fuel cooling pond and fuel rack, (x) refuel machine crane, (xi) auxiliary system and component, (xii) turbine and its pedestal, and (xiii) main power system and control instrumentation. They also examine them from another point of view, i.e. in ‘the classification of the important factor’ of items for their aseismic design.     

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.     

我国核电厂选址标准研究走向的思考

简要回顾了核动力厂选址适宜性准则的建立与发展,介绍了核动力厂厂址评价的新进展,探讨了核动力厂选址标准研究的走向。在制定我国核动力厂选址标准中,应优先建立厂址适宜性准则,包括厂址安全、环境保护和应急准备的各种要素,以指导厂址评价。     

Experimental investigation into the seismic behavior of nuclear power plant shear wall structures

Nuclear power plant structures are designed to resist large earthquakes. However, as new data are obtained on earthquake activity throughout the United States, plant design earthquake levels have increased. The U.S. Nuclear Regulatory Commission is sponsoring an analytical-experimental research program to obtain information on the structural response of nuclear power plant shear wall structures subjected to earthquake motions within and beyond their design basis. Using different size scale models constructed with microconcrete and prototypical concrete this research has demonstrated consistent results for measured values of stiffness at load levels within the design basis. Furthermore, the values are well below the theoretical stiffnesses calculated from an uncracked cross-section strength-of-materials approach. Current program emphasis is to assess the credibility of previous experimental work by beginning to resolve the ‘stiffness difference’ issue.     

Underground construction of nuclear power reactors

This paper summarizes the main findings of a comprehensive study of the underground siting of nuclear power plants carried out at the Swiss Federal Institute for Reactor Research. Main aim of the investigations made was to identify suitable siting variants and to evaluate the feasibility, the safety potential and the cost of the concept. Two of the layouts developed for the main siting alternatives — the rock cavity alternative and the pit siting — are briefly described. In these designs an accident mitigation system based on the pressure relief concept, meant to reduce the consequences for the public and the environment in the case of extreme hypothetical events, has been proposed and an evaluation of its performances has been made to quantify the achievable risk reduction. The safety potential — especially under accident conditions — of this type of siting and the possibility that an underground plant may be exposed to new accident initiating events have been evaluated.From the technical point of view, an underground nuclear power plant is considered feasible while the economic penalty associated with the underground siting lies between 11 and 15%, according to the considered alternative.     

Summary report on new regulatory codes and criteria and guidelines: IAEA guidelines for WWER reactor coolant system integrity assessment

In the frame of the nuclear safety programme to assist the countries of Central and Eastern Europe, the IAEA identified and ranked in total 263 safety issues for WWER-440/230, WWER-440/213 and WWER-1000/320 nuclear power plants, related to both design and operation. In the area of reactor coolant system integrity, 24 safety issues were identified and 15 of them ranked as of high safety significance. These include: reactor pressure-vessel integrity and related aspects, primary and secondary high-energy piping integrity, steam generator integrity and reliability of the non-destructive testing for in-service inspection. In addition to obtaining international consensus on possible solutions to address the safety issues identified and to reviewing completeness of proposed safety improvements, the IAEA initiated development of guidelines to address the issues of highest safety concern. In the area of the reactor coolant system integrity, guidelines for the leak before break concept application, for the pressurized thermal shock analysis and for the in-service inspection systems qualification were developed. Further activities of the IAEA were focused on the implementation of guidelines developed in the Member States concerned. With this objective, a Co-ordinated Research Programme ‘Round-robin Exercise on WWER-440 Reactor Pressure Vessel Embrittlement, Annealing and Re-embrittlement’, a ‘WWER-440/213 Pressurized Thermal Shock Analysis Benchmark Exercise’ and a pilot study to implement the qualification approach to a real power plant component have been initiated by the IAEA and are well under way at present. In this paper, an overview of these IAEA activities related to reactor coolant system integrity is provided and the main principles and elements of guidelines developed discussed.     

