The SPAR-H human reliability analysis method (original) (raw)
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Annals of Nuclear Energy, 2017
This study presents the quantitative part of the analysis of human actions (HAs) in human reliability analysis (HRA) for a research reactor level 1 probabilistic safety assessment (PSA). In responding to the abnormal conditions, type C HAs which are identified in the previous study are elucidated furthermore for their cognitive response element, prior to the quantification of their human error probabilities (HEPs). The HEPs are afterward quantified by mean of Standardized Plant Analysis Risk-Human Reliability Analysis (SPAR-H) method. For the type A and B HAs, their HEPs are quantified by using the Technique for Human Error Rate Prediction (THERP) method. Incorporation of the human failure events (HFEs) and their respective HEPs into the PSA models further allow the evaluation of human contribution to the risk associated with the operation of the research reactor. In this study, it is quantitatively shown that human attribute has major contribution to the risk and with the introduction of nominal level of procedureperformance shaping factor (PSF), risk can be significantly reduced.
A case study of a human reliability assessment for an existing nuclear power plant
Applied Ergonomics, 1996
This paper concerns a Human Reliability Assessment (HRA) carried out over a period of two years as part of a Nuclear Power Plant Probabilistic Safety Assessment (PSA). The HRA involved all elements of the HRA process, from problem definition and task analysis, through error representation and quantification, to impact assessment, error reduction, quality assurance and documentation. The aim of this paper is to show the methodology of HRA as applied in a real PSA, highlighting various elements of the HRA, and how the HRA interfaced with the PSA. Comments are also given on practical aspects and impacts of the various approaches, and the usefulness of a hybrid HRA/P!JA team approach.
Operator reliability study for Probabilistic Safety Analysis of an operating research reactor
Annals of Nuclear Energy, 2015
A Level 1 Probabilistic Safety Analysis (PSA) for the TRIGA Mark II research reactor of Malaysian Nuclear Agency has been developed to evaluate the potential risk in its operation. In conjunction to this PSA development, Human Reliability Analysis (HRA) is performed in order to determine human contribution to the risk. The aim of this study is to qualitatively analyze human actions (HAs) involved in the operation of this reactor according to the qualitative part of the HRA framework for PSA which is namely the identification, qualitative screening and modeling of HAs. By performing this framework, Human Failure Events (HFEs) of significant impact to the reactor safety are systematically analyzed and incorporated into the PSA structure. A part of the findings in this study will become the input for the subsequent quantitative part of the HRA framework, i.e. the Human Error Probability (HEP) quantification.
The main goal of this Nordic Nuclear Safety Research Council (NKS) project is to produce guidance for how to use human reliability analysis (HRA) to strengthen overall safety. The project consists of two sub-studies: The Nordic Point of View – A User Needs Analysis, and The American Point of View – Insights of How the US Nuclear Industry Works with HRA. The purpose of the Nordic Point of View study was a user needs analysis that aimed to survey current HRA practices in the Nordic nuclear industry, with the main focus being to connect HRA to system design. In this study, 26 Nordic (Swedish and Finnish) nuclear power plant specialists with research, practitioner, and regulatory expertise in HRA, PRA, HSI, and human performance were interviewed. This study was completed in 2009. This study concludes that HRA is an important tool when dealing with human factors in control room design or modernizations. The Nordic Point of View study showed areas where the use of HRA in the Nordic nuclea...
