Survey of seismic fragilities used in PRA studies of nuclear power plants

In recent years, seismic PRA studies have been performed on a large number of nuclear power plants in the USA. This paper presents a summary of a survey on fragility databases and the range of evaluated fragility values of various equipment categories based on past PRAs. The survey includes the use of experience data, the interpretations of available test data, and the quantification of uncertainties. The surveyed fragility databases are limited to data available in the public domain such as NUREG reports, conference proceedings and other publicly available reports. The extent of the availability of data as well as limitations are studied and tabulated for various equipment categories. The survey of the fragility values in past PRA studies includes not only the best estimate values, but also the dominant failure modes and the estimated uncertainty levels for each equipment category. The engineering judgments employed in estimating the uncertainty in the fragility values are also studied.This paper provides a perspective on the seismic fragility evaluation procedures for equipment in order to clearly identify the engineering analysis and judgment used in past seismic PRA studies.     

Dynamic modeling of the tradeoff between productivity and safety in critical engineering systems

Short-term tradeoffs between productivity and safety often exist in the operation of critical facilities such as nuclear power plants, offshore oil platforms, or simply individual cars. For example, interruption of operations for maintenance on demand can decrease short-term productivity but may be needed to ensure safety. Operations are interrupted for several reasons: scheduled maintenance, maintenance on demand, response to warnings, subsystem failure, or a catastrophic accident. The choice of operational procedures (e.g. timing and extent of scheduled maintenance) generally affects the probabilities of both production interruptions and catastrophic failures. In this paper, we present and illustrate a dynamic probabilistic model designed to describe the long-term evolution of such a system through the different phases of operation, shutdown, and possibly accident. The model's parameters represent explicitly the effects of different components' performance on the system's safety and reliability through an engineering probabilistic risk assessment (PRA). In addition to PRA, a Markov model is used to track the evolution of the system and its components through different performance phases. The model parameters are then linked to different operations strategies, to allow computation of the effects of each management strategy on the system's long-term productivity and safety. Decision analysis is then used to support the management of the short-term trade-offs between productivity and safety in order to maximize long-term performance. The value function is that of plant managers, within the constraints set by local utility commissions and national (e.g. energy) agencies. This model is illustrated by the case of outages (planned and unplanned) in nuclear power plants to show how it can be used to guide policy decisions regarding outage frequency and plant lifetime, and more specifically, the choice of a reactor tripping policy as a function of the state of the emergency core cooling subsystem.     

基于PSHA的核电厂近断层抗震设计谱构建方法

核电是一种高效、清洁的能源,随着核电厂未来向内陆区的发展,其可能会遭遇到近断层地震动的影响,但是目前我国核电厂抗震规范设计谱并未考虑近断层地震动。该文首先基于大量实际近断层脉冲型和相应无脉冲地震动记录,研究了脉冲对反应谱的放大效应,建立了修正的近断层脉冲放大系数模型;继而将地震动脉冲效应引入到近断层概率地震危险性分析中,并基于设定断层模型,给出了不同场地类型的一致危险性反应谱;通过对地震危险性结果的分解,分析了对场地最危险震级和距离,并将结果引入地震动衰减关系中得到设计谱,最后通过近断层脉冲放大系数对设计谱进行修正,得到考虑近断层脉冲效应的核电厂抗震设计谱。通过研究,建立了一种基于概率地震危险性分析框架下,考虑近断层脉冲型地震动的工程场地核电厂抗震设计谱的构建方法。     

Seismic PSA method for multiple nuclear power plants in a site

The maximum number of nuclear power plants in a site is eight and about 50% of power plants are built in sites with three or more plants in the world. Such nuclear sites have potential risks of simultaneous multiple plant damages especially at external events. Seismic probabilistic safety assessment method (Level-1 PSA) for multi-unit sites with up to 9 units has been developed. The models include Fault-tree linked Monte Carlo computation, taking into consideration multivariate correlations of components and systems from partial to complete, inside and across units. The models were programmed as a computer program CORAL reef. Sample analysis and sensitivity studies were performed to verify the models and algorithms and to understand some of risk insights and risk metrics, such as site core damage frequency (CDF per site-year) for multiple reactor plants. This study will contribute to realistic state of art seismic PSA, taking consideration of multiple reactor power plants, and to enhancement of seismic safety.     