核电厂大范围损伤管理导则研究现状

火灾、地震、水淹、极端天气条件等超设计基准外部事件可能造成核电厂大范围损伤,不仅使得电厂的系统与设备大面积失效,还导致正常的应急体系无法运转。目前,国内外都在开展应对核电厂大范围损伤的研究,以完善核电厂的纵深防御体系,降低大范围损伤事故产生的后果。本文调研了国内外核电厂大范围损伤的研究现状,分析了美国核电厂的大范围损伤管理导则、灵活多样的处理策略及台湾地区核电厂的断然处置措施,并对国内大范围损伤管理导则的研究与开发提出了一些思路与建议。     

Balancing safety and economics

The safety requirements of NPPs have always aimed at limiting societal risks. This risk approach initially resulted in deterministic design criteria and concepts. In the 1980s the paradigm ‘safety at all costs’ arose and often led to questionable backfitting measures. Conflicts between new requirements, classical design concepts and operational demands were often ignored. The design requirements for advanced reactors ensure enhanced protection against severe accidents. Still, it is questionable whether the ‘no-damage-outside-the-fence’ criteria can be achieved deterministically and at competitive costs. Market deregulation and utility privatisation call for a balance between safety and costs, without jeopardising basic safety concepts. An ideal approach must be risk-based and imply modern PSAs and new methods for cost–benefit and ALARA analyses, embed nuclear risks in a wider risk spectrum, but also make benefits transparent within the context of a broader life experience. Governments should define basic requirements, minimum standards and consistent comparison criteria, and strengthen operator responsibility. Internationally sufficient and binding safety requirements must be established and nuclear technology transfer handled in a responsible way, while existing plants, with their continuous backfitting investments, should receive particular attention.     

Main achievements of FP-4 research in reactor safety

This summary paper deals with the strategy, the organisation and main achievements of the 67 multi-partner projects cosponsored by the European Union (EU) as ‘indirect actions’ (shared-cost and concerted actions) and co-ordinated under seven clusters, each devoted to one key safety issue in nuclear reactor safety, as they were presented at FISA-99. The fundamental safety objective for nuclear power plants (NPPs) consists in protecting the public and the environment from the harmful effects resulting from ionising radiations. The 4th Euratom framework programme 1994–1998 (FP-4) has been carrying out research with this objective both through ‘indirect actions’ and through ‘direct actions’ in co-operation with the Joint Research Centre of the European Commission (EC). The total cost of this research programme was 62.8 million, out of which 34.2 million was contributed by the EU budget. Besides technological safety requirements, socio-economic aspects are becoming increasingly important due to the level of public and political acceptance and to the economic pressure of deregulated electricity markets. It is shown that research conducted in the Euratom framework may contribute to meet these requirements, thereby maintaining nuclear power as a competitive and sustainable option for the energy policy of the European Union.     

On the basic research of design analysis and testing based on the failure rate for pipings and equipment under earthquake conditions

This paper deals with the evaluation method of the failure rate of pipings and equipment of nuclear power plants under destructive earthquakes and a new design concept in this stand point of view. These researches are supported by various studies related to this subject, which have been done by the author since 1966. In this paper, the history of the development, the summaries of these studies and their significances to the practice will be described briefly.The surveys on damages of industrial facilities caused by recent destructive earthquakes are the basical study for this subject. And the continuous response observation of model structures of a plant complex to natural earthquakes is another important basic study to know the stochastic nature and significance of response analysis for the anti-earthquake design of nuclear power plants.By having the exact knowledges on these subjects, the author has been developing the evaluation procedure of the failure rate of pipings and equipment under destructive earthquake conditions, a new design method ‘counter-input design’ and others. Now his effort is going towards to establish their practical procedure after finishing the basic researchers.     

Realistic seismic qualification using the updated finite element model for in-core components of reactors

It is now mandatory to seismically qualify the safety-related structures and components used in the nuclear power plants. Among several qualification approaches the qualification by the analysis using finite element (FE) method is the most common approach used in practice. However, the estimated dynamic behaviour by FE model of a structure is known to show significant deviations from the dynamic behaviour of the ‘as-installed’ structure in many cases. Considering such limitations, few researchers have advocated re-qualification of such structures after their installation at site to enhance the confidence in qualification vis-à-vis plant safety. For such an exercise, validation of FE model with experimental modal data is important. A validated FE model can be obtained by the model updating methods in conjugation with the in situ experimental modal data. Such a model can then be used for qualification. However, for the reactor in-core components such a modal testing and FE model updating may not be straightforward. Hence, the complication involved in the reliable seismic qualification of in-core components and the advantage of using the FE model updating has been brought out in the paper through an example of a typical in-core component—a perforated horizontal tube recently installed in a nuclear reactor in India.     

