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.     

Probabilistic assessment of seismic risk for nuclear power plants

Probabilistic approaches to the design, siting, and safety analysis of nuclear power plants have been proposed by Farmer, Wall, and Garrick. Farmer and Wall established a limit line which delineates between acceptable and unacceptable risks. To implement the method, all accidental chains are systematically analyzed to determine their probability and associated fission product release magnitude; the combination is compared to the limit line. For proper implementation, the seismic risk should be evaluated in a quantified manner. Conceptually, this evaluation is made in two stages: the probability of an earthquake occurrence as a function of its intensity and, given a seismic intensity, the conditional probability of damage. This paper reports on an initial study into the latter aspect.The effect of uncertainty in several parameters which determine the response of a nuclear reactor building to earthquake forces is assessed. Probability distributions for material properties were determined from site measurements and these distributions were utilized for determining the building response and the damage criterion. A subjective probability density function for damping was assigned from the available information and the judgment of experienced engineers. Four accelerograms, El Centro N---S 1940, and three artificial earthquakes were used to represent the variability in the forcing functions. The uncertainty in the model idealization was assessed by evaluating three alternate models. A versatile computer program was developed to compute the response of the mathematical model to the forcing functions using matrix formulation and modal method of analysis. An exact solution, rather than numerical integration, was used to obtain the dynamic response of the system in generalized coordinates.The stresses within the reactor building are similar for different earthquakes considered in this study when they are normalized to ground acceleration, indicating that the shape of the accelerogram and its frequency content are less significant than the magnitude of the maximum ground acceleration for the reactor building. The variation in modulus of elasticity for concrete had a significant effect on the building response. Damping, in general, reduced the response, but in cases where the duration of an earthquake is short the effect was not very significant.A simple failure criteria for ultimate shear stress in shear walls, τ_ult = 4.75 √ƒ′_c, where ƒ′_c is the ultimate compressive strength of concrete, is used to estimate the initiation of cracking in the walls. The normal design of the reactor building is deterministic and is based on a 0.2 g design basis earthquake. Using the results obtained by the parametric study on the variation of response, the probability of damage was estimated by a Monte Carlo analysis. It was estimated that, given the occurrence of a design basis earthquake, there is less than approximately 10⁻³ probability of cracking in the shear walls of the reactor building. The initiation of cracking in the concrete should not lead to a significant release of contained fission products.     

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.     

Success criteria analysis in support of probabilistic risk assessment for nuclear power plants: application on SGTR accident

Success criteria analysis (SCA) bridges the gap between deterministic and probabilistic approaches for risk assessment of complex systems.To develop a risk model,SCA evaluates systems behaviour in response to postulated accidents using deterministic approach to provide required information for the probabilistic model.A systematic framework is proposed in this article for extracting the front line systems success criteria.In this regard,available approaches are critically reviewed and technical challenges are discussed.Application of the proposed methodology is demonstrated on a typical Westinghouse-type nuclear power plant.Steam generator tube rupture is selected as the postulated accident.The methodology is comprehensive and general;therefore,it can be implemented on the other types of plants and complex systems.     

Success criteria analysis in support of probabilistic risk assessment for nuclear power plants: application on SGTR accident

Success criteria analysis (SCA) bridges the gap between deterministic and probabilistic approaches for risk assessment of complex systems.To develop a risk model,SCA evaluates systems behaviour in response to postulated accidents using deterministic approach to provide required information for the probabilistic model.A systematic framework is proposed in this article for extracting the front line systems success criteria.In this regard,available approaches are critically reviewed and technical challenges are discussed.Application of the proposed methodology is demonstrated on a typical Westinghouse-type nuclear power plant.Steam generator tube rupture is selected as the postulated accident.The methodology is comprehensive and general;therefore,it can be implemented on the other types of plants and complex systems.     