Issues related to structural aging in probabilistic risk assessment of nuclear power plants

Structural components and systems have an important safety function in nuclear power plants. Although they are essentially passive under normal operating conditions, they play a key role in mitigating the impact of extreme environmental events such as earthquakes, winds, fire and floods on plant safety. Moreover, the importance of structural components and systems in accident mitigation is amplified by common-cause effects. Reinforced concrete structural components and systems in NPPs are subject to a phenomenon known as aging, leading to time-dependent changes in strength and stiffness that may impact their ability to withstand various challenges during their service lives from operation, the environment and accidents. Time-dependent changes in structural properties as well as challenges to the system are random in nature. Accordingly, condition assessment of existing structures should be performed within a probabilistic framework. The mathematical formalism of a probabilistic risk assessment (PRA) provides a means for identifying aging structural components that may play a significant role in mitigating plant risk. Structural condition assessments supporting a decision regarding continued service can be rendered more efficient if guided by the logic of a PRA.     

A historical overview of probabilistic risk assessment development and its use in the nuclear power industry: a tribute to the late Professor Norman Carl Rasmussen

This paper reviews the historical development of the probabilistic risk assessment (PRA) methods and applications in the nuclear industry. A review of nuclear safety and regulatory developments in the early days of nuclear power in the United States has been presented. It is argued that due to technical difficulties for measuring and characterizing uncertainties and concerns over legal challenges, safety design and regulation of nuclear power plants has primarily relied upon conservative safety assessment methods derived based on a set of design and safety principles. Further, it is noted that the conservatism adopted in safety and design assessments has allowed the use of deterministic performance assessment methods. This approach worked successfully in the early years of nuclear power epoch as the reactor design proved to be safe enough. However, it has been observed that as the conservative approach to design and safety criteria proved arbitrary, and yielded inconsistencies in the degree to which different safety measures in nuclear power plants protect safety and public heath, the urge for a more consistent assessment of safety became apparent in the late 1960s. In the early 1970s, as a result of public and political pressures, then the US Atomic Energy Commission initiated a new look at the safety of the nuclear power plants through a comprehensive study called ‘Reactor Safety Study’ (WASH-1400, or ‘Rasmussen Study’—after its charismatic study leader Professor Norman Rasmussen of MIT) to demonstrate safety of the nuclear power plants. Completed in October 1975, this landmark study introduced a novel probabilistic, systematic and holistic approach to the assessment of safety, which ultimately resulted in a sweeping paradigm shift in safety design and regulation of nuclear power in the United States in the turn of the Century. Technical issues of historic significance and concerns raised by the subsequent reviews of the Rasmussen Study have been discussed. Effect of major events and developments such as the Three Mile Island accident and the Nuclear Regulatory Commission and the Nuclear Industry sponsored studies on the tools, techniques and applications of the PRA that culminated in the present day risk-informed initiatives has been discussed.     

Development of a dynamical systems model of plant programmatic performance on nuclear power plant safety risk

Application of probabilistic risk assessment (PRA) techniques to model nuclear power plant accident sequences has provided a significant contribution to understanding the potential initiating events, equipment failures and operator errors that can lead to core damage accidents. Application of the lessons learned from these analyses has resulted in significant improvements in plant operation and safety. However, this approach has not been nearly as successful in addressing the impact of plant processes and management effectiveness on the risks of plant operation. The research described in this paper presents an alternative approach to addressing this issue. In this paper we propose a dynamical systems model that describes the interaction of important plant processes on nuclear safety risk. We discuss development of the mathematical model including the identification and interpretation of significant inter-process interactions. Next, we review the techniques applicable to analysis of nonlinear dynamical systems that are utilized in the characterization of the model. This is followed by a preliminary analysis of the model that demonstrates that its dynamical evolution displays features that have been observed at commercially operating plants. From this analysis, several significant insights are presented with respect to the effective control of nuclear safety risk. As an important example, analysis of the model dynamics indicates that significant benefits in effectively managing risk are obtained by integrating the plant operation and work management processes such that decisions are made utilizing a multidisciplinary and collaborative approach. We note that although the model was developed specifically to be applicable to nuclear power plants, many of the insights and conclusions obtained are likely applicable to other process industries.     

