Research review and development trends of human reliability analysis techniques

International Nuclear Information System (INIS)

Li Pengcheng; Chen Guohua; Zhang Li; Dai Licao

2011-01-01

Human reliability analysis (HRA) methods are reviewed. The theoretical basis of human reliability analysis, human error mechanism, the key elements of HRA methods as well as the existing HRA methods are respectively introduced and assessed. Their shortcomings,the current research hotspot and difficult problems are identified. Finally, it takes a close look at the trends of human reliability analysis methods. (authors)

Development of new HRA methods based upon operational experience

International Nuclear Information System (INIS)

Cooper, S.E.; Luckas, W.J.; Barriere, M.T.; Wreathall, J.

2004-01-01

Under the auspices of the US Nuclear Regulatory Commission (NRC), previously unaddressed human reliability issues are being investigated in order to support the development of human reliability analysis (HRA) methods for both low power and shutdown (LP and S) and full-power conditions. Actual operational experience, such as that reported in Licensee Event Reports (LERs), have been used to gain insights and provide a basis for the requirements of new HRA methods. In particular, operational experience has shown that new HRA methods for LP and S must address human-induced initiators, errors of commission, mistakes (vs. slips), dependencies, and the effects of multiple performance shaping factors (PSFs). (author)

Implications of an HRA framework for quantifying human acts of commission and dependency: Development of a methodology for conducting an integrated HRA/PRA

International Nuclear Information System (INIS)

Barriere, M.T.; Luckas, W.J.; Brown, W.S.; Cooper, S.E.; Wreathall, J.; Bley, D.C.

1994-01-01

To support the development of a refined human reliability analysis (HRA) framework, to address identified HRA user needs and improve HRA modeling, unique aspects of human performance have been identified from an analysis of actual plant-specific events. Through the use of the refined framework, relationships between the following HRA, human factors and probabilistic risk assessment (PRA) elements were described: the PRA model, plant states, plant conditions, PRA basic events, unsafe human actions, error mechanisms, and performance shaping factors (PSFs). The event analyses performed in the context of the refined HRA framework, identified the need for new HRA methods that are capable of: evaluating a range of different error mechanisms (e.g., slips as well as mistakes); addressing errors of commission (EOCs) and dependencies between human actions; and incorporating the influence of plant conditions and multiple PSFs on human actions. This report discusses the results of the assessment of user needs, the refinement of the existing HRA framework, as well as, the current status on EOCs, and human dependencies

Implications of an HRA framework for quantifying human acts of commission and dependency: Development of a methodology for conducting an integrated HRA/PRA

International Nuclear Information System (INIS)

Barriere, M.T.; Luckas, W.J.; Brown, W.S.; Cooper, S.E.; Wreathall, J.; Bley, D.C.

1993-01-01

To support the development of a refined human reliability analysis (HRA) framework, to address identified HRA user needs and improve HRA modeling, unique aspects of human performance have been identified from an analysis of actual plant-specific events. Through the use of the refined framework, relationships between the following HRA, human factors and probabilistic risk assessment (PRA) elements were described: the PRA model, plant states, plant conditions, PRA basic events, unsafe human actions, error mechanisms, and performance shaping factors (PSFs). The event analyses performed in the context of the refined HRA framework, identified the need for new HRA methods that are capable of: evaluating a range of different error mechanisms (e.g., slips as well as mistakes); addressing errors of commission (EOCs) and dependencies between human actions; and incorporating the influence of plant conditions and multiple PSFs on human actions. This report discusses the results of the assessment of user needs, the refinement of the existing HRA framework, as well as, the current status on EOCs, and human dependencies

Selecting the seismic HRA approach for Savannah River Plant PRA revision 1

International Nuclear Information System (INIS)

Papouchado, K.; Salaymeh, J.

1993-10-01

The Westinghouse Savannah River Company (WSRC) has prepared a level I probabilistic risk assessment (PRA), Rev. 0 of reactor operations for externally-initiated events including seismic events. The SRS PRA, Rev. 0 Seismic HRA received a critical review that expressed skepticism with the approach used for human reliability analysis because it had not been previously used and accepted in other published PRAs. This report provides a review of published probabilistic risk assessments (PRAs), the associated methodology guidance documents, and the psychological literature to identify parameters important to seismic human reliability analysis (HRA). It also describes a recommended approach for use in the Savannah River Site (SRS) PRA. The SRS seismic event PRA performs HRA to account for the contribution of human errors in the accident sequences. The HRA of human actions during and after a seismic event is an area subject to many uncertainties and involves significant analyst judgment. The approach recommended by this report is based on seismic HRA methods and associated issues and concerns identified from the review of these referenced documents that represent the current state-of-the- art knowledge and acceptance in the seismic HRA field

Task Analysis of Emergency Operating Procedures for Generating Quantitative HRA Data

Energy Technology Data Exchange (ETDEWEB)

Kim, Yochan; Park, Jinkyun; Kim, Seunghwan; Choi, Sun Yeong; Jung, Wondea; Jang, Inseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

In this paper, the analysis results of the emergency task in the procedures (EOPs; emergency operating procedures) that can be observed from the simulator data are introduced. The task type, component type, system type, and additional information related with the performance of the operators were described. In addition, a prospective application of the analyzed information to HEP quantification process was discussed. In the probabilistic safety analysis (PSA) field, various human reliability analyses (HRAs) have been performed to produce estimates of human error probabilities (HEPs) for significant tasks in complex socio-technical systems. To this end, Many HRA methods have provided basic or nominal HEPs for typical tasks and the quantitative relations describing how a certain performance context or performance shaping factors (PSFs) affects the HEPs. In the HRA community, however, the necessity of appropriate and sufficient human performance data has been recently indicated. This is because a wide range of quantitative estimates in the previous HRA methods are not supported by solid empirical bases. Hence, there have been attempts to collect HRA supporting data. For example, KAERI has started to collect information on both unsafe acts of operators and the relevant PSFs. A characteristic of the database that is being developed at KAERI is that human errors and related PSF surrogates that can be objectively observable are collected from full-scope simulator experiences. In this environment, to produce concretely grounded bases of the HEPs, the traits or attributes of tasks where significant human errors can be observed should be definitely determined. The determined traits should be applicable to compare the HEPs on the traits with the data in previous HRA methods or databases. In this study, task characteristics in a Westinghouse type of EOPs were analyzed with the defining task, component, and system taxonomies.

Human reliability analysis of Lingao Nuclear Power Station

International Nuclear Information System (INIS)

Zhang Li; Huang Shudong; Yang Hong; He Aiwu; Huang Xiangrui; Zheng Tao; Su Shengbing; Xi Haiying

2001-01-01

The necessity of human reliability analysis (HRA) of Lingao Nuclear Power Station are analyzed, and the method and operation procedures of HRA is briefed. One of the human factors events (HFE) is analyzed in detail and some questions of HRA are discussed. The authors present the analytical results of 61 HFEs, and make a brief introduction of HRA contribution to Lingao Nuclear Power Station

Human Reliability Analysis for Small Modular Reactors

Energy Technology Data Exchange (ETDEWEB)

Ronald L. Boring; David I. Gertman

2012-06-01

Because no human reliability analysis (HRA) method was specifically developed for small modular reactors (SMRs), the application of any current HRA method to SMRs represents tradeoffs. A first- generation HRA method like THERP provides clearly defined activity types, but these activity types do not map to the human-system interface or concept of operations confronting SMR operators. A second- generation HRA method like ATHEANA is flexible enough to be used for SMR applications, but there is currently insufficient guidance for the analyst, requiring considerably more first-of-a-kind analyses and extensive SMR expertise in order to complete a quality HRA. Although no current HRA method is optimized to SMRs, it is possible to use existing HRA methods to identify errors, incorporate them as human failure events in the probabilistic risk assessment (PRA), and quantify them. In this paper, we provided preliminary guidance to assist the human reliability analyst and reviewer in understanding how to apply current HRA methods to the domain of SMRs. While it is possible to perform a satisfactory HRA using existing HRA methods, ultimately it is desirable to formally incorporate SMR considerations into the methods. This may require the development of new HRA methods. More practicably, existing methods need to be adapted to incorporate SMRs. Such adaptations may take the form of guidance on the complex mapping between conventional light water reactors and small modular reactors. While many behaviors and activities are shared between current plants and SMRs, the methods must adapt if they are to perform a valid and accurate analysis of plant personnel performance in SMRs.

Standardization of domestic human reliability analysis and experience of human reliability analysis in probabilistic safety assessment for NPPs under design

International Nuclear Information System (INIS)

Kang, D. I.; Jung, W. D.

2002-01-01

This paper introduces the background and development activities of domestic standardization of procedure and method for Human Reliability Analysis (HRA) to avoid the intervention of subjectivity by HRA analyst in Probabilistic Safety Assessment (PSA) as possible, and the review of the HRA results for domestic nuclear power plants under design studied by Korea Atomic Energy Research Institute. We identify the HRA methods used for PSA for domestic NPPs and discuss the subjectivity of HRA analyst shown in performing a HRA. Also, we introduce the PSA guidelines published in USA and review the HRA results based on them. We propose the system of a standard procedure and method for HRA to be developed

MODIFICATION OF THE SPAR-H METHOD TO SUPPORT HRA FOR LEVEL 2 PSA

Energy Technology Data Exchange (ETDEWEB)

St. Germain, S.; Boring, R.; Banaseanu, G.; Akl, Y.; Xu, M.

2016-10-01

Currently, available Human Reliability Analysis (HRA) methods were generally developed to support Level 1 Probabilistic Safety Analysis (PSA) models. There has been an increased emphasis placed on Level 2 PSA in recent years; however, the currently used HRA methods are not ideal for this application, including the SPAR-H method. Challenges that will likely be present during a severe accident such as degraded or hazardous operating conditions, shift in control from the main control room to the technical support center, unavailability of instrumentation, and others are not routinely considered for Level 1 HRA analysis. These factors combine to create a much more uncertain condition to be accounted for in the HRA analysis. While the SPAR-H shaping factors were established to support Level 1 HRA, previous studies have shown it may be used for Level 2 HRA analysis as well. The Canadian Nuclear Safety Commission (CNSC) and Idaho National Laboratory (INL) in a joint project are investigating modifications to the SPAR-H method to create more consistency in applying the performance shaping factors used in the method for Level 2 analysis.

Inclusion of fatigue effects in human reliability analysis

Energy Technology Data Exchange (ETDEWEB)

Griffith, Candice D. [Vanderbilt University, Nashville, TN (United States); Mahadevan, Sankaran, E-mail: sankaran.mahadevan@vanderbilt.edu [Vanderbilt University, Nashville, TN (United States)

2011-11-15

The effect of fatigue on human performance has been observed to be an important factor in many industrial accidents. However, defining and measuring fatigue is not easily accomplished. This creates difficulties in including fatigue effects in probabilistic risk assessments (PRA) of complex engineering systems that seek to include human reliability analysis (HRA). Thus the objectives of this paper are to discuss (1) the importance of the effects of fatigue on performance, (2) the difficulties associated with defining and measuring fatigue, (3) the current status of inclusion of fatigue in HRA methods, and (4) the future directions and challenges for the inclusion of fatigue, specifically sleep deprivation, in HRA. - Highlights: >We highlight the need for fatigue and sleep deprivation effects on performance to be included in human reliability analysis (HRA) methods. Current methods do not explicitly include sleep deprivation effects. > We discuss the difficulties in defining and measuring fatigue. > We review sleep deprivation research, and discuss the limitations and future needs of the current HRA methods.

The quantitative failure of human reliability analysis

Energy Technology Data Exchange (ETDEWEB)

Bennett, C.T.

1995-07-01

This philosophical treatise argues the merits of Human Reliability Analysis (HRA) in the context of the nuclear power industry. Actually, the author attacks historic and current HRA as having failed in informing policy makers who make decisions based on risk that humans contribute to systems performance. He argues for an HRA based on Bayesian (fact-based) inferential statistics, which advocates a systems analysis process that employs cogent heuristics when using opinion, and tempers itself with a rational debate over the weight given subjective and empirical probabilities.

A probabilistic cognitive simulator for HRA studies (PROCOS)

International Nuclear Information System (INIS)

Trucco, P.; Leva, M.C.

2007-01-01

The paper deals with the development of a simulator for approaching human errors in complex operational frameworks (e.g., plant commissioning). The aim is to integrate the quantification capabilities of the so-called 'first-generation' human reliability assessment (HRA) methods with a cognitive evaluation of the operator. The simulator allows analysing both error prevention and error recovery. It integrates cognitive human error analysis with standard hazard analysis methods (Hazop and event tree) by means of a 'semi static approach'. The comparison between the results obtained through the proposed approach and those of a traditional HRA method such as human error assessment and reduction technique, shows the capability of the simulator to provide coherent and accurate analysis

Inclusion of fatigue effects in human reliability analysis

International Nuclear Information System (INIS)

Griffith, Candice D.; Mahadevan, Sankaran

2011-01-01

The effect of fatigue on human performance has been observed to be an important factor in many industrial accidents. However, defining and measuring fatigue is not easily accomplished. This creates difficulties in including fatigue effects in probabilistic risk assessments (PRA) of complex engineering systems that seek to include human reliability analysis (HRA). Thus the objectives of this paper are to discuss (1) the importance of the effects of fatigue on performance, (2) the difficulties associated with defining and measuring fatigue, (3) the current status of inclusion of fatigue in HRA methods, and (4) the future directions and challenges for the inclusion of fatigue, specifically sleep deprivation, in HRA. - Highlights: →We highlight the need for fatigue and sleep deprivation effects on performance to be included in human reliability analysis (HRA) methods. Current methods do not explicitly include sleep deprivation effects. → We discuss the difficulties in defining and measuring fatigue. → We review sleep deprivation research, and discuss the limitations and future needs of the current HRA methods.

Human reliability analysis methods for probabilistic safety assessment

International Nuclear Information System (INIS)

Pyy, P.

2000-11-01

Human reliability analysis (HRA) of a probabilistic safety assessment (PSA) includes identifying human actions from safety point of view, modelling the most important of them in PSA models, and assessing their probabilities. As manifested by many incidents and studies, human actions may have both positive and negative effect on safety and economy. Human reliability analysis is one of the areas of probabilistic safety assessment (PSA) that has direct applications outside the nuclear industry. The thesis focuses upon developments in human reliability analysis methods and data. The aim is to support PSA by extending the applicability of HRA. The thesis consists of six publications and a summary. The summary includes general considerations and a discussion about human actions in the nuclear power plant (NPP) environment. A condensed discussion about the results of the attached publications is then given, including new development in methods and data. At the end of the summary part, the contribution of the publications to good practice in HRA is presented. In the publications, studies based on the collection of data on maintenance-related failures, simulator runs and expert judgement are presented in order to extend the human reliability analysis database. Furthermore, methodological frameworks are presented to perform a comprehensive HRA, including shutdown conditions, to study reliability of decision making, and to study the effects of wrong human actions. In the last publication, an interdisciplinary approach to analysing human decision making is presented. The publications also include practical applications of the presented methodological frameworks. (orig.)

An Estimation of Human Error Probability of Filtered Containment Venting System Using Dynamic HRA Method

Energy Technology Data Exchange (ETDEWEB)

Jang, Seunghyun; Jae, Moosung [Hanyang University, Seoul (Korea, Republic of)

2016-10-15

The human failure events (HFEs) are considered in the development of system fault trees as well as accident sequence event trees in part of Probabilistic Safety Assessment (PSA). As a method for analyzing the human error, several methods, such as Technique for Human Error Rate Prediction (THERP), Human Cognitive Reliability (HCR), and Standardized Plant Analysis Risk-Human Reliability Analysis (SPAR-H) are used and new methods for human reliability analysis (HRA) are under developing at this time. This paper presents a dynamic HRA method for assessing the human failure events and estimation of human error probability for filtered containment venting system (FCVS) is performed. The action associated with implementation of the containment venting during a station blackout sequence is used as an example. In this report, dynamic HRA method was used to analyze FCVS-related operator action. The distributions of the required time and the available time were developed by MAAP code and LHS sampling. Though the numerical calculations given here are only for illustrative purpose, the dynamic HRA method can be useful tools to estimate the human error estimation and it can be applied to any kind of the operator actions, including the severe accident management strategy.

Development of a methodology for conducting an integrated HRA/PRA --. Task 1, An assessment of human reliability influences during LP&S conditions PWRs

Energy Technology Data Exchange (ETDEWEB)

Luckas, W.J.; Barriere, M.T.; Brown, W.S. [Brookhaven National Lab., Upton, NY (United States); Wreathall, J. [Wreathall (John) and Co., Dublin, OH (United States); Cooper, S.E. [Science Applications International Corp., McLean, VA (United States)

1993-06-01

During Low Power and Shutdown (LP&S) conditions in a nuclear power plant (i.e., when the reactor is subcritical or at less than 10--15% power), human interactions with the plant`s systems will be more frequent and more direct. Control is typically not mediated by automation, and there are fewer protective systems available. Therefore, an assessment of LP&S related risk should include a greater emphasis on human reliability than such an assessment made for power operation conditions. In order to properly account for the increase in human interaction and thus be able to perform a probabilistic risk assessment (PRA) applicable to operations during LP&S, it is important that a comprehensive human reliability assessment (HRA) methodology be developed and integrated into the LP&S PRA. The tasks comprising the comprehensive HRA methodology development are as follows: (1) identification of the human reliability related influences and associated human actions during LP&S, (2) identification of potentially important LP&S related human actions and appropriate HRA framework and quantification methods, and (3) incorporation and coordination of methodology development with other integrated PRA/HRA efforts. This paper describes the first task, i.e., the assessment of human reliability influences and any associated human actions during LP&S conditions for a pressurized water reactor (PWR).

Representing cognitive activities and errors in HRA trees

International Nuclear Information System (INIS)

Gertman, D.I.

1992-01-01

A graphic representation method is presented herein for adapting an existing technology--human reliability analysis (HRA) event trees, used to support event sequence logic structures and calculations--to include a representation of the underlying cognitive activity and corresponding errors associated with human performance. The analyst is presented with three potential means of representing human activity: the NUREG/CR-1278 HRA event tree approach; the skill-, rule- and knowledge-based paradigm; and the slips, lapses, and mistakes paradigm. The above approaches for representing human activity are integrated in order to produce an enriched HRA event tree -- the cognitive event tree system (COGENT)-- which, in turn, can be used to increase the analyst's understanding of the basic behavioral mechanisms underlying human error and the representation of that error in probabilistic risk assessment. Issues pertaining to the implementation of COGENT are also discussed

Failures without errors: quantification of context in HRA

International Nuclear Information System (INIS)

Fujita, Yushi; Hollnagel, Erik

2004-01-01

PSA-cum-human reliability analysis (HRA) has traditionally used individual human actions, hence individual 'human errors', as a meaningful unit of analysis. This is inconsistent with the current understanding of accidents, which points out that the notion of 'human error' is ill defined and that adverse events more often are the due to the working conditions than to people. Several HRA approaches, such as ATHEANA and CREAM have recognised this conflict and proposed ways to deal with it. This paper describes an improvement of the basic screening method in CREAM, whereby a rating of the performance conditions can be used to calculate a Mean Failure Rate directly without invoking the notion of human error

HRA Data Collection from the Simulations of Abnormal Situations

Energy Technology Data Exchange (ETDEWEB)

Kim, Yo Chan; Park, Jink Yun; Jung, Won Dea [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

In this study, it was revealed that the designed worksheets were feasible to collect HRA data, especially in abnormal situations. The defined taxonomy of UAs was unambiguous to distinguish actions of operators and quantify the probabilities. It is widely recognized that reliability of operators are critical to complex socio-technical systems. For this reason, human reliability analysis (HRA), which aims to identify unsafe actions (UAs) that contribute to risks of the systems and assess the failure rates of the actions, has been conducted. Although many techniques of HRA have been developed and used in many years, many reports indicated lack of database for supporting empirical bases of HRA methods. Thus, there have been recent efforts to collect data about human reliability from plant experience, simulator experiment or qualification, and laboratory experiments. As one of these efforts, KAERI also established a guideline to collect information about human reliability and performance shaping factors from simulators. This guideline particularly presented a set of worksheets that allows comprehensively gathering objectively observable information in simulations of emergency situations. This paper reports the process and preliminary results of the data collection from the simulations of abnormal situations based on the developed worksheets in KAERI database guideline. We analyzed operator behaviors of the sixteen experiments for the two kinds of abnormal situations: RCP (reactor coolant pump) cyclone filter blockage and CDP (condensate pump) valve stuck. The UAs of operators were identified and quantified. Because the number of simulations was limited and data of various situations will be obtained continuously, it is uncertain to conclude the resulted probabilities. Based on the worksheet, operator behaviors in many different kinds of scenarios will be analyzed and the relations between human reliabilities and the observed factors will be also investigated.

Human Reliability Analysis For Computerized Procedures

International Nuclear Information System (INIS)

Boring, Ronald L.; Gertman, David I.; Le Blanc, Katya

2011-01-01

This paper provides a characterization of human reliability analysis (HRA) issues for computerized procedures in nuclear power plant control rooms. It is beyond the scope of this paper to propose a new HRA approach or to recommend specific methods or refinements to those methods. Rather, this paper provides a review of HRA as applied to traditional paper-based procedures, followed by a discussion of what specific factors should additionally be considered in HRAs for computerized procedures. Performance shaping factors and failure modes unique to computerized procedures are highlighted. Since there is no definitive guide to HRA for paper-based procedures, this paper also serves to clarify the existing guidance on paper-based procedures before delving into the unique aspects of computerized procedures.

Suggestions for an improved HRA method for use in Probabilistic Safety Assessment

International Nuclear Information System (INIS)

Parry, Gareth W.

1995-01-01

This paper discusses why an improved Human Reliability Analysis (HRA) approach for use in Probabilistic Safety Assessments (PSAs) is needed, and proposes a set of requirements on the improved HRA method. The constraints imposed by the need to embed the approach into the PSA methodology are discussed. One approach to laying the foundation for an improved method, using models from the cognitive psychology and behavioral science disciplines, is outlined

Applicability of HRA to support advanced MMI design review

International Nuclear Information System (INIS)

Kim, Inn Seock

2000-01-01

More than half of all incidents in large complex technological systems, particularly in nuclear power or aviation industries, were attributable in some way to human erroneous actions. These incidents were largely due to the human engineering deficiencies of man-machine interface (MMI). In nuclear industry, advanced computer-based MMI designs are emerging as part of new reactor designs. The impact of advanced MMI technology on the operator performance, and as a result, on plant safety should be thoroughly evaluated before such technology is actually adopted in nuclear power plants. This paper discusses the applicability of human reliability analysis (HRA) to support the design review process. Both the first-generation and the second-generation HRA methods are considered focusing on a couple of promising HRA methods, i.e., ATHEANA and CREAM, with the potential to assist the design review process. (author)

Human reliability analysis of control room operators

Energy Technology Data Exchange (ETDEWEB)

Santos, Isaac J.A.L.; Carvalho, Paulo Victor R.; Grecco, Claudio H.S. [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil)

2005-07-01

Human reliability is the probability that a person correctly performs some system required action in a required time period and performs no extraneous action that can degrade the system Human reliability analysis (HRA) is the analysis, prediction and evaluation of work-oriented human performance using some indices as human error likelihood and probability of task accomplishment. Significant progress has been made in the HRA field during the last years, mainly in nuclear area. Some first-generation HRA methods were developed, as THERP (Technique for human error rate prediction). Now, an array of called second-generation methods are emerging as alternatives, for instance ATHEANA (A Technique for human event analysis). The ergonomics approach has as tool the ergonomic work analysis. It focus on the study of operator's activities in physical and mental form, considering at the same time the observed characteristics of operator and the elements of the work environment as they are presented to and perceived by the operators. The aim of this paper is to propose a methodology to analyze the human reliability of the operators of industrial plant control room, using a framework that includes the approach used by ATHEANA, THERP and the work ergonomics analysis. (author)

Assessment of modern methods of human factor reliability analysis in PSA studies

International Nuclear Information System (INIS)

Holy, J.

2001-12-01

The report is structured as follows: Classical terms and objects (Probabilistic safety assessment as a framework for human reliability assessment; Human failure within the PSA model; Basic types of operator failure modelled in a PSA study and analyzed by HRA methods; Qualitative analysis of human reliability; Quantitative analysis of human reliability used; Process of analysis of nuclear reactor operator reliability in a PSA study); New terms and objects (Analysis of dependences; Errors of omission; Errors of commission; Error forcing context); and Overview and brief assessment of human reliability analysis (Basic characteristics of the methods; Assets and drawbacks of the use of each of HRA method; History and prospects of the use of the methods). (P.A.)

User's manual of a support system for human reliability analysis

International Nuclear Information System (INIS)

Yokobayashi, Masao; Tamura, Kazuo.

1995-10-01

Many kinds of human reliability analysis (HRA) methods have been developed. However, users are required to be skillful so as to use them, and also required complicated works such as drawing event tree (ET) and calculation of uncertainty bounds. Moreover, each method is not so complete that only one method of them is not enough to evaluate human reliability. Therefore, a personal computer (PC) based support system for HRA has been developed to execute HRA practically and efficiently. The system consists of two methods, namely, simple method and detailed one. The former uses ASEP that is a simplified THERP-technique, and combined method of OAT and HRA-ET/DeBDA is used for the latter. Users can select a suitable method for their purpose. Human error probability (HEP) data were collected and a database of them was built to use for the support system. This paper describes outline of the HRA methods, support functions and user's guide of the system. (author)

Standardizing the practice of human reliability analysis

International Nuclear Information System (INIS)

Hallbert, B.P.

1993-01-01

The practice of human reliability analysis (HRA) within the nuclear industry varies greatly in terms of posited mechanisms that shape human performance, methods of characterizing and analytically modeling human behavior, and the techniques that are employed to estimate the frequency with which human error occurs. This variation has been a source of contention among HRA practitioners regarding the validity of results obtained from different HRA methods. It has also resulted in attempts to develop standard methods and procedures for conducting HRAs. For many of the same reasons, the practice of HRA has not been standardized or has been standardized only to the extent that individual analysts have developed heuristics and consistent approaches in their practice of HRA. From the standpoint of consumers and regulators, this has resulted in a lack of clear acceptance criteria for the assumptions, modeling, and quantification of human errors in probabilistic risk assessments

Task analysis and computer aid development for human reliability analysis in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Yoon, W. C.; Kim, H.; Park, H. S.; Choi, H. H.; Moon, J. M.; Heo, J. Y.; Ham, D. H.; Lee, K. K.; Han, B. T. [Korea Advanced Institute of Science and Technology, Taejeon (Korea)

2001-04-01

Importance of human reliability analysis (HRA) that predicts the error's occurrence possibility in a quantitative and qualitative manners is gradually increased by human errors' effects on the system's safety. HRA needs a task analysis as a virtue step, but extant task analysis techniques have the problem that a collection of information about the situation, which the human error occurs, depends entirely on HRA analyzers. The problem makes results of the task analysis inconsistent and unreliable. To complement such problem, KAERI developed the structural information analysis (SIA) that helps to analyze task's structure and situations systematically. In this study, the SIA method was evaluated by HRA experts, and a prototype computerized supporting system named CASIA (Computer Aid for SIA) was developed for the purpose of supporting to perform HRA using the SIA method. Additionally, through applying the SIA method to emergency operating procedures, we derived generic task types used in emergency and accumulated the analysis results in the database of the CASIA. The CASIA is expected to help HRA analyzers perform the analysis more easily and consistently. If more analyses will be performed and more data will be accumulated to the CASIA's database, HRA analyzers can share freely and spread smoothly his or her analysis experiences, and there by the quality of the HRA analysis will be improved. 35 refs., 38 figs., 25 tabs. (Author)

Development of A Standard Method for Human Reliability Analysis of Nuclear Power Plants

International Nuclear Information System (INIS)

Jung, Won Dea; Kang, Dae Il; Kim, Jae Whan

2005-12-01

According as the demand of risk-informed regulation and applications increase, the quality and reliability of a probabilistic safety assessment (PSA) has been more important. KAERI started a study to standardize the process and the rules of HRA (Human Reliability Analysis) which was known as a major contributor to the uncertainty of PSA. The study made progress as follows; assessing the level of quality of the HRAs in Korea and identifying the weaknesses of the HRAs, determining the requirements for developing a standard HRA method, developing the process and rules for quantifying human error probability. Since the risk-informed applications use the ASME PSA standard to ensure PSA quality, the standard HRA method was developed to meet the ASME HRA requirements with level of category II. The standard method was based on THERP and ASEP HRA that are widely used for conventional HRA. However, the method focuses on standardizing and specifying the analysis process, quantification rules and criteria to minimize the deviation of the analysis results caused by different analysts. Several HRA experts from different organizations in Korea participated in developing the standard method. Several case studies were interactively undertaken to verify the usability and applicability of the standard method

Development of A Standard Method for Human Reliability Analysis of Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Jung, Won Dea; Kang, Dae Il; Kim, Jae Whan

2005-12-15

According as the demand of risk-informed regulation and applications increase, the quality and reliability of a probabilistic safety assessment (PSA) has been more important. KAERI started a study to standardize the process and the rules of HRA (Human Reliability Analysis) which was known as a major contributor to the uncertainty of PSA. The study made progress as follows; assessing the level of quality of the HRAs in Korea and identifying the weaknesses of the HRAs, determining the requirements for developing a standard HRA method, developing the process and rules for quantifying human error probability. Since the risk-informed applications use the ASME PSA standard to ensure PSA quality, the standard HRA method was developed to meet the ASME HRA requirements with level of category II. The standard method was based on THERP and ASEP HRA that are widely used for conventional HRA. However, the method focuses on standardizing and specifying the analysis process, quantification rules and criteria to minimize the deviation of the analysis results caused by different analysts. Several HRA experts from different organizations in Korea participated in developing the standard method. Several case studies were interactively undertaken to verify the usability and applicability of the standard method.

Operator reliability study for Probabilistic Safety Analysis of an operating research reactor

International Nuclear Information System (INIS)

Mohamed, F.; Hassan, A.; Yahaya, R.; Rahman, I.; Maskin, M.; Praktom, P.; Charlie, F.

2015-01-01

Highlights: • Human Reliability Analysis (HRA) for Level 1 Probabilistic Safety Analysis (PSA) is performed on research nuclear reactor. • Implemented qualitative HRA framework is addressed. • Human Failure Events of significant impact to the reactor safety are derived. - Abstract: 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

Development of a methodology for conducting an integrated HRA/PRA --

Energy Technology Data Exchange (ETDEWEB)

Luckas, W.J.; Barriere, M.T.; Brown, W.S. (Brookhaven National Lab., Upton, NY (United States)); Wreathall, J. (Wreathall (John) and Co., Dublin, OH (United States)); Cooper, S.E. (Science Applications International Corp., McLean, VA (United States))

1993-01-01

During Low Power and Shutdown (LP S) conditions in a nuclear power plant (i.e., when the reactor is subcritical or at less than 10--15% power), human interactions with the plant's systems will be more frequent and more direct. Control is typically not mediated by automation, and there are fewer protective systems available. Therefore, an assessment of LP S related risk should include a greater emphasis on human reliability than such an assessment made for power operation conditions. In order to properly account for the increase in human interaction and thus be able to perform a probabilistic risk assessment (PRA) applicable to operations during LP S, it is important that a comprehensive human reliability assessment (HRA) methodology be developed and integrated into the LP S PRA. The tasks comprising the comprehensive HRA methodology development are as follows: (1) identification of the human reliability related influences and associated human actions during LP S, (2) identification of potentially important LP S related human actions and appropriate HRA framework and quantification methods, and (3) incorporation and coordination of methodology development with other integrated PRA/HRA efforts. This paper describes the first task, i.e., the assessment of human reliability influences and any associated human actions during LP S conditions for a pressurized water reactor (PWR).

Advances in human reliability analysis in Mexico

International Nuclear Information System (INIS)

Nelson, Pamela F.; Gonzalez C, M.; Ruiz S, T.; Guillen M, D.; Contreras V, A.

2010-10-01

Human Reliability Analysis (HRA) is a very important part of Probabilistic Risk Analysis (PRA), and constant work is dedicated to improving methods, guidance and data in order to approach realism in the results as well as looking for ways to use these to reduce accident frequency at plants. Further, in order to advance in these areas, several HRA studies are being performed globally. Mexico has participated in the International HRA Empirical study with the objective of -benchmarking- HRA methods by comparing HRA predictions to actual crew performance in a simulator, as well as in the empirical study on a US nuclear power plant currently in progress. The focus of the first study was the development of an understanding of how methods are applied by various analysts, and characterize the methods for their capability to guide the analysts to identify potential human failures, and associated causes and performance shaping factors. The HRA benchmarking study has been performed by using the Halden simulator, 14 European crews, and 15 HRA equipment s (NRC, EPRI, and foreign HRA equipment s using different HRA methods). This effort in Mexico is reflected through the work being performed on updating the Laguna Verde PRA to comply with the ASME PRA standard. In order to be considered an HRA with technical adequacy, that is, be considered as a capability category II, for risk-informed applications, the methodology used for the HRA in the original PRA is not considered sufficiently detailed, and the methodology had to upgraded. The HCR/CBDT/THERP method was chosen, since this is used in many nuclear plants with similar design. The HRA update includes identification and evaluation of human errors that can occur during testing and maintenance, as well as human errors that can occur during an accident using the Emergency Operating Procedures. The review of procedures for maintenance, surveillance and operation is a necessary step in HRA and provides insight into the possible

User`s manual of a support system for human reliability analysis

Energy Technology Data Exchange (ETDEWEB)

Yokobayashi, Masao [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Tamura, Kazuo

1995-10-01

Many kinds of human reliability analysis (HRA) methods have been developed. However, users are required to be skillful so as to use them, and also required complicated works such as drawing event tree (ET) and calculation of uncertainty bounds. Moreover, each method is not so complete that only one method of them is not enough to evaluate human reliability. Therefore, a personal computer (PC) based support system for HRA has been developed to execute HRA practically and efficiently. The system consists of two methods, namely, simple method and detailed one. The former uses ASEP that is a simplified THERP-technique, and combined method of OAT and HRA-ET/DeBDA is used for the latter. Users can select a suitable method for their purpose. Human error probability (HEP) data were collected and a database of them was built to use for the support system. This paper describes outline of the HRA methods, support functions and user`s guide of the system. (author).

Development of an analysis rule of diagnosis error for standard method of human reliability analysis

International Nuclear Information System (INIS)

Jeong, W. D.; Kang, D. I.; Jeong, K. S.

2003-01-01

This paper presents the status of development of Korea standard method for Human Reliability Analysis (HRA), and proposed a standard procedure and rules for the evaluation of diagnosis error probability. The quality of KSNP HRA was evaluated using the requirement of ASME PRA standard guideline, and the design requirement for the standard HRA method was defined. Analysis procedure and rules, developed so far, to analyze diagnosis error probability was suggested as a part of the standard method. And also a study of comprehensive application was performed to evaluate the suitability of the proposed rules

Development of a methodology for conducting an integrated HRA/PRA --

International Nuclear Information System (INIS)

Luckas, W.J.; Barriere, M.T.; Brown, W.S.; Wreathall, J.; Cooper, S.E.

1993-01-01

During Low Power and Shutdown (LP ampersand S) conditions in a nuclear power plant (i.e., when the reactor is subcritical or at less than 10--15% power), human interactions with the plant's systems will be more frequent and more direct. Control is typically not mediated by automation, and there are fewer protective systems available. Therefore, an assessment of LP ampersand S related risk should include a greater emphasis on human reliability than such an assessment made for power operation conditions. In order to properly account for the increase in human interaction and thus be able to perform a probabilistic risk assessment (PRA) applicable to operations during LP ampersand S, it is important that a comprehensive human reliability assessment (HRA) methodology be developed and integrated into the LP ampersand S PRA. The tasks comprising the comprehensive HRA methodology development are as follows: (1) identification of the human reliability related influences and associated human actions during LP ampersand S, (2) identification of potentially important LP ampersand S related human actions and appropriate HRA framework and quantification methods, and (3) incorporation and coordination of methodology development with other integrated PRA/HRA efforts. This paper describes the first task, i.e., the assessment of human reliability influences and any associated human actions during LP ampersand S conditions for a pressurized water reactor (PWR)

An integrated approach to human reliability analysis -- decision analytic dynamic reliability model

International Nuclear Information System (INIS)

Holmberg, J.; Hukki, K.; Norros, L.; Pulkkinen, U.; Pyy, P.

1999-01-01

The reliability of human operators in process control is sensitive to the context. In many contemporary human reliability analysis (HRA) methods, this is not sufficiently taken into account. The aim of this article is that integration between probabilistic and psychological approaches in human reliability should be attempted. This is achieved first, by adopting such methods that adequately reflect the essential features of the process control activity, and secondly, by carrying out an interactive HRA process. Description of the activity context, probabilistic modeling, and psychological analysis form an iterative interdisciplinary sequence of analysis in which the results of one sub-task maybe input to another. The analysis of the context is carried out first with the help of a common set of conceptual tools. The resulting descriptions of the context promote the probabilistic modeling, through which new results regarding the probabilistic dynamics can be achieved. These can be incorporated in the context descriptions used as reference in the psychological analysis of actual performance. The results also provide new knowledge of the constraints of activity, by providing information of the premises of the operator's actions. Finally, the stochastic marked point process model gives a tool, by which psychological methodology may be interpreted and utilized for reliability analysis

The use of empirical data sources in HRA

International Nuclear Information System (INIS)

Hallbert, Bruce; Gertman, David; Lois, Erasmia; Marble, Julie; Blackman, Harold; Byers, James

2004-01-01

This paper presents a review of available information related to human performance to support Human Reliability Analysis (HRA) performed for nuclear power plants (NPPs). A number of data sources are identified as potentially useful. These include NPP licensee event reports, augmented inspection team reports, operator requalification data, results from the literature in experimental psychology, and the Aviation Safety Reporting System. The paper discusses how utilizing such information improves our capability to model and quantify human performance. In particular, the paper discusses how information related to performance shaping factors can be extracted from empirical data to determine their size effect, their relative effects, as well as their interactions. The paper concludes that appropriate use of existing sources can help addressing some of the important issues we are currently facing in HRA

An advanced human reliability analysis methodology: analysis of cognitive errors focused on

International Nuclear Information System (INIS)

Kim, J. H.; Jeong, W. D.

2001-01-01

The conventional Human Reliability Analysis (HRA) methods such as THERP/ASEP, HCR and SLIM has been criticised for their deficiency in analysing cognitive errors which occurs during operator's decision making process. In order to supplement the limitation of the conventional methods, an advanced HRA method, what is called the 2 nd generation HRA method, including both qualitative analysis and quantitative assessment of cognitive errors has been being developed based on the state-of-the-art theory of cognitive systems engineering and error psychology. The method was developed on the basis of human decision-making model and the relation between the cognitive function and the performance influencing factors. The application of the proposed method to two emergency operation tasks is presented

Modeling human reliability analysis using MIDAS

International Nuclear Information System (INIS)

Boring, R. L.

2006-01-01

This paper documents current efforts to infuse human reliability analysis (HRA) into human performance simulation. The Idaho National Laboratory is teamed with NASA Ames Research Center to bridge the SPAR-H HRA method with NASA's Man-machine Integration Design and Analysis System (MIDAS) for use in simulating and modeling the human contribution to risk in nuclear power plant control room operations. It is anticipated that the union of MIDAS and SPAR-H will pave the path for cost-effective, timely, and valid simulated control room operators for studying current and next generation control room configurations. This paper highlights considerations for creating the dynamic HRA framework necessary for simulation, including event dependency and granularity. This paper also highlights how the SPAR-H performance shaping factors can be modeled in MIDAS across static, dynamic, and initiator conditions common to control room scenarios. This paper concludes with a discussion of the relationship of the workload factors currently in MIDAS and the performance shaping factors in SPAR-H. (authors)

Issues in benchmarking human reliability analysis methods: A literature review

International Nuclear Information System (INIS)

Boring, Ronald L.; Hendrickson, Stacey M.L.; Forester, John A.; Tran, Tuan Q.; Lois, Erasmia

2010-01-01

There is a diversity of human reliability analysis (HRA) methods available for use in assessing human performance within probabilistic risk assessments (PRA). Due to the significant differences in the methods, including the scope, approach, and underlying models, there is a need for an empirical comparison investigating the validity and reliability of the methods. To accomplish this empirical comparison, a benchmarking study comparing and evaluating HRA methods in assessing operator performance in simulator experiments is currently underway. In order to account for as many effects as possible in the construction of this benchmarking study, a literature review was conducted, reviewing past benchmarking studies in the areas of psychology and risk assessment. A number of lessons learned through these studies is presented in order to aid in the design of future HRA benchmarking endeavors.

Issues in benchmarking human reliability analysis methods : a literature review.

Energy Technology Data Exchange (ETDEWEB)

Lois, Erasmia (US Nuclear Regulatory Commission); Forester, John Alan; Tran, Tuan Q. (Idaho National Laboratory, Idaho Falls, ID); Hendrickson, Stacey M. Langfitt; Boring, Ronald L. (Idaho National Laboratory, Idaho Falls, ID)

2008-04-01

There is a diversity of human reliability analysis (HRA) methods available for use in assessing human performance within probabilistic risk assessment (PRA). Due to the significant differences in the methods, including the scope, approach, and underlying models, there is a need for an empirical comparison investigating the validity and reliability of the methods. To accomplish this empirical comparison, a benchmarking study is currently underway that compares HRA methods with each other and against operator performance in simulator studies. In order to account for as many effects as possible in the construction of this benchmarking study, a literature review was conducted, reviewing past benchmarking studies in the areas of psychology and risk assessment. A number of lessons learned through these studies are presented in order to aid in the design of future HRA benchmarking endeavors.

The Importance of HRA in Human Space Flight: Understanding the Risks

Science.gov (United States)

Hamlin, Teri

2010-01-01

Human performance is critical to crew safety during space missions. Humans interact with hardware and software during ground processing, normal flight, and in response to events. Human interactions with hardware and software can cause Loss of Crew and/or Vehicle (LOCV) through improper actions, or may prevent LOCV through recovery and control actions. Humans have the ability to deal with complex situations and system interactions beyond the capability of machines. Human Reliability Analysis (HRA) is a method used to qualitatively and quantitatively assess the occurrence of human failures that affect availability and reliability of complex systems. Modeling human actions with their corresponding failure probabilities in a Probabilistic Risk Assessment (PRA) provides a more complete picture of system risks and risk contributions. A high-quality HRA can provide valuable information on potential areas for improvement, including training, procedures, human interfaces design, and the need for automation. Modeling human error has always been a challenge in part because performance data is not always readily available. For spaceflight, the challenge is amplified not only because of the small number of participants and limited amount of performance data available, but also due to the lack of definition of the unique factors influencing human performance in space. These factors, called performance shaping factors in HRA terminology, are used in HRA techniques to modify basic human error probabilities in order to capture the context of an analyzed task. Many of the human error modeling techniques were developed within the context of nuclear power plants and therefore the methodologies do not address spaceflight factors such as the effects of microgravity and longer duration missions. This presentation will describe the types of human error risks which have shown up as risk drivers in the Shuttle PRA which may be applicable to commercial space flight. As with other large PRAs

IEEE guide for the analysis of human reliability

International Nuclear Information System (INIS)

Dougherty, E.M. Jr.