A comparison of background seismic risks and the incremental seismic risk due to nuclear power plants

The seismic risk for the continental United States, in terms of the expected annual number of deaths and severe injuries, and the expected property damage, is evaluated in this work. Probabilistic models and correlations are developed and used in the evaluations of the risks, accounting for such important variables as the variability of property values, damage factors and so on. In addition, the incremental seismic risk due to the presence of nuclear power plants is evaluated utilizing results and methods available in the literature. The results show that the incremental risk is generally very small compared to the background seismic risk, even if a very high probability for core melt is postulated.     

Probabilistic assessment of the seismic hazard for Eastern United States nuclear power plants

The purpose of the seismic hazard characterization of the Eastern United States project, for the Nuclear Regulatory Commission, was to develop a methodology and data bases to estimate the seismic hazard at all the plant sites east of the Rocky Mountains. A summary of important conclusions reached in this multi year study is presented. The magnitude and role of the uncertainty in the hazard estimates is emphasized in regard to the intended final use of the results.     

New seismic design spectra for nuclear power plants

Under a US Nuclear Regulatory Commission-sponsored project recommendations for seismic design ground motions for nuclear facilities are being developed. These recommendations will take several forms. Spectral shapes will be developed empirically and augmented as necessary by analytical models. Alternative methods of scaling the recommended shapes will be included that use a procedure that integrates over fragility curves to obtain approximately consistent risk at all sites. Site-specific soil effects will be taken into account by recommending site-specific analyses that can be used to modify rock hazard curves at a site. Also, a database of strong motion records will be archived for the project, along with recommendations on the development of artificial motions. This will aid the generation of motions for detailed soil- and structural-response studies.     

A sensitivity study of seismic structure-soil-structure interaction problems for nuclear power plants

A sensitivity study for the interaction effects of adjacent structures of nuclear power plants caused by horizontal seismic excitation has been performed. The key structural and soil parameters for linear and for nonlinear behaviour were varied within their applicable bandwidth. It has been shown that the interaction phenomena can contribute to the response of structures to such a large extent that it cannot be disregarded.     

Risk assessment of nuclear power plants

The objective of the present work is to develop recommendations for controlling the safety of nuclear power plants on the basis of risk assessments and safety certification of nuclear power plants. The Kursk nuclear power plant is considered as an example of a nuclear power plant with an RBMK reactor. The concept of risk assessment of a nuclear power plant consists in constructing a set of scenarios of the appearance and development of possible accidents followed by an evaluation of the realization frequency and determination of the scales of the consequences of each one. The result of an analysis is an evaluation of a system of risk indicators in accordance with the requirements of the safety compliance certificate of the nuclear power plant as well as the development of recommendations for increasing plant safety. In risk assessment, the consequences are divided into categories of the seriousness of the damage, for which their probability is evaluated separately. The graphical interpretation of risk due to any dangerous object consists of frequency–consequences curves. Recommendations are developed on the basis of the results of risk analysis.     

Issues in seismic isolation of nuclear power plants

The paper covers the issues involved in considering seismic isolation for nuclear plants. The application of isolation techniques to non-nuclear installations is discussed. Its potential application to nuclear components and plants is considered and the lack of actual, experimental verification of novel techniques is portrayed. Finally a cost comparison, based on certain preliminary assumptions of isolated and non-isolated nuclear plants is made.     

Seismic probabilistic risk analysis based on stochastic simulation of accelerograms for nuclear power plants in the UK

This article presents an approach to probabilistically assess the seismic risk of nuclear power plants (NPPs) in the UK. The approach proposed is based on direct stochastic simulation of the seismic input to conduct nonlinear dynamic analysis of a structural model of the NPP analysed. Therefore, it does not require the use of ground motion prediction equations and scaling/matching procedures to define suitable accelerograms as is done in conventional approaches. Additionally, as the structural response is directly calculated, it does not require the use of Monte Carlo-type algorithms to simulate the damage state of the NPP analysed. However, it demands longer use of computer resources as a relatively large number of nonlinear dynamic analyses are needed to perform. The approach is illustrated using an example of a 1000 MW Pressurised Water Reactor building located in a representative UK nuclear site. A comparison of risk assessment is made between the conventional and proposed approaches. Results obtained are reasonable and well constrained by conventional procedures; hence, it can confidently be used by the UK New Build Programme in the next two decades to generate 16 GWe of new nuclear capacity.     