Methods and requirements for seismic design in the United Kingdom

The United Kingdom is in an area of low but significant seismicity compared with the more active areas of the world where there are major active faults or tectonic plate boundaries. This paper presents the methods and requirements that are adopted to consider the extreme load in the design of nuclear facilities. In the United Kingdom, detailed procedures for demonstrating seismic adequacy are not specified by the nuclear licensing authority and as such the methods described in this paper are based on precedents arising from recent licensing applications. In presenting the method and requirements, the paper discusses the applicability of simplified methods for seismic qualification for both “new” and “existing” facilities. The paper concludes that simplified methods are applied to a significant extent for demonstrating the adequacy of existing plant. However, for new plant these methods have been limited in some cases to the evaluation of design loads and to the qualification of items where the required degree of assurance is less than that associated with formal qualification and for supporting studies which do not directly affect design. It is expected that as the body of experience in earthquake engineering develops in the United Kingdom, there will be a greater tendency to adopt more simplified procedures with a greater degree of confidence.     

Power plant siting: A paretian environmental approach

Several different interest groups are involved in the siting decision process for nuclear power plants. Their inputs and how the final decision affects them are usually not explicitly included in mathematical analyses. Paretian analysis is a technique which attempts to identify the preferences of each of the interest groups involved in the decision-making process, and to illuminate the trade-offs among these groups. The specific problem studied is the deployment by a regional power plant siting agency of nuclear base-load power plants to coastal sites in New England. Four interest groups participate in power plant siting decisions; utility functions can represent their preferences. Techniques to find Pareto-admissible decisions are described.     

Automated generation of spectrum-compatible artificial time histories

Criteria for the seismic design of nuclear power plants are usually defined in the form of response spectra, and it is often necessary to generate artificial time histories of ground motions to ‘match’ these spectra. A recently developed automated iterative procedure employing a combination of frequency- and time-domain techniques in any desired sequence for the development of spectrum-compatible artificial time histories is presented. The procedure employs four basic steps: (1) generation of an initial (staring) time history using either sinusoidal superposition with an envelope function or specification of a real time history of a recorded ground motion; (2) manipulation of the amplitude and phase of the Fourier transform representation of this time history, and generation of successive time histories which have response spectra converging to the target design spectrum; (3) manipulation of the amplitude and phase of the Fourier transform representation only locally in areas where the peaks of the computed spectrum have larger magnitudes of deviations than desirable; and (4) manipulation of local areas of the latest time history. Computer program EDAC/SEQGEN, which completely automates the above steps, is described. Results are presented for the development of two artificial time histories with corresponding spectra matching the target USNRC Regulatory Guide 1.60 response spectra within about 5–7%.     

A proposal for an aseismic design method of equipment and piping for NPPs in a low seismicity area

Recently a regulatory code for an aseismic design of high-pressure gas facilities became effective by the order of the Ministry of International Trade and Industry (MITI) in Japan. This order includes details of the aseismic design of vessels whose “factor of importance” are relatively lower than Class A (Class I) items in nuclear power plants.The author develops his idea on an aseismic design method of equipment and piping of nuclear power plants in a Low Seismicity Area (LSA) based on his experience of the new code for petro-chemical industries and oil refinaries pertaining to high pressure gas facilities mentioned above.The definition of LSA is usually the area whose maximum intensity has never exceeded MMI VI or VII. However, there are two types of LSA, one is really such a low seismicity area, and the other type is the area which has the possibility of stronger earthquake occurrence than those mentioned above, even though it is low. One of the typical examples is the area subjected to “New Madrid Earthquake-1812”. The author develops his concept along these two lines.He briefly describes the new code for high-pressure gas facilities in Japan. This code describes the design methodology of both types of aseismic design analysis, that is, rather sophisticated dynamic methods for facilities whose potential hazard is as high as those in a nuclear power plant, such as liquified chlorine gas storage, and simplified dynamic and static methods for most of the equipment and vessels in those plants. One of the features of this code is the use of design formulae and charts to simplify their design procedure as well as the set of specific computer codes by the MITI. These computer codes are prepared by the MITI or approved by the MITI for providing equivalent capability to the practice designated in the MITI order.The author's philosophy for the code of equipment and pipings in LSA is that they must be as simple as possible, and most of the analytical work for the design should be eliminated, or at least limit the use of simplified methods, such as the static seismic coefficient method or the modified seismic coefficient method with a simplified response spectrum. The use of general design criteria or a guideline of structural details may be better than a sophisticated design analysis as a result.     