1987-01-01

The Institute of Electrical and Electronics Engineers (IEEE) working group 7.4 of the Human Factors and Control Facilities Subcommittee of the Nuclear Power Engineering Committee (NPEC) has released its fifth draft of a Guide for General Principles of Human Action Reliability Analysis for Nuclear Power Generating Stations, for approval of NPEC. A guide is the least mandating in the IEEE hierarchy of standards. The purpose is to enhance the performance of an human reliability analysis (HRA) as a part of a probabilistic risk assessment (PRA), to assure reproducible results, and to standardize documentation. The guide does not recommend or even discuss specific techniques, which are too rapidly evolving today. Considerable maturation in the analysis of human reliability in a PRA context has taken place in recent years. The IEEE guide on this subject is an initial step toward bringing HRA out of the research and development arena into the toolbox of standard engineering practices

Human reliability analysis in the man-machine interface design review

International Nuclear Information System (INIS)

Kim, I.S.

2001-01-01

Advanced, computer-based man-machine interface (MMI) is emerging as part of the new design of nuclear power plants. The impact of advanced MMI on the operator performance, and as a result, on plant safety should be thoroughly evaluated before such technology is actually adopted in the plants. This paper discusses the applicability of human reliability analysis (HRA) to support the design review process. Both the first-generation and the second-generation HRA methods are considered focusing on a couple of promising HRA methods, i.e. ATHEANA and CREAM, with the potential to assist the design review process

Knowledge-base for the new human reliability analysis method, A Technique for Human Error Analysis (ATHEANA)

International Nuclear Information System (INIS)

Cooper, S.E.; Wreathall, J.; Thompson, C.M., Drouin, M.; Bley, D.C.

1996-01-01

This paper describes the knowledge base for the application of the new human reliability analysis (HRA) method, a ''A Technique for Human Error Analysis'' (ATHEANA). Since application of ATHEANA requires the identification of previously unmodeled human failure events, especially errors of commission, and associated error-forcing contexts (i.e., combinations of plant conditions and performance shaping factors), this knowledge base is an essential aid for the HRA analyst

Human reliability analysis for advanced control room of KNGR

International Nuclear Information System (INIS)

Kim, Myung-Ro; Park, Seong-Kyu

2000-01-01

There are two purposes in Human Reliability Analysis (HRA) which was performed during Korean Next Generation Reactor (KNGR) Phase 2 research project. One is to present the human error probability quantification results for Probabilistic Safety Assessment (PSA) and the other is to provide a list of the critical operator actions for Human Factor Engineering (HFE). Critical operator actions were identified from the KNGR HRA/RSA based on selection criteria and incorporated in the MMI Task Analysis, where they receive additional treatment. The use of HRA/PSA results in design, procedure development, and training was ensured by their incorporation in the MMI task analysis and MCR design such as fixed position alarms, displays and controls. Any dominant PSA sequence that takes credit for human performance to achieve acceptable results was incorporated in MMIS validation activities through the PSA-based critical operator actions. The integration of KNGR HRA into MMI design was sufficiently addressed all applicable review criteria of NUREG-0800, Chapter 18, Section 2 F and NUREG-0711. (S.Y.)

Probabilistic safety analysis and human reliability analysis. Proceedings. Working material

International Nuclear Information System (INIS)

1996-01-01

An international meeting on Probabilistic Safety Assessment (PSA) and Human Reliability Analysis (HRA) was jointly organized by Electricite de France - Research and Development (EDF DER) and SRI International in co-ordination with the International Atomic Energy Agency. The meeting was held in Paris 21-23 November 1994. A group of international and French specialists in PSA and HRA participated at the meeting and discussed the state of the art and current trends in the following six topics: PSA Methodology; PSA Applications; From PSA to Dependability; Incident Analysis; Safety Indicators; Human Reliability. For each topic a background paper was prepared by EDF/DER and reviewed by the international group of specialists who attended the meeting. The results of this meeting provide a comprehensive overview of the most important questions related to the readiness of PSA for specific uses and areas where further research and development is required. Refs, figs, tabs

Probabilistic safety analysis and human reliability analysis. Proceedings. Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-12-31

An international meeting on Probabilistic Safety Assessment (PSA) and Human Reliability Analysis (HRA) was jointly organized by Electricite de France - Research and Development (EDF DER) and SRI International in co-ordination with the International Atomic Energy Agency. The meeting was held in Paris 21-23 November 1994. A group of international and French specialists in PSA and HRA participated at the meeting and discussed the state of the art and current trends in the following six topics: PSA Methodology; PSA Applications; From PSA to Dependability; Incident Analysis; Safety Indicators; Human Reliability. For each topic a background paper was prepared by EDF/DER and reviewed by the international group of specialists who attended the meeting. The results of this meeting provide a comprehensive overview of the most important questions related to the readiness of PSA for specific uses and areas where further research and development is required. Refs, figs, tabs.

Development of a BN framework for human reliability analysis through virtual simulation

International Nuclear Information System (INIS)

Garg, Vipul; Santhosh, T.V.; Vinod, Gopika; Antony, P.D.

2017-01-01

Humans are an integral part of complex systems such as nuclear power plants and have to play a significant role in ensuring the safety and reliability of these systems. Failure to perform the intended task within the stipulated time by the operator can challenge the safety of the system. Human reliability analysis (HRA) is a widely practiced methodology to estimate the contribution of operator error towards the overall risk to the facility. HRA methods quantify this contribution in terms of human error probability (HEP) accounting for various psychological and physiological factors that influence the performance of the operator. These factors are referred to as human factors (HF), which enhance or degrade the human performance. The paper discusses the use of virtual simulation as a tool to generate the HF data from the virtual model of an in-house experimental facility. This paper also demonstrates the use of multi-attribute utility theory to determine a suitable HRA method amongst several HRA methods to quantify the HEP based on the desired set of HRA attributes. As classical HRA methods, generally, do not address the interactions among the HFs, the Bayesian network technique has been employed in this study to account for HF interactions. (author)

Adapting Human Reliability Analysis from Nuclear Power to Oil and Gas Applications

Energy Technology Data Exchange (ETDEWEB)

Boring, Ronald Laurids [Idaho National Laboratory

2015-09-01

ABSTRACT: Human reliability analysis (HRA), as currently used in risk assessments, largely derives its methods and guidance from application in the nuclear energy domain. While there are many similarities be-tween nuclear energy and other safety critical domains such as oil and gas, there remain clear differences. This paper provides an overview of HRA state of the practice in nuclear energy and then describes areas where refinements to the methods may be necessary to capture the operational context of oil and gas. Many key distinctions important to nuclear energy HRA such as Level 1 vs. Level 2 analysis may prove insignifi-cant for oil and gas applications. On the other hand, existing HRA methods may not be sensitive enough to factors like the extensive use of digital controls in oil and gas. This paper provides an overview of these con-siderations to assist in the adaptation of existing nuclear-centered HRA methods to the petroleum sector.

Human reliability analysis using event trees

International Nuclear Information System (INIS)

Heslinga, G.

1983-01-01

The shut-down procedure of a technologically complex installation as a nuclear power plant consists of a lot of human actions, some of which have to be performed several times. The procedure is regarded as a chain of modules of specific actions, some of which are analyzed separately. The analysis is carried out by making a Human Reliability Analysis event tree (HRA event tree) of each action, breaking down each action into small elementary steps. The application of event trees in human reliability analysis implies more difficulties than in the case of technical systems where event trees were mainly used until now. The most important reason is that the operator is able to recover a wrong performance; memory influences play a significant role. In this study these difficulties are dealt with theoretically. The following conclusions can be drawn: (1) in principle event trees may be used in human reliability analysis; (2) although in practice the operator will recover his fault partly, theoretically this can be described as starting the whole event tree again; (3) compact formulas have been derived, by which the probability of reaching a specific failure consequence on passing through the HRA event tree after several times of recovery is to be calculated. (orig.)

Comparison of methods for dependency determination between human failure events within human reliability analysis

International Nuclear Information System (INIS)

Cepis, M.

2007-01-01

The Human Reliability Analysis (HRA) is a highly subjective evaluation of human performance, which is an input for probabilistic safety assessment, which deals with many parameters of high uncertainty. The objective of this paper is to show that subjectivism can have a large impact on human reliability results and consequently on probabilistic safety assessment results and applications. The objective is to identify the key features, which may decrease of subjectivity of human reliability analysis. Human reliability methods are compared with focus on dependency comparison between Institute Jozef Stefan - Human Reliability Analysis (IJS-HRA) and Standardized Plant Analysis Risk Human Reliability Analysis (SPAR-H). Results show large differences in the calculated human error probabilities for the same events within the same probabilistic safety assessment, which are the consequence of subjectivity. The subjectivity can be reduced by development of more detailed guidelines for human reliability analysis with many practical examples for all steps of the process of evaluation of human performance. (author)

Comparison of Methods for Dependency Determination between Human Failure Events within Human Reliability Analysis

International Nuclear Information System (INIS)

Cepin, M.

2008-01-01

The human reliability analysis (HRA) is a highly subjective evaluation of human performance, which is an input for probabilistic safety assessment, which deals with many parameters of high uncertainty. The objective of this paper is to show that subjectivism can have a large impact on human reliability results and consequently on probabilistic safety assessment results and applications. The objective is to identify the key features, which may decrease subjectivity of human reliability analysis. Human reliability methods are compared with focus on dependency comparison between Institute Jozef Stefan human reliability analysis (IJS-HRA) and standardized plant analysis risk human reliability analysis (SPAR-H). Results show large differences in the calculated human error probabilities for the same events within the same probabilistic safety assessment, which are the consequence of subjectivity. The subjectivity can be reduced by development of more detailed guidelines for human reliability analysis with many practical examples for all steps of the process of evaluation of human performance

Human reliability guidance - How to increase the synergies between human reliability, human factors, and system design and engineering. Phase 1: The Nordic Point of View - A user needs analysis

International Nuclear Information System (INIS)

Oxstrand, J.; Boring, R.L.

2010-12-01

The main goal of this Nordic Nuclear Safety Research (NKS) council project is to produce guidance for how to use human reliability analysis (HRA) to strengthen overall safety. This project is intended to work across (and hopefully diminish) the borders that exist between human reliability analysis (HRA) and human-system interaction, human performance, human factors, and probabilistic risk assessment at Nordic nuclear power plants. This project consists of two major phases, where the initial phase (phase 1) is a study of current practices in the Nordic region, which is presented in this report. Even though the project covers the synergies between HRA and all other relevant fields, the main focus for the phase is to bridge HRA and design. Interviews with 26 Swedish and Finnish plant experts are summarized the present report, and 10 principles to improve the utilization of HRA at plants are presented. A second study, which is not documented in this preliminary report, will chronicle insights into how the US nuclear industry works with HRA. To gain this knowledge the author will conduct interviews with the US regulator, research laboratories, and utilities. (Author)

Representing cognitive activities and errors in HRA trees

International Nuclear Information System (INIS)

Gertman, D.I.

1992-01-01

This paper discusses development of a means by which to present cognitive information in human reliability assessment (HRA) event trees. The descriptions found in probabilistic risk assessments (PRAs) regarding the demands on, and the resulting performance of, nuclear power plant (NPP) crews often make use of the technique for human error rate prediction (THERP), which provides a mechanism, the HRA event tree, for presenting the analyst's conceptualization of the activities underlying performance and the errors associated with that performance. When using THERP, analysts have often omitted the more complex elements of human cognition from these trees. There has yet to be a concerted effort to take theory, principles, and data from cognitive psychology and wed it to the logic structure of the HRA event tree. This paper attempts to do so. The COGENT modeling scheme (cognitively based HRA event trees) adds two taxonomies to the HRA event tree proposed by Swain and Guttman. The first taxonomy, the one proposed by Norman and Reason, describes the type of error committed and implies something about the underlying cognition as well. The second of these, the Rasmussen taxonomy, provides description regarding the skill-based, rule-based, or knowledge-based behavior underlying the execution of tasks. It is not apparent and must be deduced from the pattern of errors exhibited by personnel

Human reliability guidance - How to increase the synergies between human reliability, human factors, and system design and engineering. Phase 2: The American Point of View - Insights of how the US nuclear industry works with human reliability analysis

International Nuclear Information System (INIS)

Oxstrand, J.

2010-12-01

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 substudies: 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 nuclear industry could be improved. To gain more knowledge about how these improvements could be made, and what improvements to focus on, the second study was conducted. The second study is focused on the American nuclear industry, which has many more years of experience with risk assessment and human reliability than the Nordic nuclear industry. Interviews were conducted to collect information to help the author understand the similarities and differences between the American and the Nordic nuclear industries, and to find data regarding the findings from the first study. The main focus of this report is to identify potential HRA improvements based on the data collected in the American Point of View survey. (Author)

Human reliability guidance - How to increase the synergies between human reliability, human factors, and system design and engineering. Phase 2: The American Point of View - Insights of how the US nuclear industry works with human reliability analysis

Energy Technology Data Exchange (ETDEWEB)

Oxstrand, J. (Vattenfall Ringhals AB, Stockholm (Sweden))

2010-12-15

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 substudies: 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 nuclear industry could be improved. To gain more knowledge about how these improvements could be made, and what improvements to focus on, the second study was conducted. The second study is focused on the American nuclear industry, which has many more years of experience with risk assessment and human reliability than the Nordic nuclear industry. Interviews were conducted to collect information to help the author understand the similarities and differences between the American and the Nordic nuclear industries, and to find data regarding the findings from the first study. The main focus of this report is to identify potential HRA improvements based on the data collected in the American Point of View survey. (Author)

Lessons Learned on Benchmarking from the International Human Reliability Analysis Empirical Study

International Nuclear Information System (INIS)

Boring, Ronald L.; Forester, John A.; Bye, Andreas; Dang, Vinh N.; Lois, Erasmia

2010-01-01

The International Human Reliability Analysis (HRA) Empirical Study is a comparative benchmark of the prediction of HRA methods to the performance of nuclear power plant crews in a control room simulator. There are a number of unique aspects to the present study that distinguish it from previous HRA benchmarks, most notably the emphasis on a method-to-data comparison instead of a method-to-method comparison. This paper reviews seven lessons learned about HRA benchmarking from conducting the study: (1) the dual purposes of the study afforded by joining another HRA study; (2) the importance of comparing not only quantitative but also qualitative aspects of HRA; (3) consideration of both negative and positive drivers on crew performance; (4) a relatively large sample size of crews; (5) the use of multiple methods and scenarios to provide a well-rounded view of HRA performance; (6) the importance of clearly defined human failure events; and (7) the use of a common comparison language to 'translate' the results of different HRA methods. These seven lessons learned highlight how the present study can serve as a useful template for future benchmarking studies.

Lessons Learned on Benchmarking from the International Human Reliability Analysis Empirical Study

Energy Technology Data Exchange (ETDEWEB)

Ronald L. Boring; John A. Forester; Andreas Bye; Vinh N. Dang; Erasmia Lois

2010-06-01

The International Human Reliability Analysis (HRA) Empirical Study is a comparative benchmark of the prediction of HRA methods to the performance of nuclear power plant crews in a control room simulator. There are a number of unique aspects to the present study that distinguish it from previous HRA benchmarks, most notably the emphasis on a method-to-data comparison instead of a method-to-method comparison. This paper reviews seven lessons learned about HRA benchmarking from conducting the study: (1) the dual purposes of the study afforded by joining another HRA study; (2) the importance of comparing not only quantitative but also qualitative aspects of HRA; (3) consideration of both negative and positive drivers on crew performance; (4) a relatively large sample size of crews; (5) the use of multiple methods and scenarios to provide a well-rounded view of HRA performance; (6) the importance of clearly defined human failure events; and (7) the use of a common comparison language to “translate” the results of different HRA methods. These seven lessons learned highlight how the present study can serve as a useful template for future benchmarking studies.

A Research Roadmap for Computation-Based Human Reliability Analysis

Energy Technology Data Exchange (ETDEWEB)

Boring, Ronald [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Joe, Jeffrey [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Groth, Katrina [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-08-01

The United States (U.S.) Department of Energy (DOE) is sponsoring research through the Light Water Reactor Sustainability (LWRS) program to extend the life of the currently operating fleet of commercial nuclear power plants. The Risk Informed Safety Margin Characterization (RISMC) research pathway within LWRS looks at ways to maintain and improve the safety margins of these plants. The RISMC pathway includes significant developments in the area of thermalhydraulics code modeling and the development of tools to facilitate dynamic probabilistic risk assessment (PRA). PRA is primarily concerned with the risk of hardware systems at the plant; yet, hardware reliability is often secondary in overall risk significance to human errors that can trigger or compound undesirable events at the plant. This report highlights ongoing efforts to develop a computation-based approach to human reliability analysis (HRA). This computation-based approach differs from existing static and dynamic HRA approaches in that it: (i) interfaces with a dynamic computation engine that includes a full scope plant model, and (ii) interfaces with a PRA software toolset. The computation-based HRA approach presented in this report is called the Human Unimodels for Nuclear Technology to Enhance Reliability (HUNTER) and incorporates in a hybrid fashion elements of existing HRA methods to interface with new computational tools developed under the RISMC pathway. The goal of this research effort is to model human performance more accurately than existing approaches, thereby minimizing modeling uncertainty found in current plant risk models.

A Research Roadmap for Computation-Based Human Reliability Analysis

International Nuclear Information System (INIS)

Boring, Ronald; Mandelli, Diego; Joe, Jeffrey; Smith, Curtis; Groth, Katrina

2015-01-01

The United States (U.S.) Department of Energy (DOE) is sponsoring research through the Light Water Reactor Sustainability (LWRS) program to extend the life of the currently operating fleet of commercial nuclear power plants. The Risk Informed Safety Margin Characterization (RISMC) research pathway within LWRS looks at ways to maintain and improve the safety margins of these plants. The RISMC pathway includes significant developments in the area of thermalhydraulics code modeling and the development of tools to facilitate dynamic probabilistic risk assessment (PRA). PRA is primarily concerned with the risk of hardware systems at the plant; yet, hardware reliability is often secondary in overall risk significance to human errors that can trigger or compound undesirable events at the plant. This report highlights ongoing efforts to develop a computation-based approach to human reliability analysis (HRA). This computation-based approach differs from existing static and dynamic HRA approaches in that it: (i) interfaces with a dynamic computation engine that includes a full scope plant model, and (ii) interfaces with a PRA software toolset. The computation-based HRA approach presented in this report is called the Human Unimodels for Nuclear Technology to Enhance Reliability (HUNTER) and incorporates in a hybrid fashion elements of existing HRA methods to interface with new computational tools developed under the RISMC pathway. The goal of this research effort is to model human performance more accurately than existing approaches, thereby minimizing modeling uncertainty found in current plant risk models.

Human Reliability Analysis in Support of Risk Assessment for Positive Train Control

Science.gov (United States)

2003-06-01

This report describes an approach to evaluating the reliability of human actions that are modeled in a probabilistic risk assessment : (PRA) of train control operations. This approach to human reliability analysis (HRA) has been applied in the case o...

Current Human Reliability Analysis Methods Applied to Computerized Procedures

Energy Technology Data Exchange (ETDEWEB)

Ronald L. Boring

2012-06-01

Computerized procedures (CPs) are an emerging technology within nuclear power plant control rooms. While CPs have been implemented internationally in advanced control rooms, to date no US nuclear power plant has implemented CPs in its main control room (Fink et al., 2009). Yet, CPs are a reality of new plant builds and are an area of considerable interest to existing plants, which see advantages in terms of enhanced ease of use and easier records management by omitting the need for updating hardcopy procedures. The overall intent of this paper is to provide a characterization of human reliability analysis (HRA) issues for computerized procedures. It is beyond the scope of this document to propose a new HRA approach or to recommend specific methods or refinements to those methods. Rather, this paper serves as a review of current HRA as it may be used for the analysis and review of computerized procedures.

Limitations in simulator time-based human reliability analysis methods

International Nuclear Information System (INIS)

Wreathall, J.

1989-01-01

Developments in human reliability analysis (HRA) methods have evolved slowly. Current methods are little changed from those of almost a decade ago, particularly in the use of time-reliability relationships. While these methods were suitable as an interim step, the time (and the need) has come to specify the next evolution of HRA methods. As with any performance-oriented data source, power plant simulator data have no direct connection to HRA models. Errors reported in data are normal deficiencies observed in human performance; failures are events modeled in probabilistic risk assessments (PRAs). Not all errors cause failures; not all failures are caused by errors. Second, the times at which actions are taken provide no measure of the likelihood of failures to act correctly within an accident scenario. Inferences can be made about human reliability, but they must be made with great care. Specific limitations are discussed. Simulator performance data are useful in providing qualitative evidence of the variety of error types and their potential influences on operating systems. More work is required to combine recent developments in the psychology of error with the qualitative data collected at stimulators. Until data become openly available, however, such an advance will not be practical

Applicability of simplified human reliability analysis methods for severe accidents

Energy Technology Data Exchange (ETDEWEB)

Boring, R.; St Germain, S. [Idaho National Lab., Idaho Falls, Idaho (United States); Banaseanu, G.; Chatri, H.; Akl, Y. [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada)

2016-03-15

Most contemporary human reliability analysis (HRA) methods were created to analyse design-basis accidents at nuclear power plants. As part of a comprehensive expansion of risk assessments at many plants internationally, HRAs will begin considering severe accident scenarios. Severe accidents, while extremely rare, constitute high consequence events that significantly challenge successful operations and recovery. Challenges during severe accidents include degraded and hazardous operating conditions at the plant, the shift in control from the main control room to the technical support center, the unavailability of plant instrumentation, and the need to use different types of operating procedures. Such shifts in operations may also test key assumptions in existing HRA methods. This paper discusses key differences between design basis and severe accidents, reviews efforts to date to create customized HRA methods suitable for severe accidents, and recommends practices for adapting existing HRA methods that are already being used for HRAs at the plants. (author)

Development of a HRA method based on Human Factor Issues for advanced NPP

International Nuclear Information System (INIS)

Lee, Seung Woo; Seong, Poong Hyun; Ha, Jun Su; Park, Jae Hyuk; Kim, Ja Kyung

2010-01-01

A design of instrumentation and control (I and C) systems for various plant systems including nuclear power plants (NPPs) is rapidly moving toward fully digital I and C and modern computer techniques have been gradually introduced into the design of advanced main control room (MCR). In advanced MCR, computer based Human-System Interfaces (HSIs) such as CRT based displays, large display panels (LDP), advanced information system, soft control and computerized procedure system (CPS) are applied in advanced MCR. Human operators in an advanced MCR still play an important role. However, various research and experiences from NPPs with an advanced MCR show that characteristics of human operators' task would be changed due to the use of inexperienced HSIs. This gives implications to the PSFs (Performance Shaping Factors) in HRA (Human Reliability Analysis). PSF in HRA is an aspect of the human's individual characteristics, environment, organization, or task that specifically decrements or improves human performance resulting in increasing or decreasing the likelihood of human error. These PSFs have been suggested in various ways depending on the HRA methods used. In most HRA methods, however, there is a lack of inconsistency for the derivation of the PSFs and a lack of considerations of how the changes implemented in advanced MCR give impact on the operators' task. In this study, a framework for the derivation of and evaluation in the PSFs to be used in HRA for advanced NPPs is suggested

Improvement of human reliability analysis method for PRA

International Nuclear Information System (INIS)

Tanji, Junichi; Fujimoto, Haruo

2013-09-01

It is required to refine human reliability analysis (HRA) method by, for example, incorporating consideration for the cognitive process of operator into the evaluation of diagnosis errors and decision-making errors, as a part of the development and improvement of methods used in probabilistic risk assessments (PRAs). JNES has been developed a HRA method based on ATHENA which is suitable to handle the structured relationship among diagnosis errors, decision-making errors and operator cognition process. This report summarizes outcomes obtained from the improvement of HRA method, in which enhancement to evaluate how the plant degraded condition affects operator cognitive process and to evaluate human error probabilities (HEPs) which correspond to the contents of operator tasks is made. In addition, this report describes the results of case studies on the representative accident sequences to investigate the applicability of HRA method developed. HEPs of the same accident sequences are also estimated using THERP method, which is most popularly used HRA method, and comparisons of the results obtained using these two methods are made to depict the differences of these methods and issues to be solved. Important conclusions obtained are as follows: (1) Improvement of HRA method using operator cognitive action model. Clarification of factors to be considered in the evaluation of human errors, incorporation of degraded plant safety condition into HRA and investigation of HEPs which are affected by the contents of operator tasks were made to improve the HRA method which can integrate operator cognitive action model into ATHENA method. In addition, the detail of procedures of the improved method was delineated in the form of flowchart. (2) Case studies and comparison with the results evaluated by THERP method. Four operator actions modeled in the PRAs of representative BWR5 and 4-loop PWR plants were selected and evaluated as case studies. These cases were also evaluated using

Human Reliability Analysis: session summary

International Nuclear Information System (INIS)

Hall, R.E.

1985-01-01

The use of Human Reliability Analysis (HRA) to identify and resolve human factors issues has significantly increased over the past two years. Today, utilities, research institutions, consulting firms, and the regulatory agency have found a common application of HRA tools and Probabilistic Risk Assessment (PRA). The ''1985 IEEE Third Conference on Human Factors and Power Plants'' devoted three sessions to the discussion of these applications and a review of the insights so gained. This paper summarizes the three sessions and presents those common conclusions that were discussed during the meeting. The paper concludes that session participants supported the use of an adequately documented ''living PRA'' to address human factors issues in design and procedural changes, regulatory compliance, and training and that the techniques can produce cost effective qualitative results that are complementary to more classical human factors methods

CONSIDERATIONS FOR THE TREATMENT OF COMPUTERIZED PROCEDURES IN HUMAN RELIABILITY ANALYSIS

Energy Technology Data Exchange (ETDEWEB)

Ronald L. Boring; David I. Gertman

2012-07-01

Computerized procedures (CPs) are an emerging technology within nuclear power plant control rooms. While CPs have been implemented internationally in advanced control rooms, to date no US nuclear power plant has implemented CPs in its main control room. Yet, CPs are a reality of new plant builds and are an area of considerable interest to existing plants, which see advantages in terms of easier records management by omitting the need for updating hardcopy procedures. The overall intent of this paper is to provide a characterization of human reliability analysis (HRA) issues for computerized procedures. It is beyond the scope of this document to propose a new HRA approach or to recommend specific methods or refinements to those methods. Rather, this paper serves as a review of current HRA as it may be used for the analysis and review of computerized procedures.

Development of a qualitative evaluation framework for performance shaping factors (PSFs) in advanced MCR HRA

International Nuclear Information System (INIS)

Lee, Seung Woo; Kim, Ar Ryum; Ha, Jun Su; Seong, Poong Hyun

2011-01-01

Highlights: → Context changes in advanced MCR have impact on PSFs. → PSFs in the 1st and 2nd generation HRA methods are reviewed. → We made a qualitative evaluation framework for PSF based on human factor issues. - Abstract: Human reliability analysis (HRA) is performed as part of the probabilistic risk assessment to identify and quantify human actions and the associated impacts on structures, systems, and components of complex facilities. In performing HRA, conditions that influence human performance have been analyzed in terms of several context factors. These context factors, which are called performance shaping factors (PSFs) are used to adjust the basic human error probability (BHEP), and PSFs have been derived in various ways depending on the HRA methods used. As the design of instrumentation and control (I and C) systems for nuclear power plants (NPPs) is rapidly moving toward fully digital I and C, and modern computer techniques have been gradually introduced into the design of advanced main control room (MCR), computer-based human-system interfaces (HSIs), such as CRT-based displays, large display panels (LDPs), advanced information systems, soft control, and computerized procedure system (CPS) will be applied in advanced MCR. Environmental changes in MCR have some implications for PSFs, and they have an influence on when PSFs should be applied in HRA because different situations might induce different internal or external factors which can lead to human errors. In this study, PSFs for advanced MCR HRA are derived, and a new qualitative evaluation framework for these PSFs is suggested. First, PSFs from various HRA methods are collected, and these PSFs are further grouped into PSFs categories to be used in advanced MCR HRA. Second, human factor (HF) issues in advanced MCR are analyzed and derived to be used as an evaluation framework for PSFs.

Study and application of human reliability analysis for digital human-system interface

International Nuclear Information System (INIS)

Jia Ming; Liu Yanzi; Zhang Jianbo

2014-01-01

The knowledge of human-orientated abilities and limitations could be used to digital human-system interface (HSI) design by human reliability analysis (HRA) technology. Further, control room system design could achieve the perfect match of man-machine-environment. This research was conducted to establish an integrated HRA method. This method identified HSI potential design flaws which may affect human performance and cause human error. Then a systematic approach was adopted to optimize HSI. It turns out that this method is practical and objective, and effectively improves the safety, reliability and economy of nuclear power plant. This method was applied to CRP1000 projects under construction successfully with great potential. (authors)

An analysis of operational experience during low power and shutdown and a plan for addressing human reliability assessment issues

International Nuclear Information System (INIS)

Barriere, M.; Luckas, W.; Whitehead, D.; Ramey-Smith, A.

1994-06-01

Recent nuclear power plant events (e.g. Chernobyl, Diablo Canyon, and Vogtle) and US Nuclear Regulatory Commission (NRC) reports (e.g. NUREG-1449) have led to concerns regarding human reliability during low power and shutdown (LP ampersand S) conditions and limitations of human reliability analysis (HRA) methodologies in adequately representing the LP ampersand S environment. As a result of these concerns, the NRC initiated two parallel research projects to assess the influence of LP ampersand S conditions on human reliability through an analysis of operational experience at pressurized water reactors (PWRs) an boiling water reactors (BWRs). These research projects, performed by Brookhaven National Laboratory for PWRS, and Sandia National Laboratories for BWRs, identified unique aspects of human performance during LP ampersand S conditions and provided a program plan for research and development necessary to improve existing HRA methodologies. This report documents the results of the analysis of LP ampersand S operating experience and describes the improved HRA program plan

A data-informed PIF hierarchy for model-based Human Reliability Analysis

International Nuclear Information System (INIS)

Groth, Katrina M.; Mosleh, Ali

2012-01-01

This paper addresses three problems associated with the use of Performance Shaping Factors in Human Reliability Analysis. (1) There are more than a dozen Human Reliability Analysis (HRA) methods that use Performance Influencing Factors (PIFs) or Performance Shaping Factors (PSFs) to model human performance, but there is not a standard set of PIFs used among the methods, nor is there a framework available to compare the PIFs used in various methods. (2) The PIFs currently in use are not defined specifically enough to ensure consistent interpretation of similar PIFs across methods. (3) There are few rules governing the creation, definition, and usage of PIF sets. This paper introduces a hierarchical set of PIFs that can be used for both qualitative and quantitative HRA. The proposed PIF set is arranged in a hierarchy that can be collapsed or expanded to meet multiple objectives. The PIF hierarchy has been developed with respect to a set fundamental principles necessary for PIF sets, which are also introduced in this paper. This paper includes definitions of the PIFs to allow analysts to map the proposed PIFs onto current and future HRA methods. The standardized PIF hierarchy will allow analysts to combine different types of data and will therefore make the best use of the limited data in HRA. The collapsible hierarchy provides the structure necessary to combine multiple types of information without reducing the quality of the information.

Structural Design of HRA Database using generic task for Quantitative Analysis of Human Performance

Energy Technology Data Exchange (ETDEWEB)

Kim, Seung Hwan; Kim, Yo Chan; Choi, Sun Yeong; Park, Jin Kyun; Jung Won Dea [KAERI, Daejeon (Korea, Republic of)

2016-05-15

This paper describes a design of generic task based HRA database for quantitative analysis of human performance in order to estimate the number of task conductions. The estimation method to get the total task conduction number using direct counting is not easy to realize and maintain its data collection framework. To resolve this problem, this paper suggests an indirect method and a database structure using generic task that enables to estimate the total number of conduction based on instructions of operating procedures of nuclear power plants. In order to reduce human errors, therefore, all information on the human errors taken by operators in the power plant should be systematically collected and examined in its management. Korea Atomic Energy Research Institute (KAERI) is carrying out a research to develop a data collection framework to establish a Human Reliability Analysis (HRA) database that could be employed as technical bases to generate human error probabilities (HEPs) and performance shaping factors (PSFs)]. As a result of the study, the essential table schema was designed to the generic task database which stores generic tasks, procedure lists and task tree structures, and other supporting tables. The number of task conduction based on the operating procedures for HEP estimation was enabled through the generic task database and framework. To verify the framework applicability, case study for the simulated experiments was performed and analyzed using graphic user interfaces developed in this study.

Structural Design of HRA Database using generic task for Quantitative Analysis of Human Performance

International Nuclear Information System (INIS)

Kim, Seung Hwan; Kim, Yo Chan; Choi, Sun Yeong; Park, Jin Kyun; Jung Won Dea

2016-01-01

This paper describes a design of generic task based HRA database for quantitative analysis of human performance in order to estimate the number of task conductions. The estimation method to get the total task conduction number using direct counting is not easy to realize and maintain its data collection framework. To resolve this problem, this paper suggests an indirect method and a database structure using generic task that enables to estimate the total number of conduction based on instructions of operating procedures of nuclear power plants. In order to reduce human errors, therefore, all information on the human errors taken by operators in the power plant should be systematically collected and examined in its management. Korea Atomic Energy Research Institute (KAERI) is carrying out a research to develop a data collection framework to establish a Human Reliability Analysis (HRA) database that could be employed as technical bases to generate human error probabilities (HEPs) and performance shaping factors (PSFs)]. As a result of the study, the essential table schema was designed to the generic task database which stores generic tasks, procedure lists and task tree structures, and other supporting tables. The number of task conduction based on the operating procedures for HEP estimation was enabled through the generic task database and framework. To verify the framework applicability, case study for the simulated experiments was performed and analyzed using graphic user interfaces developed in this study.

Model-based human reliability analysis: prospects and requirements

International Nuclear Information System (INIS)

Mosleh, A.; Chang, Y.H.

2004-01-01

Major limitations of the conventional methods for human reliability analysis (HRA), particularly those developed for operator response analysis in probabilistic safety assessments (PSA) of nuclear power plants, are summarized as a motivation for the need and a basis for developing requirements for the next generation HRA methods. It is argued that a model-based approach that provides explicit cognitive causal links between operator behaviors and directly or indirectly measurable causal factors should be at the core of the advanced methods. An example of such causal model is briefly reviewed, where due to the model complexity and input requirements can only be currently implemented in a dynamic PSA environment. The computer simulation code developed for this purpose is also described briefly, together with current limitations in the models, data, and the computer implementation

Estimation of the human error probabilities in the human reliability analysis

International Nuclear Information System (INIS)

Liu Haibin; He Xuhong; Tong Jiejuan; Shen Shifei

2006-01-01

Human error data is an important issue of human reliability analysis (HRA). Using of Bayesian parameter estimation, which can use multiple information, such as the historical data of NPP and expert judgment data to modify the human error data, could get the human error data reflecting the real situation of NPP more truly. This paper, using the numeric compute program developed by the authors, presents some typical examples to illustrate the process of the Bayesian parameter estimation in HRA and discusses the effect of different modification data on the Bayesian parameter estimation. (authors)

Establishing guidance for the review of human reliability analysis in PSA

International Nuclear Information System (INIS)

Reer, B.; Dang, V.N.; Hirschberg, S.; Meyer, P.

2000-01-01

PSI was commissioned to develop Guidelines for the Regulatory Review of the Human Reliability Analysis (HRA) within Probabilistic Safety Assessments (PSAs) for nuclear power plants. In the Guidelines, HRA quality is addressed in terms of 97 indicators. Each indicator is formulated as a question, described as a specific feature of the analysis, and then explained in detail. Two analysis stages are distinguished: the selection of the human errors to be modelled, and their quantification to determine their impact on the core damage frequency. Review findings are grouped under two headings: transparency and adequacy. An analysis is 'transparent' if an externally qualified person is able to reproduce the analysis results, and 'adequate' if such results reflect the plant-specific conditions related to safety. To allocate resources efficiently, the review is structured in two phases: (1) The Quick Review, which clarifies whether the HRA has a fundamental deficiency and, furthermore, if it points to information needs and areas of emphasis for the detailed review, and (2) The Detailed Review, which results in well-grounded findings, based on extended examinations and close-plant contacts. (authors)

HUMAN RELIABILITY ANALYSIS FOR COMPUTERIZED PROCEDURES, PART TWO: APPLICABILITY OF CURRENT METHODS

Energy Technology Data Exchange (ETDEWEB)

Ronald L. Boring; David I. Gertman

2012-10-01

Computerized procedures (CPs) are an emerging technology within nuclear power plant control rooms. While CPs have been implemented internationally in advanced control rooms, to date no U.S. nuclear power plant has implemented CPs in its main control room. Yet, CPs are a reality of new plant builds and are an area of considerable interest to existing plants, which see advantages in terms of easier records management by omitting the need for updating hardcopy procedures. The overall intent of this paper is to provide a characterization of human reliability analysis (HRA) issues for computerized procedures. It is beyond the scope of this document to propose a new HRA approach or to recommend specific methods or refinements to those methods. Rather, this paper serves as a review of current HRA as it may be used for the analysis and review of computerized procedures.

Individual Differences in Human Reliability Analysis

Energy Technology Data Exchange (ETDEWEB)

Jeffrey C. Joe; Ronald L. Boring

2014-06-01

While human reliability analysis (HRA) methods include uncertainty in quantification, the nominal model of human error in HRA typically assumes that operator performance does not vary significantly when they are given the same initiating event, indicators, procedures, and training, and that any differences in operator performance are simply aleatory (i.e., random). While this assumption generally holds true when performing routine actions, variability in operator response has been observed in multiple studies, especially in complex situations that go beyond training and procedures. As such, complexity can lead to differences in operator performance (e.g., operator understanding and decision-making). Furthermore, psychological research has shown that there are a number of known antecedents (i.e., attributable causes) that consistently contribute to observable and systematically measurable (i.e., not random) differences in behavior. This paper reviews examples of individual differences taken from operational experience and the psychological literature. The impact of these differences in human behavior and their implications for HRA are then discussed. We propose that individual differences should not be treated as aleatory, but rather as epistemic. Ultimately, by understanding the sources of individual differences, it is possible to remove some epistemic uncertainty from analyses.

A new dynamic HRA method and its application

International Nuclear Information System (INIS)

Je, Moo Sung; Park, Chang Kyoo

1995-01-01

This paper presents a new dynamic HRA (Human Reliability Analysis) method and its application for quantifying the human error probabilities in implementing an accident management action. For comparisons of current HRA methods with the new method, the characteristics of THERP, HCR, and SLIM-MAUD, which are most frequently used methods in PSAs, are discussed. The action associated with the implementation of the cavity flooding during a station blackout sequence is considered for its application. This method is based on the concepts of the quantified correlation between the performance requirement and performance achievement. The MAAP 3.0B code and Latin Hypercube sampling technique are used to determine the uncertainty of the performance achievement parameter. Meanwhile, the value of the performance requirement parameter is obtained from interviews. Based on these stochastic distributions obtained, human error probabilities are calculated with respect to the various means and variances of the timings. It is shown that this method is very flexible in that it can be applied to any kind of the operator actions, including the actions associated with the implementation of accident management strategies. 1 fig., 3 tabs., 17 refs. (Author)

Issues in cognitive reliability

International Nuclear Information System (INIS)

Woods, D.D.; Hitchler, M.J.; Rumancik, J.A.

1984-01-01

This chapter examines some problems in current methods to assess reactor operator reliability at cognitive tasks and discusses new approaches to solve these problems. The two types of human failures are errors in the execution of an intention and errors in the formation/selection of an intention. Topics considered include the types of description, error correction, cognitive performance and response time, the speed-accuracy tradeoff function, function based task analysis, and cognitive task analysis. One problem of human reliability analysis (HRA) techniques in general is the question of what are the units of behavior whose reliability are to be determined. A second problem for HRA is that people often detect and correct their errors. The use of function based analysis, which maps the problem space for plant control, is recommended

Proceedings of the international workshop on building the new HRA: errors of commission - from research to application

International Nuclear Information System (INIS)

2003-01-01

The main mission of the Working Group on Risk Assessment (RISK) is to advance the understanding and utilisation of probabilistic safety analysis (PSA) in ensuring continued safety of nuclear installations in Member countries. One of the major criticisms of current PSAs is that they do not adequately address an important class of human system interactions, namely inappropriate actions, particularly those that might occur during the response to a transient or accident, that place the plant in a situation of higher risk. This class of inappropriate actions is often referred to as 'errors of commission'. The principal characteristic of an error of commission in a PSA context is that its consequence is a state of unavailability of a component, system or function. This is in contrast to an error of omission, which is characterised by a lack of action and, therefore, preserves the status quo of a system, component, or function. In the PSA context, the most significant errors of commission are those that, in addition to resulting in failure to perform some function, also fail or make unavailable other equipment or functions needed to mitigate the accident scenario, or otherwise exacerbate the situation. The workshop reported herein is an extension of the work of the Working Group on Risk Assessment (RISK) performed to review errors of commission in probabilistic safety analysis (NEA/CSNI/R(2000)17). The main purpose of the meeting was to provide a forum for exchange of information including lessons learned, identification of gaps in our current understanding and knowledge, data needs, and research needs. This workshop also provides a perspective for another workshop, Building the New HRA: Strengthening the Link Between Experience and HRA, to be held in Munich in January of 2002. Individual speakers present a broad international perspective that summarises technical issues, lessons learned, and experiences gained through applying second-generation human reliability

Structured information analysis for human reliability analysis of emergency tasks in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Jung, Won Dea; Kim, Jae Whan; Park, Jin Kyun; Ha, Jae Joo [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-02-01

More than twenty HRA (Human Reliability Analysis) methodologies have been developed and used for the safety analysis in nuclear field during the past two decades. However, no methodology appears to have universally been accepted, as various limitations have been raised for more widely used ones. One of the most important limitations of conventional HRA is insufficient analysis of the task structure and problem space. To resolve this problem, we suggest SIA (Structured Information Analysis) for HRA. The proposed SIA consists of three parts. The first part is the scenario analysis that investigates the contextual information related to the given task on the basis of selected scenarios. The second is the goals-means analysis to define the relations between the cognitive goal and task steps. The third is the cognitive function analysis module that identifies the cognitive patterns and information flows involved in the task. Through the three-part analysis, systematic investigation is made possible from the macroscopic information on the tasks to the microscopic information on the specific cognitive processes. It is expected that analysts can attain a structured set of information that helps to predict the types and possibility of human error in the given task. 48 refs., 12 figs., 11 tabs. (Author)

HUMAN RELIABILITY ANALYSIS DENGAN PENDEKATAN COGNITIVE RELIABILITY AND ERROR ANALYSIS METHOD (CREAM

Directory of Open Access Journals (Sweden)

Zahirah Alifia Maulida

2015-01-01

Full Text Available Kecelakaan kerja pada bidang grinding dan welding menempati urutan tertinggi selama lima tahun terakhir di PT. X. Kecelakaan ini disebabkan oleh human error. Human error terjadi karena pengaruh lingkungan kerja fisik dan non fisik.Penelitian kali menggunakan skenario untuk memprediksi serta mengurangi kemungkinan terjadinya error pada manusia dengan pendekatan CREAM (Cognitive Reliability and Error Analysis Method. CREAM adalah salah satu metode human reliability analysis yang berfungsi untuk mendapatkan nilai Cognitive Failure Probability (CFP yang dapat dilakukan dengan dua cara yaitu basic method dan extended method. Pada basic method hanya akan didapatkan nilai failure probabailty secara umum, sedangkan untuk extended method akan didapatkan CFP untuk setiap task. Hasil penelitian menunjukkan faktor- faktor yang mempengaruhi timbulnya error pada pekerjaan grinding dan welding adalah kecukupan organisasi, kecukupan dari Man Machine Interface (MMI & dukungan operasional, ketersediaan prosedur/ perencanaan, serta kecukupan pelatihan dan pengalaman. Aspek kognitif pada pekerjaan grinding yang memiliki nilai error paling tinggi adalah planning dengan nilai CFP 0.3 dan pada pekerjaan welding yaitu aspek kognitif execution dengan nilai CFP 0.18. Sebagai upaya untuk mengurangi nilai error kognitif pada pekerjaan grinding dan welding rekomendasi yang diberikan adalah memberikan training secara rutin, work instrucstion yang lebih rinci dan memberikan sosialisasi alat. Kata kunci: CREAM (cognitive reliability and error analysis method, HRA (human reliability analysis, cognitive error Abstract The accidents in grinding and welding sectors were the highest cases over the last five years in PT. X and it caused by human error. Human error occurs due to the influence of working environment both physically and non-physically. This study will implement an approaching scenario called CREAM (Cognitive Reliability and Error Analysis Method. CREAM is one of human

Capturing cognitive causal paths in human reliability analysis with Bayesian network models

International Nuclear Information System (INIS)

Zwirglmaier, Kilian; Straub, Daniel; Groth, Katrina M.

