The NUREG-1150 probabilistic risk assessment for the Grand Gulf Nuclear Station

This paper summarizes the findings of the probabilistic risk assessment (PRA) for Unit 1 of the Grand Gulf Nuclear Station performed in support of NUREG-1150. The emphasis is on the “back-end” analyses, that is, the accident progression, source term, consequence analyses, and risk results obtained when the results of these analyses are combined with the accident frequency analysis. The offsite risk from internal initiating events was found to be quite low, both with respect to the safety goals and to the other plants analyzed in NUREG-1150. The offsite risk is dominated by short-term station blackout plant damage states. The long-term station blackout group and the anticipated transients without scram (ATWS) group contribute considerably less to risk. Transients in which the power conversion system is unavailable are very minor contributors to risk. The low values for risk can be attributed to low core damage frequency, good emergency response, and plant features that reduce the potential source term.     

PRAAGE88: An interactive IBM-PC code for aging analysis of NUREG-1150 systems

The effects of time dependent failure rates caused by the aging of components are becoming increasingly important in probabilistic risk assessment (PRA) and reliability analyses of nuclear power plant systems. In the NRC Nuclear Plant Aging Research (NPAR) program, the effects of aging in nuclear systems are being evaluated through the use of time varying failure rates that are determined as a function of the age of the system. These analyses involve complex systems and include various sensitivity studies; thus, the PRAAGE88 computer code was developed to facilitate these calculations. PRAAGE88 is an IBM PC based code that computes system unavailability, component unavailability, and various importance measures for use in evaluating the effect of aging on reactor systems. This paper describes the methodology utilized in the code, its capabilities and areas of application.     

Methodology and results of the seismic probabilistic safety assessment of Krško Nuclear Power Plant

A seismic IPEEE (Individual Plant Examination for External Events) was performed for the Kr

Full-size image

ko plant. The methodology adopted is the seismic PSA (probabilistic safety assessment). The Kr

Full-size image

ko NPP is located on a medium to high seismicity site. The PSA study described here includes all the steps in the PSA sequence, i.e. reassessment of site hazard, calculation of plant structures response including soil–structure interaction, seismic plant walkdowns, probabilistic seismic fragility analysis of plant structures and components, and quantification of seismic core damage frequency (CDF). Relay chatter analysis and soil stability studies were also performed. The seismic PSA described here is limited to the analysis of CDF (level 1 PSA). The subsequent determination and quantification of plant damage states, containment behaviour and radioactive releases to the outside (level 2 PSA) have been performed for the Kr

Full-size image

ko NPP but are not further described in this paper. The results of the seismic PSA study indicates that, with some upgrades suggested by the PSA team, the seismic induced CDF is comparable with most US and Western Europe NPPs located in high seismic areas.     

Decision making and probabilistic risk assessment

Several questions have been posed, in advance, to members of a panel on Decision Making held at the International Post-Conference Seminar on the Role of Data and Judgment in Probabilistic Risk and Safety Analysis, (SMiRT-8, August 1985). The relationship between probabilistic risk assessment and decision making is discussed with emphasis on: the role of judgment; verification of results of probabilistic risk assessments; the review process; safety goals and standards.     

WWER-1000 shutdown probabilistic risk assessment: An introductory insight

The probabilistic risk assessment has become an essential part of every plant’s safety/risk analysis. In the past, regulations and safety assessments of nuclear power plants were traditionally concentrated on full power operation. However, because of the events that might potentially occur during low power and shutdown modes, the assessment of risk at these operational modes is now gaining more importance.     

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.     

Uncertainty propagation in probabilistic risk assessment: A comparative study

Three uncertainty propagation techniques, namely method of moments, discrete probability distribution (DPD), and Monte Carlo simulation, generally used in probabilistic risk assessment, are compared and conclusions drawn in terms of the accuracy of the results. For small uncertainty in the basic event unavailabilities, the three methods give similar results. For large uncertainty, the method of moments is in error, and the appropriate method is to propagate uncertainty in the discrete form either by DPD method without sampling or by Monte Carlo.     

On the application of probabilistic risk assessment to reactors with inherently safe features

In this paper, it is shown that because of the public perception of the risk of nuclear power and the likelihood that in the event of a severe core damage accident in a reactor claimed to have a high degree of inherent safety, it is necessary to reconsider the basis for establishing safety objectives. It is shown that, if there were a large program of inherently safe reactors, the safety objectives would be determined more by investment risk than by the public health risk. These considerations lead to an objective on the order of 1 × 10⁻⁷ per r.yr (reactor year) for the probabability of a severe core damage accident. It is also shown that the introduction of inherently safe features leads to a considerable change in the allocation of the safety goal between the major safety functions. For these reactors, a major portion of the allocation shifts from the decay heat removal function to the scram function, with emphasis on insuring the integrity of critical structures.     

