Seismic risk analysis and decisions for nuclear power plants

Seismic risk analysis and associated sensitivity studies constitute a part of the Seismic Safety Margins Research Program being conducted by the Lawrence Livermore National Laboratory for the US Nuclear Regulatory Commission. Although seismic risks are an important contributor to the total nuclear risk, the occurrence of earthquake-related seismic phenomena is low. Safety decisions involving seismic hazards must be made, however. This paper briefly described several categories of decisions that can be made using seismic risk analysis. While risk analysis does not provide all the information required for these decisions, it is a useful tool in that it provides additional information for the decision-making process. We anticipate a growing interest in the use of seismic risk analysis in nuclear safety evaluations.     

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.     

New generation of boiling-water-reactor nuclear power plants of increased safety

Scientific-Research Construction Institute of Power Engineering. Russian Scientific Center Kurchatov Institute. Special Design Office of Industrial Organization Izhorskii Plant. All-Union Scientific-Research and Design Institute of Power Engineering Scientific-Industrial Organization. Translated from Atomnaya Énergiya, Vol. 73, No. 1, pp. 13–19, July, 1992.     

Seismic isolation of nuclear power plants — EDF's philosophy

The elastomer bearing pads used since 1963 as supports for prestressed concrete pressure vessels (PCPVs) was quickly chosen by Electricité de France (ED) to improve the capability of nuclear power plants (NPPs) to withstand strong earthquakes and to reduce the seismic loads on structures and equipment. The standardized units for 900 and 1300 MW(e) pressurizedwater reactor (PWR) plants have moderate seismic design loads of 0.2 and 0.15 g, respectively. These design loads were exceeded by the site dependent spectra of Cruas (France) and Koeberg (South Africa). To keep the plant design unchanged and to take the advantages of standardization, these units were put on laminated bearings with or without sliding plates. For the future French 1500 MW(e) fast breeder reactors (FBRs), which are more sensitive to seismic loads, the base isolation is considered by EDF at the beginning of the design, even for low ground motions of 0.1 g. The buildings are placed on laminated bearings while the reactor block is supported by springs and dampers. The isolated plant has identical costs as a conventional design such as SPX1 at Creys—Malville.     

Operating procedure automation to enhance safety of nuclear power plants

Use of logic statements and computer assist are explored as means for automation and improvement on design of operating procedures including those employed in abnormal and emergency situations. Operating procedures for downpower and loss of forced circulation are used for demonstration. Human-factors analysis is performed on generic emergency operating procedures for three strategies of control; manual, semi-automatic and automatic, using standard emergency operating procedures. Such preliminary analysis shows that automation of procedures is feasible provided that fault-tolerant software and hardware become available for design of the controllers. Recommendations are provided for tests to substantiate the promise of enhancement of plant safety. Adequate design of operating procedures through automation may alleviate several major operational problems of nuclear power plants. Also, automation of procedures is necessary for partial or overall automatic control of plants. Fully automatic operations are needed for space applications while supervised automation of land-based and offshore plants may become the thrust of new generation of nuclear power plants.     

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.     

Recommended research for improving seismic safety of light-water nuclear power plants

Recommendations for research to improve the seismic safety of Light Water Reactors (LWR) are presented in this paper, based on analysis of the answers to a questionnaire returned by 55 persons or groups working in the area of seismic safety of nuclear power plants. In addition to the questionnaire results, the recommendations also include ideas expressed at a meeting of an ad hoc group of professionals, formed by Sandia Laboratories; a review of literature, current research programs, and engineering judgment.     

Systematic analysis of the safety of power-generating units of nuclear power plants

Scientific and Engineering Center of the State Nuclear Regulatory Agency of Russia. Translated from Atomnaya énergiya, Vol. 77, No. 5, pp. 344–350, November, 1994.     

Offshore nuclear power plants

Offshore Power Systems, a joint enterprise of Westinghouse and Tenneco, has been formed to manufacture floating nuclear power plants. Commitments for the first two offshore plants have been received from the Public Service Electric and Gas Company. This paper describes the floating nuclear plant concept with special reference to its advantages and its novel features. The novel features are a consequence of the floating aspect and include the design of the platform, the safety analysis and also the analysis and specification of plant motions due to environmental effects such as wind, waves and earthquakes. Site-related aspects such as the breakwater and mooring systems are discussed. The nuclear power plants will be manufactured in a central facility and this manufacturing concept is described.     

Research needs for improved seismic safety of mechanical equipment in nuclear power plants

Areas of research on shock absorbing multiple supported and buried piping of nuclear power plants are recommended.     

Seismic probabilistic safety analysis of unit 1 of the Loviisa nuclear power plant

The seismic probabilistic safety assessment consists of five phases. In the seismic hazard analysis the seismicity of the plant site is quantified. In the second phase, the structural response of plant buildings is evaluated. On the basis of structural response, the seismic fragilities of selected plant components are developed. In the following phase, the plant logic in the form of fault trees and event trees is established. In the last step, quantification of the core damage risk on the basis of the above information is carried out. For the median value of the annual core damage frequency, a value of 4.4 × 10⁻⁷ was determined.     