2017-01-01

reIn the last decade, Bayesian networks (BNs) have been identified as a powerful tool for human reliability analysis (HRA), with multiple advantages over traditional HRA methods. In this paper we illustrate how BNs can be used to include additional, qualitative causal paths to provide traceability. The proposed framework provides the foundation to resolve several needs frequently expressed by the HRA community. First, the developed extended BN structure reflects the causal paths found in cognitive psychology literature, thereby addressing the need for causal traceability and strong scientific basis in HRA. Secondly, the use of node reduction algorithms allows the BN to be condensed to a level of detail at which quantification is as straightforward as the techniques used in existing HRA. We illustrate the framework by developing a BN version of the critical data misperceived crew failure mode in the IDHEAS HRA method, which is currently under development at the US NRC . We illustrate how the model could be quantified with a combination of expert-probabilities and information from operator performance databases such as SACADA. This paper lays the foundations necessary to expand the cognitive and quantitative foundations of HRA. - Highlights: • A framework for building traceable BNs for HRA, based on cognitive causal paths. • A qualitative BN structure, directly showing these causal paths is developed. • Node reduction algorithms are used for making the BN structure quantifiable. • BN quantified through expert estimates and observed data (Bayesian updating). • The framework is illustrated for a crew failure mode of IDHEAS.

A review of the evolution of human reliability analysis methods at nuclear industry

International Nuclear Information System (INIS)

Oliveira, Lécio N. de; Santos, Isaac José A. Luquetti dos; Carvalho, Paulo V.R.

2017-01-01

This paper reviews the status of researches on the application of human reliability analysis methods at nuclear industry and its evolution along the years. Human reliability analysis (HRA) is one of the elements used in Probabilistic Safety Analysis (PSA) and is performed as part of PSAs to quantify the likelihood that people will fail to take action, such as errors of omission and errors of commission. Although HRA may be used at lots of areas, the focus of this paper is to review the applicability of HRA methods along the years at nuclear industry, especially in Nuclear Power Plants (NPP). An electronic search on CAPES Portal of Journals (A bibliographic database) was performed. This literature review covers original papers published since the first generation of HRA methods until the ones published on March 2017. A total of 94 papers were retrieved by the initial search and 13 were selected to be fully reviewed and for data extraction after the application of inclusion and exclusion criteria, quality and suitability evaluation according to applicability at nuclear industry. Results point out that the methods from first generation are more used in practice than methods from second generation. This occurs because it is more concentrated towards quantification, in terms of success or failure of human action what make them useful for quantitative risk assessment to PSA. Although the second generation considers context and error of commission in human error prediction, they are not wider used in practice at nuclear industry to PSA. (author)

A review of the evolution of human reliability analysis methods at nuclear industry

Energy Technology Data Exchange (ETDEWEB)

Oliveira, Lécio N. de; Santos, Isaac José A. Luquetti dos; Carvalho, Paulo V.R., E-mail: lecionoliveira@gmail.com, E-mail: luquetti@ien.gov.br, E-mail: paulov@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2017-11-01

This paper reviews the status of researches on the application of human reliability analysis methods at nuclear industry and its evolution along the years. Human reliability analysis (HRA) is one of the elements used in Probabilistic Safety Analysis (PSA) and is performed as part of PSAs to quantify the likelihood that people will fail to take action, such as errors of omission and errors of commission. Although HRA may be used at lots of areas, the focus of this paper is to review the applicability of HRA methods along the years at nuclear industry, especially in Nuclear Power Plants (NPP). An electronic search on CAPES Portal of Journals (A bibliographic database) was performed. This literature review covers original papers published since the first generation of HRA methods until the ones published on March 2017. A total of 94 papers were retrieved by the initial search and 13 were selected to be fully reviewed and for data extraction after the application of inclusion and exclusion criteria, quality and suitability evaluation according to applicability at nuclear industry. Results point out that the methods from first generation are more used in practice than methods from second generation. This occurs because it is more concentrated towards quantification, in terms of success or failure of human action what make them useful for quantitative risk assessment to PSA. Although the second generation considers context and error of commission in human error prediction, they are not wider used in practice at nuclear industry to PSA. (author)

Development of advanced methods and related software for human reliability evaluation within probabilistic safety analyses

International Nuclear Information System (INIS)

Kosmowski, K.T.; Mertens, J.; Degen, G.; Reer, B.

1994-06-01

Human Reliability Analysis (HRA) is an important part of Probabilistic Safety Analysis (PSA). The first part of this report consists of an overview of types of human behaviour and human error including the effect of significant performance shaping factors on human reliability. Particularly with regard to safety assessments for nuclear power plants a lot of HRA methods have been developed. The most important of these methods are presented and discussed in the report, together with techniques for incorporating HRA into PSA and with models of operator cognitive behaviour. Based on existing HRA methods the concept of a software system is described. For the development of this system the utilization of modern programming tools is proposed; the essential goal is the effective application of HRA methods. A possible integration of computeraided HRA within PSA is discussed. The features of Expert System Technology and examples of applications (PSA, HRA) are presented in four appendices. (orig.) [de

Lessons learned from HRA and human-system modeling efforts

International Nuclear Information System (INIS)

Hallbert, B.P.

1993-01-01

Human-System modeling is not unique to the field of Human Reliability Analysis (HRA). Since human factors professionals first began their explorations of human activities, they have done so with the concept of open-quotes systemclose quotes in mind. Though the two - human and system - are distinct, they can be properly understood only in terms of each other: the system provides a context in which goals and objectives for work are defined, and the human plays either a pre-defined or ad hoc role in meeting these goals. In this sense, every intervention which attempts to evaluate or improve upon some system parameter requires that an understanding of human-system interactions be developed. It is too often the case, however, that somewhere between the inception of a system and its implementation, the human-system relationships are overlooked, misunderstood, or inadequately framed. This results in mismatches between demands versus capabilities of human operators, systems which are difficult to operate, and the obvious end product-human error. The lessons learned from human system modeling provide a valuable feedback mechanism to the process of HRA, and the technologies which employ this form of modeling

Haven't a Cue? Mapping the CUE Space as an Aid to HRA Modeling

Energy Technology Data Exchange (ETDEWEB)

David I Gertman; Ronald L Boring; Jacques Hugo; William Phoenix

2012-06-01

Advances in automation present a new modeling environment for the human reliability analysis (HRA) practitioner. Many, if not most, current day HRA methods have their origin in characterizing and quantifying human performance in analog environments where mode awareness and system status indications are potentially less comprehensive, but simpler to comprehend at a glance when compared to advanced presentation systems. The introduction of highly complex automation has the potential to lead to: decreased levels of situation awareness caused by the need for increased monitoring; confusion regarding the often non-obvious causes of automation failures, and emergent system dependencies that formerly may have been uncharacterized. Understanding the relation of incoming cues available to operators during plant upset conditions, in conjunction with operating procedures, yields insight into understanding the nature of the expected operator response in this control room environment. Static systems methods such as fault trees do not contain the appropriate temporal information or necessarily specify the relationship among cues leading to operator response. In this paper, we do not attempt to replace standard performance shaping factors commonly used in HRA nor offer a new HRA method, existing methods may suffice. In this paper we strive to enhance current understanding of the basis for operator response through a technique that can be used during the qualitative portion of the HRA analysis process. The CUE map is a means to visualize the relationship among salient cues in the control room that help influence operator response, show how the cognitive map of the operator changes as information is gained or lost, and is applicable to existing as well as advanced hybrid plants and small modular reactor designs. A brief application involving loss of condensate is presented and advantages and limitations of the modeling approach and use of the CUE map are discussed.

Multi-Unit Considerations for Human Reliability Analysis

Energy Technology Data Exchange (ETDEWEB)

St. Germain, S.; Boring, R.; Banaseanu, G.; Akl, Y.; Chatri, H.

2017-03-01

This paper uses the insights from the Standardized Plant Analysis Risk-Human Reliability Analysis (SPAR-H) methodology to help identify human actions currently modeled in the single unit PSA that may need to be modified to account for additional challenges imposed by a multi-unit accident as well as identify possible new human actions that might be modeled to more accurately characterize multi-unit risk. In identifying these potential human action impacts, the use of the SPAR-H strategy to include both errors in diagnosis and errors in action is considered as well as identifying characteristics of a multi-unit accident scenario that may impact the selection of the performance shaping factors (PSFs) used in SPAR-H. The lessons learned from the Fukushima Daiichi reactor accident will be addressed to further help identify areas where improved modeling may be required. While these multi-unit impacts may require modifications to a Level 1 PSA model, it is expected to have much more importance for Level 2 modeling. There is little currently written specifically about multi-unit HRA issues. A review of related published research will be presented. While this paper cannot answer all issues related to multi-unit HRA, it will hopefully serve as a starting point to generate discussion and spark additional ideas towards the proper treatment of HRA in a multi-unit PSA.

Bayesian belief networks for human reliability analysis: A review of applications and gaps

International Nuclear Information System (INIS)

Mkrtchyan, L.; Podofillini, L.; Dang, V.N.

2015-01-01

The use of Bayesian Belief Networks (BBNs) in risk analysis (and in particular Human Reliability Analysis, HRA) is fostered by a number of features, attractive in fields with shortage of data and consequent reliance on subjective judgments: the intuitive graphical representation, the possibility of combining diverse sources of information, the use the probabilistic framework to characterize uncertainties. In HRA, BBN applications are steadily increasing, each emphasizing a different BBN feature or a different HRA aspect to improve. This paper aims at a critical review of these features as well as at suggesting research needs. Five groups of BBN applications are analysed: modelling of organizational factors, analysis of the relationships among failure influencing factors, BBN-based extensions of existing HRA methods, dependency assessment among human failure events, assessment of situation awareness. Further, the paper analyses the process for building BBNs and in particular how expert judgment is used in the assessment of the BBN conditional probability distributions. The gaps identified in the review suggest the need for establishing more systematic frameworks to integrate the different sources of information relevant for HRA (cognitive models, empirical data, and expert judgment) and to investigate algorithms to avoid elicitation of many relationships via expert judgment. - Highlights: • We analyze BBN uses for HRA applications; but some conclusions can be generalized. • Special review focus on BBN building approaches, key for model acceptance. • Gaps relate to the transparency of the BBN building and quantification phases. • Need for more systematic framework to integrate different sources of information. • Need of ways to avoid elicitation of many relationships via expert judgment

Study on a new framework of Human Reliability Analysis to evaluate soft control execution error in advanced MCRs of NPPs

International Nuclear Information System (INIS)

Jang, Inseok; Kim, Ar Ryum; Jung, Wondea; Seong, Poong Hyun

2016-01-01

Highlights: • The operation environment of MCRs in NPPs has changed by adopting new HSIs. • The operation action in NPP Advanced MCRs is performed by soft control. • New HRA framework should be considered in the HRA for advanced MCRs. • HRA framework for evaluation of soft control execution human error is suggested. • Suggested method will be helpful to analyze human reliability in advance MCRs. - Abstract: Since the Three Mile Island (TMI)-2 accident, human error has been recognized as one of the main causes of Nuclear Power Plant (NPP) accidents, and numerous studies related to Human Reliability Analysis (HRA) have been carried out. Most of these methods were developed considering the conventional type of Main Control Rooms (MCRs). However, the operating environment of MCRs in NPPs has changed with the adoption of new Human-System Interfaces (HSIs) that are based on computer-based technologies. The MCRs that include these digital technologies, such as large display panels, computerized procedures, and soft controls, are called advanced MCRs. Among the many features of advanced MCRs, soft controls are a particularly important feature because operating actions in NPP advanced MCRs are performed by soft control. Due to the differences in interfaces between soft control and hardwired conventional type control, different Human Error Probabilities (HEPs) and a new HRA framework should be considered in the HRA for advanced MCRs. To this end, a new framework of a HRA method for evaluating soft control execution human error is suggested by performing a soft control task analysis and the literature regarding widely accepted human error taxonomies is reviewed. Moreover, since most current HRA databases deal with operation in conventional MCRs and are not explicitly designed to deal with digital HSIs, empirical analysis of human error and error recovery considering soft controls under an advanced MCR mockup are carried out to collect human error data, which is

PROOF OF CONCEPT FOR A HUMAN RELIABILITY ANALYSIS METHOD FOR HEURISTIC USABILITY EVALUATION OF SOFTWARE

International Nuclear Information System (INIS)

Ronald L. Boring; David I. Gertman; Jeffrey C. Joe; Julie L. Marble

2005-01-01

An ongoing issue within human-computer interaction (HCI) is the need for simplified or ''discount'' methods. The current economic slowdown has necessitated innovative methods that are results driven and cost effective. The myriad methods of design and usability are currently being cost-justified, and new techniques are actively being explored that meet current budgets and needs. Recent efforts in human reliability analysis (HRA) are highlighted by the ten-year development of the Standardized Plant Analysis Risk HRA (SPAR-H) method. The SPAR-H method has been used primarily for determining human centered risk at nuclear power plants. The SPAR-H method, however, shares task analysis underpinnings with HCI. Despite this methodological overlap, there is currently no HRA approach deployed in heuristic usability evaluation. This paper presents an extension of the existing SPAR-H method to be used as part of heuristic usability evaluation in HCI

Culture Representation in Human Reliability Analysis

Energy Technology Data Exchange (ETDEWEB)

David Gertman; Julie Marble; Steven Novack

2006-12-01

Understanding human-system response is critical to being able to plan and predict mission success in the modern battlespace. Commonly, human reliability analysis has been used to predict failures of human performance in complex, critical systems. However, most human reliability methods fail to take culture into account. This paper takes an easily understood state of the art human reliability analysis method and extends that method to account for the influence of culture, including acceptance of new technology, upon performance. The cultural parameters used to modify the human reliability analysis were determined from two standard industry approaches to cultural assessment: Hofstede’s (1991) cultural factors and Davis’ (1989) technology acceptance model (TAM). The result is called the Culture Adjustment Method (CAM). An example is presented that (1) reviews human reliability assessment with and without cultural attributes for a Supervisory Control and Data Acquisition (SCADA) system attack, (2) demonstrates how country specific information can be used to increase the realism of HRA modeling, and (3) discusses the differences in human error probability estimates arising from cultural differences.

Human reliability analysis of errors of commission: a review of methods and applications

Energy Technology Data Exchange (ETDEWEB)

Reer, B

2007-06-15

Illustrated by specific examples relevant to contemporary probabilistic safety assessment (PSA), this report presents a review of human reliability analysis (HRA) addressing post initiator errors of commission (EOCs), i.e. inappropriate actions under abnormal operating conditions. The review addressed both methods and applications. Emerging HRA methods providing advanced features and explicit guidance suitable for PSA are: A Technique for Human Event Analysis (ATHEANA, key publications in 1998/2000), Methode d'Evaluation de la Realisation des Missions Operateur pour la Surete (MERMOS, 1998/2000), the EOC HRA method developed by the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, 2003), the Misdiagnosis Tree Analysis (MDTA) method (2005/2006), the Cognitive Reliability and Error Analysis Method (CREAM, 1998), and the Commission Errors Search and Assessment (CESA) method (2002/2004). As a result of a thorough investigation of various PSA/HRA applications, this paper furthermore presents an overview of EOCs (termination of safety injection, shutdown of secondary cooling, etc.) referred to in predictive studies and a qualitative review of cases of EOC quantification. The main conclusions of the review of both the methods and the EOC HRA cases are: (1) The CESA search scheme, which proceeds from possible operator actions to the affected systems to scenarios, may be preferable because this scheme provides a formalized way for identifying relatively important scenarios with EOC opportunities; (2) an EOC identification guidance like CESA, which is strongly based on the procedural guidance and important measures of systems or components affected by inappropriate actions, however should pay some attention to EOCs associated with familiar but non-procedural actions and EOCs leading to failures of manually initiated safety functions. (3) Orientations of advanced EOC quantification comprise a) modeling of multiple contexts for a given scenario, b) accounting for

Human reliability analysis of errors of commission: a review of methods and applications

International Nuclear Information System (INIS)

Reer, B.

2007-06-01

Illustrated by specific examples relevant to contemporary probabilistic safety assessment (PSA), this report presents a review of human reliability analysis (HRA) addressing post initiator errors of commission (EOCs), i.e. inappropriate actions under abnormal operating conditions. The review addressed both methods and applications. Emerging HRA methods providing advanced features and explicit guidance suitable for PSA are: A Technique for Human Event Analysis (ATHEANA, key publications in 1998/2000), Methode d'Evaluation de la Realisation des Missions Operateur pour la Surete (MERMOS, 1998/2000), the EOC HRA method developed by the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, 2003), the Misdiagnosis Tree Analysis (MDTA) method (2005/2006), the Cognitive Reliability and Error Analysis Method (CREAM, 1998), and the Commission Errors Search and Assessment (CESA) method (2002/2004). As a result of a thorough investigation of various PSA/HRA applications, this paper furthermore presents an overview of EOCs (termination of safety injection, shutdown of secondary cooling, etc.) referred to in predictive studies and a qualitative review of cases of EOC quantification. The main conclusions of the review of both the methods and the EOC HRA cases are: (1) The CESA search scheme, which proceeds from possible operator actions to the affected systems to scenarios, may be preferable because this scheme provides a formalized way for identifying relatively important scenarios with EOC opportunities; (2) an EOC identification guidance like CESA, which is strongly based on the procedural guidance and important measures of systems or components affected by inappropriate actions, however should pay some attention to EOCs associated with familiar but non-procedural actions and EOCs leading to failures of manually initiated safety functions. (3) Orientations of advanced EOC quantification comprise a) modeling of multiple contexts for a given scenario, b) accounting for

Some insights from recent applications of HRA methods in PSA effort and plant operation feedback in Czech Republic

International Nuclear Information System (INIS)

Holy, Jaroslav

2004-01-01

The methods of human reliability analysis (HRA) form an integral part of probabilistic safety assessment. These methods are currently still under development and there is significant lack of data needed for applications. As a consequence, no unique formal approach for solution of many aspects related to plant staff human performance does exist. In this contribution, some examples of fruitful work are given, which reflect the needs and possibilities specific for HRA field. In the first two examples, methodology, course and results of analysis of concrete matters related to pre-accident errors and shutdown control room crew performance are described. The last example is oriented to one potential human performance related data source-observation of control room crew full-scope simulator exercise. Some conclusions of general validity are presented in the final part of the contribution

Comparison of HRA methods based on WWER-1000 NPP real and simulated accident scenarios

International Nuclear Information System (INIS)

Petkov, Gueorgui

2010-01-01

Full text: Adequate treatment of human interactions in probabilistic safety analysis (PSA) studies is a key to the understanding of accident sequences and their relative importance in overall risk. Human interactions with machines have long been recognized as important contributors to the safe operation of nuclear power plants (NPP). Human interactions affect the ordering of dominant accident sequences and hence have a significant effect on the risk of NPP. By virtue of the ability to combine the treatment of both human and hardware reliability in real accidents, NPP fullscope, multifunctional and computer-based simulators provide a unique way of developing an understanding of the importance of specific human actions for overall plant safety. Context dependent human reliability assessment (HRA) models, such as the holistic decision tree (HDT) and performance evaluation of teamwork (PET) methods, are the so-called second generation HRA techniques. The HDT model has been used for a number of PSA studies. The PET method reflects promising prospects for dealing with dynamic aspects of human performance. The paper presents a comparison of the two HRA techniques for calculation of post-accident human error probability in the PSA. The real and simulated event training scenario 'turbine's stop after loss of feedwater' based on standard PSA model assumptions is designed for WWER-1000 computer simulator and their detailed boundary conditions are described and analyzed. The error probability of post-accident individual actions will be calculated by means of each investigated technique based on student's computer simulator training archives

A study on the dependency evaluation for multiple human actions in human reliability analysis of probabilistic safety assessment

International Nuclear Information System (INIS)

Kang, D. I.; Yang, J. E.; Jung, W. D.; Sung, T. Y.; Park, J. H.; Lee, Y. H.; Hwang, M. J.; Kim, K. Y.; Jin, Y. H.; Kim, S. C.

1997-02-01

This report describes the study results on the method of the dependency evaluation and the modeling, and the limited value of human error probability (HEP) for multiple human actions in accident sequences of probabilistic safety assessment (PSA). THERP and Parry's method, which have been generally used in dependency evaluation of human reliability analysis (HRA), are introduced and their limitations are discussed. New dependency evaluation method in HRA is established to make up for the weak points of THERP and Parry's methods. The limited value of HEP is also established based on the review of several HRA related documents. This report describes the definition, the type, the evaluation method, and the evaluation example of dependency to help the reader's understanding. It is expected that this study results will give a guidance to HRA analysts in dependency evaluation of multiple human actions and enable PSA analysts to understand HRA in detail. (author). 23 refs., 3 tabs., 2 figs

Tailoring a Human Reliability Analysis to Your Industry Needs

Science.gov (United States)

DeMott, D. L.

2016-01-01

Companies at risk of accidents caused by human error that result in catastrophic consequences include: airline industry mishaps, medical malpractice, medication mistakes, aerospace failures, major oil spills, transportation mishaps, power production failures and manufacturing facility incidents. Human Reliability Assessment (HRA) is used to analyze the inherent risk of human behavior or actions introducing errors into the operation of a system or process. These assessments can be used to identify where errors are most likely to arise and the potential risks involved if they do occur. Using the basic concepts of HRA, an evolving group of methodologies are used to meet various industry needs. Determining which methodology or combination of techniques will provide a quality human reliability assessment is a key element to developing effective strategies for understanding and dealing with risks caused by human errors. There are a number of concerns and difficulties in "tailoring" a Human Reliability Assessment (HRA) for different industries. Although a variety of HRA methodologies are available to analyze human error events, determining the most appropriate tools to provide the most useful results can depend on industry specific cultures and requirements. Methodology selection may be based on a variety of factors that include: 1) how people act and react in different industries, 2) expectations based on industry standards, 3) factors that influence how the human errors could occur such as tasks, tools, environment, workplace, support, training and procedure, 4) type and availability of data, 5) how the industry views risk & reliability, and 6) types of emergencies, contingencies and routine tasks. Other considerations for methodology selection should be based on what information is needed from the assessment. If the principal concern is determination of the primary risk factors contributing to the potential human error, a more detailed analysis method may be employed

Phoenix – A model-based Human Reliability Analysis methodology: Qualitative Analysis Procedure

International Nuclear Information System (INIS)

Ekanem, Nsimah J.; Mosleh, Ali; Shen, Song-Hua

2016-01-01

Phoenix method is an attempt to address various issues in the field of Human Reliability Analysis (HRA). Built on a cognitive human response model, Phoenix incorporates strong elements of current HRA good practices, leverages lessons learned from empirical studies, and takes advantage of the best features of existing and emerging HRA methods. Its original framework was introduced in previous publications. This paper reports on the completed methodology, summarizing the steps and techniques of its qualitative analysis phase. The methodology introduces the “Crew Response Tree” which provides a structure for capturing the context associated with Human Failure Events (HFEs), including errors of omission and commission. It also uses a team-centered version of the Information, Decision and Action cognitive model and “macro-cognitive” abstractions of crew behavior, as well as relevant findings from cognitive psychology literature and operating experience, to identify potential causes of failures and influencing factors during procedure-driven and knowledge-supported crew-plant interactions. The result is the set of identified HFEs and likely scenarios leading to each. The methodology itself is generic in the sense that it is compatible with various quantification methods, and can be adapted for use across different environments including nuclear, oil and gas, aerospace, aviation, and healthcare. - Highlights: • Produces a detailed, consistent, traceable, reproducible and properly documented HRA. • Uses “Crew Response Tree” to capture context associated with Human Failure Events. • Models dependencies between Human Failure Events and influencing factors. • Provides a human performance model for relating context to performance. • Provides a framework for relating Crew Failure Modes to its influencing factors.

A Critique on the Effectiveness of Current Human Reliability Analysis Approach for the Human-Machine Interface Design in Nuclear Power Plants

International Nuclear Information System (INIS)

Lee, Yong Hee

2010-01-01

Human Reliability Analysis (HRA) in cooperation of PSA has been conducted to evaluate the safety of a system and the validity of a system design. HRA has been believed to provide a quantitative value of human error potential and the safety level of a design alternative in Nuclear Power Plants (NPPs). However, it becomes doubtful that current HRA is worth to conduct to evaluate the human factors of NPP design, since there have been many critiques upon the virtue of HRA. Inevitably, the newer the technology becomes, the larger endeavors bound for the new facilitated methods. This paper describes the limitations and the obsolescence of the current HRA, especially for the design evaluation of Human-Machine Interface (HMI) utilizing the recent digital technologies. An alternative approach to the assessment of the human error potential of HMI design is proposed

Inclusion of task dependence in human reliability analysis

International Nuclear Information System (INIS)

Su, Xiaoyan; Mahadevan, Sankaran; Xu, Peida; Deng, Yong

2014-01-01

Dependence assessment among human errors in human reliability analysis (HRA) is an important issue, which includes the evaluation of the dependence among human tasks and the effect of the dependence on the final human error probability (HEP). This paper represents a computational model to handle dependence in human reliability analysis. The aim of the study is to automatically provide conclusions on the overall degree of dependence and calculate the conditional human error probability (CHEP) once the judgments of the input factors are given. The dependence influencing factors are first identified by the experts and the priorities of these factors are also taken into consideration. Anchors and qualitative labels are provided as guidance for the HRA analyst's judgment of the input factors. The overall degree of dependence between human failure events is calculated based on the input values and the weights of the input factors. Finally, the CHEP is obtained according to a computing formula derived from the technique for human error rate prediction (THERP) method. The proposed method is able to quantify the subjective judgment from the experts and improve the transparency in the HEP evaluation process. Two examples are illustrated to show the effectiveness and the flexibility of the proposed method. - Highlights: • We propose a computational model to handle dependence in human reliability analysis. • The priorities of the dependence influencing factors are taken into consideration. • The overall dependence degree is determined by input judgments and the weights of factors. • The CHEP is obtained according to a computing formula derived from THERP

Task Decomposition in Human Reliability Analysis

Energy Technology Data Exchange (ETDEWEB)

Boring, Ronald Laurids [Idaho National Laboratory; Joe, Jeffrey Clark [Idaho National Laboratory

2014-06-01

In the probabilistic safety assessments (PSAs) used in the nuclear industry, human failure events (HFEs) are determined as a subset of hardware failures, namely those hardware failures that could be triggered by human action or inaction. This approach is top-down, starting with hardware faults and deducing human contributions to those faults. Elsewhere, more traditionally human factors driven approaches would tend to look at opportunities for human errors first in a task analysis and then identify which of those errors is risk significant. The intersection of top-down and bottom-up approaches to defining HFEs has not been carefully studied. Ideally, both approaches should arrive at the same set of HFEs. This question remains central as human reliability analysis (HRA) methods are generalized to new domains like oil and gas. The HFEs used in nuclear PSAs tend to be top-down— defined as a subset of the PSA—whereas the HFEs used in petroleum quantitative risk assessments (QRAs) are more likely to be bottom-up—derived from a task analysis conducted by human factors experts. The marriage of these approaches is necessary in order to ensure that HRA methods developed for top-down HFEs are also sufficient for bottom-up applications.

New advances in human reliability using the EPRIHRA calculator

International Nuclear Information System (INIS)

Julius, J. A.; Grobbelaar, J. F.

2006-01-01

This paper describes new advances in human reliability associated with the integration of HRA methods, lessons learned during the first few years of operation of the EPRI HRA / PRA Tools Users Group, and application of human reliability techniques in areas beyond the more traditional Level 1 internal events PRA. This paper is organized as follows. 1. EPRI HRA Users Group Overview (mission, membership, activities, approach) 2. HRA Methods Currently Used (selection, integration, and addressing dependencies) 3. New Advances in HRA Methods 4. Conclusions. (authors)

Integration of human reliability analysis into the probabilistic risk assessment process: Phase 1

International Nuclear Information System (INIS)

Bell, B.J.; Vickroy, S.C.

1984-10-01

A research program was initiated to develop a testable set of analytical procedures for integrating human reliability analysis (HRA) into the probabilistic risk assessment (PRA) process to more adequately assess the overall impact of human performance on risk. In this three-phase program, stand-alone HRA/PRA analytic procedures will be developed and field evaluated to provide improved methods, techniques, and models for applying quantitative and qualitative human error data which systematically integrate HRA principles, techniques, and analyses throughout the entire PRA process. Phase 1 of the program involved analysis of state-of-the-art PRAs to define the structures and processes currently in use in the industry. Phase 2 research will involve developing a new or revised PRA methodology which will enable more efficient regulation of the industry using quantitative or qualitative results of the PRA. Finally, Phase 3 will be to field test those procedures to assure that the results generated by the new methodologies will be usable and acceptable to the NRC. This paper briefly describes the first phase of the program and outlines the second

A Review of Human Reliability Needs in the U.S. Nuclear Industry

Energy Technology Data Exchange (ETDEWEB)

Boring, Ronald Laurids [Idaho National Laboratory

2015-08-01

In this survey, 34 subject matter experts from the U.S. nuclear industry were interviewed to determine specific needs for human reliability analysis (HRA). Conclusions from the interviews are detailed in this article. A summary of the findings includes: (1) The need for improved guidance on the use of HRA methods generally and for specific applications. (2) The need for additional training in HRA to provide more hands-on experience in the application of HRA methods. (3) Thedevelopment of HRA approaches suitable for advanced reactors, severe accident situations, and low-power and shutdown applications. (4) The refinement of HRA methods to account forfactors such as crew variability, latent errors, more sophisticated dependency modeling, and errors of commission. (5) The continued need for simplified HRA methods appropriate for field applications. (6) The need for tighter coupling of HRA and human factors. (7) The need for improvements in the quantitative basis of HRA methods. These findings suggest the field of HRA is mature but still benefits from refinements.

Simulation and Non-Simulation Based Human Reliability Analysis Approaches

Energy Technology Data Exchange (ETDEWEB)

Boring, Ronald Laurids [Idaho National Lab. (INL), Idaho Falls, ID (United States); Shirley, Rachel Elizabeth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Joe, Jeffrey Clark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-12-01

Part of the U.S. Department of Energy’s Light Water Reactor Sustainability (LWRS) Program, the Risk-Informed Safety Margin Characterization (RISMC) Pathway develops approaches to estimating and managing safety margins. RISMC simulations pair deterministic plant physics models with probabilistic risk models. As human interactions are an essential element of plant risk, it is necessary to integrate human actions into the RISMC risk model. In this report, we review simulation-based and non-simulation-based human reliability assessment (HRA) methods. Chapter 2 surveys non-simulation-based HRA methods. Conventional HRA methods target static Probabilistic Risk Assessments for Level 1 events. These methods would require significant modification for use in dynamic simulation of Level 2 and Level 3 events. Chapter 3 is a review of human performance models. A variety of methods and models simulate dynamic human performance; however, most of these human performance models were developed outside the risk domain and have not been used for HRA. The exception is the ADS-IDAC model, which can be thought of as a virtual operator program. This model is resource-intensive but provides a detailed model of every operator action in a given scenario, along with models of numerous factors that can influence operator performance. Finally, Chapter 4 reviews the treatment of timing of operator actions in HRA methods. This chapter is an example of one of the critical gaps between existing HRA methods and the needs of dynamic HRA. This report summarizes the foundational information needed to develop a feasible approach to modeling human interactions in the RISMC simulations.

Multidisciplinary framework for human reliability analysis with an application to errors of commission and dependencies

International Nuclear Information System (INIS)

Barriere, M.T.; Luckas, W.J.; Wreathall, J.; Cooper, S.E.; Bley, D.C.; Ramey-Smith, A.

1995-08-01

Since the early 1970s, human reliability analysis (HRA) has been considered to be an integral part of probabilistic risk assessments (PRAs). Nuclear power plant (NPP) events, from Three Mile Island through the mid-1980s, showed the importance of human performance to NPP risk. Recent events demonstrate that human performance continues to be a dominant source of risk. In light of these observations, the current limitations of existing HRA approaches become apparent when the role of humans is examined explicitly in the context of real NPP events. The development of new or improved HRA methodologies to more realistically represent human performance is recognized by the Nuclear Regulatory Commission (NRC) as a necessary means to increase the utility of PRAS. To accomplish this objective, an Improved HRA Project, sponsored by the NRC's Office of Nuclear Regulatory Research (RES), was initiated in late February, 1992, at Brookhaven National Laboratory (BNL) to develop an improved method for HRA that more realistically assesses the human contribution to plant risk and can be fully integrated with PRA. This report describes the research efforts including the development of a multidisciplinary HRA framework, the characterization and representation of errors of commission, and an approach for addressing human dependencies. The implications of the research and necessary requirements for further development also are discussed

Multidisciplinary framework for human reliability analysis with an application to errors of commission and dependencies

Energy Technology Data Exchange (ETDEWEB)

Barriere, M.T.; Luckas, W.J. [Brookhaven National Lab., Upton, NY (United States); Wreathall, J. [Wreathall (John) and Co., Dublin, OH (United States); Cooper, S.E. [Science Applications International Corp., Reston, VA (United States); Bley, D.C. [PLG, Inc., Newport Beach, CA (United States); Ramey-Smith, A. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Technology

1995-08-01

Since the early 1970s, human reliability analysis (HRA) has been considered to be an integral part of probabilistic risk assessments (PRAs). Nuclear power plant (NPP) events, from Three Mile Island through the mid-1980s, showed the importance of human performance to NPP risk. Recent events demonstrate that human performance continues to be a dominant source of risk. In light of these observations, the current limitations of existing HRA approaches become apparent when the role of humans is examined explicitly in the context of real NPP events. The development of new or improved HRA methodologies to more realistically represent human performance is recognized by the Nuclear Regulatory Commission (NRC) as a necessary means to increase the utility of PRAS. To accomplish this objective, an Improved HRA Project, sponsored by the NRC`s Office of Nuclear Regulatory Research (RES), was initiated in late February, 1992, at Brookhaven National Laboratory (BNL) to develop an improved method for HRA that more realistically assesses the human contribution to plant risk and can be fully integrated with PRA. This report describes the research efforts including the development of a multidisciplinary HRA framework, the characterization and representation of errors of commission, and an approach for addressing human dependencies. The implications of the research and necessary requirements for further development also are discussed.

Review of cause-based decision tree approach for the development of domestic standard human reliability analysis procedure in low power/shutdown operation probabilistic safety assessment

International Nuclear Information System (INIS)

Kang, D. I.; Jung, W. D.

2003-01-01

We review the Cause-Based Decision Tree (CBDT) approach to decide whether we incorporate it or not for the development of domestic standard Human Reliability Analysis (HRA) procedure in low power/shutdown operation Probabilistic Safety Assessment (PSA). In this paper, we introduce the cause based decision tree approach, quantify human errors using it, and identify merits and demerits of it in comparision with previously used THERP. The review results show that it is difficult to incorporate the CBDT method for the development of domestic standard HRA procedure in low power/shutdown PSA because the CBDT method need for the subjective judgment of HRA analyst like as THERP. However, it is expected that the incorporation of the CBDT method into the development of domestic standard HRA procedure only for the comparision of quantitative HRA results will relieve the burden of development of detailed HRA procedure and will help maintain consistent quantitative HRA results

Human reliability analysis data obtainment through fuzzy logic in nuclear plants

International Nuclear Information System (INIS)

Nascimento, C.S. do; Mesquita, R.N. de

2012-01-01

Highlights: ► Human Error Probability estimates from operator's reactions to emergency situations. ► Human Reliability Analysis input data obtainment through fuzzy logic inference. ► Performance Shaping Factors evaluation influence level onto the operator's actions. - Abstract: Human error has been recognized as an important factor for many industrial and nuclear accidents occurrence. Human error data is scarcely available for different reasons among which, lapses in historical database registry methodology is an important one. Human Reliability Analysis (HRA) is an usual tool employed to estimate the probability that an operator will reasonably perform a system required task in required time without degrading the system. This meta-analysis requires specific Human Error Probability estimates for most of its procedure. This work obtains Human Error Probability (HEP) estimates from operator's actions in response to emergency situations hypothesis on Research Reactor IEA-R1 from IPEN, Brazil. Through this proposed methodology HRA should be able to be performed even with shortage of related human error statistical data. A Performance Shaping Factors (PSF's) evaluation in order to classify and estimate their influence level onto the operator's actions and to determine their actual state over the plant was also done. Both HEP estimation and PSF evaluation were done based on expert judgment using interviews and questionnaires. Expert group was established based on selected IEA-R1 operators, and their evaluation were put into a knowledge representation system which used linguistic variables and group evaluation values that were obtained through Fuzzy Logic and Fuzzy Set theory. HEP obtained values show good agreement with literature published data corroborating the proposed methodology as a good alternative to be used on HRA.

Application of human reliability analysis methodology of second generation

International Nuclear Information System (INIS)

Ruiz S, T. de J.; Nelson E, P. F.

2009-10-01

The human reliability analysis (HRA) is a very important part of probabilistic safety analysis. The main contribution of HRA in nuclear power plants is the identification and characterization of the issues that are brought together for an error occurring in the human tasks that occur under normal operation conditions and those made after abnormal event. Additionally, the analysis of various accidents in history, it was found that the human component has been a contributing factor in the cause. Because of need to understand the forms and probability of human error in the 60 decade begins with the collection of generic data that result in the development of the first generation of HRA methodologies. Subsequently develop methods to include in their models additional performance shaping factors and the interaction between them. So by the 90 mid, comes what is considered the second generation methodologies. Among these is the methodology A Technique for Human Event Analysis (ATHEANA). The application of this method in a generic human failure event, it is interesting because it includes in its modeling commission error, the additional deviations quantification to nominal scenario considered in the accident sequence of probabilistic safety analysis and, for this event the dependency actions evaluation. That is, the generic human failure event was required first independent evaluation of the two related human failure events . So the gathering of the new human error probabilities involves the nominal scenario quantification and cases of significant deviations considered by the potential impact on analyzed human failure events. Like probabilistic safety analysis, with the analysis of the sequences were extracted factors more specific with the highest contribution in the human error probabilities. (Author)

A methodology to incorporate organizational factors into human reliability analysis

International Nuclear Information System (INIS)

Li Pengcheng; Chen Guohua; Zhang Li; Xiao Dongsheng

2010-01-01

A new holistic methodology for Human Reliability Analysis (HRA) is proposed to model the effects of the organizational factors on the human reliability. Firstly, a conceptual framework is built, which is used to analyze the causal relationships between the organizational factors and human reliability. Then, the inference model for Human Reliability Analysis is built by combining the conceptual framework with Bayesian networks, which is used to execute the causal inference and diagnostic inference of human reliability. Finally, a case example is presented to demonstrate the specific application of the proposed methodology. The results show that the proposed methodology of combining the conceptual model with Bayesian Networks can not only easily model the causal relationship between organizational factors and human reliability, but in a given context, people can quantitatively measure the human operational reliability, and identify the most likely root causes or the prioritization of root causes caused human error. (authors)

Comparison of the THERP quantitative tables with the human reliability analysis techniques of second generation

International Nuclear Information System (INIS)

Alvarenga, Marco Antonio Bayout; Fonseca, Renato Alves

2009-01-01

The methodology THERP is classified as a Human Reliability Analysis (HRA) technique of first generation and its emergence was an important initial step for the development of HRA techniques in the industry. Due to the fact of being a first generation technique, THERP quantification tables of human errors are based on a taxonomy that does not take into account the human errors mechanisms. Concerning the three cognitive levels in the Rasmussen framework for the cognitive information processing in human beings, THERP deals in most cases with errors that happen in the perceptual-motor level (stimulus-response). In the rules level, this technique can work better using the time dependent probabilities curves of diagnosis errors, obtained in nuclear power plants simulators. Nevertheless, this is done without processing any error mechanisms. Another deficiency is the fact that the performance shaping factors are in limited number. Furthermore, the influences (predictable or not) of operational context, arising from operational deviations of the most probable (in terms of occurrence probabilities) standard scenarios beside the consequent operational tendencies (operator actions) are not estimated. This work makes a critical analysis of these deficiencies and it points out possible solutions in order to modify the THERP tables, seeking a realistic quantification, that does not underestimate or overestimate the human errors probabilities when applying the HRA techniques to nuclear power plants. The critical analysis is accomplished through a qualitative comparison between THERP, a HRA technique of first generation, with CREAM, as well as ATHEANA, which are HRA techniques of second generation. (author)

Comparison of the THERP quantitative tables with the human reliability analysis techniques of second generation

Energy Technology Data Exchange (ETDEWEB)

Alvarenga, Marco Antonio Bayout; Fonseca, Renato Alves [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)], e-mail: bayout@cnen.gov.br, e-mail: rfonseca@cnen.gov.br

2009-07-01

The methodology THERP is classified as a Human Reliability Analysis (HRA) technique of first generation and its emergence was an important initial step for the development of HRA techniques in the industry. Due to the fact of being a first generation technique, THERP quantification tables of human errors are based on a taxonomy that does not take into account the human errors mechanisms. Concerning the three cognitive levels in the Rasmussen framework for the cognitive information processing in human beings, THERP deals in most cases with errors that happen in the perceptual-motor level (stimulus-response). In the rules level, this technique can work better using the time dependent probabilities curves of diagnosis errors, obtained in nuclear power plants simulators. Nevertheless, this is done without processing any error mechanisms. Another deficiency is the fact that the performance shaping factors are in limited number. Furthermore, the influences (predictable or not) of operational context, arising from operational deviations of the most probable (in terms of occurrence probabilities) standard scenarios beside the consequent operational tendencies (operator actions) are not estimated. This work makes a critical analysis of these deficiencies and it points out possible solutions in order to modify the THERP tables, seeking a realistic quantification, that does not underestimate or overestimate the human errors probabilities when applying the HRA techniques to nuclear power plants. The critical analysis is accomplished through a qualitative comparison between THERP, a HRA technique of first generation, with CREAM, as well as ATHEANA, which are HRA techniques of second generation. (author)

Bridging Human Reliability Analysis and Psychology, Part 1: The Psychological Literature Review for the IDHEAS Method

Energy Technology Data Exchange (ETDEWEB)

April M. Whaley; Stacey M. L. Hendrickson; Ronald L. Boring; Jeffrey C. Joe; Katya L. Le Blanc; Jing Xing

2012-06-01

In response to Staff Requirements Memorandum (SRM) SRM-M061020, the U.S. Nuclear Regulatory Commission (NRC) is sponsoring work to update the technical basis underlying human reliability analysis (HRA) in an effort to improve the robustness of HRA. The ultimate goal of this work is to develop a hybrid of existing methods addressing limitations of current HRA models and in particular issues related to intra- and inter-method variabilities and results. This hybrid method is now known as the Integrated Decision-tree Human Event Analysis System (IDHEAS). Existing HRA methods have looked at elements of the psychological literature, but there has not previously been a systematic attempt to translate the complete span of cognition from perception to action into mechanisms that can inform HRA. Therefore, a first step of this effort was to perform a literature search of psychology, cognition, behavioral science, teamwork, and operating performance to incorporate current understanding of human performance in operating environments, thus affording an improved technical foundation for HRA. However, this literature review went one step further by mining the literature findings to establish causal relationships and explicit links between the different types of human failures, performance drivers and associated performance measures ultimately used for quantification. This is the first of two papers that detail the literature review (paper 1) and its product (paper 2). This paper describes the literature review and the high-level architecture used to organize the literature review, and the second paper (Whaley, Hendrickson, Boring, & Xing, these proceedings) describes the resultant cognitive framework.