浅议某核电厂排水点的优化

主要从某核电厂温排水影响范围、排水工程总投资(包括用海费用、养殖补偿费用、工程造价等)、排水设施(如排水明渠、盾构或暗渠等)长度以及排水点所在海域的水深等几个方面阐述了该核电厂排水方案排水点的优化过程,并对这几个主要因素进行了简单的分析。     

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.     

核电厂选址阶段的核安全监督

根据核电厂选址程序以及相关的规范标准，结合前一时期核电厂选址．分析了核电厂选址阶段核安全监督的方式和特点，并对当前监督存在的问题进行了讨论。     

核电厂空气净化系统分析

核电厂空气净化系统是核岛辅助系统中的重要系统之一,通过对空气净化系统工作特点和净化原理的分析,以及对主要空气净化专属设备的研究,得到了设计选择核电厂空气净化设备的一些原则。     

压水堆核电站的用钢量及要求

文中简要介绍了西德压水堆核电站机械设备的分级,用钢量及要求。     

大亚湾核电站汽轮机特点

本文论述了全速核电汽轮机的某些特点并介绍了我国第一座大型核电站采用的全速百万千瓦级汽轮机的主要参数、热力系统、平面布置、本体设计以及调节系统等基本情况及其特点。     

ESAP, an easy-to-use expert system for the systematic analysis of operator actions within the scope of probabilistic risk assessment

The assessment of operator actions within the scope of probabilistic risk analyses is an important task; however, it is connected with considerable uncertainties. The aim of the expert system ESAP (an easy-to-use expert system for the systematic analysis of operator actions within the scope of probabilistic risk assessment) is the reduction of those uncertainties caused by subjectivities in using the technique of human error rate prediction (THERP) methodology. The transformation of THERP using the expert system shell knowledge engineering environment (KEE) and the input logic are briefly described as well as the calculation of error probabilities using KEE rules. The advantages of ESAP and further developments are summarised.     

Leak area and leakage rate prediction for probabilistic risk assessment of concrete containments under severe core conditions

The fundamental gap in knowledge for estimating release for probabilistic risk assessment of concrete containments subject to beyond design basis loads is in estimating leak areas and leakage rates. By evaluating the available literature and carefully studying the test results, several generic rules are postulated for leak areas and leakage rates of concrete containments. These rules, coupled with theory-based leakage flow equations and empirically-based crack roughness constants, provide a realistic estimate of leak rates through liner tears in concrete containments.     

APWR核电厂控制与仪表系统

本文提出压水堆核电厂控制与仪表系统的改进目标.设计反应堆一体化数字控制与保护系统,完善计算机系统监视功能,应用多路数据传输技术和标准化设备,将人因工程科学用于系统和设备的设计。是核电厂控制与仪表系统的发展趋势。它将提高核电厂运行的安全性和可用性,降低建造投资和运行费用。     

The integrated management system as an auxiliary means for providing data for the probabilistic safety assessment

The results of probabilistic safety analyses (PSA) provide plant-specific reliability parameters which reflect more precisely the plant investigated. These parameters are ascertained by a sufficiently extended study of the individual systems and components. An extension of the study period with regard to the total operational life of a plant is to be recommended in most cases. The cost of preparing the required data and processing it to obtaining plant specific reliability parameters for the components can, however, be considerable. The volume and contents of the operational documentation, including maintenance instructions and recurrent tests, are, in all German power stations, more or less the same in view of the fact that these instructions are laid down by law. The accessibility to the data acquired during operation is dependant on the auxiliary means used for the documentation. At the NPP Emsland a computer-aided integrated management system (IBFS) has being used since its start-up. In this paper the methodical procedure for data supply from the IBFS will be explained. In the IBFS all data of operational events are available, since all tasks of plant operation are carried out on the basis of the IBFS. These are optimal prerequisites to obtain from the IBFS the necessary data and information for the PSA-relevant components.     

秦山核电厂调试

本文综合介绍了秦山核电厂的调试,其中包括无核和带核调试的试验内容、进度、调试网络、调试机构、人员和管理。总结了调试经验。调试结果证明,秦山核电厂的设计、建造是成功的。