Human reliability analysis: exploring the intellectual structure of a research field

Humans play a crucial role in modern socio-technical systems. Rooted in reliability engineering, the discipline of Human Reliability Analysis (HRA) has been broadly applied in a variety of domains in order to understand, manage and prevent the potential for human errors. This paper investigates the existing literature pertaining to HRA and aims to provide clarity in the research field by synthesizing the literature in a systematic way through systematic bibliometric analyses. The multi-method approach followed in this research combines factor analysis, multi-dimensional scaling, and bibliometric mapping to identify main HRA research areas. This document reviews over 1200 contributions, with the ultimate goal of identifying current research streams and outlining the potential for future research via a large-scale analysis of contributions indexed in Scopus database

Historical review of fire safety at NPP and application of fire PSA to Westinghouse PWR NPP in the frame of risk-informed decision making by

The importance of fire as a potential initiator of multiple-system failures took on a new perspective after the cable-tray fire at Browns Ferry in 1975 The review have shown that the first generation Nuclear Power Plant (NPP) fire safety was not factored as high risk area that needed to be effectively assessed and quantified. This resulted in development of peculiar fire safety regulations, standards and expensive backfits. Lack of appropriate regulations and effective methods of fire risk assessment, prescriptive, difficult and expensive retrofit regulations were instituted in USA. The alternative risk-informed performance based regulation was established in USA to resolve the challenges of the prescriptive rules. The review have revealed that both the prescriptive and risk-informed performance based approaches will not represent adequate design basis for new Nuclear Power Plants. The Japanese were pulled in the path of renew fire safety regulations and risk quantification after the Fukushima accident. It has been recognized that effective fire safety assessment, and culture, in concert with countermeasures to prevent, detect, suppress, and mitigate the effect of fires if they occur, will minimized NPP fire risk. Among the numerous recommendation the fire safety at NPP must be planned and engineered before construction begin using the state-of-the-arts technology. Also, the methods of fire risk assessment must integrate the state-of-the-arts deterministic and probabilistic approaches. Two methods are presented which serve to incorporate the fire-related risk into the current practices in nuclear power plants with respect to the assessment of configurations. The first method is a fire protection systems and key safety functions Unavailability Matrix (UM) which is developed to identify structures, systems, and components significant for fire-related risk. The second method is a fire zones and key safety functions (KSFs) fire risk matrix which is useful to identify fire zones which are candidates for risk management actions. The UM is an innovative tool to communicate fire risk. The Monte Carlo method has been used to assess the uncertainty of the UM. The analysis shows that the uncertainty is sufficiently bounded. The significant fire-related risk is localized in six KSF representative components and one fire protection system which should be included in the maintenance rule. The unavailability of fire protection systems does not significantly affect the risk. The fire risk matrix identifies the fire zones that contribute the most to the fire-related risk. These zones belong to the control building and electric penetrations building. The aggregation of Internal Events PSA model and Fire PSA model have shown that the Fire PSA contributes 38.4% to the Risk increase. The feasibility of developing Fire-related Risk Monitor from the FIRE PSA for the Spanish NPP was carried out. One of the main challenges is that RiskSpectrum® fire PSA has 384 fire cases and 384 CDF but in Risk Monitor one CDF is required. However, CAFTA is unable to convert a Sequential Fault Tree structure of the internal Event tree in the Fire PSA. The conversion fails to implement neither all of the sequences leading to core damage nor the Fault Tree selection of the frequency of fire. The proposal is to suppress exchange events and introduce the alignment of the consequences so that a unique result of core damage can be quantified. The detection and fire suppression Event Trees in the reference model were replaced by detection and fire extinction Fault trees. The frequency of each Fire Case of the conversion model and the reference model are quantified and the frequencies compared. The results shows that 90% of the cases are valid, however, the rest have challenges with MCS. A unique CDF of 7.65x10-7 is quantified compared with 9.83×10-6 of the reference. The conversion of the new model in CAFTA was not successful due to software incompatibility.La importància del incendi com un potencial iniciador de sistema múltiples fallides van agafar una nova perspectiva després del incendi al cable-safata de Browns Ferry el 1975. La revisió ha mostrat que la primera generació de seguretat