Human reliability analysis of performing tasks in plants based on fuzzy integral

International Nuclear Information System (INIS)

Washio, Takashi; Kitamura, Yutaka; Takahashi, Hideaki

1991-01-01

The effective improvement of the human working conditions in nuclear power plants might be a solution for the enhancement of the operation safety. The human reliability analysis (HRA) gives a methodological basis of the improvement based on the evaluation of human reliability under various working conditions. This study investigates some difficulties of the human reliability analysis using conventional linear models and recent fuzzy integral models, and provides some solutions to the difficulties. The following practical features of the provided methods are confirmed in comparison with the conventional methods: (1) Applicability to various types of tasks (2) Capability of evaluating complicated dependencies among working condition factors (3) A priori human reliability evaluation based on a systematic task analysis of human action processes (4) A conversion scheme to probability from indices representing human reliability. (author)

Guidelines for the regulatory review of the human reliability analysis in PSAs

International Nuclear Information System (INIS)

Reer, Bernhard; Dang, V.N.; Hirschberg, Stefan; Meyer, Patrick

2000-01-01

In the review guidelines recently developed for the Swiss Federal Nuclear Inspectorate, the Human Reliability Analysis (HRA) is reviewed in two stages. The preliminary review is aimed at identifying major shortcomings and potential issues to be examined in the detailed review. The detailed review comprehensively addresses the overall adequacy and transparency of the HRA. For the two review stages, 97 indicators are defined in terms of questions focusing on verifiable features of the methodology, implementation and results. The guidelines provide steps for information gathering and present examples of acceptable practices as well as of potential deficiencies. Both review stages may result in requests for clarification, additional documentation or analyses. The first applications of the guidelines consist of the preliminary reviews of two HRAs. (author)

Human-centered modeling in human reliability analysis: some trends based on case studies

International Nuclear Information System (INIS)

Mosneron-Dupin, F.; Reer, B.; Heslinga, G.; Straeter, O.; Gerdes, V.; Saliou, G.; Ullwer, W.

1997-01-01

As an informal working group of researchers from France, Germany and The Netherlands created in 1993, the EARTH association is investigating significant subjects in the field of human reliability analysis (HRA). Our initial review of cases from nuclear operating experience showed that decision-based unrequired actions (DUA) contribute to risk significantly on the one hand. On the other hand, our evaluation of current HRA methods showed that these methods do not cover such actions adequately. Especially, practice-oriented guidelines for their predictive identification are lacking. We assumed that a basic cause for such difficulties was that these methods actually use a limited representation of the stimulus-organism-response (SOR) paradigm. We proposed a human-centered model, which better highlights the active role of the operators and the importance of their culture, attitudes and goals. This orientation was encouraged by our review of current HRA research activities. We therefore decided to envisage progress by identifying cognitive tendencies in the context of operating and simulator experience. For this purpose, advanced approaches for retrospective event analysis were discussed. Some orientations for improvements were proposed. By analyzing cases, various cognitive tendencies were identified, together with useful information about their context. Some of them match psychological findings already published in the literature, some of them are not covered adequately by the literature that we reviewed. Finally, this exploratory study shows that contextual and case-illustrated findings about cognitive tendencies provide useful help for the predictive identification of DUA in HRA. More research should be carried out to complement our findings and elaborate more detailed and systematic guidelines for using them in HRA studies

Integration of human reliability analysis into the probabilistic risk assessment process: phase 1

International Nuclear Information System (INIS)

Bell, B.J.; Vickroy, S.C.

1985-01-01

The US Nuclear Regulatory Commission and Pacific Northwest Laboratory initiated a research program in 1984 to develop a testable set of analytical procedures for integrating human reliability analysis (HRA) into the probabilistic risk assessment (PRA) process to more adequately assess the overall impact of human performance on risk. In this three phase program, stand-alone HRA/PRA analytic procedures will be developed and field evaluated to provide improved methods, techniques, and models for applying quantitative and qualitative human error data which systematically integrate HRA principles, techniques, and analyses throughout the entire PRA process. Phase 1 of the program involved analysis of state-of-the-art PRAs to define the structures and processes currently in use in the industry. Phase 2 research will involve developing a new or revised PRA methodology which will enable more efficient regulation of the industry using quantitative or qualitative results of the PRA. Finally, Phase 3 will be to field test those procedures to assure that the results generated by the new methodologies will be usable and acceptable to the NRC. This paper briefly describes the first phase of the program and outlines the second

Human reliability analysis data obtainment through fuzzy logic in nuclear plants

Energy Technology Data Exchange (ETDEWEB)

Nascimento, C.S. do, E-mail: claudio.souza@ctmsp.mar.mil.br [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Av. Professor Lineu Prestes 2468, 05508-000 Sao Paulo, SP (Brazil); Mesquita, R.N. de, E-mail: rnavarro@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN - SP), Av. Professor Lineu Prestes 2242, 05508-000 Sao Paulo, SP (Brazil)

2012-09-15

Highlights: Black-Right-Pointing-Pointer Human Error Probability estimates from operator's reactions to emergency situations. Black-Right-Pointing-Pointer Human Reliability Analysis input data obtainment through fuzzy logic inference. Black-Right-Pointing-Pointer Performance Shaping Factors evaluation influence level onto the operator's actions. - Abstract: Human error has been recognized as an important factor for many industrial and nuclear accidents occurrence. Human error data is scarcely available for different reasons among which, lapses in historical database registry methodology is an important one. Human Reliability Analysis (HRA) is an usual tool employed to estimate the probability that an operator will reasonably perform a system required task in required time without degrading the system. This meta-analysis requires specific Human Error Probability estimates for most of its procedure. This work obtains Human Error Probability (HEP) estimates from operator's actions in response to emergency situations hypothesis on Research Reactor IEA-R1 from IPEN, Brazil. Through this proposed methodology HRA should be able to be performed even with shortage of related human error statistical data. A Performance Shaping Factors (PSF's) evaluation in order to classify and estimate their influence level onto the operator's actions and to determine their actual state over the plant was also done. Both HEP estimation and PSF evaluation were done based on expert judgment using interviews and questionnaires. Expert group was established based on selected IEA-R1 operators, and their evaluation were put into a knowledge representation system which used linguistic variables and group evaluation values that were obtained through Fuzzy Logic and Fuzzy Set theory. HEP obtained values show good agreement with literature published data corroborating the proposed methodology as a good alternative to be used on HRA.

A shortened version of the THERP/Handbook approach to human reliability analysis for probabilistic risk assessment

International Nuclear Information System (INIS)

Swain, A.D.

1986-01-01

The approach to human reliability analysis (HRA) known as THERP/Handbook has been applied to several probabilistic risk assessments (PRAs) of nuclear power plants (NPPs) and other complex systems. The approach is based on a thorough task analysis of the man-machine interfaces, including the interactions among the people, involved in the operations being assessed. The idea is to assess fully the underlying performance shaping factors (PSFs) and dependence effects which result either in reliable or unreliable human performance

Review of advances in human reliability analysis of errors of commission, Part 1: EOC identification

International Nuclear Information System (INIS)

Reer, Bernhard

2008-01-01

In close connection with examples relevant to contemporary probabilistic safety assessment (PSA), a review of advances in human reliability analysis (HRA) of post-initiator errors of commission (EOCs), i.e. inappropriate actions under abnormal operating conditions, has been carried out. The review comprises both EOC identification (part 1) and quantification (part 2); part 1 is presented in this article. Emerging HRA methods addressing the problem of EOC identification are: A Technique for Human Event Analysis (ATHEANA), the EOC HRA method developed by Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS), the Misdiagnosis Tree Analysis (MDTA) method, and the Commission Errors Search and Assessment (CESA) method. Most of the EOCs referred to in predictive studies comprise the stop of running or the inhibition of anticipated functions; a few comprise the start of a function. The CESA search scheme-which proceeds from possible operator actions to the affected systems to scenarios and uses procedures and importance measures as key sources of input information-provides a formalized way for identifying relatively important scenarios with EOC opportunities. In the implementation however, attention should be paid regarding EOCs associated with familiar but non-procedural actions and EOCs leading to failures of manually initiated safety functions

Human Factors Reliability Analysis for Assuring Nuclear Safety Using Fuzzy Fault Tree

International Nuclear Information System (INIS)

Eisawy, E.A.-F. I.; Sallam, H.

2016-01-01

In order to ensure effective prevention of harmful events, the risk assessment process cannot ignore the role of humans in the dynamics of accidental events and thus the seriousness of the consequences that may derive from them. Human reliability analysis (HRA) involves the use of qualitative and quantitative methods to assess the human contribution to risk. HRA techniques have been developed in order to provide human error probability values associated with operators’ tasks to be included within the broader context of system risk assessment, and are aimed at reducing the probability of accidental events. Fault tree analysis (FTA) is a graphical model that displays the various combinations of equipment failures and human errors that can result in the main system failure of interest. FTA is a risk analysis technique to assess likelihood (in a probabilistic context) of an event. The objective data available to estimate the likelihood is often missing, and even if available, is subject to incompleteness and imprecision or vagueness. Without addressing incompleteness and imprecision in the available data, FTA and subsequent risk analysis give a false impression of precision and correctness that undermines the overall credibility of the process. To solve this problem, qualitative justification in the context of failure possibilities can be used as alternative for quantitative justification. In this paper, we introduce the approach of fuzzy reliability as solution for fault tree analysis drawbacks. A new fuzzy fault tree method is proposed for the analysis of human reliability based on fuzzy sets and fuzzy operations t-norms, co-norms, defuzzification, and fuzzy failure probability. (author)

Benchmarking HRA methods against different NPP simulator data

International Nuclear Information System (INIS)

Petkov, Gueorgui; Filipov, Kalin; Velev, Vladimir; Grigorov, Alexander; Popov, Dimiter; Lazarov, Lazar; Stoichev, Kosta

2008-01-01

The paper presents both international and Bulgarian experience in assessing HRA methods, underlying models approaches for their validation and verification by benchmarking HRA methods against different NPP simulator data. The organization, status, methodology and outlooks of the studies are described

Features of an advanced human reliability analysis method, AGAPE-ET

Energy Technology Data Exchange (ETDEWEB)

Kim, Jae Whan; Jung, Won Dea; Park, Jin Kyun [Korea Atomic Energy Research Institute, Taejeon (Korea, Republic of)

2005-11-15

This paper presents the main features of an advanced human reliability analysis (HRA) method, AGAPE-ET. It has the capabilities to deal with the diagnosis failures and the errors of commission (EOC), which have not been normally treated in the conventional HRAs. For the analysis of the potential for diagnosis failures, an analysis framework, which is called the misdiagnosis tree analysis (MDTA), and a taxonomy of the misdiagnosis causes with appropriate quantification schemes are provided. For the identification of the EOC events from the misdiagnosis, some procedural guidance is given. An example of the application of the method is also provided.

Features of an advanced human reliability analysis method, AGAPE-ET

International Nuclear Information System (INIS)

Kim, Jae Whan; Jung, Won Dea; Park, Jin Kyun

2005-01-01

This paper presents the main features of an advanced human reliability analysis (HRA) method, AGAPE-ET. It has the capabilities to deal with the diagnosis failures and the errors of commission (EOC), which have not been normally treated in the conventional HRAs. For the analysis of the potential for diagnosis failures, an analysis framework, which is called the misdiagnosis tree analysis (MDTA), and a taxonomy of the misdiagnosis causes with appropriate quantification schemes are provided. For the identification of the EOC events from the misdiagnosis, some procedural guidance is given. An example of the application of the method is also provided

Integration of Human Reliability Analysis Models into the Simulation-Based Framework for the Risk-Informed Safety Margin Characterization Toolkit

International Nuclear Information System (INIS)

Boring, Ronald; Mandelli, Diego; Rasmussen, Martin; Ulrich, Thomas; Groth, Katrina; Smith, Curtis

2016-01-01

This report presents an application of a computation-based human reliability analysis (HRA) framework called the Human Unimodel for Nuclear Technology to Enhance Reliability (HUNTER). HUNTER has been developed not as a standalone HRA method but rather as framework that ties together different HRA methods to model dynamic risk of human activities as part of an overall probabilistic risk assessment (PRA). While we have adopted particular methods to build an initial model, the HUNTER framework is meant to be intrinsically flexible to new pieces that achieve particular modeling goals. In the present report, the HUNTER implementation has the following goals: • Integration with a high fidelity thermal-hydraulic model capable of modeling nuclear power plant behaviors and transients • Consideration of a PRA context • Incorporation of a solid psychological basis for operator performance • Demonstration of a functional dynamic model of a plant upset condition and appropriate operator response This report outlines these efforts and presents the case study of a station blackout scenario to demonstrate the various modules developed to date under the HUNTER research umbrella.

Integration of Human Reliability Analysis Models into the Simulation-Based Framework for the Risk-Informed Safety Margin Characterization Toolkit

Energy Technology Data Exchange (ETDEWEB)

Boring, Ronald [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rasmussen, Martin [Norwegian Univ. of Science and Technology, Trondheim (Norway). Social Research; Herberger, Sarah [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ulrich, Thomas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Groth, Katrina [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-06-01

This report presents an application of a computation-based human reliability analysis (HRA) framework called the Human Unimodel for Nuclear Technology to Enhance Reliability (HUNTER). HUNTER has been developed not as a standalone HRA method but rather as framework that ties together different HRA methods to model dynamic risk of human activities as part of an overall probabilistic risk assessment (PRA). While we have adopted particular methods to build an initial model, the HUNTER framework is meant to be intrinsically flexible to new pieces that achieve particular modeling goals. In the present report, the HUNTER implementation has the following goals: • Integration with a high fidelity thermal-hydraulic model capable of modeling nuclear power plant behaviors and transients • Consideration of a PRA context • Incorporation of a solid psychological basis for operator performance • Demonstration of a functional dynamic model of a plant upset condition and appropriate operator response This report outlines these efforts and presents the case study of a station blackout scenario to demonstrate the various modules developed to date under the HUNTER research umbrella.

Human reliability assessment and probabilistic risk assessment

International Nuclear Information System (INIS)

Embrey, D.E.; Lucas, D.A.

1989-01-01

Human reliability assessment (HRA) is used within Probabilistic Risk Assessment (PRA) to identify the human errors (both omission and commission) which have a significant effect on the overall safety of the system and to quantify the probability of their occurrence. There exist a variey of HRA techniques and the selection of an appropriate one is often difficult. This paper reviews a number of available HRA techniques and discusses their strengths and weaknesses. The techniques reviewed include: decompositional methods, time-reliability curves and systematic expert judgement techniques. (orig.)

Dependence assessment in human reliability analysis based on D numbers and AHP

International Nuclear Information System (INIS)

Zhou, Xinyi; Deng, Xinyang; Deng, Yong; Mahadevan, Sankaran

2017-01-01

Highlights: • D numbers and AHP are combined to implement dependence assessment in HRA. • A new tool, called D numbers, is used to deal with the uncertainty in HRA. • The proposed method can well address the fuzziness and subjectivity in linguistic assessment. • The proposed method is well applicable in dependence assessment which inherently has a linguistic assessment process. - Abstract: Since human errors always cause heavy loss especially in nuclear engineering, human reliability analysis (HRA) has attracted more and more attention. Dependence assessment plays a vital role in HRA, measuring the dependence degree of human errors. Many researches have been done while still have improvement space. In this paper, a dependence assessment model based on D numbers and analytic hierarchy process (AHP) is proposed. Firstly, identify the factors used to measure the dependence level of two human operations. Besides, in terms of the suggested dependence level, determine and quantify the anchor points for each factor. Secondly, D numbers and AHP are adopted in model. Experts evaluate the dependence level of human operations for each factor. Then, the evaluation results are presented as D numbers and fused by D number’s combination rule that can obtain the dependence probability of human operations for each factor. The weights of factors can be determined by AHP. Thirdly, based on the dependence probability for each factor and its corresponding weight, the dependence probability of two human operations and its confidence can be obtained. The proposed method can well address the fuzziness and subjectivity in linguistic assessment. The proposed method is well applicable to assess the dependence degree of human errors in HRA which inherently has a linguistic assessment process.

Dependence assessment in human reliability analysis based on D numbers and AHP

Energy Technology Data Exchange (ETDEWEB)

Zhou, Xinyi; Deng, Xinyang [School of Computer and Information Science, Southwest University, Chongqing 400715 (China); Deng, Yong, E-mail: ydeng@swu.edu.cn [School of Computer and Information Science, Southwest University, Chongqing 400715 (China); Institute of Fundamental and Frontier Sciences, University of Electronic Science and Technology of China, Chengdu, Sichuan 610054 (China); Mahadevan, Sankaran [School of Engineering, Vanderbilt University, Nashville, TN 37235 (United States)

2017-03-15

Highlights: • D numbers and AHP are combined to implement dependence assessment in HRA. • A new tool, called D numbers, is used to deal with the uncertainty in HRA. • The proposed method can well address the fuzziness and subjectivity in linguistic assessment. • The proposed method is well applicable in dependence assessment which inherently has a linguistic assessment process. - Abstract: Since human errors always cause heavy loss especially in nuclear engineering, human reliability analysis (HRA) has attracted more and more attention. Dependence assessment plays a vital role in HRA, measuring the dependence degree of human errors. Many researches have been done while still have improvement space. In this paper, a dependence assessment model based on D numbers and analytic hierarchy process (AHP) is proposed. Firstly, identify the factors used to measure the dependence level of two human operations. Besides, in terms of the suggested dependence level, determine and quantify the anchor points for each factor. Secondly, D numbers and AHP are adopted in model. Experts evaluate the dependence level of human operations for each factor. Then, the evaluation results are presented as D numbers and fused by D number’s combination rule that can obtain the dependence probability of human operations for each factor. The weights of factors can be determined by AHP. Thirdly, based on the dependence probability for each factor and its corresponding weight, the dependence probability of two human operations and its confidence can be obtained. The proposed method can well address the fuzziness and subjectivity in linguistic assessment. The proposed method is well applicable to assess the dependence degree of human errors in HRA which inherently has a linguistic assessment process.

DEPEND-HRA-A method for consideration of dependency in human reliability analysis

International Nuclear Information System (INIS)

Cepin, Marko

2008-01-01

A consideration of dependencies between human actions is an important issue within the human reliability analysis. A method was developed, which integrates the features of existing methods and the experience from a full scope plant simulator. The method is used on real plant-specific human reliability analysis as a part of the probabilistic safety assessment of a nuclear power plant. The method distinguishes dependency for pre-initiator events from dependency for initiator and post-initiator events. The method identifies dependencies based on scenarios, where consecutive human actions are modeled, and based on a list of minimal cut sets, which is obtained by running the minimal cut set analysis considering high values of human error probabilities in the evaluation. A large example study, which consisted of a large number of human failure events, demonstrated the applicability of the method. Comparative analyses that were performed show that both selection of dependency method and selection of dependency levels within the method largely impact the results of probabilistic safety assessment. If the core damage frequency is not impacted much, the listings of important basic events in terms of risk increase and risk decrease factors may change considerably. More efforts are needed on the subject, which will prepare the background for more detailed guidelines, which will remove the subjectivity from the evaluations as much as it is possible

Human reliability analysis for steam generator feed-and-bleed accident in Bushehr NPP-1

International Nuclear Information System (INIS)

Jafarian, Reza; Sepanloo, Kamran

2006-01-01

According to the incident/accident reports, unsuccessful implementation of steam generator feed-and-bleed procedure is one of the most important events in nuclear power plants operation which greatly contributes to the level of risk of the plants. Generally, the loss of all feed water pumps flow (as one of the precursors) results in failure to maintain adequate cooling of the reactor core unless the operating crew initiate and follow the feed-and-bleed procedure correctly and timely. In this paper, firstly, a Human Reliability Analysis (HRA) event tree is presented delineating the major human activities and errors in the implementation of the steam generator (SG) feed-and-bleed procedure following the loss of (both normal and emergency) water feed to four SGs of Bushehr Nuclear Power Plant Unit 1 (BNPP-1). Secondly, the graphical method of task analysis as a part of HRA is used as a means of delineating correct and incorrect human actions. To be used in the probabilistic risk assessment (PRA), the outputs of the HRA event trees are fed into the system event trees, functional event trees or system fault trees. As a part of a probabilistic risk assessment of BNPP-1 and to assess the reliability of control room operators, a human reliability analysis model is applied based on the THERP (Technique for Human Error Rate Prediction) technique. The THERP method is used in the form of event trees named as the probability tree diagrams. In this research the Human Reliability Analysis event tree is constructed based on the background information and assumptions made and on a similar NPP task analysis. It is done so because the BNPP-1 is not an operational nuclear power plant. Thirdly, based on NUREG/CR-1278 Handbook, a computer program has been developed in Visual Basic language and used to illustrate the major human activities and determination of error rates of operators in the course of the implementation of the steam generator feed-and-bleed procedure. Finally, total

Human Reliability Analysis for steam generator feed-and-bleed accident in Bushehr NPP-1

International Nuclear Information System (INIS)

Jafarian, R.; Sepanloo, K.

2005-01-01

According to the incident/accident reports, unsuccessful implementation of steam generator feed-and-bleed procedure is one of the most important events in nuclear power plants operation which greatly contributes to the level of risk of the plants. Generally, the loss of all feed water pumps flow (as one of the precursors) results in failure to maintain adequate cooling of the reactor core unless the operating crew initiate and follow the feed-and-bleed procedure correctly and timely. In this paper, firstly, a Human Reliability Analysis (HRA) event tree is presented delineating the major human activities and errors in the implementation of the steam generator (SG) feed-and-bleed procedure following the loss of (both normal and emergency) water feed to four SGs of Bushehr Nuclear Power Plant unit1 (BNPP-1). Secondly, the graphical method of task analysis as a part of HRA is used as a means of delineating correct and incorrect human actions. To be used in the probabilistic risk assessment (PRA), the outputs of the HRA event trees are fed into the system event trees, functional event trees or system fault trees. As a part of a probabilistic risk assessment of BNPP-1 and to assess the reliability of control room operators, a human reliability analysis model is applied based on the THERP (Technique for Human Error Rate Prediction) technique. The THERP method is used in the form of event trees named as the probability tree diagrams. In this research the Human Reliability Analysis event tree is constructed based on the background information and assumptions made and on a similar NPP task analysis. It is done so because the BNPP-1 is not an operational nuclear power plant. Thirdly, based on NUREG/CR-1278 Handbook, a computer program has been developed in Visual Basic language and used to illustrate the major human activities and determination of error rates of operators in the course of the implementation of the steam generator feed-and-bleed procedure. Finally, total

A Conceptual Framework of Human Reliability Analysis for Execution Human Error in NPP Advanced MCRs

Energy Technology Data Exchange (ETDEWEB)

Jang, In Seok; Kim, Ar Ryum; Seong, Poong Hyun [KAIST, Daejeon (Korea, Republic of); Jung, Won Dea [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-08-15

The operation environment of Main Control Rooms (MCRs) in Nuclear Power Plants (NPPs) has changed with the adoption of new human-system interfaces that are based on computer-based technologies. The MCRs that include these digital and computer technologies, such as large display panels, computerized procedures, and soft controls, are called Advanced MCRs. Among the many features of Advanced MCRs, soft controls are a particularly important feature because the operation action in NPP Advanced MCRs is performed by soft control. Using soft controls such as mouse control, and touch screens, operators can select a specific screen, then choose the controller, and finally manipulate the given devices. Due to the different interfaces between soft control and hardwired conventional type control, different human error probabilities and a new Human Reliability Analysis (HRA) framework should be considered in the HRA for advanced MCRs. In other words, new human error modes should be considered for interface management tasks such as navigation tasks, and icon (device) selection tasks in monitors and a new framework of HRA method taking these newly generated human error modes into account should be considered. In this paper, a conceptual framework for a HRA method for the evaluation of soft control execution human error in advanced MCRs is suggested by analyzing soft control tasks.

A Conceptual Framework of Human Reliability Analysis for Execution Human Error in NPP Advanced MCRs

International Nuclear Information System (INIS)

Jang, In Seok; Kim, Ar Ryum; Seong, Poong Hyun; Jung, Won Dea

2014-01-01

The operation environment of Main Control Rooms (MCRs) in Nuclear Power Plants (NPPs) has changed with the adoption of new human-system interfaces that are based on computer-based technologies. The MCRs that include these digital and computer technologies, such as large display panels, computerized procedures, and soft controls, are called Advanced MCRs. Among the many features of Advanced MCRs, soft controls are a particularly important feature because the operation action in NPP Advanced MCRs is performed by soft control. Using soft controls such as mouse control, and touch screens, operators can select a specific screen, then choose the controller, and finally manipulate the given devices. Due to the different interfaces between soft control and hardwired conventional type control, different human error probabilities and a new Human Reliability Analysis (HRA) framework should be considered in the HRA for advanced MCRs. In other words, new human error modes should be considered for interface management tasks such as navigation tasks, and icon (device) selection tasks in monitors and a new framework of HRA method taking these newly generated human error modes into account should be considered. In this paper, a conceptual framework for a HRA method for the evaluation of soft control execution human error in advanced MCRs is suggested by analyzing soft control tasks

Collection and classification of human error and human reliability data from Indian nuclear power plants for use in PSA

International Nuclear Information System (INIS)

Subramaniam, K.; Saraf, R.K.; Sanyasi Rao, V.V.S.; Venkat Raj, V.; Venkatraman, R.

2000-01-01

Complex systems such as NPPs involve a large number of Human Interactions (HIs) in every phase of plant operations. Human Reliability Analysis (HRA) in the context of a PSA, attempts to model the HIs and evaluate/predict their impact on safety and reliability using human error/human reliability data. A large number of HRA techniques have been developed for modelling and integrating HIs into PSA but there is a significant lack of HAR data. In the face of insufficient data, human reliability analysts have had to resort to expert judgement methods in order to extend the insufficient data sets. In this situation, the generation of data from plant operating experience assumes importance. The development of a HRA data bank for Indian nuclear power plants was therefore initiated as part of the programme of work on HRA. Later, with the establishment of the coordinated research programme (CRP) on collection of human reliability data and use in PSA by IAEA in 1994-95, the development was carried out under the aegis of the IAEA research contract No. 8239/RB. The work described in this report covers the activities of development of a data taxonomy and a human error reporting form (HERF) based on it, data structuring, review and analysis of plant event reports, collection of data on human errors, analysis of the data and calculation of human error probabilities (HEPs). Analysis of plant operating experience does yield a good amount of qualitative data but obtaining quantitative data on human reliability in the form of HEPs is seen to be more difficult. The difficulties have been highlighted and some ways to bring about improvements in the data situation have been discussed. The implementation of a data system for HRA is described and useful features that can be incorporated in future systems are also discussed. (author)

Preliminary Human Reliability Issues in Reviewing SMART PSA

International Nuclear Information System (INIS)

Lee, Chang Ju; Sheen, Cheol

2010-01-01

Human reliability analysis (HRA) identifies the human failure events (HFEs) that can negatively impact normal or emergency plant operations, and systematically estimates probabilities of HFEs using data (when available), models, or expert judgment. In case of newly-conceptualized reactors like SMART (System-integrated Modular Advanced Reactor), HRA results must be provided by first evaluating the applicability of a set of human errors that has been typically applied in PSAs for existing PWRs. Additional human errors should also be identified reflecting its unique design and operational features. The objective of this paper is double-folded: to discuss a direction of HRA used in confirming risk level of SAMRT-type reactors; and to extract preliminarily considerable points or issues for regulatory verification, referred to available safety guides

Human reliability analysis as an evaluation tool of the emergency evacuation process on industrial installation

Energy Technology Data Exchange (ETDEWEB)

Santos, Isaac J.A.L. dos; Grecco, Claudio H.S.; Mol, Antonio C.A.; Carvalho, Paulo V.R.; Oliveira, Mauro V.; Botelho, Felipe Mury [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)]. E-mail: luquetti@ien.gov.br; grecco@ien.gov.br; mol@ien.gov.br; paulov@ien.gov.br; mvitor@ien.gov.br; felipemury@superig.com.br

2007-07-01

Human reliability is the probability that a person correctly performs some required activity by the system in a required time period and performs no extraneous activity that can degrade the system. Human reliability analysis (HRA) is the analysis, prediction and evaluation of work-oriented human performance using some indices as human error likelihood and probability of task accomplishment. The human error concept must not have connotation of guilt and punishment, having to be treated as a natural consequence, that emerges due to the not continuity between the human capacity and the system demand. The majority of the human error is a consequence of the work situation and not of the responsibility lack of the worker. The anticipation and the control of potentially adverse impacts of human action or interactions between the humans and the system are integral parts of the process safety, where the factors that influence the human performance must be recognized and managed. The aim of this paper is to propose a methodology to evaluate the emergency evacuation process on industrial installations including SLIM-MAUD, a HRA first-generation method, and using virtual reality and simulation software to build and to simulate the chosen emergency scenes. (author)

Human reliability analysis as an evaluation tool of the emergency evacuation process on industrial installation

International Nuclear Information System (INIS)

Santos, Isaac J.A.L. dos; Grecco, Claudio H.S.; Mol, Antonio C.A.; Carvalho, Paulo V.R.; Oliveira, Mauro V.; Botelho, Felipe Mury

2007-01-01

Human reliability is the probability that a person correctly performs some required activity by the system in a required time period and performs no extraneous activity that can degrade the system. Human reliability analysis (HRA) is the analysis, prediction and evaluation of work-oriented human performance using some indices as human error likelihood and probability of task accomplishment. The human error concept must not have connotation of guilt and punishment, having to be treated as a natural consequence, that emerges due to the not continuity between the human capacity and the system demand. The majority of the human error is a consequence of the work situation and not of the responsibility lack of the worker. The anticipation and the control of potentially adverse impacts of human action or interactions between the humans and the system are integral parts of the process safety, where the factors that influence the human performance must be recognized and managed. The aim of this paper is to propose a methodology to evaluate the emergency evacuation process on industrial installations including SLIM-MAUD, a HRA first-generation method, and using virtual reality and simulation software to build and to simulate the chosen emergency scenes. (author)

Science-Based Simulation Model of Human Performance for Human Reliability Analysis

International Nuclear Information System (INIS)

Kelly, Dana L.; Boring, Ronald L.; Mosleh, Ali; Smidts, Carol

2011-01-01

Human reliability analysis (HRA), a component of an integrated probabilistic risk assessment (PRA), is the means by which the human contribution to risk is assessed, both qualitatively and quantitatively. However, among the literally dozens of HRA methods that have been developed, most cannot fully model and quantify the types of errors that occurred at Three Mile Island. Furthermore, all of the methods lack a solid empirical basis, relying heavily on expert judgment or empirical results derived in non-reactor domains. Finally, all of the methods are essentially static, and are thus unable to capture the dynamics of an accident in progress. The objective of this work is to begin exploring a dynamic simulation approach to HRA, one whose models have a basis in psychological theories of human performance, and whose quantitative estimates have an empirical basis. This paper highlights a plan to formalize collaboration among the Idaho National Laboratory (INL), the University of Maryland, and The Ohio State University (OSU) to continue development of a simulation model initially formulated at the University of Maryland. Initial work will focus on enhancing the underlying human performance models with the most recent psychological research, and on planning follow-on studies to establish an empirical basis for the model, based on simulator experiments to be carried out at the INL and at the OSU.

Human reliability analysis for steam generator feed-and-bleed accident in Bushehr NPP-1

Energy Technology Data Exchange (ETDEWEB)

Jafarian, Reza [Valiasr University of Rafsanjan, Rafsanjan, 28 (Iran, Islamic Republic of); Sepanloo, Kamran [Atomic Energy Organization of Iran (AEOI), external link End of North Karegar Av., Tehran 14155-1339 (Iran, Islamic Republic of)

2006-07-01

According to the incident/accident reports, unsuccessful implementation of steam generator feed-and-bleed procedure is one of the most important events in nuclear power plants operation which greatly contributes to the level of risk of the plants. Generally, the loss of all feed water pumps flow (as one of the precursors) results in failure to maintain adequate cooling of the reactor core unless the operating crew initiate and follow the feed-and-bleed procedure correctly and timely. In this paper, firstly, a Human Reliability Analysis (HRA) event tree is presented delineating the major human activities and errors in the implementation of the steam generator (SG) feed-and-bleed procedure following the loss of (both normal and emergency) water feed to four SGs of Bushehr Nuclear Power Plant Unit 1 (BNPP-1). Secondly, the graphical method of task analysis as a part of HRA is used as a means of delineating correct and incorrect human actions. To be used in the probabilistic risk assessment (PRA), the outputs of the HRA event trees are fed into the system event trees, functional event trees or system fault trees. As a part of a probabilistic risk assessment of BNPP-1 and to assess the reliability of control room operators, a human reliability analysis model is applied based on the THERP (Technique for Human Error Rate Prediction) technique. The THERP method is used in the form of event trees named as the probability tree diagrams. In this research the Human Reliability Analysis event tree is constructed based on the background information and assumptions made and on a similar NPP task analysis. It is done so because the BNPP-1 is not an operational nuclear power plant. Thirdly, based on NUREG/CR-1278 Handbook, a computer program has been developed in Visual Basic language and used to illustrate the major human activities and determination of error rates of operators in the course of the implementation of the steam generator feed-and-bleed procedure. Finally, total

Human Performance Modeling for Dynamic Human Reliability Analysis

Energy Technology Data Exchange (ETDEWEB)

Boring, Ronald Laurids [Idaho National Laboratory; Joe, Jeffrey Clark [Idaho National Laboratory; Mandelli, Diego [Idaho National Laboratory

2015-08-01

Part of the U.S. Department of Energy’s (DOE’s) Light Water Reac- tor Sustainability (LWRS) Program, the Risk-Informed Safety Margin Charac- terization (RISMC) Pathway develops approaches to estimating and managing safety margins. RISMC simulations pair deterministic plant physics models with probabilistic risk models. As human interactions are an essential element of plant risk, it is necessary to integrate human actions into the RISMC risk framework. In this paper, we review simulation based and non simulation based human reliability analysis (HRA) methods. This paper summarizes the founda- tional information needed to develop a feasible approach to modeling human in- teractions in RISMC simulations.

Human reliability analysis approach to level 1 PSA - shutdown and low power operation of Mochovce NPP, Unit 1, Slovakia

International Nuclear Information System (INIS)

Stojka, Tibor; Holy, Jaroslav

2003-01-01

The paper presents general approach, used methods and form of documentation of the results as have been applied within the Human Reliability Analysis (HRA) task of the shutdown and low power PSA (SPSA) study for Mochovce nuclear power plant, Unit 1, Slovakia. The paper describes main goals of the HRA task within the SPSA project, applied methods and data sources. Basic steps of the HRA task and human errors (HE) classification are also specified in its first part. The main part of the paper deals with pre-initiator human errors, human-induced initiators and response to initiator human errors. Since the expert judgment method (SLIM) was used for the last type of human errors probability assessment, also related activities are described including preparation works (performance shaping factors (PSFs) selection, development of PSF classification tables, preparation of aid tools for interview with plant experts), qualitative analysis (sources of information and basic steps) and quantitative analysis itself (human errors classification for final quantification including criteria used for the classification, description of structure of the spreadsheet used for quantification and treatment with dependencies). The last part of the paper describes form of documentation of the final results and provides some findings. (author)

Impact of Advanced HSIs on Human Reliability

International Nuclear Information System (INIS)

Duc, Duy Le; Kim, Jonghyun

2013-01-01

This study investigated how a digitalized control room may influence operators' performance. The new HSI system is highly supportive of knowledge-based works and during complex scenarios. The most noticeable enhancement and gained improvement came from the utilization of the CPS. The results also showed that for different task types, the effects of distinctive features are diverse. Since there is large flexibility in the design of advanced HSI systems, HRA should also consider the detailed design analysis for the plant of interest. Current designs of advanced Main Control Room (MCR) apply digital technology whose features include the Advanced Alarm System (AAS), Digital Information Display System (DIDS), Computerized Procedure System (CPS), and Soft Controls (SCs). Despite the significant improvements made to these features, the full impact have yet to be thoroughly assessed using Human Reliability Analysis (HRA). Furthermore, the evaluation criteria for these new features have not been provided; and there are no available data to perform adjustments for human error probabilities (HEPs), which have been developed for conventional control rooms. The aim of this study is to examine the potential effects of the new Human-System Interface (HSI) features on human reliability. Firstly, the characteristics and functions of the AAS, DIDS, CPS and SCs are assessed and categorized. Secondly, tasks related to the features are discussed, focusing on the differences between conventional and digital control rooms. Qualitative investigation of the impacts is performed by reviewing available literatures. Finally, a new model for the quantitative estimation of HEPs based on the Korean Standard HRA (K-HRA) method is proposed

Impact of Advanced HSIs on Human Reliability

Energy Technology Data Exchange (ETDEWEB)

Duc, Duy Le; Kim, Jonghyun [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2013-10-15

This study investigated how a digitalized control room may influence operators' performance. The new HSI system is highly supportive of knowledge-based works and during complex scenarios. The most noticeable enhancement and gained improvement came from the utilization of the CPS. The results also showed that for different task types, the effects of distinctive features are diverse. Since there is large flexibility in the design of advanced HSI systems, HRA should also consider the detailed design analysis for the plant of interest. Current designs of advanced Main Control Room (MCR) apply digital technology whose features include the Advanced Alarm System (AAS), Digital Information Display System (DIDS), Computerized Procedure System (CPS), and Soft Controls (SCs). Despite the significant improvements made to these features, the full impact have yet to be thoroughly assessed using Human Reliability Analysis (HRA). Furthermore, the evaluation criteria for these new features have not been provided; and there are no available data to perform adjustments for human error probabilities (HEPs), which have been developed for conventional control rooms. The aim of this study is to examine the potential effects of the new Human-System Interface (HSI) features on human reliability. Firstly, the characteristics and functions of the AAS, DIDS, CPS and SCs are assessed and categorized. Secondly, tasks related to the features are discussed, focusing on the differences between conventional and digital control rooms. Qualitative investigation of the impacts is performed by reviewing available literatures. Finally, a new model for the quantitative estimation of HEPs based on the Korean Standard HRA (K-HRA) method is proposed.

IDHEAS – A NEW APPROACH FOR HUMAN RELIABILITY ANALYSIS

Energy Technology Data Exchange (ETDEWEB)

G. W. Parry; J.A Forester; V.N. Dang; S. M. L. Hendrickson; M. Presley; E. Lois; J. Xing

2013-09-01

This paper describes a method, IDHEAS (Integrated Decision-Tree Human Event Analysis System) that has been developed jointly by the US NRC and EPRI as an improved approach to Human Reliability Analysis (HRA) that is based on an understanding of the cognitive mechanisms and performance influencing factors (PIFs) that affect operator responses. The paper describes the various elements of the method, namely the performance of a detailed cognitive task analysis that is documented in a crew response tree (CRT), and the development of the associated time-line to identify the critical tasks, i.e. those whose failure results in a human failure event (HFE), and an approach to quantification that is based on explanations of why the HFE might occur.

Case study on the use of PSA methods: Human reliability analysis

International Nuclear Information System (INIS)

1991-04-01

The overall objective of treating human reliability in a probabilistic safety analysis is to ensure that the key human interactions of typical crews are accurately and systematically incorporated into the study in a traceable manner. An additional objective is to make the human reliability analysis (HRA) as realistic as possible, taking into account the emergency procedures, the man-machine interface, the focus of training process, and the knowledge and experience of the crews. Section 3 of the paper describes an overview of this analytical process which leads to three more detailed example problems described in Section 4. Section 5 discusses a peer review process. References are presented that are useful in performing HRAs. In addition appendices are provided for definitions, selected data and a generic list of performance shaping factors. 35 refs, figs and tabs

EDF/EPRI collaborative program on operator reliability experiments

International Nuclear Information System (INIS)

Villemeur, A.; Meslin, T.; Mosneron, F.; Worledge, D.H.; Joksimovich, V.; Spurgin, A.J.

1988-01-01

Electricite de France (EDF) and Electric Power Research Institute (EPRI) have been involved in human reliability studies over the last few years, in the context of improvements in human reliability assessment (HRA) methodologies, and have been following a systematic process since 1982 which consists of addressing the following five ingredients: - First, classify human interactions into a limited number of classes. - Second, introduce an acceptable framework to organize the application of HRA to PRA studies. - Third, select approach(es) to quantification. - Fourth, test promising models. - Fifth, establish an appropriate data base for tested model(s) with regard to specific applications. EPRI has just recently completed Phase I of the fourth topic. This primarily focused on testing the fundamental hypotheses behing the human cognitive reliability (HCR) correlation, using power plant simulators. EDF has been carrying out simulator studies since 1980, both for man-machine interface validation and HRA data collection. This background of experience provided a stepping stone for the EPRI project. On the other hand, before 1986, EDF had mainly been concentrating on getting qualitative insights from the tests and lacked experience in quantitative analysis and modeling, while EPRI had made advances in this latter area. Before the EPRI Operator Reliability Experiments (ORE) project was initiated, it was abundantly clear to EPRI and EDF that cooperation between the two could be useful and that both parties could gain from the cooperation

An overview of the evolution of human reliability analysis in the context of probabilistic risk assessment

International Nuclear Information System (INIS)

Bley, Dennis C.; Lois, Erasmia; Kolaczkowski, Alan M.; Forester, John Alan; Wreathall, John; Cooper, Susan E.

