Determination of design alternatives and performance criteria for safety systems in a nuclear power plant via simulated annealing

This study presents an efficient methodology that derives design alternatives and performance criteria for safety functions/systems in commercial nuclear power plants. Determination of the design alternatives and intermediate-level performance criteria is posed as a reliability allocation problem. The reliability allocation is performed in a single step by means of the concept of two-tier noninferior solutions in the objective and risk spaces within the top-level probabilistic safety criteria (PSC). Two kinds of two-tier noninferior solutions are obtained: desirable design alternatives and intolerable intermediate-level PSC of safety functions/systems.The weighted Chebyshev norm (WCN) approach with an improved Metropolis algorithm in simulated annealing is used to find the two-tier noninferior solutions. This is very efficient in searching for the global minimum of the difficult multiobjective optimization problem (MOP) which results from strong nonlinearity of a probabilistic safety assessment (PSA) model and nonconvexity of the problem. The methodology developed in this study can be used as an efficient design tool for desirable safety function/system alternatives and for the determination of intermediate-level performance criteria.The methodology is applied to a realistic streamlined PSA model that is developed based on the PSA results of the Surry Unit 1 nuclear power plant. The methodology developed in this study is very efficient in providing the intolerable intermediate-level PSC and desirable design alternatives of safety functions/systems.     

An overview of PSA importance measures

The present paper deals with the use of probabilistic safety assessment (PSA) importance measures to optimise the performance of a nuclear power plant. This article is intended to give an overview on the subject for PSA practitioners. The most frequently used importance measures are shortly addressed. It is shown that two importance measures are sufficient to describe the character of the coredamage-equation. The most often used are the risk achievement and Fussell–Vesely importance, in combination with each other. In the field of nuclear power plant test and maintenance activities the Birnbaum importance is advocated.     

Examination of issues related to the development and implementation of real-time operational safety monitoring tools in the nuclear power industry

In recent years, risk and reliability techniques have been increasingly used to optimize deterministic requirements and to improve the operational safety of nuclear power stations. This paper discusses the historical development and current status of implementation of real-time operational safety monitoring tools in the nuclear power industry worldwide. A safety monitor is defined as a PC-based risk management tool, based on a plant specific PSA, which can be used to manage plant safety during the day-to-day operation of a nuclear power plant by planning maintenance activities and providing advisory information to plant operational staff in order to avoid high risk plant configurations. As this technique has only been applied in a few plants worldwide, the technology is still evolving and there are several technical and implementation-related issues which still need to be resolved. This paper attempts to summarize all such issues and describe how they have been addressed in several different applications of this technology around the world.     

Methods and practices used in incident analysis in the Finnish nuclear power industry

According to the Finnish Nuclear Energy Act it is licensee's responsibility to ensure safe use of nuclear energy. Radiation and Nuclear Safety Authority (STUK) is the regulatory body responsible for the state supervision of the safe use of nuclear power in Finland. One essential prerequisite for the safe and reliable operation of nuclear power plants is that lessons are learned from the operational experience. It is utility's prime responsibility to assess the operational events and implement appropriate corrective actions. STUK controls licensees' operational experience feedback arrangements and implementation as part of its inspection activities. In addition to this in Finland, the regulatory body performs its own assessment of the operational experience. Review and investigation of operational events is a part of the regulatory oversight of operational safety. Review of operational events is done by STUK basically at three different levels. First step is to perform a general review of all operational events, transients and reactor scram reports, which the licensees submit for information to STUK. The second level activities are related to the clarification of events at site and entering of events' specific data into the event register database of STUK. This is done for events which meet the set criteria for the operator to submit a special report to STUK for approval. Safety significance of operational events is determined using probabilistic safety assessment (PSA) techniques. Risk significance of events and the number of safety significant events are followed by STUK indicators. The final step in operational event assessment performed by STUK is to assign STUK's own investigation team for events deemed to have special importance, especially when the licensee's organisation has not operated as planned. STUK launches its own detail investigation once a year on average. An analysis and evaluation of event investigation methods applied at STUK, and at the two Finnish nuclear power plant operators Teollisuuden Voima Oy (TVO) and Fortum Power and Heat Oy (Fortum) was carried out by the Technical Research Centre (VTT) on request of STUK at the end of 1990s. The study aimed at providing a broad overview and suggestions for improvement of the whole organisational framework to support event investigation practices at the regulatory body and at the utilities. The main objective of the research was to evaluate the adequacy and reliability of event investigation analysis methods and practices in the Finnish nuclear power industry and based on the results to further develop them. The results and suggestions of the research are reviewed in the paper and the corrective actions implemented in event investigation and operating experience procedures both at STUK and at utilities are discussed as well. STUK has developed its own procedure for the risk-informed analysis of nuclear power plant events. The PSA based event analysis method is used to assess the safety significance and importance measures associated with the unavailability of components and systems subject to Technical Specifications. The insights from recently performed PSA based analyses are also briefly discussed in the paper.     