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.     

The swedish underground containment studies: State of art

Since 1973 studies of underground siting for nuclear power plants have been going on in Sweden. War protection, being the primary aim in accordance with the instructions, the first containment study has lead to siting in rock or in a pit. Rock siting gives better war protection than pit siting and also has less effect on the landscape, the cost being about equal. The second study was aimed at surveying the advantages and disadvantages of a rock sited 1000 MW BWA nuclear power plant from a reactor safety standpoint, compared to a plant above ground.Based on the instructions and considerations within the study group, the following criteria for the plant design have been established. (1) The plant should be designed to give protection against external acts of war with conventional weapons. (2) The plant should have a safety level equal to that of an above ground plant. It should fulfil the demands set by the authorities for above ground plants with respect to normal operation and accidents. No accidents that can be dealt with above ground may be permitted to result in more serious consequences, nor may they have a higher probability in a plant sited in rock. (3) The design of the plant should moreover utilize the possibilities of improving the safety afforded by rock siting. The criterion about war protection leads to siting in rock or pit, as shown in a previous CDL study. The study group has concentrated its work on rock siting.To clarify the two other criteria, the study group has outlined four alternative designs of a rock sited plant: (1) Reactor with complete pressure suppression (PS) containment placed together with central auxiliary equipment in a closed cavern with 48 m span that is placed about 50 m under the rock surface. (2) Reactor with complete PS containment placed together with auxiliary equipment in similar cavern as alternative (1) but open to the atmosphere. (3) Reactor placed ment. (4) Reactor placed in a containment directly surrounded by the rock. Auxiliary equipment placed in separate caverns.Standardization and quality improvement today are preferred to choosing new systems and more advanced technical solutions. Also, considering the desire from a safety standpoint, a rock sited plant should as much as possible exploit established technology. A consequence of the desire to use established technology is that the reactor cavern should be open to the atmosphere. If the cavern is closed, certain pipe rupture accidents may give overpressures that are difficult to master without large design alterations. The criterion about utilizing the possibilities for increasing the safety leads the interest to extreme improbable accidents, where certain advantages seem to be attainable with rock siting.A tight, strong cavern around the reactor would thereby be an ideal solution. This design appears, however, difficult to combine with the criterion about equal safety level as above ground. This criterion that controls the safety in the circumstances normally considered for nuclear power plants must be satisfied primarily, since extreme accidents have such very low probability. The tight cavern has therefore had to stand back for the open. The study shows, however, that an open reactor cavern can also be designed to significantly increase the protection of the surroundings in extreme improbable accidents compared to above ground plants.The chosen technical design of the reactor plant demands a cavern with a 45–50 m span. Caverns without strengthening efforts with such spans are used in mines, but have not previously been used for industrial plants. Studies of the stability of such caverns show that a safety level is attainable corresponding to the safety required for the other parts of the nuclear power plant. The conditions are that the rock is of high quality, that necessary strengthening measures are taken and that careful studies of the rock are made before and during the blasting, and also during operation of the plant.The third study was delivered to the government in 1977. One part in this study is going deeper in certain questions (safety, operation, maintenance, sabotage, war protection, cost and decommissioning). Another part aims to a broader view of risks and consequences in peace and war and also advantages and disadvantages of nuclear power plant for district heating.