2009-01-01

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

An overview of the evolution of human reliability analysis in the context of probabilistic risk assessment.

Energy Technology Data Exchange (ETDEWEB)

Bley, Dennis C. (Buttonwood Consulting Inc., Oakton, VA); Lois, Erasmia (U.S. Nuclear Regulatory Commission, Washington, DC); Kolaczkowski, Alan M. (Science Applications International Corporation, Eugene, OR); Forester, John Alan; Wreathall, John (John Wreathall and Co., Dublin, OH); Cooper, Susan E. (U.S. Nuclear Regulatory Commission, Washington, DC)

2009-01-01

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.

Science Based Human Reliability Analysis: Using Digital Nuclear Power Plant Simulators for Human Reliability Research

Science.gov (United States)

Shirley, Rachel Elizabeth

Nuclear power plant (NPP) simulators are proliferating in academic research institutions and national laboratories in response to the availability of affordable, digital simulator platforms. Accompanying the new research facilities is a renewed interest in using data collected in NPP simulators for Human Reliability Analysis (HRA) research. An experiment conducted in The Ohio State University (OSU) NPP Simulator Facility develops data collection methods and analytical tools to improve use of simulator data in HRA. In the pilot experiment, student operators respond to design basis accidents in the OSU NPP Simulator Facility. Thirty-three undergraduate and graduate engineering students participated in the research. Following each accident scenario, student operators completed a survey about perceived simulator biases and watched a video of the scenario. During the video, they periodically recorded their perceived strength of significant Performance Shaping Factors (PSFs) such as Stress. This dissertation reviews three aspects of simulator-based research using the data collected in the OSU NPP Simulator Facility: First, a qualitative comparison of student operator performance to computer simulations of expected operator performance generated by the Information Decision Action Crew (IDAC) HRA method. Areas of comparison include procedure steps, timing of operator actions, and PSFs. Second, development of a quantitative model of the simulator bias introduced by the simulator environment. Two types of bias are defined: Environmental Bias and Motivational Bias. This research examines Motivational Bias--that is, the effect of the simulator environment on an operator's motivations, goals, and priorities. A bias causal map is introduced to model motivational bias interactions in the OSU experiment. Data collected in the OSU NPP Simulator Facility are analyzed using Structural Equation Modeling (SEM). Data include crew characteristics, operator surveys, and time to recognize

Review of advances in human reliability analysis of errors of commission-Part 2: EOC quantification

International Nuclear Information System (INIS)

Reer, Bernhard

2008-01-01

In close connection with examples relevant to contemporary probabilistic safety assessment (PSA), a review of advances in human reliability analysis (HRA) of post-initiator errors of commission (EOCs), i.e. inappropriate actions under abnormal operating conditions, has been carried out. The review comprises both EOC identification (part 1) and quantification (part 2); part 2 is presented in this article. Emerging HRA methods in this field are: ATHEANA, MERMOS, the EOC HRA method developed by Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS), the MDTA method and CREAM. The essential advanced features are on the conceptual side, especially to envisage the modeling of multiple contexts for an EOC to be quantified (ATHEANA, MERMOS and MDTA), in order to explicitly address adverse conditions. There is promising progress in providing systematic guidance to better account for cognitive demands and tendencies (GRS, CREAM), and EOC recovery (MDTA). Problematic issues are associated with the implementation of multiple context modeling and the assessment of context-specific error probabilities. Approaches for task or error opportunity scaling (CREAM, GRS) and the concept of reference cases (ATHEANA outlook) provide promising orientations for achieving progress towards data-based quantification. Further development work is needed and should be carried out in close connection with large-scale applications of existing approaches

Modeling and Quantification of Team Performance in Human Reliability Analysis for Probabilistic Risk Assessment

Energy Technology Data Exchange (ETDEWEB)

Jeffrey C. JOe; Ronald L. Boring

2014-06-01

Probabilistic Risk Assessment (PRA) and Human Reliability Assessment (HRA) are important technical contributors to the United States (U.S.) Nuclear Regulatory Commission’s (NRC) risk-informed and performance based approach to regulating U.S. commercial nuclear activities. Furthermore, all currently operating commercial NPPs in the U.S. are required by federal regulation to be staffed with crews of operators. Yet, aspects of team performance are underspecified in most HRA methods that are widely used in the nuclear industry. There are a variety of "emergent" team cognition and teamwork errors (e.g., communication errors) that are 1) distinct from individual human errors, and 2) important to understand from a PRA perspective. The lack of robust models or quantification of team performance is an issue that affects the accuracy and validity of HRA methods and models, leading to significant uncertainty in estimating HEPs. This paper describes research that has the objective to model and quantify team dynamics and teamwork within NPP control room crews for risk informed applications, thereby improving the technical basis of HRA, which improves the risk-informed approach the NRC uses to regulate the U.S. commercial nuclear industry.

Proof-of-Concept Demonstrations for Computation-Based Human Reliability Analysis. Modeling Operator Performance During Flooding Scenarios

Energy Technology Data Exchange (ETDEWEB)

Joe, Jeffrey Clark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Boring, Ronald Laurids [Idaho National Lab. (INL), Idaho Falls, ID (United States); Herberger, Sarah Elizabeth Marie [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis Lee [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-09-01

The United States (U.S.) Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) program has the overall objective to help sustain the existing commercial nuclear power plants (NPPs). To accomplish this program objective, there are multiple LWRS “pathways,” or research and development (R&D) focus areas. One LWRS focus area is called the Risk-Informed Safety Margin and Characterization (RISMC) pathway. Initial efforts under this pathway to combine probabilistic and plant multi-physics models to quantify safety margins and support business decisions also included HRA, but in a somewhat simplified manner. HRA experts at Idaho National Laboratory (INL) have been collaborating with other experts to develop a computational HRA approach, called the Human Unimodel for Nuclear Technology to Enhance Reliability (HUNTER), for inclusion into the RISMC framework. The basic premise of this research is to leverage applicable computational techniques, namely simulation and modeling, to develop and then, using RAVEN as a controller, seamlessly integrate virtual operator models (HUNTER) with 1) the dynamic computational MOOSE runtime environment that includes a full-scope plant model, and 2) the RISMC framework PRA models already in use. The HUNTER computational HRA approach is a hybrid approach that leverages past work from cognitive psychology, human performance modeling, and HRA, but it is also a significant departure from existing static and even dynamic HRA methods. This report is divided into five chapters that cover the development of an external flooding event test case and associated statistical modeling considerations.

Proof-of-Concept Demonstrations for Computation-Based Human Reliability Analysis. Modeling Operator Performance During Flooding Scenarios

International Nuclear Information System (INIS)

Joe, Jeffrey Clark; Boring, Ronald Laurids; Herberger, Sarah Elizabeth Marie; Mandelli, Diego; Smith, Curtis Lee

2015-01-01

The United States (U.S.) Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) program has the overall objective to help sustain the existing commercial nuclear power plants (NPPs). To accomplish this program objective, there are multiple LWRS 'pathways,' or research and development (R&D) focus areas. One LWRS focus area is called the Risk-Informed Safety Margin and Characterization (RISMC) pathway. Initial efforts under this pathway to combine probabilistic and plant multi-physics models to quantify safety margins and support business decisions also included HRA, but in a somewhat simplified manner. HRA experts at Idaho National Laboratory (INL) have been collaborating with other experts to develop a computational HRA approach, called the Human Unimodel for Nuclear Technology to Enhance Reliability (HUNTER), for inclusion into the RISMC framework. The basic premise of this research is to leverage applicable computational techniques, namely simulation and modeling, to develop and then, using RAVEN as a controller, seamlessly integrate virtual operator models (HUNTER) with 1) the dynamic computational MOOSE runtime environment that includes a full-scope plant model, and 2) the RISMC framework PRA models already in use. The HUNTER computational HRA approach is a hybrid approach that leverages past work from cognitive psychology, human performance modeling, and HRA, but it is also a significant departure from existing static and even dynamic HRA methods. This report is divided into five chapters that cover the development of an external flooding event test case and associated statistical modeling considerations.

A Multi-Methods Approach to HRA and Human Performance Modeling: A Field Assessment

Energy Technology Data Exchange (ETDEWEB)

Jacques Hugo; David I Gertman

2012-06-01

The Advanced Test Reactor (ATR) is a research reactor at the Idaho National Laboratory is primarily designed and used to test materials to be used in other, larger-scale and prototype reactors. The reactor offers various specialized systems and allows certain experiments to be run at their own temperature and pressure. The ATR Canal temporarily stores completed experiments and used fuel. It also has facilities to conduct underwater operations such as experiment examination or removal. In reviewing the ATR safety basis, a number of concerns were identified involving the ATR canal. A brief study identified ergonomic issues involving the manual handling of fuel elements in the canal that may increase the probability of human error and possible unwanted acute physical outcomes to the operator. In response to this concern, that refined the previous HRA scoping analysis by determining the probability of the inadvertent exposure of a fuel element to the air during fuel movement and inspection was conducted. The HRA analysis employed the SPAR-H method and was supplemented by information gained from a detailed analysis of the fuel inspection and transfer tasks. This latter analysis included ergonomics, work cycles, task duration, and workload imposed by tool and workplace characteristics, personal protective clothing, and operational practices that have the potential to increase physical and mental workload. Part of this analysis consisted of NASA-TLX analyses, combined with operational sequence analysis, computational human performance analysis (CHPA), and 3D graphical modeling to determine task failures and precursors to such failures that have safety implications. Experience in applying multiple analysis techniques in support of HRA methods is discussed.

A Multi-Methods Approach to HRA and Human Performance Modeling: A Field Assessment

International Nuclear Information System (INIS)

Hugo, Jacques; Gertman, David I.

2012-01-01

The Advanced Test Reactor (ATR) is a research reactor at the Idaho National Laboratory is primarily designed and used to test materials to be used in other, larger-scale and prototype reactors. The reactor offers various specialized systems and allows certain experiments to be run at their own temperature and pressure. The ATR Canal temporarily stores completed experiments and used fuel. It also has facilities to conduct underwater operations such as experiment examination or removal. In reviewing the ATR safety basis, a number of concerns were identified involving the ATR canal. A brief study identified ergonomic issues involving the manual handling of fuel elements in the canal that may increase the probability of human error and possible unwanted acute physical outcomes to the operator. In response to this concern, that refined the previous HRA scoping analysis by determining the probability of the inadvertent exposure of a fuel element to the air during fuel movement and inspection was conducted. The HRA analysis employed the SPAR-H method and was supplemented by information gained from a detailed analysis of the fuel inspection and transfer tasks. This latter analysis included ergonomics, work cycles, task duration, and workload imposed by tool and workplace characteristics, personal protective clothing, and operational practices that have the potential to increase physical and mental workload. Part of this analysis consisted of NASA-TLX analyses, combined with operational sequence analysis, computational human performance analysis (CHPA), and 3D graphical modeling to determine task failures and precursors to such failures that have safety implications. Experience in applying multiple analysis techniques in support of HRA methods is discussed.

Human reliability assessment in context

International Nuclear Information System (INIS)

Hollnagel, Erik

2005-01-01

Human Reliability Assessment (HRA) is conducted on the unspoken premise that 'human error' is a meaningful concept and that it can be associated with individual actions. The basis for this assumption it found in the origin of HRA, as a necessary extension of PSA to account for the impact of failures emanating from human actions. Although it was natural to model HRA on PSA, a large number of studies have shown that the premises are wrong, specifically that human and technological functions cannot be decomposed in the same manner. The general experience from accident studies also indicates that action failures are a function of the context, and that it is the variability of the context rather than the 'human error probability' that is the much sought for signal. Accepting this will have significant consequences for the way in which HRA, and ultimately also PSA, should be pursued

Establishing the Appropriate Attributes in Current Human Reliability Assessment Techniques for Nuclear Safety

International Nuclear Information System (INIS)

Bowie, Jane; Munley, Gary; Dang, Vinh; Wreathall, John; Bye, Andreas; Cooper, Susan; Marble, Julie; Peters, Sean; Xing, Jing; Fauchille, Veronique; Fiset, Jean Yves; Haage, Monica; Johanson, Gunnar; Jung, Won Dae; Kim, Jaewhan; Lee, Seung Jung; Kubicek, Jan; Le Bot, Pierre; Pesme, Helene; Preischl, Wolfgang; Salway, Alice; Amri, Abdallah; Lamarre, Greg; White, Andrew; )

2015-03-01

This report presents the results of a joint task of the Working Groups on Risk Assessment (WGRISK) and on Human and Organisational Factors (WGHOF) of the OECD/NEA CSNI, to identify desirable attributes of Human Reliability Assessment (HRA) methods, and to evaluate a range of HRA methods used in OECD member countries against those attributes. The purpose of this project is to provide information that will support regulators and operators of nuclear facilities when making judgements about the appropriateness of HRA methods for conducting assessments in support of Probabilistic Safety Assessments (PSA). The task was performed by an international team of Human Factors, HRA and PSA experts from a broad range of OECD member countries. As in other reviews of HRA methods, the study did not set out to recommend or promote the use of any particular HRA method. Rather the study aims to identify the strengths and limitations of commonly used and developing methods to aid those responsible for production of HRAs in selecting appropriate tools for specific HRA applications. The study also aims to assist regulators when making judgements on the appropriateness of the application of an HRA technique within nuclear-related probabilistic safety assessments. The report is aimed at practitioners in the field of human reliability assessment, human factors, and risk assessment more generally

A method and application study on holistic decision tree for human reliability analysis in nuclear power plant

International Nuclear Information System (INIS)

Sun Feng; Zhong Shan; Wu Zhiyu

2008-01-01

The paper introduces a human reliability analysis method mainly used in Nuclear Power Plant Safety Assessment and the Holistic Decision Tree (HDT) method and how to apply it. The focus is primarily on providing the basic framework and some background of HDT method and steps to perform it. Influence factors and quality descriptors are formed by the interview with operators in Qinshan Nuclear Power Plant and HDT analysis performed for SGTR and SLOCA based on this information. The HDT model can use a graphic tree structure to indicate that error rate is a function of influence factors. HDT method is capable of dealing with the uncertainty in HRA, and it is reliable and practical. (authors)

A Human Reliability Analysis of Post- Accident Human Errors in the Low Power and Shutdown PSA of KSNP

Energy Technology Data Exchange (ETDEWEB)

Kang, Daeil; Kim, J. H.; Jang, S. C

2007-03-15

Korea Atomic Energy Research Institute, using the ANS low power and shutdown (LPSD) probabilistic risk assessment (PRA) Standard, evaluated the LPSD PSA model of the KSNP, Yonggwang Units 5 and 6, and identified the items to be improved. The evaluation results of human reliability analysis (HRA) of the post-accident human errors in the LPSD PSA model for the KSNP showed that 10 items among 19 items of supporting requirements for those in the ANS PRA Standard were identified as them to be improved. Thus, we newly carried out a HRA for post-accident human errors in the LPSD PSA model for the KSNP. Following tasks are the improvements in the HRA of post-accident human errors of the LPSD PSA model for the KSNP compared with the previous one: Interviews with operators in the interpretation of the procedure, modeling of operator actions, and the quantification results of human errors, site visit. Applications of limiting value to the combined post-accident human errors. Documentation of information of all the input and bases for the detailed quantifications and the dependency analysis using the quantification sheets The assessment results for the new HRA results of post-accident human errors using the ANS LPSD PRA Standard show that above 80% items of its supporting requirements for post-accident human errors were graded as its Category II. The number of the re-estimated human errors using the LPSD Korea Standard HRA method is 385. Among them, the number of individual post-accident human errors is 253. The number of dependent post-accident human errors is 135. The quantification results of the LPSD PSA model for the KSNP with new HEPs show that core damage frequency (CDF) is increased by 5.1% compared with the previous baseline CDF It is expected that this study results will be greatly helpful to improve the PSA quality for the domestic nuclear power plants because they have sufficient PSA quality to meet the Category II of Supporting Requirements for the post

A Human Reliability Analysis of Post- Accident Human Errors in the Low Power and Shutdown PSA of KSNP

International Nuclear Information System (INIS)

Kang, Daeil; Kim, J. H.; Jang, S. C.

2007-03-01

Korea Atomic Energy Research Institute, using the ANS low power and shutdown (LPSD) probabilistic risk assessment (PRA) Standard, evaluated the LPSD PSA model of the KSNP, Yonggwang Units 5 and 6, and identified the items to be improved. The evaluation results of human reliability analysis (HRA) of the post-accident human errors in the LPSD PSA model for the KSNP showed that 10 items among 19 items of supporting requirements for those in the ANS PRA Standard were identified as them to be improved. Thus, we newly carried out a HRA for post-accident human errors in the LPSD PSA model for the KSNP. Following tasks are the improvements in the HRA of post-accident human errors of the LPSD PSA model for the KSNP compared with the previous one: Interviews with operators in the interpretation of the procedure, modeling of operator actions, and the quantification results of human errors, site visit. Applications of limiting value to the combined post-accident human errors. Documentation of information of all the input and bases for the detailed quantifications and the dependency analysis using the quantification sheets The assessment results for the new HRA results of post-accident human errors using the ANS LPSD PRA Standard show that above 80% items of its supporting requirements for post-accident human errors were graded as its Category II. The number of the re-estimated human errors using the LPSD Korea Standard HRA method is 385. Among them, the number of individual post-accident human errors is 253. The number of dependent post-accident human errors is 135. The quantification results of the LPSD PSA model for the KSNP with new HEPs show that core damage frequency (CDF) is increased by 5.1% compared with the previous baseline CDF It is expected that this study results will be greatly helpful to improve the PSA quality for the domestic nuclear power plants because they have sufficient PSA quality to meet the Category II of Supporting Requirements for the post

Use of Human Reliability Insights to Improve Decision-Making

International Nuclear Information System (INIS)

Julius, J. A.; Moieni, P.; Grobbelaar, J.; Kohlhepp, K.

2016-01-01

This paper describes the use of insights obtained during the development and application of human reliability analysis (HRA) as part of a probabilistic risk assessment (PRA) to support decision-making, including improvements to operations, training, and safety culture. Insights have been gained from the development and application of HRA as part of a PRA for nuclear power plants in the USA, Europe and Asia over the last two decades. These models consist of Level 1 and Level 2 PRA models of internal and external events, during full power and shutdown modes of plant operation. These insights include the use of human factors information to improve the qualitative portion of the HRA. The subsequent quantification in the HRA effectively prioritises the contributors to the unreliability of operator actions, and the process facilitates the identification of the factors that are important to the success of the operator actions. Additionally, the tools and techniques also allow for the evaluation of key assumptions and sources of uncertainty. The end results have been used to effectively support decision-making for day-to-day plant operations as well as licensing issues. HRA results have been used to provide feedback and improvements to plant procedures, operator training and other areas contributing the plant safety culture. Examples of the types of insights are presented in this paper. (author)

Top-down and bottom-up definitions of human failure events in human reliability analysis

International Nuclear Information System (INIS)

Boring, Ronald Laurids

2014-01-01

In the probabilistic risk assessments (PRAs) used in the nuclear industry, human failure events (HFEs) are determined as a subset of hardware failures, namely those hardware failures that could be triggered by human action or inaction. This approach is top-down, starting with hardware faults and deducing human contributions to those faults. Elsewhere, more traditionally human factors driven approaches would tend to look at opportunities for human errors first in a task analysis and then identify which of those errors is risk significant. The intersection of top-down and bottom-up approaches to defining HFEs has not been carefully studied. Ideally, both approaches should arrive at the same set of HFEs. This question is crucial, however, as human reliability analysis (HRA) methods are generalized to new domains like oil and gas. The HFEs used in nuclear PRAs tend to be top-down - defined as a subset of the PRA - whereas the HFEs used in petroleum quantitative risk assessments (QRAs) often tend to be bottom-up - derived from a task analysis conducted by human factors experts. The marriage of these approaches is necessary in order to ensure that HRA methods developed for top-down HFEs are also sufficient for bottom-up applications.

3D hra s technologií Leap Motion

OpenAIRE

Mainuš, Matěj

2014-01-01

Cílem této bakalářské práce bylo navrhnout a implementovat 3D hru labyrint ovládanou pomocí gest rukou. Pro rozpoznávání pohybu a gest hra využívá technologii Leap Motion, aplikace samotná je vytvořena v herním enginu Unity. Výsledkem práce je multiplatformní 3D hra s vlastní knihovnou, která integruje Leap Motion SDK do Unity a eliminuje chyby v detekci rukou. The goal of this bachelor's thesis is to design and create a 3D labyrinth game controlled by hand gestures. This is achieved by us...

Quantification of human reliability in probabilistic safety assessment

International Nuclear Information System (INIS)

Hirschberg, S.; Dankg, Vinh N.

1996-01-01

Human performance may substantially influence the reliability and safety of complex technical systems. For this reason, Human Reliability Analysis (HRA) constitutes an important part of Probabilistic Safety Assessment (PSAs) or Quantitative Risk Analyses (QRAs). The results of these studies as well as analyses of past accidents and incidents clearly demonstrate the importance of human interactions. The contribution of human errors to the core damage frequency (CDF), as estimated in the Swedish nuclear PSAs, are between 15 and 88%. A survey of the FRAs in the Swiss PSAs shows that also for the Swiss nuclear power plants the estimated HE contributions are substantial (49% of the CDF due to internal events in the case of Beznau and 70% in the case of Muehleberg; for the total CDF, including external events, 25% respectively 20%). Similar results can be extracted from the PSAs carried out for French, German, and US plants. In PSAs or QRAs, the adequate treatment of the human interactions with the system is a key to the understanding of accident sequences and their relative importance to overall risk. The main objectives of HRA are: first, to ensure that the key human interactions are systematically identified and incorporated into the safety analysis in a traceable manner, and second, to quantify the probabilities of their success and failure. Adopting a structured and systematic approach to the assessment of human performance makes it possible to provide greater confidence that the safety and availability of human-machine systems is not unduly jeopardized by human performance problems. Section 2 discusses the different types of human interactions analysed in PSAs. More generally, the section presents how HRA fits in the overall safety analysis, that is, how the human interactions to be quantified are identified. Section 3 addresses the methods for quantification. Section 4 concludes the paper by presenting some recommendations and pointing out the limitations of the

The cognitive environment simulation as a tool for modeling human performance and reliability

International Nuclear Information System (INIS)

Woods, D.D.; Pople, H. Jr.; Roth, E.M.

1990-01-01

The US Nuclear Regulatory Commission is sponsoring a research program to develop improved methods to model the cognitive behavior of nuclear power plant (NPP) personnel. Under this program, a tool for simulating how people form intentions to act in NPP emergency situations was developed using artificial intelligence (AI) techniques. This tool is called Cognitive Environment Simulation (CES). The Cognitive Reliability Assessment Technique (or CREATE) was also developed to specify how CBS can be used to enhance the measurement of the human contribution to risk in probabilistic risk assessment (PRA) studies. The next step in the research program was to evaluate the modeling tool and the method for using the tool for Human Reliability Analysis (HRA) in PRAs. Three evaluation activities were conducted. First, a panel of highly distinguished experts in cognitive modeling, AI, PRA and HRA provided a technical review of the simulation development work. Second, based on panel recommendations, CES was exercised on a family of steam generator tube rupture incidents where empirical data on operator performance already existed. Third, a workshop with HRA practitioners was held to analyze a worked example of the CREATE method to evaluate the role of CES/CREATE in HRA. The results of all three evaluations indicate that CES/CREATE represents a promising approach to modeling operator intention formation during emergency operations

A mid-layer model for human reliability analysis: understanding the cognitive causes of human failure events

International Nuclear Information System (INIS)

Shen, Song-Hua; Chang, James Y.H.; Boring, Ronald L.; Whaley, April M.; Lois, Erasmia; Langfitt Hendrickson, Stacey M.; Oxstrand, Johanna H.; Forester, John Alan; Kelly, Dana L.; Mosleh, Ali

2010-01-01

The Office of Nuclear Regulatory Research (RES) at the US Nuclear Regulatory Commission (USNRC) is sponsoring work in response to a Staff Requirements Memorandum (SRM) directing an effort to establish a single human reliability analysis (HRA) method for the agency or guidance for the use of multiple methods. As part of this effort an attempt to develop a comprehensive HRA qualitative approach is being pursued. This paper presents a draft of the method's middle layer, a part of the qualitative analysis phase that links failure mechanisms to performance shaping factors. Starting with a Crew Response Tree (CRT) that has identified human failure events, analysts identify potential failure mechanisms using the mid-layer model. The mid-layer model presented in this paper traces the identification of the failure mechanisms using the Information-Diagnosis/Decision-Action (IDA) model and cognitive models from the psychological literature. Each failure mechanism is grouped according to a phase of IDA. Under each phase of IDA, the cognitive models help identify the relevant performance shaping factors for the failure mechanism. The use of IDA and cognitive models can be traced through fault trees, which provide a detailed complement to the CRT.

A Mid-Layer Model for Human Reliability Analysis: Understanding the Cognitive Causes of Human Failure Events

Energy Technology Data Exchange (ETDEWEB)

Stacey M. L. Hendrickson; April M. Whaley; Ronald L. Boring; James Y. H. Chang; Song-Hua Shen; Ali Mosleh; Johanna H. Oxstrand; John A. Forester; Dana L. Kelly; Erasmia L. Lois

2010-06-01

The Office of Nuclear Regulatory Research (RES) is sponsoring work in response to a Staff Requirements Memorandum (SRM) directing an effort to establish a single human reliability analysis (HRA) method for the agency or guidance for the use of multiple methods. As part of this effort an attempt to develop a comprehensive HRA qualitative approach is being pursued. This paper presents a draft of the method’s middle layer, a part of the qualitative analysis phase that links failure mechanisms to performance shaping factors. Starting with a Crew Response Tree (CRT) that has identified human failure events, analysts identify potential failure mechanisms using the mid-layer model. The mid-layer model presented in this paper traces the identification of the failure mechanisms using the Information-Diagnosis/Decision-Action (IDA) model and cognitive models from the psychological literature. Each failure mechanism is grouped according to a phase of IDA. Under each phase of IDA, the cognitive models help identify the relevant performance shaping factors for the failure mechanism. The use of IDA and cognitive models can be traced through fault trees, which provide a detailed complement to the CRT.

Foundations for a time reliability correlation system to quantify human reliability

International Nuclear Information System (INIS)

Dougherty, E.M. Jr.; Fragola, J.R.

1988-01-01

Time reliability correlations (TRCs) have been used in human reliability analysis (HRA) in conjunction with probabilistic risk assessment (PRA) to quantify post-initiator human failure events. The first TRCs were judgmental but recent data taken from simulators have provided evidence for development of a system of TRCs. This system has the equational form: t = tau R X tau U , where the first factor is the lognormally distributed random variable of successful response time, derived from the simulator data, and the second factor is a unitary lognormal random variable to account for uncertainty in the model. The first random variable is further factored into a median response time and a factor to account for the dominant type of behavior assumed to be involved in the response and a second factor to account for other influences on the reliability of the response

A survey on the human reliability analysis methods for the design of Korean next generation reactor

International Nuclear Information System (INIS)

Lee, Yong Hee; Lee, J. W.; Park, J. C.; Kwack, H. Y.; Lee, K. Y.; Park, J. K.; Kim, I. S.; Jung, K. W.

2000-03-01

Enhanced features through applying recent domestic technologies may characterize the safety and efficiency of KNGR(Korea Next Generation Reactor). Human engineered interface and control room environment are expected to be beneficial to the human aspects of KNGR design. However, since the current method for human reliability analysis is not up to date after THERP/SHARP, it becomes hard to assess the potential of human errors due to both of the positive and negative effect of the design changes in KNGR. This is a state of the art report on the human reliability analysis methods that are potentially available for the application to the KNGR design. We surveyed every technical aspects of existing HRA methods, and compared them in order to obtain the requirements for the assessment of human error potentials within KNGR design. We categorized the more than 10 methods into the first and the second generation according to the suggestion of Dr. Hollnagel. THERP was revisited in detail. ATHEANA proposed by US NRC for an advanced design and CREAM proposed by Dr. Hollnagel were reviewed and compared. We conclude that the key requirements might include the enhancement in the early steps for human error identification and the quantification steps with considerations of more extended error shaping factors over PSFs(performance shaping factors). The utilization of the steps and approaches of ATHEANA and CREAM will be beneficial to the attainment of an appropriate HRA method for KNGR. However, the steps and data from THERP will be still maintained because of the continuity with previous PSA activities in KNGR design

A survey on the human reliability analysis methods for the design of Korean next generation reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Hee; Lee, J. W.; Park, J. C.; Kwack, H. Y.; Lee, K. Y.; Park, J. K.; Kim, I. S.; Jung, K. W

2000-03-01

Enhanced features through applying recent domestic technologies may characterize the safety and efficiency of KNGR(Korea Next Generation Reactor). Human engineered interface and control room environment are expected to be beneficial to the human aspects of KNGR design. However, since the current method for human reliability analysis is not up to date after THERP/SHARP, it becomes hard to assess the potential of human errors due to both of the positive and negative effect of the design changes in KNGR. This is a state of the art report on the human reliability analysis methods that are potentially available for the application to the KNGR design. We surveyed every technical aspects of existing HRA methods, and compared them in order to obtain the requirements for the assessment of human error potentials within KNGR design. We categorized the more than 10 methods into the first and the second generation according to the suggestion of Dr. Hollnagel. THERP was revisited in detail. ATHEANA proposed by US NRC for an advanced design and CREAM proposed by Dr. Hollnagel were reviewed and compared. We conclude that the key requirements might include the enhancement in the early steps for human error identification and the quantification steps with considerations of more extended error shaping factors over PSFs(performance shaping factors). The utilization of the steps and approaches of ATHEANA and CREAM will be beneficial to the attainment of an appropriate HRA method for KNGR. However, the steps and data from THERP will be still maintained because of the continuity with previous PSA activities in KNGR design.

Correlation Relationship of Performance Shaping Factors (PSFs) for Human Reliability Analysis

Energy Technology Data Exchange (ETDEWEB)

Bheka, M. Khumalo; Kim, Jonghyun [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2014-10-15

At TMI-2, operators permitted thousands of gallons of water to escape from the reactor plant before realizing that the coolant pumps were behaving abnormally. The coolant pumps were then turned off, which in turn led to the destruction of the reactor itself as cooling was completely lost within the core. Human also plays a role in many aspects of complex systems e.g. in design and manufacture of hardware, interface between human and system and also in maintaining such systems as well as for coping with unusual events that place the NPP system at a risk. This is why human reliability analysis (HRA) - an aspect of risk assessments which systematically identifies and analyzes the causes and consequences of human decisions and actions - is important in nuclear power plant operations. It either upgrades or degrades human performance; therefore it has an impact on the possibility of error. These PSFs can be used in various HRA methods to estimate Human Error Probabilities (HEPs). There are many current HRA methods who propose sets of PSFs for normal operation mode of NPP. Some of these PSFs in the sets have some degree of dependency and overlap. Overlapping PSFs introduce error in HEP evaluations due to the fact that some elements are counted more than once in data; this skews the relationship amongst PSF and masks the way that the elements interact to affect performance. This study uses a causal model that represents dependencies and relationships amongst PSFs for HEP evaluation during normal NPP operational states. The model is built taking into consideration the dependencies among PSFs and thus eliminating overlap. The use of an interdependent model of PSFs is expected to produce more accurate HEPs compared to other current methods. PSF sets produced in this study can be further used as nodes (variables) and directed arcs (causal influence between nodes) in HEP evaluation methods such as Bayesian belief (BN) networks. This study was done to estimate the relationships

Comparison of the Utility of Two Assessments for Explaining and Predicting Productivity Change: Well-Being Versus an HRA.

Science.gov (United States)

Gandy, William M; Coberley, Carter; Pope, James E; Rula, Elizabeth Y

2016-01-01

To compare utility of employee well-being to health risk assessment (HRA) as predictors of productivity change. Panel data from 2189 employees who completed surveys 2 years apart were used in hierarchical models comparing the influence of well-being and health risk on longitudinal changes in presenteeism and job performance. Absenteeism change was evaluated in a nonexempt subsample. Change in well-being was the most significant independent predictor of productivity change across all three measures. Comparing hierarchical models, well-being models performed significantly better than HRA models. The HRA added no incremental explanatory power over well-being in combined models. Alone, nonphysical health well-being components outperformed the HRA for all productivity measures. Well-being offers a more comprehensive measure of factors that influence productivity and can be considered preferential to HRA in understanding and addressing suboptimal productivity.

Handbook of human-reliability analysis with emphasis on nuclear power plant applications. Final report

Energy Technology Data Exchange (ETDEWEB)

Swain, A D; Guttmann, H E

1983-08-01

The primary purpose of the Handbook is to present methods, models, and estimated human error probabilities (HEPs) to enable qualified analysts to make quantitative or qualitative assessments of occurrences of human errors in nuclear power plants (NPPs) that affect the availability or operational reliability of engineered safety features and components. The Handbook is intended to provide much of the modeling and information necessary for the performance of human reliability analysis (HRA) as a part of probabilistic risk assessment (PRA) of NPPs. Although not a design guide, a second purpose of the Handbook is to enable the user to recognize error-likely equipment design, plant policies and practices, written procedures, and other human factors problems so that improvements can be considered. The Handbook provides the methodology to identify and quantify the potential for human error in NPP tasks.

Handbook of human-reliability analysis with emphasis on nuclear power plant applications. Final report

International Nuclear Information System (INIS)

Swain, A.D.; Guttmann, H.E.

1983-08-01

The primary purpose of the Handbook is to present methods, models, and estimated human error probabilities (HEPs) to enable qualified analysts to make quantitative or qualitative assessments of occurrences of human errors in nuclear power plants (NPPs) that affect the availability or operational reliability of engineered safety features and components. The Handbook is intended to provide much of the modeling and information necessary for the performance of human reliability analysis (HRA) as a part of probabilistic risk assessment (PRA) of NPPs. Although not a design guide, a second purpose of the Handbook is to enable the user to recognize error-likely equipment design, plant policies and practices, written procedures, and other human factors problems so that improvements can be considered. The Handbook provides the methodology to identify and quantify the potential for human error in NPP tasks

Models and data requirements for human reliability analysis

International Nuclear Information System (INIS)

1989-03-01

It has been widely recognised for many years that the safety of the nuclear power generation depends heavily on the human factors related to plant operation. This has been confirmed by the accidents at Three Mile Island and Chernobyl. Both these cases revealed how human actions can defeat engineered safeguards and the need for special operator training to cover the possibility of unexpected plant conditions. The importance of the human factor also stands out in the analysis of abnormal events and insights from probabilistic safety assessments (PSA's), which reveal a large proportion of cases having their origin in faulty operator performance. A consultants' meeting, organized jointly by the International Atomic Energy Agency (IAEA) and the International Institute for Applied Systems Analysis (IIASA) was held at IIASA in Laxenburg, Austria, December 7-11, 1987, with the aim of reviewing existing models used in Probabilistic Safety Assessment (PSA) for Human Reliability Analysis (HRA) and of identifying the data required. The report collects both the contributions offered by the members of the Expert Task Force and the findings of the extensive discussions that took place during the meeting. Refs, figs and tabs

The performance shaping factors influence analysis on the human reliability for NPP operation

International Nuclear Information System (INIS)

Farcasiu, M.; Nitoi, M.; Apostol, M.; Florescu, G.

2008-01-01

The Human Reliability Analysis (HRA) is an important step in Probabilistic Safety Assessment (PSA) studies and offers an advisability for concrete improvement of the man - machine - organization interfaces, reliability and safety. The goals of this analysis are to obtain sufficient details in order to understand and document all-important factors that affect human performance. The purpose of this paper is to estimate the human errors probabilities in view of the negative or positive effect of the human performance shaping factors (PSFs) for the mitigation of the initiating events which could occur in Nuclear Power Plant (NPP). Using THERP and SPAR-H methods, an analysis model of PSFs influence on the human reliability is performed. This model is applied to more important activities, that are necessary to mitigate 'one steam generator tube failure' event at Cernavoda NPP. The results are joint human error probabilities (JHEP) values estimated for the following situations: without regarding to PSFs influence; with PSFs in specific conditions; with PSFs which could have only positive influence and with PSFs which could have only negative influence. In addition, PSFs with negative influence were identified and using the DOE method, the necessary activities for changing negative influence were assigned. (authors)

Comparing the treatment of uncertainty in Bayesian networks and fuzzy expert systems used for a human reliability analysis application

International Nuclear Information System (INIS)

Baraldi, Piero; Podofillini, Luca; Mkrtchyan, Lusine; Zio, Enrico; Dang, Vinh N.

2015-01-01

The use of expert systems can be helpful to improve the transparency and repeatability of assessments in areas of risk analysis with limited data available. In this field, human reliability analysis (HRA) is no exception, and, in particular, dependence analysis is an HRA task strongly based on analyst judgement. The analysis of dependence among Human Failure Events refers to the assessment of the effect of an earlier human failure on the probability of the subsequent ones. This paper analyses and compares two expert systems, based on Bayesian Belief Networks and Fuzzy Logic (a Fuzzy Expert System, FES), respectively. The comparison shows that a BBN approach should be preferred in all the cases characterized by quantifiable uncertainty in the input (i.e. when probability distributions can be assigned to describe the input parameters uncertainty), since it provides a satisfactory representation of the uncertainty and its output is directly interpretable for use within PSA. On the other hand, in cases characterized by very limited knowledge, an analyst may feel constrained by the probabilistic framework, which requires assigning probability distributions for describing uncertainty. In these cases, the FES seems to lead to a more transparent representation of the input and output uncertainty. - Highlights: • We analyse treatment of uncertainty in two expert systems. • We compare a Bayesian Belief Network (BBN) and a Fuzzy Expert System (FES). • We focus on the input assessment, inference engines and output assessment. • We focus on an application problem of interest for human reliability analysis. • We emphasize the application rather than math to reach non-BBN or FES specialists

Probabilistic risk assessment for a loss of coolant accident in McMaster Nuclear Reactor and application of reliability physics model for modeling human reliability

Science.gov (United States)

Ha, Taesung

A probabilistic risk assessment (PRA) was conducted for a loss of coolant accident, (LOCA) in the McMaster Nuclear Reactor (MNR). A level 1 PRA was completed including event sequence modeling, system modeling, and quantification. To support the quantification of the accident sequence identified, data analysis using the Bayesian method and human reliability analysis (HRA) using the accident sequence evaluation procedure (ASEP) approach were performed. Since human performance in research reactors is significantly different from that in power reactors, a time-oriented HRA model (reliability physics model) was applied for the human error probability (HEP) estimation of the core relocation. This model is based on two competing random variables: phenomenological time and performance time. The response surface and direct Monte Carlo simulation with Latin Hypercube sampling were applied for estimating the phenomenological time, whereas the performance time was obtained from interviews with operators. An appropriate probability distribution for the phenomenological time was assigned by statistical goodness-of-fit tests. The human error probability (HEP) for the core relocation was estimated from these two competing quantities: phenomenological time and operators' performance time. The sensitivity of each probability distribution in human reliability estimation was investigated. In order to quantify the uncertainty in the predicted HEPs, a Bayesian approach was selected due to its capability of incorporating uncertainties in model itself and the parameters in that model. The HEP from the current time-oriented model was compared with that from the ASEP approach. Both results were used to evaluate the sensitivity of alternative huinan reliability modeling for the manual core relocation in the LOCA risk model. This exercise demonstrated the applicability of a reliability physics model supplemented with a. Bayesian approach for modeling human reliability and its potential

Screening, sensitivity, and uncertainty for the CREAM method of Human Reliability Analysis

International Nuclear Information System (INIS)

Bedford, Tim; Bayley, Clare; Revie, Matthew

2013-01-01

This paper reports a sensitivity analysis of the Cognitive Reliability and Error Analysis Method for Human Reliability Analysis. We consider three different aspects: the difference between the outputs of the Basic and Extended methods, on the same HRA scenario; the variability in outputs through the choices made for common performance conditions (CPCs); and the variability in outputs through the assignment of choices for cognitive function failures (CFFs). We discuss the problem of interpreting categories when applying the method, compare its quantitative structure to that of first generation methods and discuss also how dependence is modelled with the approach. We show that the control mode intervals used in the Basic method are too narrow to be consistent with the Extended method. This motivates a new screening method that gives improved accuracy with respect to the Basic method, in the sense that (on average) halves the uncertainty associated with the Basic method. We make some observations on the design of a screening method that are generally applicable in Risk Analysis. Finally, we propose a new method of combining CPC weights with nominal probabilities so that the calculated probabilities are always in range (i.e. between 0 and 1), while satisfying sensible properties that are consistent with the overall CREAM method

The dependence level analysis between the human actions in NPP Operation

International Nuclear Information System (INIS)

Farcasiu, M.; Nitoi, M.; Apostol, M.; Florescu, G.; Prisecaru, Ilie

2009-01-01

The Human Reliability Analysis (HRA) is an important method in Probabilistic Safety Assessment (PSA) studies and offers desirability for concrete improvement of the man - machine - organization interfaces, reliability and safety. An important step in HRA is the dependence level analysis between the human actions performed by the same person or between the actions performed by different persons, step in quantitative analysis of the human errors probabilities. The purpose of this paper is to develop a model to analyze the dependence level between human actions for Nuclear Power Plant (NPP) operation. The model estimates the conditional human error probabilities (CHEP) and joint human error probabilities (JHEP). The achieved sensitivity analyses determine human performance sensibility to systematic variations for dependence level between human actions. The human error probabilities estimated in this paper are adequate values for integration both in HRA and in PSA realized for NPP. This type of analysis helps in finding and analyzing the ways of reducing the likelihood of human errors, so that the impact of human factor to systems availability, reliability and safety can be realistically estimated. In order to demonstrate the usability of this model an analysis is performed upon the dependences between the necessary human actions in mitigating the consequences of LOCA events, particularly for the case of Cernavoda NPP. (authors)

The SACADA database for human reliability and human performance

International Nuclear Information System (INIS)

James Chang, Y.; Bley, Dennis; Criscione, Lawrence; Kirwan, Barry; Mosleh, Ali; Madary, Todd; Nowell, Rodney; Richards, Robert; Roth, Emilie M.; Sieben, Scott; Zoulis, Antonios

2014-01-01

Lack of appropriate and sufficient human performance data has been identified as a key factor affecting human reliability analysis (HRA) quality especially in the estimation of human error probability (HEP). The Scenario Authoring, Characterization, and Debriefing Application (SACADA) database was developed by the U.S. Nuclear Regulatory Commission (NRC) to address this data need. An agreement between NRC and the South Texas Project Nuclear Operating Company (STPNOC) was established to support the SACADA development with aims to make the SACADA tool suitable for implementation in the nuclear power plants' operator training program to collect operator performance information. The collected data would support the STPNOC's operator training program and be shared with the NRC for improving HRA quality. This paper discusses the SACADA data taxonomy, the theoretical foundation, the prospective data to be generated from the SACADA raw data to inform human reliability and human performance, and the considerations on the use of simulator data for HRA. Each SACADA data point consists of two information segments: context and performance results. Context is a characterization of the performance challenges to task success. The performance results are the results of performing the task. The data taxonomy uses a macrocognitive functions model for the framework. At a high level, information is classified according to the macrocognitive functions of detecting the plant abnormality, understanding the abnormality, deciding the response plan, executing the response plan, and team related aspects (i.e., communication, teamwork, and supervision). The data are expected to be useful for analyzing the relations between context, error modes and error causes in human performance

Fifty Years of THERP and Human Reliability Analysis

Energy Technology Data Exchange (ETDEWEB)

Ronald L. Boring

2012-06-01

In 1962 at a Human Factors Society symposium, Alan Swain presented a paper introducing a Technique for Human Error Rate Prediction (THERP). This was followed in 1963 by a Sandia Laboratories monograph outlining basic human error quantification using THERP and, in 1964, by a special journal edition of Human Factors on quantification of human performance. Throughout the 1960s, Swain and his colleagues focused on collecting human performance data for the Sandia Human Error Rate Bank (SHERB), primarily in connection with supporting the reliability of nuclear weapons assembly in the US. In 1969, Swain met with Jens Rasmussen of Risø National Laboratory and discussed the applicability of THERP to nuclear power applications. By 1975, in WASH-1400, Swain had articulated the use of THERP for nuclear power applications, and the approach was finalized in the watershed publication of the NUREG/CR-1278 in 1983. THERP is now 50 years old, and remains the most well known and most widely used HRA method. In this paper, the author discusses the history of THERP, based on published reports and personal communication and interviews with Swain. The author also outlines the significance of THERP. The foundations of human reliability analysis are found in THERP: human failure events, task analysis, performance shaping factors, human error probabilities, dependence, event trees, recovery, and pre- and post-initiating events were all introduced in THERP. While THERP is not without its detractors, and it is showing signs of its age in the face of newer technological applications, the longevity of THERP is a testament of its tremendous significance. THERP started the field of human reliability analysis. This paper concludes with a discussion of THERP in the context of newer methods, which can be seen as extensions of or departures from Swain’s pioneering work.