A methodology for the prioritization of operating experience in nuclear power plants

The processing of operating experience at nuclear power plants is critically important to safe and reliable operations because it represents the process by which important external information is incorporated into the organization. Nuclear power plants typically receive between 800 and 1000 pieces of operating experience every year, of which 70–80 meet the criteria for a more extensive review. This paper deals with the prioritization of these items. Specifically, a prioritization methodology utilizing multiattribute utility theory has been developed. What sets this methodology apart from other techniques employing multiattribute methods is its emphasis on deliberations to achieve consensus among objectives and preferences among those objectives. Along with an explanation of this methodology, the results of its application to the prioritization of operating experience at a nuclear power plant are presented. Lastly, the results of a workshop that was held at MIT are presented. The workshop demonstrated the feasibility of the prioritization methodology and the validity of the case study results.     

Probabilistic analysis of accident precursors in the nuclear industry

Feedback of operating experience has always been an important issue in the nuclear industry. A probabilistic safety analysis (PSA) can be used as a tool to analyse how an operational event might have developed adversely in order to obtain a quantitative assessment of the safety significance of the event. This process is called PSA-based event analysis (PSAEA). A comprehensive set of PSAEA guidelines was developed by an international project. The main characteristics of this methodology are summarised. This approach to analyse incidents can be used to meet different objectives of utilities or nuclear regulators. The paper describes the main objectives and the experiences of the Belgian nuclear regulatory organisation AVN with the application of PSA-based event analysis. Some interesting aspects of the process of PSAEA are further developed and underlined. Several case studies are discussed and an overview of the obtained results is given. Finally, the interest of a broad and interactive forum on PSAEA is highlighted.     

A bounded-search iterated greedy algorithm for the distributed permutation flowshop scheduling problem

As the interest of practitioners and researchers in scheduling in a multi-factory environment is growing, there is an increasing need to provide efficient algorithms for this type of decision problems, characterised by simultaneously addressing the assignment of jobs to different factories/workshops and their subsequent scheduling. Here we address the so-called distributed permutation flowshop scheduling problem, in which a set of jobs has to be scheduled over a number of identical factories, each one with its machines arranged as a flowshop. Several heuristics have been designed for this problem, although there is no direct comparison among them. In this paper, we propose a new heuristic which exploits the specific structure of the problem. The computational experience carried out on a well-known testbed shows that the proposed heuristic outperforms existing state-of-the-art heuristics, being able to obtain better upper bounds for more than one quarter of the problems in the testbed.     

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.     

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.     

A review of the ‘taguchi methods’ for off-line quality control

The Taguchi methods have recently become popular in the U.S.A following a realization of their importance in Japanese quality design. This review is an initial attempt to extract the important ideas while drawing on the ‘Western’ experience with response surface methodology and experimental design.     