RiskA计算引擎在核电站概率安全评价中的应用

研究分析了R＆R Workstation平台的计算引擎配置问题,实现了该平台下调用FDS团队自主研发的RiskA计算引擎.基于上述工作,采用真实核电站的概率安全评价模型,对RiskA计算引擎进行了正确性校核,并与CQUANT引擎进行了计算速度比较.测试表明正确实现了在PRAQUANT下对于RiskA计算引擎调用,Ris...     

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.     

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.     

Safeguard assessment for life extension in nuclear power plants (NPPs) using a production function

Life extension is investigated as a safeguard assessment for the stability on the operation of the nuclear power plants (NPPs). The Cobb-Douglas function, one of the production functions, is modified for the nuclear safeguard in NPPs, which was developed for the life quality of the social and natural objects. Nuclear Safeguard Estimator Function (NSEF) is developed for the application in NPPs. The cases of NPPs are compared with each other in the aspect of the secure performance. The results are obtained by the standard productivity comparisons with the designed power operations. The range of secure life extension is between 1.008 and 5.353 in 2000 MW_e and the range is between 0.302 and 0.994 in 600 MW_e. So, the successfulness of the power operation increases about 5 times higher than that of the interested power in this study, which means that the safeguard assessment has been performed in the life extension of the NPPs. The technology assessment (TA) is suggested for the safe operation which is an advanced method comparing conventional probabilistic safety assessment (PSA).     

Fuzzy probability on reliability study of nuclear power plant probabilistic safety assessment: A review

Fault tree analysis (FTA) is a graphical model which has been widely used as a deductive tool for nuclear power plant (NPP) probabilistic safety assessment (PSA). The conventional one assumes that basic events of fault trees always have precise failure probabilities or failure rates. However, in real-world applications, this assumption is still arguable. For example, there is a case where an extremely hazardous accident has never happened or occurs infrequently. Therefore, reasonable historical failure data are unavailable or insufficient to be used for statistically estimating the reliability characteristics of their components. To deal with this problem, fuzzy probability approaches have been proposed and implemented. However, those existing approaches still have limitations, such as lack of fuzzy gate representations and incapability to generate probabilities greater than 1.0E-3. Therefore, a review on the current implementations of fuzzy probabilities in the NPP PSA is necessary. This study has categorized two types of fuzzy probability approaches, i.e. fuzzy based FTA and fuzzy hybrid FTA. This study also confirms that the fuzzy based FTA should be used when the uncertainties are the main focus of the FTA. Meanwhile, the fuzzy hybrid FTA should be used when the reliability of basic events of fault trees can only be expressed by qualitative linguistic terms rather than numerical values.     

Application of thermal comfort theory in probabilistic safety assessment of a nuclear power plant

Human factor errors in probabilistic safety assessment(PSA) of a nuclear power plant(NPP) can be prevented using thermal comfort analysis.In this paper,the THERP+HCR model is modified by using PMV (Predicted Mean Vote) and PPD(Predicted Percentage Dissatisfied) index system,so as to obtain the operator cognitive reliability,and to reflect and analyze human perception,thermal comfort status,and cognitive ability in a specific NPP environment.The mechanism of human factors in the PSA is analyzed by operators of skill,rule and knowledge types.The THERP+HCR model modified by thermal comfort theory can reflect the conditions in actual environment,and optimize reliability analysis of human factors.Improving human thermal comfort for different types of operators reduces adverse factors due to human errors,and provides a safe and optimum decision-making for NPPs.