Human reliability under sleep deprivation: Derivation of performance shaping factor multipliers from empirical data

International Nuclear Information System (INIS)

Griffith, Candice D.; Mahadevan, Sankaran

2015-01-01

This paper develops a probabilistic approach that could use empirical data to derive values of performance shaping factor (PSF) multipliers for use in quantitative human reliability analysis (HRA). The proposed approach is illustrated with data on sleep deprivation effects on performance. A review of existing HRA methods reveals that sleep deprivation is not explicitly included at present, and expert opinion is frequently used to inform HRA model multipliers. In this paper, quantitative data from empirical studies regarding the effect of continuous hours of wakefulness on performance measures (reaction time, accuracy, and number of lapses) are used to develop a method to derive PSF multiplier values for sleep deprivation, in the context of the SPAR-H model. Data is extracted from the identified studies according to the meta-analysis research synthesis method and used to investigate performance trends and error probabilities. The error probabilities in test and control conditions are compared, and the resulting probability ratios are suggested for use in informing the selection of PSF multipliers in HRA methods. Although illustrated for sleep deprivation, the proposed methodology is general, and can be applied to other performance shaping factors. - Highlights: • Method proposed to derive performance shaping factor multipliers from empirical data. • Studies reporting the effect of sleep deprivation on performance are analyzed. • Test data using psychomotor vigilance tasks are analyzed. • Error probability multipliers computed for reaction time, lapses, and accuracy measures.

"Trojitá hra" - FTTH

OpenAIRE

Bobkovič, Peter

2008-01-01

Tento projekt sa zaoberá možnosťami realizácie pripojenia poslednej míle k užívatežovi pomocou FTTH, čo je optika až do domu. Celá práca je rozdelená na 2 časti a to všeobecný prehžad FTTH a vlastné meranie. Vo všeobecnom prehžade sú postupne uvedené druhy sietí FTTH, problematika s FTTH spojená, zmapovaná situácia vo svete a trojitá hra (cenník, ponuka TV a pod.). Vo druhej časti je vypracované postupné zmeranie a výsledky troch vykonaných meraní (PPM, OTDR, priama metóda) ako aj vzhžad čist...

HuRECA: Human Reliability Evaluator for Computer-based Control Room Actions

International Nuclear Information System (INIS)

Kim, Jae Whan; Lee, Seung Jun; Jang, Seung Cheol

2011-01-01

As computer-based design features such as computer-based procedures (CBP), soft controls (SCs), and integrated information systems are being adopted in main control rooms (MCR) of nuclear power plants, a human reliability analysis (HRA) method capable of dealing with the effects of these design features on human reliability is needed. From the observations of human factors engineering verification and validation experiments, we have drawn some major important characteristics on operator behaviors and design-related influencing factors (DIFs) from the perspective of human reliability. Firstly, there are new DIFs that should be considered in developing an HRA method for computer-based control rooms including especially CBP and SCs. In the case of the computer-based procedure rather than the paper-based procedure, the structural and managerial elements should be considered as important PSFs in addition to the procedural contents. In the case of the soft controllers, the so-called interface management tasks (or secondary tasks) should be reflected in the assessment of human error probability. Secondly, computer-based control rooms can provide more effective error recovery features than conventional control rooms. Major error recovery features for computer-based control rooms include the automatic logic checking function of the computer-based procedure and the information sharing feature of the general computer-based designs

Science-based HRA: experimental comparison of operator performance to IDAC (Information-Decision-Action Crew) simulations

Energy Technology Data Exchange (ETDEWEB)

Shirley, Rachel [The Ohio State Univ., Columbus, OH (United States); Smidts, Carol [The Ohio State Univ., Columbus, OH (United States); Boring, Ronald [Idaho National Lab. (INL), Idaho Falls, ID (United States); Li, Yuandan [Univ. of Maryland, College Park, MD (United States); Mosleh, Ali [Univ. of Maryland, College Park, MD (United States)

2015-02-01

Information-Decision-Action Crew (IDAC) operator model simulations of a Steam Generator Tube Rupture are compared to student operator performance in studies conducted in the Ohio State University’s Nuclear Power Plant Simulator Facility. This study is presented as a prototype for conducting simulator studies to validate key aspects of Human Reliability Analysis (HRA) methods. Seven student operator crews are compared to simulation results for crews designed to demonstrate three different decision-making strategies. The IDAC model used in the simulations is modified slightly to capture novice behavior rather that expert operators. Operator actions and scenario pacing are compared. A preliminary review of available performance shaping factors (PSFs) is presented. After the scenario in the NPP Simulator Facility, student operators review a video of the scenario and evaluate six PSFs at pre-determined points in the scenario. This provides a dynamic record of the PSFs experienced by the OSU student operators. In this preliminary analysis, Time Constraint Load (TCL) calculated in the IDAC simulations is compared to TCL reported by student operators. We identify potential modifications to the IDAC model to develop an “IDAC Student Operator Model.” This analysis provides insights into how similar experiments could be conducted using expert operators to improve the fidelity of IDAC simulations.

The Development of Marine Accidents Human Reliability Assessment Approach: HEART Methodology and MOP Model

OpenAIRE

Ludfi Pratiwi Bowo; Wanginingastuti Mutmainnah; Masao Furusho

2017-01-01

Humans are one of the important factors in the assessment of accidents, particularly marine accidents. Hence, studies are conducted to assess the contribution of human factors in accidents. There are two generations of Human Reliability Assessment (HRA) that have been developed. Those methodologies are classified by the differences of viewpoints of problem-solving, as the first generation and second generation. The accident analysis can be determined using three techniques of analysis; sequen...

A Human Performance Analysis on Emergency Tasks of a Nuclear Power Plant

International Nuclear Information System (INIS)

Jung, Wondea; Park, Jinkyun; Kim, Jae W.

2007-01-01

Considering risk-informed activities that require the probabilistic safety assessment (PSA) quality to be as high as possible, an HRA should be performed by using a systematic method with realistic plant specific data to meet the requirements for risk-informed applications. In order to obtain more objective HRA results, data extracted from real experiences or simulators is essential. To support HRA activities and researches, we have developed a human performance database, OPERA (Operator Performance and Reliability Analysis). This paper introduces a study to analyze an operators' performance time, which is the most crucial input for estimating a human error probability of a post-initiating human failure event

Impact of Advanced Alarm Systems and Information Displays on Human Reliability in the Digital Control Room of Nuclear Power Plants

International Nuclear Information System (INIS)

Kim, Jong Hyun; Dang, Vinh N

2011-01-01

This paper discusses the potential impacts of two advanced features of digital control rooms, alarm systems and information display systems, on the Human Reliability Analysis (HRA) in nuclear power plants. Although the features of digital control rooms have already been implemented in new or upgraded nuclear power plants, HRAs have so far not taken much credit for these features. In this circumstance, this paper aims at examining the potential effects of these features on human performance and discussing how these effects can be addressed with existing HRA methods. A conclusion derivable from past experimental studies is that those features are supportive in the severe conditions such as complex scenarios and knowledge-based works. However, in the less complex scenarios and rule-based work, they may have no difference with or sometimes negative impacts on operator performance. The discussion about the impact on the HRA is provided on the basis on the THERP method

Čertovy obrázky a královská hra

Czech Academy of Sciences Publication Activity Database

Opavská, Zdeňka

2017-01-01

Roč. 96, č. 10 (2017), s. 600-600 ISSN 0042-4544 Institutional support: RVO:68378092 Keywords : Czech * lexicology * phraseology Subject RIV: AI - Linguistics OBOR OECD: Linguistics https://vesmir.cz/cz/casopis/archiv-casopisu/2017/cislo-10/certovy-obrazky-kralovska-hra. html

The role of human reliability analysis for enhancing crew performance

International Nuclear Information System (INIS)

Hannaman, G.W.; Joksimovich, V.; Worledge, D.H.; Spurgin, A.J.

1986-01-01

This paper summarizes some aspects of EPRI-sponsored research undertaken in support of improving the PRA technology. In particular, the consideration of how human actions that impact accident sequences can be analyzed in a systematic way to supplement the type of ergonomic studies normally carried out in support of control room design. The HRA/PRA approach described not only identifies the operator information and interface needs, but also helps to identify issues and areas for additional research. The process includes a link to data collections. Preliminary collections of data and analytical benchmark support the idea that such analytical frameworks and models provide support for ranking the importance of various human reliability issues

Probabilistic Safety Assessment: An Effective Tool to Support “Systemic Approach” to Nuclear Safety and Analysis of Human and Organizational Aspects

International Nuclear Information System (INIS)

Kuzmina, I.

2016-01-01

The Probabilistic Safety Assessment (PSA) represents a comprehensive conceptual and analytical tool for quantitative evaluation of risk of undesirable consequences from nuclear facilities and drawing on qualitative insights for nuclear safety. PSA considers various technical, human, and organizational factors in an integral manner thus explicitly pursuing a true ‘systemic approach’ to safety and enabling holistic insights for further safety improvement. Human Reliability Analysis (HRA) is one of the major tasks within PSA. The poster paper provides an overview of the objectives and scope of PSA and HRA and discusses on further needs in the area of HRA. (author)

Design-related influencing factors of the computerized procedure system for inclusion into human reliability analysis of the advanced control room

International Nuclear Information System (INIS)

Kim, Jaewhan; Lee, Seung Jun; Jang, Seung Cheol; Ahn, Kwang-Il; Shin, Yeong Cheol

2013-01-01

This paper presents major design factors of the computerized procedure system (CPS) by task characteristics/requirements, with individual relative weight evaluated by the analytic hierarchy process (AHP) technique, for inclusion into human reliability analysis (HRA) of the advanced control rooms. Task characteristics/requirements of an individual procedural step are classified into four categories according to the dynamic characteristics of an emergency situation: (1) a single-static step, (2) a single-dynamic and single-checking step, (3) a single-dynamic and continuous-monitoring step, and (4) a multiple-dynamic and continuous-monitoring step. According to the importance ranking evaluation by the AHP technique, ‘clearness of the instruction for taking action’, ‘clearness of the instruction and its structure for rule interpretation’, and ‘adequate provision of requisite information’ were rated as of being higher importance for all the task classifications. Importance of ‘adequacy of the monitoring function’ and ‘adequacy of representation of the dynamic link or relationship between procedural steps’ is dependent upon task characteristics. The result of the present study gives a valuable insight on which design factors of the CPS should be incorporated, with relative importance or weight between design factors, into HRA of the advanced control rooms. (author)

Absorption, distribution, and excretion of 8-methoxypsoralen in HRA/Skh mice

International Nuclear Information System (INIS)

Muni, I.A.; Schneider, F.H.; Olsson, T.A. III; King, M.

1984-01-01

The tissue distribution and excretion of [ 3 H]8-methoxypsoralen (8-MOP), a well-accepted therapeutic agent for the treatment of psoriasis, was studied in hairless HRA/Skh female mice. Mice were given single oral doses of 6 mg of [ 3 H]8-MOP or 5-[ 14 C]8-MOP/kg in corn oil. Radiochemical analyses of tissues and excreta were accomplished by liquid scintillation counting. The 8-MOP appeared to be rapidly absorbed through the gastrointestinal tract, where the tritium levels were highest, followed by skin, blood, and liver; levels were lowest in fat (adipose tissue). In female HRA/Skh mice which had not been irradiated with UVA (320-400 nm), 84% of the carbon-14 and 58% of the tritium were recovered in the urine and feces within 24 hours of oral administration of 5-[ 14 C]8-MOP or [ 3 H]8-MOP, respectively. Animals that were exposed to UVA and received [3H]8-MOP excreted approximately 12% less tritium in the urine and feces compared with the animals which received no UVA

Using Evidence Credibility Decay Model for dependence assessment in human reliability analysis

International Nuclear Information System (INIS)

Guo, Xingfeng; Zhou, Yanhui; Qian, Jin; Deng, Yong

2017-01-01

Highlights: • A new computational model is proposed for dependence assessment in HRA. • We combined three factors of “CT”, “TR” and “SP” within Dempster–Shafer theory. • The BBA of “SP” is reconstructed by discounting rate based on the ECDM. • Simulation experiments are illustrated to show the efficiency of the proposed method. - Abstract: Dependence assessment among human errors plays an important role in human reliability analysis. When dependence between two sequent tasks exists in human reliability analysis, if the preceding task fails, the failure probability of the following task is higher than success. Typically, three major factors are considered: “Closeness in Time” (CT), “Task Relatedness” (TR) and “Similarity of Performers” (SP). Assume TR is not changed, both SP and CT influence the degree of dependence level and SP is discounted by the time as the result of combine two factors in this paper. In this paper, a new computational model is proposed based on the Dempster–Shafer Evidence Theory (DSET) and Evidence Credibility Decay Model (ECDM) to assess the dependence between tasks in human reliability analysis. First, the influenced factors among human tasks are identified and the basic belief assignments (BBAs) of each factor are constructed based on expert evaluation. Then, the BBA of SP is discounted as the result of combining two factors and reconstructed by using the ECDM, the factors are integrated into a fused BBA. Finally, the dependence level is calculated based on fused BBA. Experimental results demonstrate that the proposed model not only quantitatively describe the fact that the input factors influence the dependence level, but also exactly show how the dependence level regular changes with different situations of input factors.

The recovery factors analysis of the human errors for research reactors

International Nuclear Information System (INIS)

Farcasiu, M.; Nitoi, M.; Apostol, M.; Turcu, I.; Florescu, Ghe.

2006-01-01

The results of many Probabilistic Safety Assessment (PSA) studies show a very significant contribution of human errors to systems unavailability of the nuclear installations. The treatment of human interactions is considered one of the major limitations in the context of PSA. To identify those human actions that can have an effect on system reliability or availability applying the Human Reliability Analysis (HRA) is necessary. The recovery factors analysis of the human action is an important step in HRA. This paper presents how can be reduced the human errors probabilities (HEP) using those elements that have the capacity to recovery human error. The recovery factors modeling is marked to identify error likelihood situations or situations that conduct at development of the accident. This analysis is realized by THERP method. The necessary information was obtained from the operating experience of the research reactor TRIGA of the INR Pitesti. The required data were obtained from generic databases. (authors)

Considerations on the elements of quantifying human reliability

International Nuclear Information System (INIS)

Straeter, Oliver

2004-01-01

This paper attempts to provide a contribution for the discussion of what the term 'data' means and how the qualitative perspective can be linked with the quantitative one. It will argue that the terms 'quantitative data' and 'qualitative data' are not distinct but a continuum that spans over the entire spectrum of the expertise that has to be involved in the HRA process. It elaborates the rational behind any human reliability quantification figure and suggests a scientific way forward to better data for human reliability assessment

Human error data collection as a precursor to the development of a human reliability assessment capability in air traffic management

International Nuclear Information System (INIS)

Kirwan, Barry; Gibson, W. Huw; Hickling, Brian

2008-01-01

Quantified risk and safety assessments are now required for safety cases for European air traffic management (ATM) services. Since ATM is highly human-dependent for its safety, this suggests a need for formal human reliability assessment (HRA), as carried out in other industries such as nuclear power. Since the fundamental aspect of HRA is human error data, in the form of human error probabilities (HEPs), it was decided to take a first step towards development of an ATM HRA approach by deriving some HEPs in an ATM context. This paper reports a study, which collected HEPs via analysing the results of a real-time simulation involving air traffic controllers (ATCOs) and pilots, with a focus on communication errors. This study did indeed derive HEPs that were found to be concordant with other known communication human error data. This is a first step, and shows promise for HRA in ATM, since HEPs have been derived which could be used in safety assessments, although these HEPs are for only one (albeit critical) aspect of ATCOs' tasks (communications). The paper discusses options and potential ways forward for the development of a full HRA capability in ATM

Franchisingový projekt Hračky Modré z nebe

OpenAIRE

Maňasová, Petra

2008-01-01

The diploma thesis contains a description of the situation in franchising in the Czech Republic. The main part is focused on a franchising system and future strategy of the project Hračky Modré z nebe. It presents the company Eltsen a.s. which wants to realize the project in the future, its historical development and current situation. The thesis describes the way which the company would like to choose to be able to realize whole project.

Application of Bayesian Belief networks to the human reliability analysis of an oil tanker operation focusing on collision accidents

International Nuclear Information System (INIS)

Martins, Marcelo Ramos; Maturana, Marcos Coelho

2013-01-01

During the last three decades, several techniques have been developed for the quantitative study of human reliability. In the 1980s, techniques were developed to model systems by means of binary trees, which did not allow for the representation of the context in which human actions occur. Thus, these techniques cannot model the representation of individuals, their interrelationships, and the dynamics of a system. These issues make the improvement of methods for Human Reliability Analysis (HRA) a pressing need. To eliminate or at least attenuate these limitations, some authors have proposed modeling systems using Bayesian Belief Networks (BBNs). The application of these tools is expected to address many of the deficiencies in current approaches to modeling human actions with binary trees. This paper presents a methodology based on BBN for analyzing human reliability and applies this method to the operation of an oil tanker, focusing on the risk of collision accidents. The obtained model was used to determine the most likely sequence of hazardous events and thus isolate critical activities in the operation of the ship to study Internal Factors (IFs), Skills, and Management and Organizational Factors (MOFs) that should receive more attention for risk reduction.

Collection and classification of human reliability data for use in probabilistic safety assessments. Final report of a co-ordinated research programme 1995-1998

International Nuclear Information System (INIS)

1998-10-01

One of the most important lessons from abnormal events in NPPs is that they often result from incorrect human action. The awareness of the importance of human factors and human reliability has increased significantly over 10-15 years primarily owing to the fact that some major incidents (nuclear or non-nuclear) have had significant human error contributions. Each of these incidents have revealed different types of human errors, some of which were not generally recognized prior to the incident. The analysis of these events led to wide recognition of the fact that more information about human actions and errors is needed to improve the safety and operation of nuclear power plants. At the same time, the need or proper human reliability data was recognised in view of probabilistic safety assessment (PSA). No PSA study can be regarded as complete and accurate without adequate incorporation of human reliability analysis (HRA). In order to support incorporation of human reliability data into PSA the IAEA established a coordinated research programme with the objective to develop a common data base structure for human errors that might have important contributions to risk in different types of reactors. This report is a product of four years of coordinated research and describes the data collection and classification schemes currently in use in Member States as well as an outlook into future, discussing what types of data might be needed to support the new improved HRA methods which are currently under development

A limited assessment of the ASEP human reliability analysis procedure using simulator examination results

International Nuclear Information System (INIS)

Gore, B.R.; Dukelow, J.S. Jr.; Mitts, T.M.; Nicholson, W.L.

1995-10-01

This report presents a limited assessment of the conservatism of the Accident Sequence Evaluation Program (ASEP) human reliability analysis (HRA) procedure described in NUREG/CR-4772. In particular, the, ASEP post-accident, post-diagnosis, nominal HRA procedure is assessed within the context of an individual's performance of critical tasks on the simulator portion of requalification examinations administered to nuclear power plant operators. An assessment of the degree to which operator perforn:Lance during simulator examinations is an accurate reflection of operator performance during actual accident conditions was outside the scope of work for this project; therefore, no direct inference can be made from this report about such performance. The data for this study are derived from simulator examination reports from the NRC requalification examination cycle. A total of 4071 critical tasks were identified, of which 45 had been failed. The ASEP procedure was used to estimate human error probability (HEP) values for critical tasks, and the HEP results were compared with the failure rates observed in the examinations. The ASEP procedure was applied by PNL operator license examiners who supplemented the limited information in the examination reports with expert judgment based upon their extensive simulator examination experience. ASEP analyses were performed for a sample of 162 critical tasks selected randomly from the 4071, and the results were used to characterize the entire population. ASEP analyses were also performed for all of the 45 failed critical tasks. Two tests were performed to assess the bias of the ASEP HEPs compared with the data from the requalification examinations. The first compared the average of the ASEP HEP values with the fraction of the population actually failed and it found a statistically significant factor of two bias on the average

A technique for human error analysis (ATHEANA)

International Nuclear Information System (INIS)

Cooper, S.E.; Ramey-Smith, A.M.; Wreathall, J.; Parry, G.W.

1996-05-01

Probabilistic risk assessment (PRA) has become an important tool in the nuclear power industry, both for the Nuclear Regulatory Commission (NRC) and the operating utilities. Human reliability analysis (HRA) is a critical element of PRA; however, limitations in the analysis of human actions in PRAs have long been recognized as a constraint when using PRA. A multidisciplinary HRA framework has been developed with the objective of providing a structured approach for analyzing operating experience and understanding nuclear plant safety, human error, and the underlying factors that affect them. The concepts of the framework have matured into a rudimentary working HRA method. A trial application of the method has demonstrated that it is possible to identify potentially significant human failure events from actual operating experience which are not generally included in current PRAs, as well as to identify associated performance shaping factors and plant conditions that have an observable impact on the frequency of core damage. A general process was developed, albeit in preliminary form, that addresses the iterative steps of defining human failure events and estimating their probabilities using search schemes. Additionally, a knowledge- base was developed which describes the links between performance shaping factors and resulting unsafe actions

A technique for human error analysis (ATHEANA)

Energy Technology Data Exchange (ETDEWEB)

Cooper, S.E.; Ramey-Smith, A.M.; Wreathall, J.; Parry, G.W. [and others

1996-05-01

Probabilistic risk assessment (PRA) has become an important tool in the nuclear power industry, both for the Nuclear Regulatory Commission (NRC) and the operating utilities. Human reliability analysis (HRA) is a critical element of PRA; however, limitations in the analysis of human actions in PRAs have long been recognized as a constraint when using PRA. A multidisciplinary HRA framework has been developed with the objective of providing a structured approach for analyzing operating experience and understanding nuclear plant safety, human error, and the underlying factors that affect them. The concepts of the framework have matured into a rudimentary working HRA method. A trial application of the method has demonstrated that it is possible to identify potentially significant human failure events from actual operating experience which are not generally included in current PRAs, as well as to identify associated performance shaping factors and plant conditions that have an observable impact on the frequency of core damage. A general process was developed, albeit in preliminary form, that addresses the iterative steps of defining human failure events and estimating their probabilities using search schemes. Additionally, a knowledge- base was developed which describes the links between performance shaping factors and resulting unsafe actions.

Human reliability analysis for venting a BWR Mark I during a severe accident

International Nuclear Information System (INIS)

Nelson, W.R.; Blackman, H.S.

1986-01-01

A Human Reliability Analysis (HRA) was performed for the operator actions necessary to achieve containment venting for the Peach Bottom Atomic Power Station. This study was funded by the United States Nuclear Regulatory Commission (USNRC) and performed by the Idaho National Engineering Laboratory (INEL). The goal of the analysis was to estimate Human Error Probabilities (HEPs) to determine the likelihood that operators would fail to complete the venting process. The analysis was performed for two generic accident sequences: anticipated transient without scram (ATWS) and station blackout. Two major methods were used to estimate the HEPs: Technique for Human Error rate Prediction (THERP) and Success Likelihood Index Methodology (SLIM). For the ATWS scenarios analyzed, the calculated HEPs ranged from 0.23 to 0.35, depending on the number of vent paths that are required to reduce the containment pressure. It should be noted that the confidence bounds around these HEPs are large, However, even when considering the large confidence range, the failure probabilities are larger than what is typical for normal operator actions. For station blackout, the HEP is 1.0, resulting from the dangerous environmental conditions that are present, assuming that plant management would not deliberately expose personnel to a potentially fatal environment. These results are based on the analysis of draft procedures for containment venting. It is probable that careful revision of the procedures could reduce the human error probabilities

Using three-dimension virtual reality main control room for integrated system validation and human reliability analysis

International Nuclear Information System (INIS)

Yang Chihwei; Cheng Tsungchieh

2011-01-01

This study proposes the performance assessment in three-dimension virtual reality (3D-VR) main control room (MCR). The assessment is conducted for integrated system validation (ISV) purposes, and also for human reliability analyses (HRA). This paper describes the latest developments in 3D-VR applications, designated for the familiarization with MCR, specially taking into account the ISV and HRA. The experiences in 3D-VR application, the benefits and advantages of use of VR in training and maintenances of MCR operators in the target NPP are equally presented in this paper. Results gathered from the performance measurement lead to hazard mitigation and reduces the risk of human error in the operation and maintenance of nuclear equipments. The latest developments in simulation techniques, including 3D presentation enhances the above mentioned benefits, brings the MCR simulators closer to reality. In the near future, this type of 3D solutions should be applied more and more often in the design of MCR simulators. The presented 3D-VR are related to the MCR in NPPs, but the concept of composition and navigation through the system's elements can be easily applied for the purpose of any type of technical equipment and shall contribute in a similar manner to hazard prevention. (author)

A review of the models for evaluating organizational factors in human reliability analysis

International Nuclear Information System (INIS)

Alvarenga, Marco Antonio Bayout; Fonseca, Renato Alves da; Melo, Paulo Fernando Ferreira Frutuoso e

2009-01-01

Human factors should be evaluated in three hierarchical levels. The first level should concern the cognitive behavior of human beings during the control of processes that occur through the man-machine interface. Here, one evaluates human errors through human reliability models of first and second generation, like THERP, ASEP and HCR (first generation) and ATHEANA and CREAM (second generation). In the second level, the focus is in the cognitive behavior of human beings when they work in groups, as in nuclear power plants. The focus here is in the anthropological aspects that govern the interaction among human beings. In the third level, one is interested in the influence that the organizational culture exerts on human beings as well as on the tasks being performed. Here, one adds to the factors of the second level the economical and political aspects that shape the company organizational culture. Nowadays, the methodologies of HRA incorporate organizational factors in the group and organization levels through performance shaping factors. This work makes a critical evaluation of the deficiencies concerning human factors and evaluates the potential of quantitative techniques that have been proposed in the last decade to model organizational factors, including the interaction among groups, with the intention of eliminating this chronic deficiency of HRA models. Two important techniques will be discussed in this context: STAMP, based on system theory and FRAM, which aims at modeling the nonlinearities of socio-technical systems. (author)

A review of the models for evaluating organizational factors in human reliability analysis

Energy Technology Data Exchange (ETDEWEB)

Alvarenga, Marco Antonio Bayout; Fonseca, Renato Alves da [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)], e-mail: bayout@cnen.gov.br, e-mail: rfonseca@cnen.gov.br; Melo, Paulo Fernando Ferreira Frutuoso e [Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear], e-mail: frutuoso@con.ufrj.br

2009-07-01

Human factors should be evaluated in three hierarchical levels. The first level should concern the cognitive behavior of human beings during the control of processes that occur through the man-machine interface. Here, one evaluates human errors through human reliability models of first and second generation, like THERP, ASEP and HCR (first generation) and ATHEANA and CREAM (second generation). In the second level, the focus is in the cognitive behavior of human beings when they work in groups, as in nuclear power plants. The focus here is in the anthropological aspects that govern the interaction among human beings. In the third level, one is interested in the influence that the organizational culture exerts on human beings as well as on the tasks being performed. Here, one adds to the factors of the second level the economical and political aspects that shape the company organizational culture. Nowadays, the methodologies of HRA incorporate organizational factors in the group and organization levels through performance shaping factors. This work makes a critical evaluation of the deficiencies concerning human factors and evaluates the potential of quantitative techniques that have been proposed in the last decade to model organizational factors, including the interaction among groups, with the intention of eliminating this chronic deficiency of HRA models. Two important techniques will be discussed in this context: STAMP, based on system theory and FRAM, which aims at modeling the nonlinearities of socio-technical systems. (author)

The research history of the human behaviour from the probabilistic safety analysis viewpoint

International Nuclear Information System (INIS)

Pyy, P.

1993-01-01

The so called human errors have always been apart of the everyday life of the mankind. In that sense, the discussion on man has a contributor to the operational safety of nuclear power plants is nothing new. It is interesting, that there do not exist widely accepted definitions of the human error nor the human reliability. Some of them are discussed at the beginning of this article. The second Chapter discusses the past and today of the research of man as a contributor to safety. Similarly, the development of Human Reliability Analysis (HRA) is described. The article, then, discusses the methods used in the contemporary HRA. The division between the identification of important human activities and their probability estimation is made. Especially, the pros and cons of the approaches and data sources used in the HRA are reviewed on a coarce level. At the end, a view on the use of expert judgment is given. The human behaviour has been an endless topic of research in the history - and will be it in future as well. In the conclusion of the article an opinion is given on the development during the past 30 years. Then, a rapid view on the possible future of the area is given. (orig.)

Loss of coolant accident (LOCA) analysis for McMaster Nuclear Reactor through probabilistic risk assessment (PRA)

Energy Technology Data Exchange (ETDEWEB)

Ha, T.; Garland, W.J. [McMaster Univ., Dept. of Engineering Physics, Hamilton, Ontario (Canada)]. E-mail: hats@mcmaster.ca

2006-07-01

A probabilistic risk assessment (PRA) was conducted for the loss of coolant accident (LOCA) sequence in the McMaster Nuclear Reactor (MNR). A level 1 PRA was completed including event sequence modeling, system modeling, and quantification. To support the quantification of the accident sequence identified, data analysis using the Bayesian method and human reliability analysis (HRA) using the ASEP approach were performed. Since human performance in research reactors is significantly different from that in power reactors, a different time-oriented HRA model was proposed and applied for the estimation of the human error probability (HEP) of core relocation. This HEP estimate was less than that by the ASEP approach by a factor of about 2. These two HEP estimates were used for sensitivity analysis, and modeling uncertainty in the PRA models was quantified. This showed the necessity of appropriate human reliability models in PRA for research reactors. This method could be implemented for the operators' actions which require extensive manual execution with little cognitive load, as might be the case for some maintenance operations in power reactors. (author)

Human reliability data collection and modelling

International Nuclear Information System (INIS)

1991-09-01

The main purpose of this document is to review and outline the current state-of-the-art of the Human Reliability Assessment (HRA) used for quantitative assessment of nuclear power plants safe and economical operation. Another objective is to consider Human Performance Indicators (HPI) which can alert plant manager and regulator to departures from states of normal and acceptable operation. These two objectives are met in the three sections of this report. The first objective has been divided into two areas, based on the location of the human actions being considered. That is, the modelling and data collection associated with control room actions are addressed first in chapter 1 while actions outside the control room (including maintenance) are addressed in chapter 2. Both chapters 1 and 2 present a brief outline of the current status of HRA for these areas, and major outstanding issues. Chapter 3 discusses HPI. Such performance indicators can signal, at various levels, changes in factors which influence human performance. The final section of this report consists of papers presented by the participants of the Technical Committee Meeting. A separate abstract was prepared for each of these papers. Refs, figs and tabs

Human reliability analysis

International Nuclear Information System (INIS)

Dougherty, E.M.; Fragola, J.R.

1988-01-01

The authors present a treatment of human reliability analysis incorporating an introduction to probabilistic risk assessment for nuclear power generating stations. They treat the subject according to the framework established for general systems theory. Draws upon reliability analysis, psychology, human factors engineering, and statistics, integrating elements of these fields within a systems framework. Provides a history of human reliability analysis, and includes examples of the application of the systems approach

Analysis and Application of Reliability

International Nuclear Information System (INIS)

Jeong, Hae Seong; Park, Dong Ho; Kim, Jae Ju

1999-05-01

This book tells of analysis and application of reliability, which includes definition, importance and historical background of reliability, function of reliability and failure rate, life distribution and assumption of reliability, reliability of unrepaired system, reliability of repairable system, sampling test of reliability, failure analysis like failure analysis by FEMA and FTA, and cases, accelerated life testing such as basic conception, acceleration and acceleration factor, and analysis of accelerated life testing data, maintenance policy about alternation and inspection.

Human reliability analysis in probabilistic safety assessment for nuclear power plants. A Safety Practice. A publication within the NUSS programme

International Nuclear Information System (INIS)

1995-01-01

Probabilistic safety assessment (PSA) is playing an increasingly important role in the safe operation of nuclear power plants throughout the world. In order to establish a consistent framework for conducting PSA studies, for promoting technology transfer of the state of the art, and for encouraging uniformity in the way PSA is carried out, the IAEA is preparing a set of publications which gives guidance on various aspects of PSA. This document presents a practical approach for incorporating human reliability analysis (HRA) into PSA. It describes the steps needed and the documentation that should be provided both to support the PSA itself and to ensure effective communication of important information arising from the studies. It also describes a framework for analysing those human actions which could affect safety and for relating such human influences to specific parts of a PSA. This Safety Practice also addresses the limitations of PSA in taking account of human factors in relation to safety and risk. Refs, figs and tabs

Discrete dynamic event tree modeling and analysis of nuclear power plant crews for safety assessment

International Nuclear Information System (INIS)

Mercurio, D.

2011-01-01

Current Probabilistic Risk Assessment (PRA) and Human Reliability Analysis (HRA) methodologies model the evolution of accident sequences in Nuclear Power Plants (NPPs) mainly based on Logic Trees. The evolution of these sequences is a result of the interactions between the crew and plant; in current PRA methodologies, simplified models of these complex interactions are used. In this study, the Accident Dynamic Simulator (ADS), a modeling framework based on the Discrete Dynamic Event Tree (DDET), has been used for the simulation of crew-plant interactions during potential accident scenarios in NPPs. In addition, an operator/crew model has been developed to treat the response of the crew to the plant. The 'crew model' is made up of three operators whose behavior is guided by a set of rules-of-behavior (which represents the knowledge and training of the operators) coupled with written and mental procedures. In addition, an approach for addressing the crew timing variability in DDETs has been developed and implemented based on a set of HRA data from a simulator study. Finally, grouping techniques were developed and applied to the analysis of the scenarios generated by the crew-plant simulation. These techniques support the post-simulation analysis by grouping similar accident sequences, identifying the key contributing events, and quantifying the conditional probability of the groups. These techniques are used to characterize the context of the crew actions in order to obtain insights for HRA. The model has been applied for the analysis of a Small Loss Of Coolant Accident (SLOCA) event for a Pressurized Water Reactor (PWR). The simulation results support an improved characterization of the performance conditions or context of operator actions, which can be used in an HRA, in the analysis of the reliability of the actions. By providing information on the evolution of system indications, dynamic of cues, crew timing in performing procedure steps, situation

Derivation of main drivers affecting the possibility of human errors during low power and shutdown operation

Energy Technology Data Exchange (ETDEWEB)

Kim, Ar Ryum; Seong, Poong Hyun [KAIST, Daejeon (Korea, Republic of); Park, Jin Kyun; Kim, Jae Whan [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In order to estimate the possibility of human error and identify its nature, human reliability analysis (HRA) methods have been implemented. For this, various HRA methods have been developed so far: techniques for human error rate prediction (THERP), cause based decision tree (CBDT), the cognitive reliability and error analysis method (CREAM) and so on. Most HRA methods have been developed with a focus on full power operation of NPPs even though human performance may more largely affect the safety of the system during low power and shutdown (LPSD) operation than it would when the system is in full power operation. In this regard, it is necessary to conduct a research for developing HRA method to be used in LPSD operation. For the first step of the study, main drivers which affect the possibility of human error have been developed. Drivers which are commonly called as performance shaping factors (PSFs) are aspects of the human's individual characteristics, environment, organization, or task that specifically decrements or improves human performance, thus respectively increasing or decreasing the likelihood of human errors. In order to estimate the possibility of human error and identify its nature, human reliability analysis (HRA) methods have been implemented. For this, various HRA methods have been developed so far: techniques for human error rate prediction (THERP), cause based decision tree (CBDT), the cognitive reliability and error analysis method (CREAM) and so on. Most HRA methods have been developed with a focus on full power operation of NPPs even though human performance may more largely affect the safety of the system during low power and shutdown (LPSD) operation than it would when the system is in full power operation. In this regard, it is necessary to conduct a research for developing HRA method to be used in LPSD operation. For the first step of the study, main drivers which affect the possibility of human error have been developed. Drivers

Derivation of main drivers affecting the possibility of human errors during low power and shutdown operation

International Nuclear Information System (INIS)

Kim, Ar Ryum; Seong, Poong Hyun; Park, Jin Kyun; Kim, Jae Whan

2016-01-01

In order to estimate the possibility of human error and identify its nature, human reliability analysis (HRA) methods have been implemented. For this, various HRA methods have been developed so far: techniques for human error rate prediction (THERP), cause based decision tree (CBDT), the cognitive reliability and error analysis method (CREAM) and so on. Most HRA methods have been developed with a focus on full power operation of NPPs even though human performance may more largely affect the safety of the system during low power and shutdown (LPSD) operation than it would when the system is in full power operation. In this regard, it is necessary to conduct a research for developing HRA method to be used in LPSD operation. For the first step of the study, main drivers which affect the possibility of human error have been developed. Drivers which are commonly called as performance shaping factors (PSFs) are aspects of the human's individual characteristics, environment, organization, or task that specifically decrements or improves human performance, thus respectively increasing or decreasing the likelihood of human errors. In order to estimate the possibility of human error and identify its nature, human reliability analysis (HRA) methods have been implemented. For this, various HRA methods have been developed so far: techniques for human error rate prediction (THERP), cause based decision tree (CBDT), the cognitive reliability and error analysis method (CREAM) and so on. Most HRA methods have been developed with a focus on full power operation of NPPs even though human performance may more largely affect the safety of the system during low power and shutdown (LPSD) operation than it would when the system is in full power operation. In this regard, it is necessary to conduct a research for developing HRA method to be used in LPSD operation. For the first step of the study, main drivers which affect the possibility of human error have been developed. Drivers which

Power electronics reliability analysis.

Energy Technology Data Exchange (ETDEWEB)

Smith, Mark A.; Atcitty, Stanley

2009-12-01

This report provides the DOE and industry with a general process for analyzing power electronics reliability. The analysis can help with understanding the main causes of failures, downtime, and cost and how to reduce them. One approach is to collect field maintenance data and use it directly to calculate reliability metrics related to each cause. Another approach is to model the functional structure of the equipment using a fault tree to derive system reliability from component reliability. Analysis of a fictitious device demonstrates the latter process. Optimization can use the resulting baseline model to decide how to improve reliability and/or lower costs. It is recommended that both electric utilities and equipment manufacturers make provisions to collect and share data in order to lay the groundwork for improving reliability into the future. Reliability analysis helps guide reliability improvements in hardware and software technology including condition monitoring and prognostics and health management.

Reliability of drivers in urban intersections.