Pipe failure probability—the Thomas paper revisited

Almost twenty years ago, in Volume 2 of Reliability Engineering (the predecessor of Reliability Engineering and System Safety), a paper by H. M. Thomas of Rolls Royce & Associates Ltd. presented a generalized approach to the estimation of piping and vessel failure probability. The ‘Thomas-approach’ used insights from actual failure statistics to calculate the probability of leakage and conditional probability of rupture given leakage. It was intended for practitioners without access to data on the service experience with piping and piping system components.This article revisits the Thomas paper by drawing on insights from development of a new database on piping failures in commercial nuclear power plants worldwide (SKI-PIPE). Partially sponsored by the Swedish Nuclear Power Inspectorate (SKI), the R&D leading up to this note was performed during 1994–1999. Motivated by data requirements of reliability analysis and probabilistic safety assessment (PSA), the new database supports statistical analysis of piping failure data. Against the background of this database development program, the article reviews the applicability of the ‘Thomas approach’ in applied risk and reliability analysis. It addresses the question whether a new and expanded database on the service experience with piping systems would alter the original piping reliability correlation as suggested by H. M. Thomas.     

Use of risk assessment in the nuclear industry with specific reference to the Australian situation

The use of risk assessment in the nuclear industry began in the 1970s as a complementary approach to the deterministic methods used to assess the safety of nuclear facilities. As experience with the theory and application of probabilistic methods has grown, so too has its application. In the last decade, the use of probabilistic safety assessment has become commonplace for all phases of the life of a plant, including siting, design, construction, operation and decommissioning. In the particular case of operation of plant, the use of a ‘living’ safety case or probabilistic safety assessment, building upon operational experience, is becoming more widespread, both as an operational tool and as a basis for communication with the regulator. In the case of deciding upon a site for a proposed reactor, use is also being made of probabilistic methods in defining the effect of design parameters. Going hand in hand with this increased use of risk based methods has been the development of assessment criteria against which to judge the results being obtained from the risk analyses. This paper reviews the use of risk assessment in the light of the need for acceptability criteria and shows how these tools are applied in the Australian nuclear industry, with specific reference to the probabilistic safety assessment (PSA) performed of HIFAR.     

An analysis of safety-critical digital systems for risk-informed design

This paper quantitatively presents the results of a case study which examines the fault tree analysis framework of the safety of digital systems. The case study is performed for the digital reactor protection system of nuclear power plants. The broader usage of digital equipment in nuclear power plants gives rise to the need for assessing safety and reliability because it plays an important role in proving the safety of a designed system in the nuclear industry. We quantitatively explain the relationship between the important characteristics of digital systems and the PSA result using mathematical expressions. We also demonstrate the effect of critical factors on the system safety by sensitivity study and the result which is quantified using the fault tree method shows that some factors remarkably affect the system safety. They are the common cause failure, the coverage of fault tolerant mechanisms and software failure probability.     

Risk implications of digital reactor protection system operating experience

This paper summarizes an in-depth review of the US nuclear operating experience with the first generation of digital reactor protection systems. The accumulated operating experience from 1984 to 2006 on these first generation digital reactor protection system functions exceeds 1.27 million hours (145.5 yr). A review of failure event reports identified 141 specific events associated with these systems on seven US nuclear power plants. Twenty-six of these events involved some type of common cause failure mechanism (predominantly redundant sensors/channels being out of calibration), which temporarily rendered redundant portions of the overall trip function degraded. Most of these failures were found not to be unique to digital systems. Six of the common cause failure events were more severe and involved situations where incorrect addressable constant data sets were systematically loaded into all redundant computer channels due to personnel errors. One of these events involved a latent software design change error introduced during a software update, which would prevent proper operation, given an unlikely event involving failure of three out of four sensors of one type.Based upon this review of digital system operating experience, a series of risk assessment calculations were performed to evaluate the safety significance of the observed failure events. From the insights gained in this work, it is possible to develop a framework for establishing digital reactor protection system reliability requirements that can be related back to regulatory safety goal objectives and operating experience.     