Seismic isolation for nuclear power plants: Technical and non-technical aspects in decision making

Technical aspects of seismic isolation systems show merit for their use in nuclear power plants. Less quantifiable non-technical aspects must be evaluated in the decision to employ a seismic isolation system.First, non-technical aspects are discussed. An historical and applications perspective is given, and it is suggested that the number of applications of seismic isolation systems is correlated with the amount of research activity in this area. For nuclear plants, it is suggested that application of seismic isolation systems is in part related to standardized plant designs in high seismic regions. Also, for nuclear plants, it is suggested that direct capital cost, enhanced seismic safety, regulatory licensing and unknown locations of nearby active faults are all factors which can weigh in favor and/or not in favor for seismic isolation application.Second, technical aspects are discussed. The technical results show that seismic isolation reduces building response, and reduces floor response spectra/equipment response. These results combine in application to reduce seismic risk and thus enhance safety for nuclear plants.     

Seismic response prediction for cabinets of nuclear power plants by using impact hammer test

An effective method to predict the seismic response of electrical cabinets of nuclear power plants is developed. This method consists of three steps: (1) identification of the earthquake-equivalent force based on the idealized lumped-mass system of the cabinet, (2) identification of the state-space equation (SSE) model of the system using input-output measurements from impact hammer tests, and (3) seismic response prediction by calculating the output of the identified SSE model under the identified earthquake-equivalent force. A three-dimensional plate model of cabinet structures is presented for the numerical verification of the proposed method. Experimental validation of the proposed method is carried out on a three-story frame which represents the structure of a cabinet. The SSE model of the frame is accurately identified by impact hammer tests with high fitness values over 85% of the actual frame characteristics. Shaking table tests are performed using El Centro, Kobe, and Northridge earthquakes as input motions and the acceleration responses are measured. The responses of the model under the three earthquakes are predicted and then compared with the measured responses. The predicted and measured responses agree well with each other with fitness values of 65-75%. The proposed method is more advantageous over other methods that are based on finite element (FE) model updating since it is free from FE modeling errors. It will be especially effective for cabinet structures in nuclear power plants where conducting shaking table tests may not be feasible. Limitations of the proposed method are also discussed.     

Evaluation of response factors for seismic probabilistic safety assessment of nuclear power plants

This paper presents a method for evaluating “response factors” of components in nuclear power plants for use in a seismic probabilistic safety assessment (PSA). The response factor here is a measure of conservatism included in response calculations in seismic design analysis of components and is defined as a ratio of conservative design response to actual response. This method has the following characteristic features: (1) the components are classified into several groups based on the differences in their location and in the vibration models used in design response analyses; (2) the response factors are decomposed into subfactors corresponding to the stages of the seismic response analyses in the design practices; (3) the response factors for components are calculated as products of subfactors; (4) the subfactors are expressed either as a single value or as a function of parameters that influence the response of components.This paper describes the outline of this method and results from an application to a sample problem in which response factors were quantified for examples of components selected from the groups.     

Managing aging in nuclear power plants

Aging degradation in nuclear power plants must be controlled to prevent safety margins from declining below limits provided in plant design bases. The NPAR Program and other aging-related programs conducted under the auspices of the NRC Office of Research are developing needed technical guidance for control of aging. Results from these programs, together with relevant information developed by industry and elsewhere, are implemented through various ongoing NRC and industry programs and initiatives as well as by means of conventional regulatory instruments. The aging control process central to these efforts consists of three key elements: (1) selection of components, systems, and structures (CSS) in which aging must be controlled, (2) understanding of the mechanisms and rates of degradation in these CSS, and (3) managing degradation through effective surveillance and maintenance. These elements are addressed in Recommended Practices Guidance that integrates information developed under NPAR and other studies of aging into a systems-oriented format that tracks directly with the Safety Analysis Reports and with the NRC Standard Review Plan (NUREG-0800).     

Selected safety aspects of containments for nuclear power plants with WWER-440 reactors

Considerable attention has been and continues to be focused on the design and operational features that prevent the release of radioactive materials to the environment for a spectrum of accidents for the two classes of WWER-440 reactors: the older 230 model and the more recently designed 213 models.This paper, based on published and unpublished information, aims to clarify the perceptions of the Russian WWER-440 models 230 and 213 Nuclear Power Plant containment system designs and their relevance to selected aspects of accident mitigation. It should be noted that these are unclearly and often negatively perceived, primarily because of a lack of reliable information and a poorly assembled experimental database. Conflicting statements have been made regarding the nature and the features of the plant's containment system. The paper presents a brief outline of the design of both WWER-440 models with respect to their confinement functions. Selected safety-related aspects of the accident localization systems are discussed, and the recognized shortcomings and safety merits are pointed out. The older 230 units experience high leak rates and are designed to withstand medium-size pipe breaks. The possible implications for safety are pointed out in the paper. The on going studies that concentrate on improving the system are highlighted. Some of the proposed modifications of the system, which would significantly decrease risks associated with accidents that are beyond the original design basis, are discussed. The design of the newer 213 model differs in many aspects. It incorporates the simple and original application of passive natural processes to limit the large-break loss-of-coolant accident post accident pressure. Other features of this containment design, such as complicated geometry, dependence on several mechanical devices and interlocks, and insufficient experimental evidence, lead to doubts concerning the operation of this containment under accident conditions. For the newer 213 model, current work is devoted mainly to safety assessment and verification of the containment design. Some information concerning the on-going work is provided in the paper.