contra incendis de centrals d'Energia Nuclear (NPP) no va ser àrea de alt risc, àrea que necessitava ser efectivament avaluada i quantificada. Això va resultar en el desenvolupament de normes de seguretat de incendi peculiar, estàndards i cares revisions. La manca d'una reglamentació adequada i mètodes eficaços d'avaluació de risc d'incendi, va fer que als USA foren instituïts mètodes d'adaptació de normativa preceptius, difícils i costós. L'alternativa de regulació informada per el risc es va establir als USA per resoldre els reptes de la regulació preceptiva. La revisió ha mostrat que tant als enfocaments de normativa preceptiva i regulació informada per el risc no representen bases de disseny adequades per a noves NPP. Ha estat reconeguda que la efectiva avaluació de seguretat al incendi i la cultura en concert amb mesures per prevenir, detectar, suprimir i mitigar l'efecte d'incendis, si es produeixen, minimitzarà el risc d'incendi en una NPP. Entre les nombroses recomanacions la seguretat contra incendis a una NPP s'hauran previst i dissenyat abans de començar la construcció i utilitzant estat del art de la tecnologia. També, els mètodes d'avaluació del risc d'incendi tindran que integrar el estat del art en els enfocaments de determinista i probabilístics. Dos mètodes són presentats que serveixen per incorporar el risc relacionats amb el foc a les pràctiques actuals en centrals nuclears en respecte a l'avaluació de configuracions. El primer mètode és un sistema de protecció contra incendis i una matriu de indisponiblitats de les funcions clau de seguretat (MU) que es desenvolupa per a identificar estructures, sistemes i components significatius per riscos relacionats amb els incendis. El segon mètode és zones de focs i matriu de risc d'incendi i funcions (KSFs) clau de seguretat que és útil identificar les zones de foc que són candidats per a les accions de gestió de risc. La MU és una eina innovadora per comunicar el risc d'incendi. El risc significatiu relacionats amb el incendi està localitzat en sis components representatius KSF i un sistema de protecció de foc que cal que figuri en la regla de manteniment. La manca de sistemes de protecció contra incendis no afecta significativament al risc. La matriu de risc d'incendi identifica les zones de foc que mes contribueixen al risc relacionats amb el incendi. Aquestes zones pertanyen a l'edifici de control i edifici de penetracions elèctriques. L'agregació del model de PSA de esdeveniments interns i model de incendis PSA han demostrat que el PSA de incendis aporta 38.4% a l'augment de risc. S'ha desenvolupat la viabilitat del Monitor de risc de incendis a partir del PSA de incendis per a una central nuclear espanyola. Un dels reptes principals és que RiskSpectrum® incendis PSA te 384 casos de incendis i te 384 CDF però en risc Monitor és necessària una CDF. Tanmateix, el CAFTA és incapaç de convertir una estructura seqüencial de arbre de fallida de l'arbre esdeveniment interna en el PSA de incendis. La conversió fracassa al posar en pràctica totes les seqüències de danys al nucli i la selecció de l'arbre de fallida de la freqüència de incendi. La descoberta i supressió de arbres de l'esdeveniment de incendi en el model de referència es van substituir per detecció i els arbres de fallades d'extinció d'incendi. La freqüència de cada cas de incendi del model de conversió i el model de referència son quantificades i les freqüències son comparades. Els resultats demostra que el 90% dels casos són vàlid, no obstant això, la resta té reptes amb MCS. Un únic CDF de 7.65x10-7 s'ha quantificat en comparació amb 9.83 × 10-6 de la referència. La conversió del nou model a CAFTA no va tenir èxit a causa de la incompatibilitat del programari

Analysis and Estimation of Human Errors From Major Accident Investigation Reports

Abstract Risk analyses require proper consideration and quantification of the interaction between humans, organization, and technology in high-hazard industries. Quantitative human reliability analysis approaches require the estimation of human error probabilities (HEPs), often obtained from human performance data on different tasks in specific contexts (also known as performance shaping factors (PSFs)). Data on human errors are often collected from simulated scenarios, near-misses report systems, and experts with operational knowledge. However, these techniques usually miss the realistic context where human errors occur. The present research proposes a realistic and innovative approach for estimating HEPs using data from major accident investigation reports. The approach is based on Bayesian Networks used to model the relationship between performance shaping factors and human errors. The proposed methodology allows minimizing the expert judgment of HEPs, by using a strategy that is able to accommodate the possibility of having no information to represent some conditional dependencies within some variables. Therefore, the approach increases the transparency about the uncertainties of the human error probability estimations. The approach also allows identifying the most influential performance shaping factors, supporting assessors to recommend improvements or extra controls in risk assessments. Formal verification and validation processes are also presented.</jats:p