Quantification of epistemic and aleatory uncertainties in level-1 probabilistic safety assessment studies

There will be simplifying assumptions and idealizations in the availability models of complex processes and phenomena. These simplifications and idealizations generate uncertainties which can be classified as aleatory (arising due to randomness) and/or epistemic (due to lack of knowledge). The problem of acknowledging and treating uncertainty is vital for practical usability of reliability analysis results. The distinction of uncertainties is useful for taking the reliability/risk informed decisions with confidence and also for effective management of uncertainty. In level-1 probabilistic safety assessment (PSA) of nuclear power plants (NPP), the current practice is carrying out epistemic uncertainty analysis on the basis of a simple Monte-Carlo simulation by sampling the epistemic variables in the model. However, the aleatory uncertainty is neglected and point estimates of aleatory variables, viz., time to failure and time to repair are considered. Treatment of both types of uncertainties would require a two-phase Monte-Carlo simulation, outer loop samples epistemic variables and inner loop samples aleatory variables. A methodology based on two-phase Monte-Carlo simulation is presented for distinguishing both the kinds of uncertainty in the context of availability/reliability evaluation in level-1 PSA studies of NPP.     

Probabilitically based decision support

Probabilistic Safety Assessment, usually referred to by the acronym PSA, has by now become a recognized tool for safety analysis of nuclear power plants. In recent years, an increasing number of plants have been analysed, and as the technique has matured, the area of application of PSA based analyses has been expanded. Thus, probabilistic methods are now used increasingly in the day-to-day work concerning the safety, maintenance and operation of plants. In this context, the question of interpretation and application of analysis results in various decision situations has become crucial. This paper gives some comments concerning the basis for decision making involving probabilistic analyses.     

The german precursor study—methodology and insights

Probabilistic Safety Assessment (PSA) yields a systematic and quantitative prediction of possible accident scenarios at technical installations on the basis of data gained from the past experience on similar technical installations. Precursor studies are performed in order to make operational experience, as far as possible, available for support of PSAs. An Accident Sequence Precursor in a Nuclear Power Plant (NPP) is defined as an observed event scenario which could result, in coincidence with additional postulated events, in a potential severe core damage accident. In this paper, the methodology and the insights of the plant-specific German Precursor Study are explained in detail. As the results have demonstrated, the Precursor methodology is applicable for ranking of the safety significance of the observed events and for trending the plant risk level (described by the frequency of potential severe core damage accidents) versus operating time.     

A PSA-based vital area identification methodology development

This paper presents a vital area identification method based on the current probabilistic safety assessment (PSA) techniques. The vital area identification method in this paper is focused on core melt rather than radioactive material release. Furthermore, it describes a conceptual framework with which the risk from sabotage-induced events could be assessed.Location minimal cut sets (MCSs) are evaluated after developing a core melt location fault tree (LFT). LFT is a fault tree whose basic events are sabotage-induced damages on the locations within which various safety-related components are located. The core melt LFT is constructed by combining all sequence LFTs of various event trees with OR gates. Each sequence LFT is constructed by combining the initiating event LFT and the mitigating event LFTs with an AND gate. The vital area could be identified by using the location importance measures on the core melt location MCSs. An application was made to a typical 1000 MWe pressurized water reactor power plant located at the Korean seashore.The methodology suggested in the present paper is believed to be very consistent and most complete in identifying the vital areas in a nuclear power plant because it is based on the well-proven PSA technology.     

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.     

International survey on PSA figures of merit

Probabilistic safety assessment (PSA) figures of merit represent one of the most interesting topics in the use of PSA for safety-related applications. This is all the more interesting as the viewpoints and opinions of different experts on the use of importance factors could diverge. Even when the experts agree, the use of these measures is likely to involve a number of difficulties or pitfalls, especially if they are used too “mechanically”, without proper care. EDF considered the topic to be sufficiently important and controversial to justify exploratory studies as part of the scientific and technical monitoring program and to get advice, comments, and critical viewpoints from a number of international PSA experts and practitioners. This report presents a synthesis of the results of this international survey.