桥梁系统地震多维易损性分析

综合考虑地震地面运动以及性能极限状态的不确定性,提出了基于多地震需求参数分析的桥梁系统易损性评估方法,将易损性概念从一维扩展到多维。该方法首次提出服从多元对数正态分布的概率地震需求模型探讨桥梁体系各构件响应相关性,同时考虑各构件性能极限状态的相关性建立多维性能极限状态方程,确定结构失效域,通过MonteCarlo模拟计算系统多维地震易损性。以某一钢筋混凝土多跨连续梁高速公路桥为算例,通过非线性动力分析法获得最大响应样本,利用最大似然估计求得概率地震需求模型未知参数,计算体系多维易损性,并与构件易损性相比较。结果表明:桥梁体系多维易损性较构件易损性偏大,可避免用单一构件易损性代替系统易损性产生的非保守估计,预测结果更利于工程安全,为桥梁修复加固和交通系统可靠性分析提供理论依据。     

Serviceability of earthquake-damaged water systems: Effects of electrical power availability and power backup systems on system vulnerability

Civil infrastructure systems, such as water, electrical power, natural gas, and transportation systems, are essential to the smooth functioning of modern society. Because of their inter-connected nature, once one infrastructure system is damaged by an earthquake or other natural hazard, other infrastructure systems may malfunction as well. A number of previous studies have assessed vulnerability of infrastructure systems to earthquakes, but seldom have failures due to infrastructure system interactions been considered. In its assessment of the earthquake-induced damage of a municipal water system, this paper includes the impact of damage to the supporting electrical power system using a fault tree analysis and a shortest-path algorithm. The effect of uncertainty of seismic intensity and component fragility on network integrity is evaluated. A case study involving a simple model of the electrical power system and water system in Shelby County, TN, which includes the city of Memphis, demonstrates the importance of taking infrastructure interactions into account when evaluating the seismic vulnerability and risk to a networked system, as well as the utility of back-up power systems in electric power facilities.     

A dynamic fault tree

The fault tree analysis is a widely used method for evaluation of systems reliability and nuclear power plants safety. This paper presents a new method, which represents extension of the classic fault tree with the time requirements. The dynamic fault tree offers a range of risk informed applications. The results show that application of dynamic fault tree may reduce the system unavailability, e.g. by the proper arrangement of outages of safety equipment. The findings suggest that dynamic fault tree is a useful tool to expand and upgrade the existing models and knowledge obtained from probabilistic safety assessment with additional and time dependent information to further reduce the plant risk.     

Structural integrity aspects of reactor safety

The overall goal of nuclear power plant safety is to protect individuals, society and the environment from undue radiological hazard so that nuclear power production does not significantly add to the health risks to which individuals and society are already exposed. This paper addresses the safety principles followed during the design phase of life cycle of a nuclear power plant. The principles followed such as safety classification, design rules based on failure modes, detailed stress analysis, stress categorization, consideration of design basis events, failure probability, flaw tolerance, leak-before-break are described. Engineering structures always contain flaws, albeit of very small size. Fatigue and fracture are the two important failure modes affected by flaws. Thus flaw tolerance becomes very important. This is assessed by applying fracture mechanics principles. The R6 procedure, which is used for evaluation of structures containing flaws, has been incorporated in the software BARC-R6. Improvements by way of shell-nozzle junction pull-out, adoption of hot wire GTAW with narrow gap technique have been brought out. Post Fukushima incidence, resistance to seismic loading and containment design have assumed great importance. The paper describes these aspects in detail. Regulatory aspects of seismic design regarding siting, Seismic margin assessment, base isolation, retrofitting are the aspects covered under seismic design. Under the action of seismic loading, the piping in a nuclear power plant piping is vulnerable to a phenomenon called ratcheting. The process of seismic margin assessment and consideration of ratchetting has been backed up by a large experimental data. The experiments carried out on structures and piping components form a part of the paper.     

基于多个响应量的拱坝地震易损性分析

以白鹤滩拱坝为研究对象,选取成组强震记录,同时考虑地震动和材料的不确定性,采用增量法对白鹤滩拱坝进行了地震易损性分析。统计连续调幅地震动作用下拱坝损伤破坏过程,直观的划分了拱坝地震破坏等级,确定了拱冠位移、横缝开度和损伤体积比这三个响应量在各破坏等级间的界限值,从而可通过这三个响应量定量描述拱坝的破坏等级。通过拟合增量动力分析结果,分别建立了三个响应量的概率地震需求模型,进而求得地震易损性曲线,并综合比较了不同破坏等级下基于三个响应量的易损性曲线,全面反映了拱坝易损性。利用易损性曲线可以预测不同强度地震作用下拱坝达到各级破坏的概率,为基于性能的拱坝抗震安全评价提供了理论依据。     