Nuclear Power

The world of the twenty first century is an energy consuming society. Due to increasing population and living standards, each year the world requires more energy and new efficient systems for delivering it. Furthermore, the new systems must be inherently safe and environmentally benign. These realities of today's world are among the reasons that lead to serious interest in deploying nuclear power as a sustainable energy source. Today's nuclear reactors are safe and highly efficient energy systems that offer electricity and a multitude of co-generation energy products ranging from potable water to heat for industrial applications. The goal of the book is to show the current state-of-the-art in the covered technical areas as well as to demonstrate how general engineering principles and methods can be applied to nuclear power systems

A review of the emergency electric power supply systems at PWR nuclear power plants

Bibliography: pages 168-174.The Emergency Electric Power Supply Systems at Pressurized Water Reactor Nuclear Power Plants are reviewed, problem areas are identified, and recommendations are made for existing and future Nuclear Power Plants. A simplified introduction to a typical Pressurized Water Nuclear Reactor is given and the problems associated with the commercial use of nuclear power are discussed. An overview of the Nuclear industry's solutions is presented and covers the Reliability of equipment and the American Regulatory requirements. The alternating and direct current power supply systems are examined in terms of plant operational state and equipment type (Diesel generators, Grid network, Lead-acid batteries, Battery chargers, Inverters, and Power Distribution networks). The trends in the design of Emergency Electric Power supply systems at Nuclear Power Plants are presented. The loss of all alternating current power, known as Station Blackout, is discussed and the American and European response to this. problem is presented. Problems experienced in the direct current systems are discussed and solutions are presented. The experience at Koeberg Nuclear Power station with Lead-acid batteries is included in the discussion. The thesis concludes with recommendations for designers and operators of the Electric Power Supply Systems at Nuclear Power Stations

Comparative Analysis of Nuclear Event Investigation Methods, Tools and Techniques

Feedback from operating experience is one of the key means of enhancing nuclear safety and operational risk management. The effectiveness of learning from experience at NPPs could be maximised, if the best event investigation practices available from a series of methodologies, methods and tools in the form of a ‘toolbox’ approach were promoted. Based on available sources of technical, scientific, normative and regulatory information, an inventory, review and brief comparative analysis of information concerning event investigation methods, tools and techniques, either indicated or already used in the nuclear industry (with some examples from other high risk industry areas), was performed in this study. Its results, including the advantages and drawbacks identified from the different instruments, preliminary recommendations and conclusions, are covered in this report. The results of comparative analysis of nuclear event investigation methods, tools and techniques, presented in this interim report, are of a preliminary character. It is assumed that, for the generation of more concrete recommendations concerning the selection of the most effective and appropriate methods and tools for event investigation, new data, from experienced practitioners in the nuclear industry and/or regulatory institutions are needed. It is planned to collect such data, using the questionnaire prepared and performing the survey currently underway. This is the second step in carrying out an inventory of, reviewing, comparing and evaluating the most recent data on developments and systematic approaches in event investigation, used by organisations (mainly utilities) in the EU Member States. Once the data from this survey are collected and analysed, the final recommendations and conclusions will be developed and presented in the final report on this topic. This should help current and prospective investigators to choose the most suitable and efficient event investigation methods and tools for their particular needs.JRC.DDG.F.5-Safety of present nuclear reactor

Operating and maintenance cost reduction using probabilistic risk assessment (PRA)