A Model-Based Human Reliability Data Collection
In response to a Staff Requirements Memorandum (SRM) to the Advisory Committee on Reactor Safeguards (ACRS), the US Nuclear Regulatory Commission (NRC) has undertaken a research effort to create a consensus approach to human reliability analysis (HRA). This paper provides an overview of the approach being developed. The approach introduces the " crew response tree " (CRT) concept, which depicts the human failure events in a manner parallel to the PRA event tree process, provides a structure for capturing the " context " associated with the human failure events under analysis, and uses the Information Processing Model as a platform to identify potential failures. It incorporates behavioral science knowledge by providing the decompositions of human failures/failure mechanisms/failure factors built from a top-down and bottom-up approach, the latter reflecting those findings from scientific papers that document theories and data of interest. The structure provides a roadmap for incorporating the phenomena with which crews would be dealing, the plant characteristics (e.g., design, indications, procedures, training), and human performance capabilities (awareness, decision, action). In terms of quantification, the approach uses the typical PRA conditional probability expression, which is delineated to a level adequate for associating the probability of a human failure event with conditional probabilities of the associated contexts, failure mechanisms, and the underlying factors (e.g., performance shaping factors). Such mathematical formulation can be used to directly estimate HEPs using various data sources (e.g., expert estimations, anchor values, simulator or historical data), or can be modified to interface with existing quantification approaches. 1
A Human Reliability Approach Developed for Data Collection
In response to a Staff Requirements Memorandum (SRM) to the Advisory Committee on Reactor Safeguards (ACRS), the US Nuclear Regulatory Commission (NRC) has undertaken a research effort to create a consensus approach to human reliability analysis (HRA). This paper provides an overview of the approach being developed. The approach introduces the " crew response tree " (CRT) concept, which depicts the human failure events in a manner parallel to the PRA event tree process, provides a structure for capturing the " context " associated with the human failure events under analysis, and uses the Information Processing Model as a platform to identify potential failures. It incorporates behavioral science knowledge by providing the decompositions of human failures/failure mechanisms/failure factors built from a top-down and bottom-up approach, the latter reflecting those findings from scientific papers that document theories and data of interest. The structure provides a roadmap for incorporating the phenomena with which crews would be dealing, the plant characteristics (e.g., design, indications, procedures, training), and human performance capabilities (awareness, decision, action). In terms of quantification, the approach uses the typical PRA conditional probability expression, which is delineated to a level adequate for associating the probability of a human failure event with conditional probabilities of the associated contexts, failure mechanisms, and the underlying factors (e.g., performance shaping factors). Such mathematical formulation can be used to directly estimate HEPs using various data sources (e.g., expert estimations, anchor values, simulator or historical data), or can be modified to interface with existing quantification approaches. 1
PsycEXTRA Dataset, 2000
A variety of methods have been developed to generate human error probabilities for use in the US nuclear power industry. When actual operations data are not available, it is necessary for an analyst to estimate these probabilities. Most approaches, including THERP, ASEP, SLIM-MAUD, and SPAR-H, feature an atomistic approach to characterizing and estimating error. The atomistic approach is based on the notion that events and their causes can be decomposed and individually quantified. In contrast, in the holistic approach, such as found in ATHEANA, the analysis centers on the entire event, which is typically quantified as an indivisible whole. The distinction between atomistic and holistic approaches is important in understanding the nature of human reliability analysis quantification and the utility and shortcomings associated with each approach.
2009
Since the Reactor Safety Study in the early 1970's, human reliability analysis (HRA) has been evolving towards a better ability to account for the factors and conditions that can lead humans to take unsafe actions and thereby provide better estimates of the likelihood of human error for probabilistic risk assessments (PRAs). The purpose of this paper is to provide an overview of recent reviews of operational events and advances in the behavioral sciences that have impacted the evolution of HRA methods and contributed to improvements. The paper discusses the importance of human errors in complex human-technical systems, examines why humans contribute to accidents and unsafe conditions, and discusses how lessons learned over the years have changed the perspective and approach for modeling human behavior in PRAs of complicated domains such as nuclear power plants. It is argued that it has become increasingly more important to understand and model the more cognitive aspects of human performance and to address the broader range of factors that have been shown to influence human performance in complex domains. The paper concludes by addressing the current ability of HRA to adequately predict human failure events and their likelihood.
A Model-Based Human Reliability Analysis Framework
In response to a Staff Requirements Memorandum (SRM) to the Advisory Committee on Reactor Safeguards (ACRS), the US Nuclear Regulatory Commission (NRC) has undertaken a research effort to create a consensus approach to human reliability analysis (HRA). This paper provides an overview of the approach being developed. The approach introduces the "crew response tree" (CRT) concept, which depicts the human failure events in a manner parallel to the PRA event tree process, provides a structure for capturing the "context" associated with the human failure events under analysis, and uses the Information Processing Model as a platform to identify potential failures. It incorporates behavioral science knowledge by providing the decompositions of human failures/failure mechanisms/failure factors built from a top-down and bottom-up approach, the latter reflecting those findings from scientific papers that document theories and data of interest. The structure provides a road...