Science.gov (United States)

Gstalter, Herbert; Fastenmeier, Wolfgang

2010-01-01

The concept of human reliability has been widely used in industrial settings by human factors experts to optimise the person-task fit. Reliability is estimated by the probability that a task will successfully be completed by personnel in a given stage of system operation. Human Reliability Analysis (HRA) is a technique used to calculate human error probabilities as the ratio of errors committed to the number of opportunities for that error. To transfer this notion to the measurement of car driver reliability the following components are necessary: a taxonomy of driving tasks, a definition of correct behaviour in each of these tasks, a list of errors as deviations from the correct actions and an adequate observation method to register errors and opportunities for these errors. Use of the SAFE-task analysis procedure recently made it possible to derive driver errors directly from the normative analysis of behavioural requirements. Driver reliability estimates could be used to compare groups of tasks (e.g. different types of intersections with their respective regulations) as well as groups of drivers' or individual drivers' aptitudes. This approach was tested in a field study with 62 drivers of different age groups. The subjects drove an instrumented car and had to complete an urban test route, the main features of which were 18 intersections representing six different driving tasks. The subjects were accompanied by two trained observers who recorded driver errors using standardized observation sheets. Results indicate that error indices often vary between both the age group of drivers and the type of driving task. The highest error indices occurred in the non-signalised intersection tasks and the roundabout, which exactly equals the corresponding ratings of task complexity from the SAFE analysis. A comparison of age groups clearly shows the disadvantage of older drivers, whose error indices in nearly all tasks are significantly higher than those of the other groups

Sensitivity Analysis on LOCCW of Westinghouse typed Reactors Considering WOG2000 RCP Seal Leakage Model

International Nuclear Information System (INIS)

Na, Jang-Hwan; Jeon, Ho-Jun; Hwang, Seok-Won

2015-01-01

In this paper, we focus on risk insights of Westinghouse typed reactors. We identified that Reactor Coolant Pump (RCP) seal integrity is the most important contributor to Core Damage Frequency (CDF). As we reflected the latest technical report; WCAP-15603(Rev. 1-A), 'WOG2000 RCP Seal Leakage Model for Westinghouse PWRs' instead of the old version, RCP seal integrity became more important to Westinghouse typed reactors. After Fukushima accidents, Korea Hydro and Nuclear Power (KHNP) decided to develop Low Power and Shutdown (LPSD) Probabilistic Safety Assessment (PSA) models and upgrade full power PSA models of all operating Nuclear Power Plants (NPPs). As for upgrading full power PSA models, we have tried to standardize the methodology of CCF (Common Cause Failure) and HRA (Human Reliability Analysis), which are the most influential factors to risk measures of NPPs. Also, we have reviewed and reflected the latest operating experiences, reliability data sources and technical methods to improve the quality of PSA models. KHNP has operating various types of reactors; Optimized Pressurized Reactor (OPR) 1000, CANDU, Framatome and Westinghouse. So, one of the most challengeable missions is to keep the balance of risk contributors of all types of reactors. This paper presents the method of new RCP seal leakage model and the sensitivity analysis results from applying the detailed method to PSA models of Westinghouse typed reference reactors. To perform the sensitivity analysis on LOCCW of the reference Westinghouse typed reactors, we reviewed WOG2000 RCP seal leakage model and developed the detailed event tree of LOCCW considering all scenarios of RCP seal failures. Also, we performed HRA based on the T/H analysis by using the leakage rates for each scenario. We could recognize that HRA was the sensitive contributor to CDF, and the RCP seal failure scenario of 182gpm leakage rate was estimated as the most important scenario

Sensitivity Analysis on LOCCW of Westinghouse typed Reactors Considering WOG2000 RCP Seal Leakage Model

Energy Technology Data Exchange (ETDEWEB)

Na, Jang-Hwan; Jeon, Ho-Jun; Hwang, Seok-Won [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2015-10-15

In this paper, we focus on risk insights of Westinghouse typed reactors. We identified that Reactor Coolant Pump (RCP) seal integrity is the most important contributor to Core Damage Frequency (CDF). As we reflected the latest technical report; WCAP-15603(Rev. 1-A), 'WOG2000 RCP Seal Leakage Model for Westinghouse PWRs' instead of the old version, RCP seal integrity became more important to Westinghouse typed reactors. After Fukushima accidents, Korea Hydro and Nuclear Power (KHNP) decided to develop Low Power and Shutdown (LPSD) Probabilistic Safety Assessment (PSA) models and upgrade full power PSA models of all operating Nuclear Power Plants (NPPs). As for upgrading full power PSA models, we have tried to standardize the methodology of CCF (Common Cause Failure) and HRA (Human Reliability Analysis), which are the most influential factors to risk measures of NPPs. Also, we have reviewed and reflected the latest operating experiences, reliability data sources and technical methods to improve the quality of PSA models. KHNP has operating various types of reactors; Optimized Pressurized Reactor (OPR) 1000, CANDU, Framatome and Westinghouse. So, one of the most challengeable missions is to keep the balance of risk contributors of all types of reactors. This paper presents the method of new RCP seal leakage model and the sensitivity analysis results from applying the detailed method to PSA models of Westinghouse typed reference reactors. To perform the sensitivity analysis on LOCCW of the reference Westinghouse typed reactors, we reviewed WOG2000 RCP seal leakage model and developed the detailed event tree of LOCCW considering all scenarios of RCP seal failures. Also, we performed HRA based on the T/H analysis by using the leakage rates for each scenario. We could recognize that HRA was the sensitive contributor to CDF, and the RCP seal failure scenario of 182gpm leakage rate was estimated as the most important scenario.

ATHEANA: open-quotes a technique for human error analysisclose quotes entering the implementation phase

International Nuclear Information System (INIS)

Taylor, J.; O'Hara, J.; Luckas, W.

1997-01-01

Probabilistic Risk Assessment (PRA) has become an increasingly important tool in the nuclear power industry, both for the Nuclear Regulatory Commission (NRC) and the operating utilities. The NRC recently published a final policy statement, SECY-95-126, encouraging the use of PRA in regulatory activities. Human reliability analysis (HRA), while a critical element of PRA, has limitations in the analysis of human actions in PRAs that have long been recognized as a constraint when using PRA. In fact, better integration of HRA into the PRA process has long been a NRC issue. Of particular concern, has been the omission of errors of commission - those errors that are associated with inappropriate interventions by operators with operating systems. To address these concerns, the NRC identified the need to develop an improved HRA method, so that human reliability can be better represented and integrated into PRA modeling and quantification. The purpose of the Brookhaven National Laboratory (BNL) project, entitled 'Improved HRA Method Based on Operating Experience' is to develop a new method for HRA which is supported by the analysis of risk-significant operating experience. This approach will allow a more realistic assessment and representation of the human contribution to plant risk, and thereby increase the utility of PRA. The project's completed, ongoing, and future efforts fall into four phases: (1) Assessment phase (FY 92/93); (2) Analysis and Characterization phase (FY 93/94); (3) Development phase (FY 95/96); and (4) Implementation phase (FY 96/97 ongoing)

Human Reliability Assessments: Using the Past (Shuttle) to Predict the Future (Orion)

Science.gov (United States)

DeMott, Diana L.; Bigler, Mark A.

2017-01-01

NASA (National Aeronautics and Space Administration) Johnson Space Center (JSC) Safety and Mission Assurance (S&MA) uses two human reliability analysis (HRA) methodologies. The first is a simplified method which is based on how much time is available to complete the action, with consideration included for environmental and personal factors that could influence the human's reliability. This method is expected to provide a conservative value or placeholder as a preliminary estimate. This preliminary estimate or screening value is used to determine which placeholder needs a more detailed assessment. The second methodology is used to develop a more detailed human reliability assessment on the performance of critical human actions. This assessment needs to consider more than the time available, this would include factors such as: the importance of the action, the context, environmental factors, potential human stresses, previous experience, training, physical design interfaces, available procedures/checklists and internal human stresses. The more detailed assessment is expected to be more realistic than that based primarily on time available. When performing an HRA on a system or process that has an operational history, we have information specific to the task based on this history and experience. In the case of a Probabilistic Risk Assessment (PRA) that is based on a new design and has no operational history, providing a "reasonable" assessment of potential crew actions becomes more challenging. To determine what is expected of future operational parameters, the experience from individuals who had relevant experience and were familiar with the system and process previously implemented by NASA was used to provide the "best" available data. Personnel from Flight Operations, Flight Directors, Launch Test Directors, Control Room Console Operators, and Astronauts were all interviewed to provide a comprehensive picture of previous NASA operations. Verification of the

Extracting HEPs from Event Reports of Domestic Nuclear Power Plants-Case Study

Energy Technology Data Exchange (ETDEWEB)

Park, Jinkyun; Kim, Yochan; Jung, Wondea [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

it is very important to soundly estimate the human error probability (HEP) of required tasks that could degrade the operational safety of systems. To this end, it is necessary to provide the HRA practitioners with the reliable catalog of HEPs. Unfortunately, one of the common issues raised by HRA practitioners is a lack of HRA data including HEPs. Therefore, many researchers are trying to provide reliable HRA data from diverse sources, such as (1) event reports reflecting the operational experience of domestic nuclear power plants (NPPs), and (2) human performance data observed from full- and/or partial-scope simulator exercises. It is evident that the contribution of human errors to the safety of socio-technical systems is very critical. For this reason, it is important for HRA practitioners to provide reliable HRA data including HEPs. Although a full-scope simulator can be used to collect valuable HRA data, it is still necessary to extract HRA data from the review of operational experience data. If so, it is possible to expect several benefits, such as the use of HRA data gathered from the operational experience of domestic NPPs as reference information to clarify the appropriateness of those collected from full-scope simulators.

Discussion of comments from a peer review of a technique for human event analysis (ATHEANA)

International Nuclear Information System (INIS)

Forester, J.A.; Ramey-Smith, A.; Bley, D.C.; Kolaczkowski, A.M.; Cooper, S.E.; Wreathall, J.

1998-01-01

In May of 1998, a technical basis and implementation guidelines document for A Technique for Human Event Analysis (ATHEANA) was issued as a draft report for public comment (NUREG-1624). In conjunction with the release of the draft NUREG, a paper review of the method, its documentation, and the results of an initial test of the method was held over a two-day period in Seattle, Washington, in June of 1998. Four internationally-known and respected experts in human reliability analysis (HRA) were selected to serve as the peer reviewers and were paid for their services. In addition, approximately 20 other individuals with an interest in HRA and ATHEANA also attended the peer review meeting and were invited to provide comments. The peer review team was asked to comment on any aspect of the method or the report in which improvements could be made and to discuss its strengths and weaknesses. All of the reviewers thought the ATEANA method had made significant contributions to the field of PRA/HRA, in particular by addressing the most important open questions and issues in HRA, by attempting to develop an integrated approach, and by developing a framework capable of identifying types of unsafe actions that generally have not been considered using existing methods. The reviewers had many concerns about specific aspects of the methodology and made many recommendations for ways to improve and extend the method, and to make its application more cost effective and useful to PRA in general. Details of the reviewers' comments and the ATHEANA team's responses to specific criticisms will be discussed

FACTORS INFLUENCING HUMAN RELIABILITY OF HIGH TEMPERATURE GAS COOLED REACTOR OPERATION

Directory of Open Access Journals (Sweden)

Sigit Santoso

2016-10-01

ABSTRAK Peran dan tindakan operator pada reaktor berpendingin gas akan berbeda dengan peran operator pada operasi tipe reaktor lain. Analisis unjuk kerja operator dan faktor yang berpengaruh dapat dilakukan secara komprehensif melalui analisis keandalan manusia(HRA. Melalui HRA dampak dari kesalahan manusia pada sistem maupun cara untuk mengurangi dampak dan frekuensi kesalahan dapat diketahui. Makalah membahas faktor yang berpengaruh pada tindakan operator, yaitu pada kejadian kecelakaan pendingin reaktor gas bersuhu tinggi-HTGR. Analisis untuk kualifikasi faktor pembentuk kinerja(PSF dilakukan berdasarkan kurva keandalan fungsi waktu, dan metode keandalan manusia yang dikembangkan berdasar pada aspek kognitif yaitu Cognitive Reliability and Error Analysis Method (CREAM. Hasil analisis berdasar kurva keandalan fungsi waktu menunjukkan komponen waktu berkontribusi positif pada peningkatan keandalan operator (PSF<1 pada kondisi semua fitur keselamatan berfungsi sesuai rancangan. Sedangkan pada metoda analisis dengan pendekatan kognitif CREAM diketahui selain faktor ketersediaan waktu, faktor pelatihan dan rancangan HMI juga berkontribusi meningkatkan keandalan operator. Faktor pembentuk kinerja keseluruhan diketahui sebesar 0,25 dengan faktor kontribusi positif dominan atau berpengaruh pada penurunan kesalahan manusia adalah ketersediaan waktu (PSF=0,01, dan faktor kontribusi negatif dominan adalah prosedur dan siklus kerja (PSF=5. Nilai PSF tersebut sebagai faktor pengali dalam perhitungan probabilitas kesalahan manusia. Analisis faktor pembentuk kinerja perlu dikembangkan pada skenario kejadian lain untuk selanjutnya digunakan untuk perhitungan dan analisis keandalan manusia yang komprehensif dan perancangan sistem interaksi manusia mesin di ruang kendali. Kata kunci: PSF, HTGR, operator, ruang kendali, keandalan manusia

A human reliability assessment screening method for the NRU upgrade project

International Nuclear Information System (INIS)

Bremner, F.M.; Alsop, C.J.

1997-01-01

The National Research Universal (NRU) reactor is a 130MW, low pressure, heavy water cooled and moderated research reactor. The reactor is used for research, both in support of Canada's CANDU development program, and for a wide variety of other research applications. In addition, NRU plays an important part in the production of medical isotopes, e.g., generating 80% of worldwide supplies of Molybdenum-99. NRU is owned and operated by Atomic Energy of Canada Ltd. (AECL), and is currently undergoing upgrading as part of AECL's continuing commitment to operate their facilities in a safe manner. As part of these upgrades both deterministic and probabilistic safety assessments are being carried out. It was recognized that the assignment of Human Error Probabilities (HEPs) is an important part of the Probabilistic Safety Assessment (PSA) studies, particularly for a facility whose design predates modern ergonomic practices, and which will undergo a series of backfitted modifications whilst continuing to operate. A simple Human Reliability Assessment (HRA) screening method, looking at both pre- and post-accident errors, was used in the initial safety studies. However, following review of this method within AECL and externally by the regulator, it was judged that benefits could be gained for future error reduction by including additional features, as later described in this document. The HRA development project consisted of several stages; needs analysis, literature review, development of method (including testing and evaluation), and implementation. This paper discusses each of these stages in further detail. (author)

Integrated sequence analysis. Final report

International Nuclear Information System (INIS)

Andersson, K.; Pyy, P.

1998-02-01

The NKS/RAK subprojet 3 'integrated sequence analysis' (ISA) was formulated with the overall objective to develop and to test integrated methodologies in order to evaluate event sequences with significant human action contribution. The term 'methodology' denotes not only technical tools but also methods for integration of different scientific disciplines. In this report, we first discuss the background of ISA and the surveys made to map methods in different application fields, such as man machine system simulation software, human reliability analysis (HRA) and expert judgement. Specific event sequences were, after the surveys, selected for application and testing of a number of ISA methods. The event sequences discussed in the report were cold overpressure of BWR, shutdown LOCA of BWR, steam generator tube rupture of a PWR and BWR disturbed signal view in the control room after an external event. Different teams analysed these sequences by using different ISA and HRA methods. Two kinds of results were obtained from the ISA project: sequence specific and more general findings. The sequence specific results are discussed together with each sequence description. The general lessons are discussed under a separate chapter by using comparisons of different case studies. These lessons include areas ranging from plant safety management (design, procedures, instrumentation, operations, maintenance and safety practices) to methodological findings (ISA methodology, PSA,HRA, physical analyses, behavioural analyses and uncertainty assessment). Finally follows a discussion about the project and conclusions are presented. An interdisciplinary study of complex phenomena is a natural way to produce valuable and innovative results. This project came up with structured ways to perform ISA and managed to apply the in practice. The project also highlighted some areas where more work is needed. In the HRA work, development is required for the use of simulators and expert judgement as

Integrated sequence analysis. Final report

Energy Technology Data Exchange (ETDEWEB)

Andersson, K.; Pyy, P

1998-02-01

The NKS/RAK subprojet 3 `integrated sequence analysis` (ISA) was formulated with the overall objective to develop and to test integrated methodologies in order to evaluate event sequences with significant human action contribution. The term `methodology` denotes not only technical tools but also methods for integration of different scientific disciplines. In this report, we first discuss the background of ISA and the surveys made to map methods in different application fields, such as man machine system simulation software, human reliability analysis (HRA) and expert judgement. Specific event sequences were, after the surveys, selected for application and testing of a number of ISA methods. The event sequences discussed in the report were cold overpressure of BWR, shutdown LOCA of BWR, steam generator tube rupture of a PWR and BWR disturbed signal view in the control room after an external event. Different teams analysed these sequences by using different ISA and HRA methods. Two kinds of results were obtained from the ISA project: sequence specific and more general findings. The sequence specific results are discussed together with each sequence description. The general lessons are discussed under a separate chapter by using comparisons of different case studies. These lessons include areas ranging from plant safety management (design, procedures, instrumentation, operations, maintenance and safety practices) to methodological findings (ISA methodology, PSA,HRA, physical analyses, behavioural analyses and uncertainty assessment). Finally follows a discussion about the project and conclusions are presented. An interdisciplinary study of complex phenomena is a natural way to produce valuable and innovative results. This project came up with structured ways to perform ISA and managed to apply the in practice. The project also highlighted some areas where more work is needed. In the HRA work, development is required for the use of simulators and expert judgement as

Hybrid instrument applied to human reliability study in event of loss of external electric power in a nuclear power plant

International Nuclear Information System (INIS)

Martins, Eduardo Ferraz

2015-01-01

The study projects in highly complex installations involves robust modeling, supported by conceptual and mathematical tools, to carry out systematic research and structured the different risk scenarios that can lead to unwanted events from occurring equipment failures or human errors. In the context of classical modeling, the Probabilistic Safety Analysis (PSA) seeks to provide qualitative and quantitative information about the project particularity and their operational facilities, including the identification of factors or scenarios that contribute to the risk and consequent comparison options for increasing safety. In this context, the aim of the thesis is to develop a hybrid instrument (CPP-HI) innovative, from the integrated modeling techniques of Failure Mode and Effect Analysis (FMEA), concepts of Human Reliability Analysis and Probabilistic Composition of Preferences (PCP). In support of modeling and validation of the CPP-HI, a simulation was performed on a triggering event 'Loss of External Electric Power' - PEEE, in a Nuclear Power plant. The results were simulated in a virtual environment (sensitivity analysis) and are robust to the study of Human Reliability Analysis (HRA) in the context of the PSA. (author)

A strategy to the development of a human error analysis method for accident management in nuclear power plants using industrial accident dynamics

International Nuclear Information System (INIS)

Lee, Yong Hee; Kim, Jae Whan; Jung, Won Dae; Ha, Jae Ju

1998-06-01

This technical report describes the early progress of he establishment of a human error analysis method as a part of a human reliability analysis(HRA) method for the assessment of the human error potential in a given accident management strategy. At first, we review the shortages and limitations of the existing HRA methods through an example application. In order to enhance the bias to the quantitative aspect of the HRA method, we focused to the qualitative aspect, i.e., human error analysis(HEA), during the proposition of a strategy to the new method. For the establishment of a new HEA method, we discuss the basic theories and approaches to the human error in industry, and propose three basic requirements that should be maintained as pre-requisites for HEA method in practice. Finally, we test IAD(Industrial Accident Dynamics) which has been widely utilized in industrial fields, in order to know whether IAD can be so easily modified and extended to the nuclear power plant applications. We try to apply IAD to the same example case and develop new taxonomy of the performance shaping factors in accident management and their influence matrix, which could enhance the IAD method as an HEA method. (author). 33 refs., 17 tabs., 20 figs

The importance of human cognitive models in the safety analysis report of nuclear power plants - a comparative review

International Nuclear Information System (INIS)

Alvarenga, Marco A.B.; Araujo Goes, Alexandre G. de

1997-01-01

The chapter 18 of the Brazilian NPPs Safety Analysis Report (SAR) deals with Human Factor Engineering (HFE). The chapter evaluation is distributed among ten topics. One of them, the HRA (Human Reliability Analysis) becomes the central subject of the whole analysis, generating information to the other topics, as for example, high risk operational critical sequences. The HRA methods used in the past concerned the approach of modeling the human being as a component (hardware), based in a failure or success bivalent logic. In the last ten years, several human cognitive models were developed to be used in the nuclear field as well as in the conventional industry, mainly in the military aviation. In this paper, we describe their main features, comparing some models to each other, with the main purpose of determining the minimal characteristics acceptable for NPPs licensing, being part of these cognitive models, to be used mainly in the evaluation of HRAs from SARs in the NPPs. (author). 10 refs

Human reliability-based MC and A models for detecting insider theft

International Nuclear Information System (INIS)

Duran, Felicia Angelica; Wyss, Gregory Dane

2010-01-01

Material control and accounting (MC and A) safeguards operations that track and account for critical assets at nuclear facilities provide a key protection approach for defeating insider adversaries. These activities, however, have been difficult to characterize in ways that are compatible with the probabilistic path analysis methods that are used to systematically evaluate the effectiveness of a site's physical protection (security) system (PPS). MC and A activities have many similar characteristics to operator procedures performed in a nuclear power plant (NPP) to check for anomalous conditions. This work applies human reliability analysis (HRA) methods and models for human performance of NPP operations to develop detection probabilities for MC and A activities. This has enabled the development of an extended probabilistic path analysis methodology in which MC and A protections can be combined with traditional sensor data in the calculation of PPS effectiveness. The extended path analysis methodology provides an integrated evaluation of a safeguards and security system that addresses its effectiveness for attacks by both outside and inside adversaries.

Multidisciplinary System Reliability Analysis

Science.gov (United States)

Mahadevan, Sankaran; Han, Song; Chamis, Christos C. (Technical Monitor)

2001-01-01

The objective of this study is to develop a new methodology for estimating the reliability of engineering systems that encompass multiple disciplines. The methodology is formulated in the context of the NESSUS probabilistic structural analysis code, developed under the leadership of NASA Glenn Research Center. The NESSUS code has been successfully applied to the reliability estimation of a variety of structural engineering systems. This study examines whether the features of NESSUS could be used to investigate the reliability of systems in other disciplines such as heat transfer, fluid mechanics, electrical circuits etc., without considerable programming effort specific to each discipline. In this study, the mechanical equivalence between system behavior models in different disciplines are investigated to achieve this objective. A new methodology is presented for the analysis of heat transfer, fluid flow, and electrical circuit problems using the structural analysis routines within NESSUS, by utilizing the equivalence between the computational quantities in different disciplines. This technique is integrated with the fast probability integration and system reliability techniques within the NESSUS code, to successfully compute the system reliability of multidisciplinary systems. Traditional as well as progressive failure analysis methods for system reliability estimation are demonstrated, through a numerical example of a heat exchanger system involving failure modes in structural, heat transfer and fluid flow disciplines.

Specification of a Human Reliability Data Bank for conducting HRA segments of PRAs for nuclear power plants

International Nuclear Information System (INIS)

Comer, M.K.; Donovan, M.D.

1985-02-01

The US Nuclear Regulatory Commission (NRC), Sandia National Laboratories (SNL), and General Physics Corporation have conducted a research program to develop a Human Reliability Data Bank for nuclear power industry probabilistic risk assessment (PRA). As part of this program, a survey of existing human reliability data banks from other industries was conducted and a concept of a Data Bank for the nuclear industry was developed. The results of these efforts were published in the two volumes of NUREG/CR-2744: ''Human Reliability Data Bank for Nuclear Power Plant Operations: Volume 1, A Review of Existing Human Reliability Data Banks, and Volume 2, A Data Bank Concept and System Description.'' This document, NUREG/CR-4010, is the revised technical specification for the Human Reliability Data Bank. The organization of the Data Bank and a description of a data publication, the Human Reliability Data Manual, are provided. Details of the administration and operation of the Data Bank are discussed. Appendices present the detailed procedures for processing data, revising the Data Manual, operating the Data Bank, and reviewing data for the Data Bank. The final appendix is a skeleton version (structure only) of the Data Manual

Development of an integrated system for estimating human error probabilities

Energy Technology Data Exchange (ETDEWEB)

Auflick, J.L.; Hahn, H.A.; Morzinski, J.A.

1998-12-01

This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). This project had as its main objective the development of a Human Reliability Analysis (HRA), knowledge-based expert system that would provide probabilistic estimates for potential human errors within various risk assessments, safety analysis reports, and hazard assessments. HRA identifies where human errors are most likely, estimates the error rate for individual tasks, and highlights the most beneficial areas for system improvements. This project accomplished three major tasks. First, several prominent HRA techniques and associated databases were collected and translated into an electronic format. Next, the project started a knowledge engineering phase where the expertise, i.e., the procedural rules and data, were extracted from those techniques and compiled into various modules. Finally, these modules, rules, and data were combined into a nearly complete HRA expert system.

The use of the SACADA taxonomy to analyze simulation records: Insights and suggestions

International Nuclear Information System (INIS)

Park, J.; Chang, Y.J.; Kim, Y.; Choi, S.; Kim, S.; Jung, W.

2017-01-01

It is evident that diverse human reliability analysis (HRA) methods are effective for enhancing the safety of socio-technical systems through identifying the most vulnerable tasks to human errors with the associated human error probabilities. This means that reliable human performance data is an important factor affecting HRA quality. Therefore, many researchers have developed technical underpinnings (such as guidelines and taxonomies) that specify what and how HRA data can be collected from simulator experiments. Here, SACADA (Scenario Authoring, Characterization, and Debriefing Application) taxonomy recently developed by US NRC (Nuclear Regulatory Commission) is worth emphasizing, because it is constructed on the basis of a cognitive model (i.e., a top-down approach) while most of the technical underpinnings are developed by a bottom-up approach (i.e., the intensive review of existing literature). For this reason, in this study, the SACADA taxonomy is used to analyze several audio-visual records collected from the full scope simulators of nuclear power plants in the Republic of Korea. The results indicate that the SACADA taxonomy is useful to collect operator performance data in simulator training for HRA. Certain human performance information that can be provided by SACADA data provided are difficult to be covered by the bottom-up approach. - Highlights: • HRA is important for enhancing the safety of socio-technical systems. • HRA quality is largely dependent on the availability of HRA data. • One of the HRA data sources is to use a full-scope simulator. • Data contents to be collected from the full-scope simulator should be clarified. • SACADA is helpful for clarifying data contents from the cognitive perspective.

The Development of Marine Accidents Human Reliability Assessment Approach: HEART Methodology and MOP Model

Directory of Open Access Journals (Sweden)

Ludfi Pratiwi Bowo

2017-06-01

Full Text Available Humans are one of the important factors in the assessment of accidents, particularly marine accidents. Hence, studies are conducted to assess the contribution of human factors in accidents. There are two generations of Human Reliability Assessment (HRA that have been developed. Those methodologies are classified by the differences of viewpoints of problem-solving, as the first generation and second generation. The accident analysis can be determined using three techniques of analysis; sequential techniques, epidemiological techniques and systemic techniques, where the marine accidents are included in the epidemiological technique. This study compares the Human Error Assessment and Reduction Technique (HEART methodology and the 4M Overturned Pyramid (MOP model, which are applied to assess marine accidents. Furthermore, the MOP model can effectively describe the relationships of other factors which affect the accidents; whereas, the HEART methodology is only focused on human factors.

RELIABILITY ANALYSIS OF BENDING ELIABILITY ANALYSIS OF ...

African Journals Online (AJOL)

eobe

Reliability analysis of the safety levels of the criteria slabs, have been .... was also noted [2] that if the risk level or β < 3.1), the ... reliability analysis. A study [6] has shown that all geometric variables, ..... Germany, 1988. 12. Hasofer, A. M and ...

Human reliability-based MC and A methods for evaluating the effectiveness of protecting nuclear material - 59379

International Nuclear Information System (INIS)

Duran, Felicia A.; Wyss, Gregory D.

2012-01-01

Material control and accountability (MC and A) operations that track and account for critical assets at nuclear facilities provide a key protection approach for defeating insider adversaries. MC and A activities, from monitoring to inventory measurements, provide critical information about target materials and define security elements that are useful against insider threats. However, these activities have been difficult to characterize in ways that are compatible with the path analysis methods that are used to systematically evaluate the effectiveness of a site's protection system. The path analysis methodology focuses on a systematic, quantitative evaluation of the physical protection component of the system for potential external threats, and often calculates the probability that the physical protection system (PPS) is effective (PE) in defeating an adversary who uses that attack pathway. In previous work, Dawson and Hester observed that many MC and A activities can be considered a type of sensor system with alarm and assessment capabilities that provide recurring opportunities for 'detecting' the status of critical items. This work has extended that characterization of MC and A activities as probabilistic sensors that are interwoven within each protection layer of the PPS. In addition, MC and A activities have similar characteristics to operator tasks performed in a nuclear power plant (NPP) in that the reliability of these activities depends significantly on human performance. Many of the procedures involve human performance in checking for anomalous conditions. Further characterization of MC and A activities as operational procedures that check the status of critical assets provides a basis for applying human reliability analysis (HRA) models and methods to determine probabilities of detection for MC and A protection elements. This paper will discuss the application of HRA methods used in nuclear power plant probabilistic risk assessments to define detection

Reliability analysis techniques in power plant design

International Nuclear Information System (INIS)

Chang, N.E.

1981-01-01

An overview of reliability analysis techniques is presented as applied to power plant design. The key terms, power plant performance, reliability, availability and maintainability are defined. Reliability modeling, methods of analysis and component reliability data are briefly reviewed. Application of reliability analysis techniques from a design engineering approach to improving power plant productivity is discussed. (author)

Reliability analysis of shutdown system

International Nuclear Information System (INIS)

Kumar, C. Senthil; John Arul, A.; Pal Singh, Om; Suryaprakasa Rao, K.

2005-01-01

This paper presents the results of reliability analysis of Shutdown System (SDS) of Indian Prototype Fast Breeder Reactor. Reliability analysis carried out using Fault Tree Analysis predicts a value of 3.5 x 10 -8 /de for failure of shutdown function in case of global faults and 4.4 x 10 -8 /de for local faults. Based on 20 de/y, the frequency of shutdown function failure is 0.7 x 10 -6 /ry, which meets the reliability target, set by the Indian Atomic Energy Regulatory Board. The reliability is limited by Common Cause Failure (CCF) of actuation part of SDS and to a lesser extent CCF of electronic components. The failure frequency of individual systems is -3 /ry, which also meets the safety criteria. Uncertainty analysis indicates a maximum error factor of 5 for the top event unavailability

A classification scheme of erroneous behaviors for human error probability estimations based on simulator data

International Nuclear Information System (INIS)

Kim, Yochan; Park, Jinkyun; Jung, Wondea

2017-01-01

Because it has been indicated that empirical data supporting the estimates used in human reliability analysis (HRA) is insufficient, several databases have been constructed recently. To generate quantitative estimates from human reliability data, it is important to appropriately sort the erroneous behaviors found in the reliability data. Therefore, this paper proposes a scheme to classify the erroneous behaviors identified by the HuREX (Human Reliability data Extraction) framework through a review of the relevant literature. A case study of the human error probability (HEP) calculations is conducted to verify that the proposed scheme can be successfully implemented for the categorization of the erroneous behaviors and to assess whether the scheme is useful for the HEP quantification purposes. Although continuously accumulating and analyzing simulator data is desirable to secure more reliable HEPs, the resulting HEPs were insightful in several important ways with regard to human reliability in off-normal conditions. From the findings of the literature review and the case study, the potential and limitations of the proposed method are discussed. - Highlights: • A taxonomy of erroneous behaviors is proposed to estimate HEPs from a database. • The cognitive models, procedures, HRA methods, and HRA databases were reviewed. • HEPs for several types of erroneous behaviors are calculated as a case study.

'Violation' - does HRA need the concept?

International Nuclear Information System (INIS)

Dougherty, Ed

1995-01-01

Violations are part of a complex matrix of judgmental behavior. The term violation is an indicator of the possibility of recent judgment errors but more so of latent errors in management and/or design. The effect of investigations in this arena do indicate a need for an extension of the classic slip/mistake taxonomy of human reliability analysis. The note attempts to initiate this extension

Integrating reliability analysis and design

International Nuclear Information System (INIS)

Rasmuson, D.M.

1980-10-01

This report describes the Interactive Reliability Analysis Project and demonstrates the advantages of using computer-aided design systems (CADS) in reliability analysis. Common cause failure problems require presentations of systems, analysis of fault trees, and evaluation of solutions to these. Results have to be communicated between the reliability analyst and the system designer. Using a computer-aided design system saves time and money in the analysis of design. Computer-aided design systems lend themselves to cable routing, valve and switch lists, pipe routing, and other component studies. At EG and G Idaho, Inc., the Applicon CADS is being applied to the study of water reactor safety systems

Analysis of Operators Comments on the PSF Questionnaire of the Task Complexity Experiment 2003/2004

Energy Technology Data Exchange (ETDEWEB)

Torralba, B.; Martinez-Arias, R.

2007-07-01

Human Reliability Analysis (HRA) methods usually take into account the effect of Performance Shaping Factors (PSF). Therefore, the adequate treatment of PSFs in HRA of Probabilistic Safety Assessment (PSA) models has a crucial importance. There is an important need for collecting PSF data based on simulator experiments. During the task complexity experiment 2003-2004, carried out in the BWR simulator of Halden Man-Machine Laboratory (HAMMLAB), there was a data collection on PSF by means of a PSF Questionnaire. Seven crews (composed of shift supervisor, reactor operator and turbine operator) from Swedish Nuclear Power Plants participated in the experiment. The PSF Questionnaire collected data on the factors: procedures, training and experience, indications, controls, team management, team communication, individual work practice, available time for the tasks, number of tasks or information load, masking and seriousness. The main statistical significant results are presented on Performance Shaping Factors data collection and analysis of the task complexity experiment 2003/2004 (HWR-810). The analysis of the comments about PSFs, which were provided by operators on the PSF Questionnaire, is described. It has been summarised the comments provided for each PSF on the scenarios, using a content analysis technique. (Author)

Analysis of Operators Comments on the PSF Questionnaire of the Task Complexity Experiment 2003/2004

International Nuclear Information System (INIS)

Torralba, B.; Martinez-Arias, R.

2007-01-01

Human Reliability Analysis (HRA) methods usually take into account the effect of Performance Shaping Factors (PSF). Therefore, the adequate treatment of PSFs in HRA of Probabilistic Safety Assessment (PSA) models has a crucial importance. There is an important need for collecting PSF data based on simulator experiments. During the task complexity experiment 2003-2004, carried out in the BWR simulator of Halden Man-Machine Laboratory (HAMMLAB), there was a data collection on PSF by means of a PSF Questionnaire. Seven crews (composed of shift supervisor, reactor operator and turbine operator) from Swedish Nuclear Power Plants participated in the experiment. The PSF Questionnaire collected data on the factors: procedures, training and experience, indications, controls, team management, team communication, individual work practice, available time for the tasks, number of tasks or information load, masking and seriousness. The main statistical significant results are presented on Performance Shaping Factors data collection and analysis of the task complexity experiment 2003/2004 (HWR-810). The analysis of the comments about PSFs, which were provided by operators on the PSF Questionnaire, is described. It has been summarised the comments provided for each PSF on the scenarios, using a content analysis technique. (Author)

Multi-Disciplinary System Reliability Analysis

Science.gov (United States)

Mahadevan, Sankaran; Han, Song

1997-01-01

The objective of this study is to develop a new methodology for estimating the reliability of engineering systems that encompass multiple disciplines. The methodology is formulated in the context of the NESSUS probabilistic structural analysis code developed under the leadership of NASA Lewis Research Center. The NESSUS code has been successfully applied to the reliability estimation of a variety of structural engineering systems. This study examines whether the features of NESSUS could be used to investigate the reliability of systems in other disciplines such as heat transfer, fluid mechanics, electrical circuits etc., without considerable programming effort specific to each discipline. In this study, the mechanical equivalence between system behavior models in different disciplines are investigated to achieve this objective. A new methodology is presented for the analysis of heat transfer, fluid flow, and electrical circuit problems using the structural analysis routines within NESSUS, by utilizing the equivalence between the computational quantities in different disciplines. This technique is integrated with the fast probability integration and system reliability techniques within the NESSUS code, to successfully compute the system reliability of multi-disciplinary systems. Traditional as well as progressive failure analysis methods for system reliability estimation are demonstrated, through a numerical example of a heat exchanger system involving failure modes in structural, heat transfer and fluid flow disciplines.

Fundamentals and applications of systems reliability analysis

International Nuclear Information System (INIS)

Boesebeck, K.; Heuser, F.W.; Kotthoff, K.

1976-01-01

The lecture gives a survey on the application of methods of reliability analysis to assess the safety of nuclear power plants. Possible statements of reliability analysis in connection with specifications of the atomic licensing procedure are especially dealt with. Existing specifications of safety criteria are additionally discussed with the help of reliability analysis by the example of the reliability analysis of a reactor protection system. Beyond the limited application to single safety systems, the significance of reliability analysis for a closed risk concept is explained in the last part of the lecture. (orig./LH) [de

Performance Shaping Factors Assessments and Application to PHWR Outages

Energy Technology Data Exchange (ETDEWEB)

Lee, Seung Woo

2007-02-15

Human reliability analysis is definitely related to the quality of PSA because human errors have been identified as major contributors to PSA. According to NRC's 'Office of analysis and evaluation of operational data (AEOD)',82% of the reactor trips and accident during outage is caused by the events related to human errors. There is, however, no one HRA method universally accepted. Furthermore, HRA during PHWR outages has not been performed around the world yet. HRA during PHWR outages is especially important since manual management of operator is more required during PHWR. In this study, accident scenarios which HYU developed are used to perform a quantification of human error probability. In this study, overall procedures of standard HRA methodology are introduced and follows the quantification of 10 possible selected human actions during PHWR outages based on standard HRA methodology. To see the verification, quantified values were compared with the values from 'Generic CANDU Probabilistic Safety Assessment' and the values estimated by ASEP.Core Damage Frequency was estimated 3.35 x 10{sup -4} more higher than CDF estimated by AECL data. It was considered that the differences between the HEPs for OPAFW and OPECC3 make CDF higher. Therefore, complementary study of reestimating HEP for OPAFW and OPECC3 in detail is required for increasing the qualities of HRA and PSA. Moreover, one of the difficulties in performing human reliability analysis is to evaluate performance shaping factors which represent the characteristics and circumstances. For assessing a specific human action more exactly, it is necessary to consider all of the PSFs at the same time which makes an effect on the human action. Also, it requires the effect comparison among PSFs to minimize the uncertainties which are usually caused by the subjective judgements of HRA analysts. To see the sensitivity, performance shaping factors of each decision rule are changed which resulted

Performance Shaping Factors Assessments and Application to PHWR Outages

International Nuclear Information System (INIS)

Lee, Seung Woo

2007-02-01

Human reliability analysis is definitely related to the quality of PSA because human errors have been identified as major contributors to PSA. According to NRC's 'Office of analysis and evaluation of operational data (AEOD)',82% of the reactor trips and accident during outage is caused by the events related to human errors. There is, however, no one HRA method universally accepted. Furthermore, HRA during PHWR outages has not been performed around the world yet. HRA during PHWR outages is especially important since manual management of operator is more required during PHWR. In this study, accident scenarios which HYU developed are used to perform a quantification of human error probability. In this study, overall procedures of standard HRA methodology are introduced and follows the quantification of 10 possible selected human actions during PHWR outages based on standard HRA methodology. To see the verification, quantified values were compared with the values from 'Generic CANDU Probabilistic Safety Assessment' and the values estimated by ASEP.Core Damage Frequency was estimated 3.35 x 10 -4 more higher than CDF estimated by AECL data. It was considered that the differences between the HEPs for OPAFW and OPECC3 make CDF higher. Therefore, complementary study of reestimating HEP for OPAFW and OPECC3 in detail is required for increasing the qualities of HRA and PSA. Moreover, one of the difficulties in performing human reliability analysis is to evaluate performance shaping factors which represent the characteristics and circumstances. For assessing a specific human action more exactly, it is necessary to consider all of the PSFs at the same time which makes an effect on the human action. Also, it requires the effect comparison among PSFs to minimize the uncertainties which are usually caused by the subjective judgements of HRA analysts. To see the sensitivity, performance shaping factors of each decision rule are changed which resulted in changes of core damage

Risk informed analysis of training effectiveness for mitigating accidents of nuclear power plants

International Nuclear Information System (INIS)

Lee, Chang Ju

2012-01-01

A critical area for deriving expected benefits from training and exercise is the measurement of 'training effectiveness'-how well the training inputs are serving the intended purpose. This aspect is often neglected by nuclear organizations, saying that measurement is difficult. However, I believe that a technique in nuclear society has developed sufficiently to measure most important aspects of training by way of human reliability analysis (HRA) used in probabilistic safety assessment (PSA) of nuclear power plants (NPPs). The consequences of errors caused by lack of training can be evaluated in terms of the overall vulnerability to human error of the facility under consideration. This study presents current situation and considerations for measures of robustness on nuclear accidents and HRA technique on the training effectiveness. In view of risk informed approach with this consideration and an example case, I'd like to identify appropriate relationship between risk measures of robustness and training effectiveness

Reliability analysis of software based safety functions

International Nuclear Information System (INIS)

Pulkkinen, U.

1993-05-01

The methods applicable in the reliability analysis of software based safety functions are described in the report. Although the safety functions also include other components, the main emphasis in the report is on the reliability analysis of software. The check list type qualitative reliability analysis methods, such as failure mode and effects analysis (FMEA), are described, as well as the software fault tree analysis. The safety analysis based on the Petri nets is discussed. The most essential concepts and models of quantitative software reliability analysis are described. The most common software metrics and their combined use with software reliability models are discussed. The application of software reliability models in PSA is evaluated; it is observed that the recent software reliability models do not produce the estimates needed in PSA directly. As a result from the study some recommendations and conclusions are drawn. The need of formal methods in the analysis and development of software based systems, the applicability of qualitative reliability engineering methods in connection to PSA and the need to make more precise the requirements for software based systems and their analyses in the regulatory guides should be mentioned. (orig.). (46 refs., 13 figs., 1 tab.)

Critical operator actions: human reliability modeling and data issues. Principal Working Group No. 5 - Task 94-1. Final Task Report prepared by a Group of Experts of the NEA Committee on the Safety of Nuclear Installations

International Nuclear Information System (INIS)

Wilmart, P.; Grant, A.; Raina, V.M.; Patrik, M.; Cacciabue, P.C.; Cojazzi, G.; Reiman, L.; Virolainen, R.; Lanore, J.M.; Poidevin, S.; Herttrich, P.M.; Mertens, J.; Reer, B.; Straeter, O.; Bareith, A.; Hollo, E.; Traini, E.; Fukuda, M.; Hirano, M.; Kani, Y.; Muramatsu, K.; Versteeg, M.F.; Kim, T.W.; Calvo, J.; Gil, B.; Dang, V.N.; Hirschberg, S.; Meyer, P.; Schmocker, U.; Andrews, R.; Coxson, B.; Shepherd, C.H.; Murphy, J.A.; Parry, G.W.; Ramey-Smith, A.; Siu, N.O.

1998-01-01

The treatment of human interactions is considered one of the major limitations in the context of Probabilistic Safety Assessment (PSA).While the results of many PSAs show a very significant contribution of human errors, large uncertainties are normally associated with the quantitative estimates of these contributors. This problem becomes even more significant when analysing human interactions under special conditions, for example in accident scenarios for external events or for the shutdown and low power conditions. Any improvement in the current state of knowledge with respect to the data for human interactions would have a positive impact on the confidence in PSA results, including correct ranking of the dominant accident scenarios. At the same time many PSAs have been successful at identifying critical operator actions; in most cases the benefits of these qualitative insights are not jeopardised by lack of numerical precision in the estimates. The present HRA approaches as generally applied in PSAs are also limited in scope; for instance, they either ignore errors of commissions or treat these superficially. New, dynamic methods, primarily aiming at the resolution of the issues of cognitive errors including errors of commission are emerging but their full-scope applications within the PSA framework belong to the future. In the context of data, some progress has been observed partially due to use of simulators to support the human reliability analysis (HRA). These applications have been rather concentrated (but not limited) to France and USA. Recently, a very promising program has been established in Hungary. The experiences from such applications are not widely known and dissemination of the relevant insights to the PSA community has some definite merits. With respect to the identification of critical operator actions there is in some cases clear evidence and in others a good potential that the existing PSA studies may provide useful, partially generic

Philosophy of ATHEANA

International Nuclear Information System (INIS)

Bley, D.C.; Cooper, S.E.; Forester, J.A.; Kolaczkowski, A.M.; Ramey-Smith, A.; Thompson, C.M.; Whitehead, D.W.; Wreathall, J.