"January 1992."Includes bibliographical references (pages 129-132)Final report, "Operating and maintenance cost reduction using probabilistic risk assessment (PRA)"This study quantifies the change in one measure of plant risk, the frequency of loss of long-term decay heat removal, due to changes in maintenance at the James A. Fitzpatrick (JAF) plant. Quantification is accomplished in two steps. First, the effects of maintenance are quantified in terms of changes in: a) the frequency of common cause failure of residual heat removal (RHR) pumps and b) the frequency with which operators fail to correctly restore the RHR system following maintenance. These parameters are selected as the result of an importance analysis for the plant. Second, the changes in these two parameters are propagated through a simple plant model to obtain the associated change in plant risk. Based on this study's assessment of the current maintenance program at JAF, it appears that the potential for significant risk reduction due to improved maintenance is not extremely large; an optimal program might lead to an 80% reduction. The optimal program would place a stronger emphasis on predictive maintenance, and would employ improved procedures for RHR pump maintenance. There is potential for significant risk increase (around a factor of 70) if the maintenance program is significantly degraded (e.g., if post-maintenance is deemphasized). This study shows how, at a simple level, maintenance program changes can be quantified without explicit modeling of the details of a plant's management and organizational structure. However, such modeling may be required: a) to more strongly justify the quantitative factors used in the analysis and b) to quantify the effect of other program changes not yet treated (e.g., the strengthening of program elements ensuring feedback of information to organization). In addition, failure data specific to the JAF plant are also needed to increase the confidence in the quantitative results of this study.Sponsored by New York Power Authority, White Plains, NY under contract no. S-90-0019

Lessons learned from past accidents - The integration of human and organizational factors with the technical aspect

It is of prime importance to ensure the safety of chemical process plants due to volatile nature of the industry and drastic consequences of the accidents. A number of parameters can affect the safety of the process plants. One of the main parameters that has the influence on the safety of operations is the Human and Organizational Factors (HOF) as suggested by numbers of existing studies. Therefore, in order to enhance the safety of operations it is required to improve the HOF. These factors can be improved by an integrated approach as proposed in this work, instead looking at these factors in an isolation. A number of existing risk assessment approaches have been analysed in this work and their compliance requirements to the relevant International Standards with respect to the HOF. A new quantitative methodology “Method for Error Deduction and Incident Analysis (MEDIA)” has been developed in this work. During the development of this methodology, practicality; consistency; integration with other risk assessment techniques and efficient use of information were explicitly ensured. The MEDIA can help to integrate the HOF around the technical aspect and can prioritize the follow up actions based on risk. The quantification of this methodology is based on results of the accident analysis, that has been carried out in this work. The accidents of 25 years (1988-2012) in the Seveso establishments and that were reported to the European Commission’s Major Accident Reporting System (eMARS) have been studied. The results from the accident analysis have further used in order to learn lessons and to propose future recommendations. These recommendations are mainly aimed at further integration of the HOF and to improve the overall safety of chemical process plants. More specifically, these recommendations are addressed to the use of organizational checklist during the Hazard Identification (HAZID) study; improvement of existing eMARS reporting structure and the legal obligation towards the EU Member States to report their accidents to the European Commission

Human reliability analysis-accounting for human actions and external factors through the project life cycle

Airplanes, ships, nuclear power plants and chemical production plants (including oil & gas facilities) are examples of industries that depend upon the interaction between operators and machines. Consequently, to assess the risks of those systems, not only the reliability of the technological components has to be accounted for, but also the ‘human model’. For this reason, engineers have been working together with psychologists and sociologists to understand cognitive functions and how the organisational context influences individual actions. Human Reliability Analysis (HRA) identifies and analyses the causes, consequences and contributions of human performance (including failures) in complex sociotechnical systems. Generally, HRA research is concentrated in modelling workers’ performance in the “sharp-end”, assessing the ones directly involved in handling the system, especially operators. However, in theory, a reliability analysis can be applied to any kind of human action, including those from designers and managers. This research will evaluate a way of conducting HRA in the design process, as previous research has demonstrated that design failure is the predominant contributor to human errors (Moura et al., 2016). Bayesian Network (BN) – a systematic way of learning from experience and incorporating new evidence (deterministic or probabilistic) – is proposed to model the complex relationships within cognitive functions, organisational and technological factors. Conditional probability tables have been obtained from a dataset of major accidents from different industry sectors (Moura et al. 2017), using a classification scheme developed by Hollnagel (1998) for an HRA method called CREAM – Cognitive Reliability and Error Analysis Method. The model allows to infer which factors most influence human performance in different scenarios. Also, we will discuss if the model can be applied to any human actions through the project life cycle— since the design phase to the operational phase, including their management