Seismic fragility analysis for non-seismically qualified steel structure

The cooling water (C/W) pumphouse of CANDU nuclear power plants is non-safety related and is only designed to meet the seismic requirements of the National Building Code of Canada (NBCC) for normal industrial plants and public buildings. Based on a feasibility study for CANDU seismic Probabilistic Safety Assessment (PSA), it was decided to include non-seismically qualified systems in the PSA models. Thus, it was necessary to evaluate the seismic fragility of the C/W pumphouse which houses the raw service water system. The design basis seismic ground motion at CANDU site in Korea was evaluated by AECL in 1977, based on a probabilistic approach, which resulted in Design Basis Earthquake with a peak horizontal ground acceleration of 0.2g. The peak ground acceleration considered for the design of the C/W pumphouse is 0.1g. The seismic fragility evaluation accounted for conservatism in the design, both in the seismic response and in the member capacity. The median-centered seismic capacity and the high confidence and low probability of failure (HCLPF) capacity of the pumphouse steel superstructure are 0.89g and 0.33g, respectively. The ratio of the HCLPF capacity to the design peak ground acceleration is 3.3.     

NAROAS: a neural network-based advanced operator support system for the assessment of systems reliability

We have developed and implemented a computerized reliability monitoring system for nuclear power plant applications, based on a neural network. The developed computer program is a new tool related to operator decision support systems, in case of component failures, for the determination of test and maintenance policies during normal operation or to follow an incident sequence in a nuclear power plant. The NAROAS (Neural Network Advanced Reliability Advisory System) computer system has been developed as a modularized integrated system in a C++ Builder environment, using a Hopfield neural network instead of fault trees, to follow and control the different system configurations, for interventions as quickly as possible at the plant. The observed results are comparable and similar to those of other computer system results. As shown, the application of this neural network contributes to the state of the art of risk monitoring systems by turning it easier to perform online reliability calculations in the context of probabilistic safety assessments of nuclear power plants.     

State of the art in reliability of thermal-hydraulic passive systems

In order to address the issues posed by the development of advanced nuclear technologies, this article endeavours to analyse the current state of the art in reliability of passive systems, for their extensive use in future nuclear power plants. Inclusion of failure modes and reliability estimates of passive components for all systems is recommended in probabilistic safety assessment (PSA) studies. This has aroused the need for the development and demonstration of consistent methodologies and approaches for their reliability evaluation, within the community of the nuclear safety research. This report provides the insights resulting from the survey on the technical issues associated with assessing the reliability of passive systems in the context of nuclear safety, regulatory practices and probabilistic safety analysis. Special emphasis is placed on the reliability of the systems based on thermal-hydraulics, for which methods are still in a developing phase. The main achievements of these studies are presented and a viable path towards the implementation of the research efforts is delineated as well.     

Risk-significant functional dependencies in pressurized water reactors

This work provides evidence that functional dependencies among nuclear plant systems, particularly those among frontline safety systems and support systems, are often major contributors to the calculated risks in completed probabilistic risk assessments (PRAs). The study, furthermore, determined how the risk impact of functional dependencies could be reduced in future nuclear power plant designs. The risk reduction insights are summarized by a set of nine generalized design approaches, which we call design principles. These approaches are not new to the nuclear industry nor are the dependencies they address. The contribution made by this study is the use of PRA insights in pointing out the relative importance of the dependencies in terms of their contributions to risk.     

Quantitative risk analysis of oil storage facilities in seismic areas

Quantitative risk analysis (QRA) of industrial facilities has to take into account multiple hazards threatening critical equipment. Nevertheless, engineering procedures able to evaluate quantitatively the effect of seismic action are not well established. Indeed, relevant industrial accidents may be triggered by loss of containment following ground shaking or other relevant natural hazards, either directly or through cascade effects ('domino effects'). The issue of integrating structural seismic risk into quantitative probabilistic seismic risk analysis (QpsRA) is addressed in this paper by a representative study case regarding an oil storage plant with a number of atmospheric steel tanks containing flammable substances. Empirical seismic fragility curves and probit functions, properly defined both for building-like and non building-like industrial components, have been crossed with outcomes of probabilistic seismic hazard analysis (PSHA) for a test site located in south Italy. Once the seismic failure probabilities have been quantified, consequence analysis has been performed for those events which may be triggered by the loss of containment following seismic action. Results are combined by means of a specific developed code in terms of local risk contour plots, i.e. the contour line for the probability of fatal injures at any point (x, y) in the analysed area. Finally, a comparison with QRA obtained by considering only process-related top events is reported for reference.     

第二代基于性能地震工程中的地震易损性模型及正逆概率风险分析

第二代基于性能地震工程理论中的地震易损性主要是指结构构件以及非结构构件的抗震能力,与传统地震风险理论中的地震易损性定义和内涵并不相同。为了澄清二者的不一致性,首先介绍传统地震风险理论中地震易损性的定义和概率模型,然后指出第二代基于性能地震工程理论存在五个层次的地震易损性模型:地震需求易损性模型、抗震能力易损性模型、地震损伤易损性模型、地震损失易损性模型和抗震决策易损性模型,指出了这五种模型的区别及其相互关系,推导得到了地震需求易损性模型和地震损伤易损性模型分布参数的解析表达式。在此基础上,根据不同的不确定性传递路径,提出了正向PBEE和逆向PBEE的概念,以通过不同方式求解第二代基于性能地震工程理论的风险积分公式。基于地震危险性函数的近似表达式以及地震易损性模型及其分布参数的解析表达式,通过正向PBEE和逆向PBEE方法,分别得到了具有相同表达形式的工程需求参数EDP、地震损伤DM和决策变量DV三个层次的概率地震风险表达式。通过该文的研究,将传统地震风险分析理论与第二代基于性能地震工程理论统一在一致的理论框架之中。     

The impact of assessing lower head integrity on boiling water reactor probabilistic risk assessments

Over the last several decades, much effort has been directed at estimating the likelihood of a large early release of radioactivity during a nuclear accident. This effort has culminated in the Individual Plant Examinations (IPEs) for the over 100 US nuclear power plants and the NUREG 1150 study. The large early release of radioactivity requires core damage with loss of primary containment integrity during the accident. Given a successful reactor scram, early containment failure coupled with a large release of radioactivity will only occur if the reactor core vessel is breached by core debris. Most IPE/PRA studies performed to date have not considered the possibility of quenching core debris in the lower plenum. Consequently, lower head failure is presumed to closely follow the onset of core damage. Therefore, these assessments did not address the role that in-vessel debris retention plays in preserving primary containment integrity, nor do they propose a criterion for evaluating the integrity of the vessel lower head given that core damage has occurred. Yet preserving the vessel lower head integrity is a necessary condition for satisfying the plant design and licensing basis. Therefore, a more complete treatment of the risk associated with nuclear plant operation includes an evaluation of the ability to retain the core debris in-vessel. This paper presents a performance requirement for vessel integrity to be used in probabilistic risk assessments; evaluates the impact the core damage progression and lower plenum quenching models have on the likelihood of terminating the damage progression in-vessel; documents the significant reduction in BWR containment failure probability that can occur when appropriate core damage and lower head quenching models are used; reviews the implications of core debris quenching in the lower head on BWR PRA modeling; argues why crediting the capability to maintain vessel integrity is necessary from a safety point of view. These results and conclusions are derived from consideration of a BWR 4 plant with a 251 inch vessel. However, the concepts are generally applicable and results specific to other BWR designs can be developed using the methodology presented in this paper.     

考虑土-结构相互作用的核电厂楼层反应谱研究

楼层谱是核电厂设备、管道抗震设计和抗震裕度评估的重要依据。为了研究不同烈度下场地土对核电厂楼层谱的影响,该文使用叠层剪切土箱模拟土体及其边界条件,对某新型核电厂房进行1∶25缩尺模型地震模拟振动台实验。选取10组水平加速度地震动记录,按照运行安全地震动(OBE)0.15 g、极限安全地震动(SSE)0.30 g和超设计基准地震动(ULE)0.75 g作为输入。在考虑土-结构相互作用条件下,研究不同地震动强度引起楼层谱的变化规律,并对《核电厂抗震设计规范》中楼层谱确定方法的输入基准地震动进行讨论分析。根据设备响应比分析,揭示了现有核电设备抗震裕度评估方法具有一定的保守性,特别是设备响应比为1/1.5~1.5,采用现有抗震裕度评估方法得到的设备抗震裕度可能小于审核地震动。为了得到该设备的真实抗震裕度,建议对这些设备做更详细分析。