1999-01-01

ATHEANA, a second-generation Human Reliability Analysis (HRA) method integrates advances in psychology with engineering, human factors, and Probabilistic Risk Analysis (PRA) disciplines to provide an HRA quantification process and PRA modeling interface that can accommodate and represent human performance in real nuclear power plant events. The method uses the characteristics of serious accidents identified through retrospective analysis of serious operational events to set priorities in a search process for significant human failure events, unsafe acts, and error-forcing context (unfavorable plant conditions combined with negative performance-shaping factors). ATHEANA has been tested in a demonstration project at an operating pressurized water reactor

A framework to assess diagnosis error probabilities in the advanced MCR

Energy Technology Data Exchange (ETDEWEB)

Kim, Ar Ryum; Seong, Poong Hyun [KAIST, Daejeon (Korea, Republic of); Kim, Jong Hyun [Chosun University, Gwangju (Korea, Republic of); Jang, Inseok; Park, Jinkyun [Korea Atomic Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The Institute of Nuclear Power Operations (INPO)’s operating experience database revealed that about 48% of the total events in world NPPs for 2 years (2010-2011) happened due to human errors. The purposes of human reliability analysis (HRA) method are to evaluate the potential for, and mechanism of, human errors that may affect plant safety. Accordingly, various HRA methods have been developed such as technique for human error rate prediction (THERP), simplified plant analysis risk human reliability assessment (SPAR-H), cognitive reliability and error analysis method (CREAM) and so on. Many researchers have asserted that procedure, alarm, and display are critical factors to affect operators’ generic activities, especially for diagnosis activities. None of various HRA methods was explicitly designed to deal with digital systems. SCHEME (Soft Control Human error Evaluation MEthod) considers only for the probability of soft control execution error in the advanced MCR. The necessity of developing HRA methods in various conditions of NPPs has been raised. In this research, the framework to estimate diagnosis error probabilities in the advanced MCR was suggested. The assessment framework was suggested by three steps. The first step is to investigate diagnosis errors and calculate their probabilities. The second step is to quantitatively estimate PSFs’ weightings in the advanced MCR. The third step is to suggest the updated TRC model to assess the nominal diagnosis error probabilities. Additionally, the proposed framework was applied by using the full-scope simulation. Experiments conducted in domestic full-scope simulator and HAMMLAB were used as data-source. Total eighteen tasks were analyzed and twenty-three crews participated in.

Worker doses and potential health effects resulting from the accident at the Fukushima nuclear power plant in 2011.

Science.gov (United States)

Etherington, George; Zhang, Wei; Harrison, John; Walsh, Linda

2014-11-01

The reliability of exposure scenarios used in the World Health Organization's Health Risk Assessment (HRA) for Fukushima workers was examined. HRA risk estimates for cancer incidence in these workers were then reviewed. The HRA constructed four exposure scenarios to estimate worker radiation doses; recent individual dosimetric data were used to assess their reliability. Risks of specified cancer types attributable to radiation exposure were estimated in the HRA by calculating Lifetime Attributable Risks (LAR) for each scenario, and comparisons of LAR are provided. The scenarios were confirmed as reliable for the purposes of the HRA. For the lowest dose scenario (Scenario 1: 69% of the workforce), the HRA found that any elevated cancer risk was insignificant. Significantly elevated cancer risks were, however, found for the three higher dose scenarios. For the highest dose scenario (Scenario 4: 13 workers), LAR values for thyroid cancer up to 3.5% were estimated, but a radiation-related increase in thyroid cancer incidence is unlikely to be observed because of the small number of workers. For the two intermediate dose scenarios, a small number of cancer cases may occur, but these are unlikely to be observed because the variability in baseline rates of cancer incidence is much larger than the predicted radiation-related incidence rates.

Power system reliability analysis using fault trees

International Nuclear Information System (INIS)

Volkanovski, A.; Cepin, M.; Mavko, B.

2006-01-01

The power system reliability analysis method is developed from the aspect of reliable delivery of electrical energy to customers. The method is developed based on the fault tree analysis, which is widely applied in the Probabilistic Safety Assessment (PSA). The method is adapted for the power system reliability analysis. The method is developed in a way that only the basic reliability parameters of the analysed power system are necessary as an input for the calculation of reliability indices of the system. The modeling and analysis was performed on an example power system consisting of eight substations. The results include the level of reliability of current power system configuration, the combinations of component failures resulting in a failed power delivery to loads, and the importance factors for components and subsystems. (author)

Assessment of the human factor in the quantification of technical system reliability taking into consideration cognitive-causal aspects. Partial project 2. Modeling of the human behavior for reliability considerations. Final report

International Nuclear Information System (INIS)

Jennerich, Marco; Imbsweiler, Jonas; Straeter, Oliver; Arenius, Marcus

2015-03-01

This report presents the findings of the project for the consideration of human factor in the quantification of the reliability of technical systems, taking into account cognitive-causal aspects concerning the modeling of human behavior of reliability issues (funded by the Federal Ministry of Economics and Technology; grant number 15014328). This project is part of a joint project with the University of Applied Sciences Zittau / Goerlitz for assessing the human factor in the quantification of the reliability of technical systems. The concern of the University of Applied Sciences Zittau / Goerlitz is the mathematical modeling of human reliability by means of a fuzzy set approach (grant number 1501432A). The part of the project presented here provides the necessary data basis for the evaluation of the mathematical modeling using fuzzy set approach. At the appropriate places in this report, the interfaces and data bases between the two projects are outlined accordingly. HRA-methods (Human Reliability Analysis) are an essential component to analyze the reliability of socio-technical systems. Various methods have been established and are used in different areas of application. The established HRA methods have been checked on their congruence. In particular the underlying models and their parameters such as performance-influencing factors and situational influences have been investigated. The elaborated parameters were combined into a hierarchical class structure. Cross-domain incidents were studied. The specific performance-influencing factors have been worked out and have been integrated into a cross-domain database. The dominant (critical) situational factors and their interactions within the event data were identified using the CAHR method (connectionism Assessment of Human Reliability). Task dependent cognitive load profiles have been defined. Within these profiles qualitative and quantitative data of the possibility of emergence of errors have been acquired. This

Reliability analysis under epistemic uncertainty

International Nuclear Information System (INIS)

Nannapaneni, Saideep; Mahadevan, Sankaran

2016-01-01

This paper proposes a probabilistic framework to include both aleatory and epistemic uncertainty within model-based reliability estimation of engineering systems for individual limit states. Epistemic uncertainty is considered due to both data and model sources. Sparse point and/or interval data regarding the input random variables leads to uncertainty regarding their distribution types, distribution parameters, and correlations; this statistical uncertainty is included in the reliability analysis through a combination of likelihood-based representation, Bayesian hypothesis testing, and Bayesian model averaging techniques. Model errors, which include numerical solution errors and model form errors, are quantified through Gaussian process models and included in the reliability analysis. The probability integral transform is used to develop an auxiliary variable approach that facilitates a single-level representation of both aleatory and epistemic uncertainty. This strategy results in an efficient single-loop implementation of Monte Carlo simulation (MCS) and FORM/SORM techniques for reliability estimation under both aleatory and epistemic uncertainty. Two engineering examples are used to demonstrate the proposed methodology. - Highlights: • Epistemic uncertainty due to data and model included in reliability analysis. • A novel FORM-based approach proposed to include aleatory and epistemic uncertainty. • A single-loop Monte Carlo approach proposed to include both types of uncertainties. • Two engineering examples used for illustration.

Reliability analysis techniques for the design engineer

International Nuclear Information System (INIS)

Corran, E.R.; Witt, H.H.

1982-01-01

This paper describes a fault tree analysis package that eliminates most of the housekeeping tasks involved in proceeding from the initial construction of a fault tree to the final stage of presenting a reliability analysis in a safety report. It is suitable for designers with relatively little training in reliability analysis and computer operation. Users can rapidly investigate the reliability implications of various options at the design stage and evolve a system which meets specified reliability objectives. Later independent review is thus unlikely to reveal major shortcomings necessitating modification and project delays. The package operates interactively, allowing the user to concentrate on the creative task of developing the system fault tree, which may be modified and displayed graphically. For preliminary analysis, system data can be derived automatically from a generic data bank. As the analysis proceeds, improved estimates of critical failure rates and test and maintenance schedules can be inserted. The technique is applied to the reliability analysis of the recently upgraded HIFAR Containment Isolation System. (author)

A reliability analysis tool for SpaceWire network

Science.gov (United States)

Zhou, Qiang; Zhu, Longjiang; Fei, Haidong; Wang, Xingyou

2017-04-01

A SpaceWire is a standard for on-board satellite networks as the basis for future data-handling architectures. It is becoming more and more popular in space applications due to its technical advantages, including reliability, low power and fault protection, etc. High reliability is the vital issue for spacecraft. Therefore, it is very important to analyze and improve the reliability performance of the SpaceWire network. This paper deals with the problem of reliability modeling and analysis with SpaceWire network. According to the function division of distributed network, a reliability analysis method based on a task is proposed, the reliability analysis of every task can lead to the system reliability matrix, the reliability result of the network system can be deduced by integrating these entire reliability indexes in the matrix. With the method, we develop a reliability analysis tool for SpaceWire Network based on VC, where the computation schemes for reliability matrix and the multi-path-task reliability are also implemented. By using this tool, we analyze several cases on typical architectures. And the analytic results indicate that redundancy architecture has better reliability performance than basic one. In practical, the dual redundancy scheme has been adopted for some key unit, to improve the reliability index of the system or task. Finally, this reliability analysis tool will has a directive influence on both task division and topology selection in the phase of SpaceWire network system design.

Analysis of information security reliability: A tutorial

International Nuclear Information System (INIS)

Kondakci, Suleyman

2015-01-01

This article presents a concise reliability analysis of network security abstracted from stochastic modeling, reliability, and queuing theories. Network security analysis is composed of threats, their impacts, and recovery of the failed systems. A unique framework with a collection of the key reliability models is presented here to guide the determination of the system reliability based on the strength of malicious acts and performance of the recovery processes. A unique model, called Attack-obstacle model, is also proposed here for analyzing systems with immunity growth features. Most computer science curricula do not contain courses in reliability modeling applicable to different areas of computer engineering. Hence, the topic of reliability analysis is often too diffuse to most computer engineers and researchers dealing with network security. This work is thus aimed at shedding some light on this issue, which can be useful in identifying models, their assumptions and practical parameters for estimating the reliability of threatened systems and for assessing the performance of recovery facilities. It can also be useful for the classification of processes and states regarding the reliability of information systems. Systems with stochastic behaviors undergoing queue operations and random state transitions can also benefit from the approaches presented here. - Highlights: • A concise survey and tutorial in model-based reliability analysis applicable to information security. • A framework of key modeling approaches for assessing reliability of networked systems. • The framework facilitates quantitative risk assessment tasks guided by stochastic modeling and queuing theory. • Evaluation of approaches and models for modeling threats, failures, impacts, and recovery analysis of information systems

Findings from analysing and quantifying human error using current methods

International Nuclear Information System (INIS)

Dang, V.N.; Reer, B.

1999-01-01

In human reliability analysis (HRA), the scarcity of data means that, at best, judgement must be applied to transfer to the domain of the analysis what data are available for similar tasks. In particular for the quantification of tasks involving decisions, the analyst has to choose among quantification approaches that all depend to a significant degree on expert judgement. The use of expert judgement can be made more reliable by eliciting relative judgements rather than absolute judgements. These approaches, which are based on multiple criterion decision theory, focus on ranking the tasks to be analysed by difficulty. While these approaches remedy at least partially the poor performance of experts in the estimation of probabilities, they nevertheless require the calibration of the relative scale on which the actions are ranked in order to obtain the probabilities of interest. This paper presents some results from a comparison of some current HRA methods performed in the frame of a study of SLIM calibration options. The HRA quantification methods THERP, HEART, and INTENT were applied to derive calibration human error probabilities for two groups of operator actions. (author)

Reliability Analysis of Adhesive Bonded Scarf Joints

DEFF Research Database (Denmark)

Kimiaeifar, Amin; Toft, Henrik Stensgaard; Lund, Erik

2012-01-01

element analysis (FEA). For the reliability analysis a design equation is considered which is related to a deterministic code-based design equation where reliability is secured by partial safety factors together with characteristic values for the material properties and loads. The failure criteria......A probabilistic model for the reliability analysis of adhesive bonded scarfed lap joints subjected to static loading is developed. It is representative for the main laminate in a wind turbine blade subjected to flapwise bending. The structural analysis is based on a three dimensional (3D) finite...... are formulated using a von Mises, a modified von Mises and a maximum stress failure criterion. The reliability level is estimated for the scarfed lap joint and this is compared with the target reliability level implicitly used in the wind turbine standard IEC 61400-1. A convergence study is performed to validate...

Development of a methodology for the application of the analysis of human reliability to individualized temporary storage facility; Desarrollo de una metodologia de aplicacion del Analisis de Fiabilidad Humana a una instalacion de Almacen Temporal Individualizado

Energy Technology Data Exchange (ETDEWEB)

Diaz, P.; Dies, J.; Tapia, C.; Blas, A. de

2014-07-01

The paper aims to present the methodology that has been developed with the purpose of applying an ATI without the need of having experts during the process of modelling and quantification analysis of HRA. The developed methodology is based on ATHEANA and relies on the use of other methods of analysis of human action and in-depth analysis. (Author)

Study on HRA-based method for assessing digital man-machine interface

International Nuclear Information System (INIS)

Li Pengcheng; Dai Licao; Zhang Li; Zhao Ming; Hu Hong

2014-01-01

In order to identify the design flaws of digital man-machine interface (MMI) that may trigger human errors or weaken the performance of operators, a HRA-based method (namely HCR + CREAM + HEC) for assessing digital MMI was established. Firstly, the HCR method was used to identify the risk scenarios of high human error probability from the overall event as a whole perspective. Then, for the identified high-risk scenarios, the CREAM was adopted to determine the various error modes and its error probability, and the failure probability was ranked. Finally, the human factors engineering checklist of digital MMI was established according to the characteristics of digital MMI, it was used to check the digital MMI with high error probability in order to identify the design flaws of digital MMI, and the suggestions of optimization were provided. The results show that the provided assessment method can quickly and efficiently identify the design flaws of digital MMI which easily trigger human errors, and the safety of operation of the digital control system for nuclear power plants can be enhanced by optimization of design. (authors)

Selection of the important performance influencing factors for the assessment of human error under accident management situations in nuclear power plants

International Nuclear Information System (INIS)

Kim, J. H.; Jung, W. J.

1999-01-01

This paper introduces the process and final results of selection of the important Performance Influencing Factors (PIFs) under emergency operation and accident management situations in nuclear power plants for use in the assessment of human errors. We collected two types of PIF taxonomies, one is the full set PIF list mainly developed for human error analysis, and the other is the PIFs for human reliability analysis (HRA) in probabilistic safety assessment (PSA). 5 PIF taxonomies among the full set PIF list and 10 PIF taxonomies among HRA methodologies (CREAM, SLIM, INTENT, were collected in this research. By reviewing and analyzing PIFs selected for HRA methodologies, the criterion could be established for the selection of appropriate PIFs under emergency operation and accident management situations. Based on this selection criteria, a new PIF taxonomy was proposed for the assessment of human error under emergency operation and accident management situations in nuclear power plants

Reliability Analysis of a Steel Frame

Directory of Open Access Journals (Sweden)

M. Sýkora

2002-01-01

Full Text Available A steel frame with haunches is designed according to Eurocodes. The frame is exposed to self-weight, snow, and wind actions. Lateral-torsional buckling appears to represent the most critical criterion, which is considered as a basis for the limit state function. In the reliability analysis, the probabilistic models proposed by the Joint Committee for Structural Safety (JCSS are used for basic variables. The uncertainty model coefficients take into account the inaccuracy of the resistance model for the haunched girder and the inaccuracy of the action effect model. The time invariant reliability analysis is based on Turkstra's rule for combinations of snow and wind actions. The time variant analysis describes snow and wind actions by jump processes with intermittencies. Assuming a 50-year lifetime, the obtained values of the reliability index b vary within the range from 3.95 up to 5.56. The cross-profile IPE 330 designed according to Eurocodes seems to be adequate. It appears that the time invariant reliability analysis based on Turkstra's rule provides considerably lower values of b than those obtained by the time variant analysis.

ATHEANA: A Technique for Human Error Analysis: An Overview of Its Methodological Basis

International Nuclear Information System (INIS)

Wreathall, John; Ramey-Smith, Ann

1998-01-01

The U.S. NRC has developed a new human reliability analysis (HRA) method, called A Technique for Human Event Analysis (ATHEANA), to provide a way of modeling the so-called 'errors of commission' - that is, situations in which operators terminate or disable engineered safety features (ESFs) or similar equipment during accident conditions, thereby putting the plant at an increased risk of core damage. In its reviews of operational events, NRC has found that these errors of commission occur with a relatively high frequency (as high as 2 or 3 per year), but are noticeably missing from the scope of most current probabilistic risk assessments (PRAs). This new method was developed through a formalized approach that describes what can occur when operators behave rationally but have inadequate knowledge or poor judgement. In particular, the method is based on models of decision-making and response planning that have been used extensively in the aviation field, and on the analysis of major accidents in both the nuclear and non-nuclear fields. Other papers at this conference present summaries of these event analyses in both the nuclear and non-nuclear fields. This paper presents an overview of ATHEANA and summarizes how the method structures the analysis of operationally significant events, and helps HRA analysts identify and model potentially risk-significant errors of commission in plant PRAs. (authors)

Evaluation of human error estimation for nuclear power plants

International Nuclear Information System (INIS)

Haney, L.N.; Blackman, H.S.

1987-01-01

The dominant risk for severe accident occurrence in nuclear power plants (NPPs) is human error. The US Nuclear Regulatory Commission (NRC) sponsored an evaluation of Human Reliability Analysis (HRA) techniques for estimation of human error in NPPs. Twenty HRA techniques identified by a literature search were evaluated with criteria sets designed for that purpose and categorized. Data were collected at a commercial NPP with operators responding in walkthroughs of four severe accident scenarios and full scope simulator runs. Results suggest a need for refinement and validation of the techniques. 19 refs

Time-dependent reliability sensitivity analysis of motion mechanisms

International Nuclear Information System (INIS)

Wei, Pengfei; Song, Jingwen; Lu, Zhenzhou; Yue, Zhufeng

2016-01-01

Reliability sensitivity analysis aims at identifying the source of structure/mechanism failure, and quantifying the effects of each random source or their distribution parameters on failure probability or reliability. In this paper, the time-dependent parametric reliability sensitivity (PRS) analysis as well as the global reliability sensitivity (GRS) analysis is introduced for the motion mechanisms. The PRS indices are defined as the partial derivatives of the time-dependent reliability w.r.t. the distribution parameters of each random input variable, and they quantify the effect of the small change of each distribution parameter on the time-dependent reliability. The GRS indices are defined for quantifying the individual, interaction and total contributions of the uncertainty in each random input variable to the time-dependent reliability. The envelope function method combined with the first order approximation of the motion error function is introduced for efficiently estimating the time-dependent PRS and GRS indices. Both the time-dependent PRS and GRS analysis techniques can be especially useful for reliability-based design. This significance of the proposed methods as well as the effectiveness of the envelope function method for estimating the time-dependent PRS and GRS indices are demonstrated with a four-bar mechanism and a car rack-and-pinion steering linkage. - Highlights: • Time-dependent parametric reliability sensitivity analysis is presented. • Time-dependent global reliability sensitivity analysis is presented for mechanisms. • The proposed method is especially useful for enhancing the kinematic reliability. • An envelope method is introduced for efficiently implementing the proposed methods. • The proposed method is demonstrated by two real planar mechanisms.

Reliability and validity of risk analysis

International Nuclear Information System (INIS)

Aven, Terje; Heide, Bjornar

2009-01-01

In this paper we investigate to what extent risk analysis meets the scientific quality requirements of reliability and validity. We distinguish between two types of approaches within risk analysis, relative frequency-based approaches and Bayesian approaches. The former category includes both traditional statistical inference methods and the so-called probability of frequency approach. Depending on the risk analysis approach, the aim of the analysis is different, the results are presented in different ways and consequently the meaning of the concepts reliability and validity are not the same.

A Review of Various Performance Shaping Factors for Use in Advanced Control Rooms

International Nuclear Information System (INIS)

Lee, Seung Woo; Ha, Jun Su; Seong, Poong Hyun; Park, Jae Hyuk; Kim, Ja Kyung

2009-01-01

Human reliability analysis(HRA) has been performed as part of the probabilistic risk assessment to identify and quantify human actions and the associated impacts on structures, systems and components for a complex facility. Currently, representative HRA methods such as THERP, ASEP HRA and HCR are being used in Korea. In performing HRA, such conditions that influence human performances have been represented via several context factors. These context factors are referred to by different terms according to method: PSF(Performance Shaping Factors), PIF(Performance Influencing Factors), PAF(Performance Affecting Factors, EPC(Error Producing Conditions), CPC(Common Performance Conditions), and so on. These context factors which will be called PSFs in this study are used in adjusting the basic human error probability(BHEP). However, these PSFs need to be re-assessed since the context is expected to change due to the implementation of computer technologies in NPP. In this study, various PSFs used in different HRA methods are reviewed and PSFs which are frequently mentioned as important factors are derived. Also, HF(Human Factor) issues with one of the design characteristics of advanced NPP are identified

Structural Reliability Analysis of Wind Turbines: A Review

Directory of Open Access Journals (Sweden)

Zhiyu Jiang

2017-12-01

Full Text Available The paper presents a detailed review of the state-of-the-art research activities on structural reliability analysis of wind turbines between the 1990s and 2017. We describe the reliability methods including the first- and second-order reliability methods and the simulation reliability methods and show the procedure for and application areas of structural reliability analysis of wind turbines. Further, we critically review the various structural reliability studies on rotor blades, bottom-fixed support structures, floating systems and mechanical and electrical components. Finally, future applications of structural reliability methods to wind turbine designs are discussed.

Reliability analysis of reactor pressure vessel intensity

International Nuclear Information System (INIS)

Zheng Liangang; Lu Yongbo

2012-01-01

This paper performs the reliability analysis of reactor pressure vessel (RPV) with ANSYS. The analysis method include direct Monte Carlo Simulation method, Latin Hypercube Sampling, central composite design and Box-Behnken Matrix design. The RPV integrity reliability under given input condition is proposed. The result shows that the effects on the RPV base material reliability are internal press, allowable basic stress and elasticity modulus of base material in descending order, and the effects on the bolt reliability are allowable basic stress of bolt material, preload of bolt and internal press in descending order. (authors)

System reliability analysis with natural language and expert's subjectivity

International Nuclear Information System (INIS)

Onisawa, T.

1996-01-01

This paper introduces natural language expressions and expert's subjectivity to system reliability analysis. To this end, this paper defines a subjective measure of reliability and presents the method of the system reliability analysis using the measure. The subjective measure of reliability corresponds to natural language expressions of reliability estimation, which is represented by a fuzzy set defined on [0,1]. The presented method deals with the dependence among subsystems and employs parametrized operations of subjective measures of reliability which can reflect expert 's subjectivity towards the analyzed system. The analysis results are also expressed by linguistic terms. Finally this paper gives an example of the system reliability analysis by the presented method

Risk and reliability analyses (LURI) and expert judgement techniques

International Nuclear Information System (INIS)

Pyy, P.; Pulkkinen, U.

1998-01-01

Probabilistic safety analysis (PSA) is currently used as a regulatory licensing tool in risk informed and plant performance based regulation. More often also utility safety improvements are based on PSA calculations as one criterion. PSA attempts to comprehensively identify all important risk contributors, compare them with each other, assess the safety level and suggest improvements based on its findings. The strength of PSA is that it is capable to provide decision makers with numerical estimates of risks. This makes decision making easier than the comparison of purely qualitative results. PSA is the only comprehensive tool that compactly attempts to include all the important risk contributors in its scope. Despite the demonstrated strengths of PSA, there are some features that have reduced its uses. For example, the PSA scope has been limited to the power operation and process internal events (transients and LOCAs). Only lately, areas such as shutdown, external events and severe accidents have been included in PSA models in many countries. Problems related to modelling are, e.g., that rather static fault and event tree models are commonly used in PSA to model dynamic event sequences. Even if a valid model may be generated, there may not be any other data sources to be used than expert judgement. Furthermore, there are a variety of different techniques for human reliability assessment (HRA) giving varying results. In the project Reliability and Risk Analyses (LURI) these limitations and shortcomings have been studied. In the decision making area, case studies on the application of decision analysis and a doctoral thesis have been published. Further, practical aid has been given to utilities and regulatory decision making. Model uncertainty effect on PSA results has been demonstrated by two case studies. Human reliability has been studied both in the integrated safety analysis study and in the study of maintenance originated NPP component faults based on the

Reliability analysis in intelligent machines

Science.gov (United States)

Mcinroy, John E.; Saridis, George N.

1990-01-01

Given an explicit task to be executed, an intelligent machine must be able to find the probability of success, or reliability, of alternative control and sensing strategies. By using concepts for information theory and reliability theory, new techniques for finding the reliability corresponding to alternative subsets of control and sensing strategies are proposed such that a desired set of specifications can be satisfied. The analysis is straightforward, provided that a set of Gaussian random state variables is available. An example problem illustrates the technique, and general reliability results are presented for visual servoing with a computed torque-control algorithm. Moreover, the example illustrates the principle of increasing precision with decreasing intelligence at the execution level of an intelligent machine.

A Description of the Revised ATHEANA (A Technique for Human Event Analysis)

International Nuclear Information System (INIS)

FORESTER, JOHN A.; BLEY, DENNIS C.; COOPER, SUSANE; KOLACZKOWSKI, ALAN M.; THOMPSON, CATHERINE; RAMEY-SMITH, ANN; WREATHALL, JOHN

2000-01-01

This paper describes the most recent version of a human reliability analysis (HRA) method called ''A Technique for Human Event Analysis'' (ATHEANA). The new version is documented in NUREG-1624, Rev. 1 [1] and reflects improvements to the method based on comments received from a peer review that was held in 1998 (see [2] for a detailed discussion of the peer review comments) and on the results of an initial trial application of the method conducted at a nuclear power plant in 1997 (see Appendix A in [3]). A summary of the more important recommendations resulting from the peer review and trial application is provided and critical and unique aspects of the revised method are discussed

STARS software tool for analysis of reliability and safety

International Nuclear Information System (INIS)

Poucet, A.; Guagnini, E.

1989-01-01

This paper reports on the STARS (Software Tool for the Analysis of Reliability and Safety) project aims at developing an integrated set of Computer Aided Reliability Analysis tools for the various tasks involved in systems safety and reliability analysis including hazard identification, qualitative analysis, logic model construction and evaluation. The expert system technology offers the most promising perspective for developing a Computer Aided Reliability Analysis tool. Combined with graphics and analysis capabilities, it can provide a natural engineering oriented environment for computer assisted reliability and safety modelling and analysis. For hazard identification and fault tree construction, a frame/rule based expert system is used, in which the deductive (goal driven) reasoning and the heuristic, applied during manual fault tree construction, is modelled. Expert system can explain their reasoning so that the analyst can become aware of the why and the how results are being obtained. Hence, the learning aspect involved in manual reliability and safety analysis can be maintained and improved

Results of a nuclear power plant Application of a new technique for human error analysis (ATHEANA)

International Nuclear Information System (INIS)

Forester, J.A.; Whitehead, D.W.; Kolaczkowski, A.M.; Thompson, C.M.

1997-01-01

A new method to analyze human errors has been demonstrated at a pressurized water reactor (PWR) nuclear power plant. This was the first application of the new method referred to as A Technique for Human Error Analysis (ATHEANA). The main goals of the demonstration were to test the ATHEANA process as described in the frame-of-reference manual and the implementation guideline, test a training package developed for the method, test the hypothesis that plant operators and trainers have significant insight into the error-forcing-contexts (EFCs) that can make unsafe actions (UAs) more likely, and to identify ways to improve the method and its documentation. A set of criteria to evaluate the open-quotes successclose quotes of the ATHEANA method as used in the demonstration was identified. A human reliability analysis (HRA) team was formed that consisted of an expert in probabilistic risk assessment (PRA) with some background in HRA (not ATHEANA) and four personnel from the nuclear power plant. Personnel from the plant included two individuals from their PRA staff and two individuals from their training staff. Both individuals from training are currently licensed operators and one of them was a senior reactor operator open-quotes on shiftclose quotes until a few months before the demonstration. The demonstration was conducted over a 5 month period and was observed by members of the Nuclear Regulatory Commission's ATHEANA development team, who also served as consultants to the HRA team when necessary. Example results of the demonstration to date, including identified human failure events (HFEs), UAs, and EFCs are discussed. Also addressed is how simulator exercises are used in the ATHEANA demonstration project

A Human Error Analysis with Physiological Signals during Utilizing Digital Devices

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Hee; Oh, Yeon Ju; Shin, Kwang Hyeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-10-15

The introduction of advanced MCR is accompanied with lots of changes and different forms and features through the virtue of new digital technologies. There are various kinds of digital devices such as flat panel displays, touch screens, and so on. The characteristics of these digital devices give many chances to the interface management, and can be integrated into a compact single workstation in an advanced MCR so that workers can operate the plant with minimum burden during any operating condition. However, these devices may introduce new types of human errors, and thus we need a means to evaluate and prevent such error, especially those related to the digital devices. Human errors have been retrospectively assessed for accident reviews and quantitatively evaluated through HRA for PSA. However, the ergonomic verification and validation is an important process to defend all human error potential in the NPP design. HRA is a crucial part of a PSA, and helps in preparing a countermeasure for design by drawing potential human error items that affect the overall safety of NPPs. Various HRA techniques are available however: they reveal shortages of the HMI design in the digital era. - HRA techniques depend on PSFs: this means that the scope dealing with human factors is previously limited, and thus all attributes of new digital devices may not be considered in HRA. - The data used to HRA are not close to the evaluation items. So, human error analysis is not easy to apply to design by several individual experiments and cases. - The results of HRA are not statistically meaningful because accidents including human errors in NPPs are rare and have been estimated as having an extremely low probability

Reliability analysis of reactor inspection robot(RIROB)

International Nuclear Information System (INIS)

Eom, H. S.; Kim, J. H.; Lee, J. C.; Choi, Y. R.; Moon, S. S.

2002-05-01

This report describes the method and the result of the reliability analysis of RIROB developed in Korea Atomic Energy Research Institute. There are many classic techniques and models for the reliability analysis. These techniques and models have been used widely and approved in other industries such as aviation and nuclear industry. Though these techniques and models have been approved in real fields they are still insufficient for the complicated systems such RIROB which are composed of computer, networks, electronic parts, mechanical parts, and software. Particularly the application of these analysis techniques to digital and software parts of complicated systems is immature at this time thus expert judgement plays important role in evaluating the reliability of the systems at these days. In this report we proposed a method which combines diverse evidences relevant to the reliability to evaluate the reliability of complicated systems such as RIROB. The proposed method combines diverse evidences and performs inference in formal and in quantitative way by using the benefits of Bayesian Belief Nets (BBN)

Reliability analysis techniques for the design engineer

International Nuclear Information System (INIS)

Corran, E.R.; Witt, H.H.

1980-01-01

A fault tree analysis package is described that eliminates most of the housekeeping tasks involved in proceeding from the initial construction of a fault tree to the final stage of presenting a reliability analysis in a safety report. It is suitable for designers with relatively little training in reliability analysis and computer operation. Users can rapidly investigate the reliability implications of various options at the design stage, and evolve a system which meets specified reliability objectives. Later independent review is thus unlikely to reveal major shortcomings necessitating modification and projects delays. The package operates interactively allowing the user to concentrate on the creative task of developing the system fault tree, which may be modified and displayed graphically. For preliminary analysis system data can be derived automatically from a generic data bank. As the analysis procedes improved estimates of critical failure rates and test and maintenance schedules can be inserted. The computations are standard, - identification of minimal cut-sets, estimation of reliability parameters, and ranking of the effect of the individual component failure modes and system failure modes on these parameters. The user can vary the fault trees and data on-line, and print selected data for preferred systems in a form suitable for inclusion in safety reports. A case history is given - that of HIFAR containment isolation system. (author)

A process for application of ATHEANA - a new HRA method

International Nuclear Information System (INIS)

Parry, G.W.; Bley, D.C.; Cooper, S.E.

1996-01-01

This paper describes the analytical process for the application of ATHEANA, a new approach to the performance of human reliability analysis as part of a PRA. This new method, unlike existing methods, is based upon an understanding of the reasons why people make errors, and was developed primarily to address the analysis of errors of commission

Human Reliability Analysis for Design: Using Reliability Methods for Human Factors Issues

Energy Technology Data Exchange (ETDEWEB)

Ronald Laurids Boring

2010-11-01

This paper reviews the application of human reliability analysis methods to human factors design issues. An application framework is sketched in which aspects of modeling typically found in human reliability analysis are used in a complementary fashion to the existing human factors phases of design and testing. The paper provides best achievable practices for design, testing, and modeling. Such best achievable practices may be used to evaluate and human system interface in the context of design safety certifications.

Human Reliability Analysis for Design: Using Reliability Methods for Human Factors Issues

International Nuclear Information System (INIS)

Boring, Ronald Laurids

2010-01-01

This paper reviews the application of human reliability analysis methods to human factors design issues. An application framework is sketched in which aspects of modeling typically found in human reliability analysis are used in a complementary fashion to the existing human factors phases of design and testing. The paper provides best achievable practices for design, testing, and modeling. Such best achievable practices may be used to evaluate and human system interface in the context of design safety certifications.

Systems reliability analysis for the national ignition facility

International Nuclear Information System (INIS)

Majumdar, K.C.; Annese, C.E.; MacIntyre, A.T.; Sicherman, A.

1996-01-01

A Reliability, Availability and Maintainability (RAM) analysis was initiated for the National Ignition Facility (NIF). The NIF is an inertial confinement fusion research facility designed to achieve controlled thermonuclear reaction; the preferred site for the NIF is the Lawrence Livermore National Laboratory (LLNL). The NIF RAM analysis has three purposes: (1) to allocate top level reliability and availability goals for the systems, (2) to develop an operability model for optimum maintainability, and (3) to determine the achievability of the allocated goals of the RAM parameters for the NIF systems and the facility operation as a whole. An allocation model assigns the reliability and availability goals for front line and support systems by a top-down approach; reliability analysis uses a bottom-up approach to determine the system reliability and availability from component level to system level

Mechanical reliability analysis of tubes intended for hydrocarbons

Energy Technology Data Exchange (ETDEWEB)

Nahal, Mourad; Khelif, Rabia [Badji Mokhtar University, Annaba (Algeria)

2013-02-15

Reliability analysis constitutes an essential phase in any study concerning reliability. Many industrialists evaluate and improve the reliability of their products during the development cycle - from design to startup (design, manufacture, and exploitation) - to develop their knowledge on cost/reliability ratio and to control sources of failure. In this study, we obtain results for hardness, tensile, and hydrostatic tests carried out on steel tubes for transporting hydrocarbons followed by statistical analysis. Results obtained allow us to conduct a reliability study based on resistance request. Thus, index of reliability is calculated and the importance of the variables related to the tube is presented. Reliability-based assessment of residual stress effects is applied to underground pipelines under a roadway, with and without active corrosion. Residual stress has been found to greatly increase probability of failure, especially in the early stages of pipe lifetime.

A quantitative impact analysis of sensor failures on human operator's decision making in nuclear power plants

International Nuclear Information System (INIS)

Seong, Poong Hyun

2004-01-01

In emergency or accident situations in nuclear power plants, human operators take important roles in generating appropriate control signals to mitigate accident situation. In human reliability analysis (HRA) in the framework of probabilistic safety assessment (PSA), the failure probabilities of such appropriate actions are estimated and used for the safety analysis of nuclear power plants. Even though understanding the status of the plant is basically the process of information seeking and processing by human operators, it seems that conventional HRA methods such as THERP, HCR, and ASEP does not pay a lot of attention to the possibilities of providing wrong information to human operators. In this paper, a quantitative impact analysis of providing wrong information to human operators due to instrument faults or sensor failures is performed. The quantitative impact analysis is performed based on a quantitative situation assessment model. By comparing the situation in which there are sensor failures and the situation in which there are not sensor failures, the impact of sensor failures can be evaluated quantitatively. It is concluded that the impact of sensor failures are quite significant at the initial stages, but the impact is gradually reduced as human operators make more and more observations. Even though the impact analysis is highly dependent on the situation assessment model, it is expected that the conclusions made based on other situation assessment models with be consistent with the conclusion made in this paper. (author)

Reliability Analysis of Wind Turbines

DEFF Research Database (Denmark)

Toft, Henrik Stensgaard; Sørensen, John Dalsgaard

2008-01-01

In order to minimise the total expected life-cycle costs of a wind turbine it is important to estimate the reliability level for all components in the wind turbine. This paper deals with reliability analysis for the tower and blades of onshore wind turbines placed in a wind farm. The limit states...... consideres are in the ultimate limit state (ULS) extreme conditions in the standstill position and extreme conditions during operating. For wind turbines, where the magnitude of the loads is influenced by the control system, the ultimate limit state can occur in both cases. In the fatigue limit state (FLS......) the reliability level for a wind turbine placed in a wind farm is considered, and wake effects from neighbouring wind turbines is taken into account. An illustrative example with calculation of the reliability for mudline bending of the tower is considered. In the example the design is determined according...

Component reliability analysis for development of component reliability DB of Korean standard NPPs

International Nuclear Information System (INIS)

Choi, S. Y.; Han, S. H.; Kim, S. H.

2002-01-01

The reliability data of Korean NPP that reflects the plant specific characteristics is necessary for PSA and Risk Informed Application. We have performed a project to develop the component reliability DB and calculate the component reliability such as failure rate and unavailability. We have collected the component operation data and failure/repair data of Korean standard NPPs. We have analyzed failure data by developing a data analysis method which incorporates the domestic data situation. And then we have compared the reliability results with the generic data for the foreign NPPs

Reliability Analysis for Safety Grade PLC(POSAFE-Q)

International Nuclear Information System (INIS)

Choi, Kyung Chul; Song, Seung Whan; Park, Gang Min; Hwang, Sung Jae

2012-01-01

Safety Grade PLC(Programmable Logic Controller), POSAFE-Q, was developed recently in accordance with nuclear regulatory and requirements. In this paper, describe reliability analysis for digital safety grade PLC (especially POSAFE-Q). Reliability analysis scope is Prediction, Calculation of MTBF (Mean Time Between Failure), FMEA (Failure Mode Effect Analysis), PFD (Probability of Failure on Demand). (author)

Weibull distribution in reliability data analysis in nuclear power plant

International Nuclear Information System (INIS)

Ma Yingfei; Zhang Zhijian; Zhang Min; Zheng Gangyang

2015-01-01

Reliability is an important issue affecting each stage of the life cycle ranging from birth to death of a product or a system. The reliability engineering includes the equipment failure data processing, quantitative assessment of system reliability and maintenance, etc. Reliability data refers to the variety of data that describe the reliability of system or component during its operation. These data may be in the form of numbers, graphics, symbols, texts and curves. Quantitative reliability assessment is the task of the reliability data analysis. It provides the information related to preventing, detect, and correct the defects of the reliability design. Reliability data analysis under proceed with the various stages of product life cycle and reliability activities. Reliability data of Systems Structures and Components (SSCs) in Nuclear Power Plants is the key factor of probabilistic safety assessment (PSA); reliability centered maintenance and life cycle management. The Weibull distribution is widely used in reliability engineering, failure analysis, industrial engineering to represent manufacturing and delivery times. It is commonly used to model time to fail, time to repair and material strength. In this paper, an improved Weibull distribution is introduced to analyze the reliability data of the SSCs in Nuclear Power Plants. An example is given in the paper to present the result of the new method. The Weibull distribution of mechanical equipment for reliability data fitting ability is very strong in nuclear power plant. It's a widely used mathematical model for reliability analysis. The current commonly used methods are two-parameter and three-parameter Weibull distribution. Through comparison and analysis, the three-parameter Weibull distribution fits the data better. It can reflect the reliability characteristics of the equipment and it is more realistic to the actual situation. (author)

Some aspects of statistical modeling of human-error probability

International Nuclear Information System (INIS)

Prairie, R.R.

1982-01-01

Human reliability analyses (HRA) are often performed as part of risk assessment and reliability projects. Recent events in nuclear power have shown the potential importance of the human element. There are several on-going efforts in the US and elsewhere with the purpose of modeling human error such that the human contribution can be incorporated into an overall risk assessment associated with one or more aspects of nuclear power. An effort that is described here uses the HRA (event tree) to quantify and model the human contribution to risk. As an example, risk analyses are being prepared on several nuclear power plants as part of the Interim Reliability Assessment Program (IREP). In this process the risk analyst selects the elements of his fault tree that could be contributed to by human error. He then solicits the HF analyst to do a HRA on this element

Methods development to evaluate the risk of upgrading to DCS: The human factor

Energy Technology Data Exchange (ETDEWEB)

Ostrom, L.T.; Wilhelmsen, C.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

1995-04-01

The NRC recognizes that a more complete technical basis for understanding and regulating advanced digital technologies in commercial nuclear power plants is needed. A concern is that the introduction of digital safety systems may have an impact on risk. There is currently no standard methodology for measuring digital system reliability. A tool currently used to evaluate NPP risk in analog systems is the probabilistic risk assessment (PRA). The use of this tool to evaluate the digital system risk was considered to be a potential methodology for determining the risk. To test this hypothesis, it was decided to perform a limited PRA on a single dominant accident sequence. However, a review of existing human reliability analysis (HRA) methods showed that they were inadequate to analyze systems utilizing digital technology. A four step process was used to adapt existing HRA methodologies to digital environments and to develop new techniques. The HRA methods were then used to analyze an NPP that had undergone a backfit to digital technology in order to determine, as a first step, whether the methods were effective. The very small-break loss of coolant accident sequence was analyzed to determine whether the upgrade to the Eagle-21 process protection system had an effect on risk. The analysis of the very small-break LOCA documented in the Sequoyah PRA was used as the basis of the analysis. The analysis of the results of the HRA showed that the mean human error probabilities for the Eagle-21 PPS were slightly less than those for the analog system it replaced. One important observation from the analysis is that the operators have increased confidence steming from the better level of control provided by the digital system. The analysis of the PRA results, which included the human error component and the Eagle-21 PPS, disclosed that the reactor protection system had a higher failure rate than the analog system, although the difference was not statistically significant.

Reliability Analysis of Tubular Joints in Offshore Structures

DEFF Research Database (Denmark)

Thoft-Christensen, Palle; Sørensen, John Dalsgaard

1987-01-01

Reliability analysis of single tubular joints and offshore platforms with tubular joints is" presented. The failure modes considered are yielding, punching, buckling and fatigue failure. Element reliability as well as systems reliability approaches are used and illustrated by several examples....... Finally, optimal design of tubular.joints with reliability constraints is discussed and illustrated by an example....

Waste package reliability analysis

International Nuclear Information System (INIS)

Pescatore, C.; Sastre, C.

1983-01-01

Proof of future performance of a complex system such as a high-level nuclear waste package over a period of hundreds to thousands of years cannot be had in the ordinary sense of the word. The general method of probabilistic reliability analysis could provide an acceptable framework to identify, organize, and convey the information necessary to satisfy the criterion of reasonable assurance of waste package performance according to the regulatory requirements set forth in 10 CFR 60. General principles which may be used to evaluate the qualitative and quantitative reliability of a waste package design are indicated and illustrated with a sample calculation of a repository concept in basalt. 8 references, 1 table

Modelling human interactions in the assessment of man-made hazards

International Nuclear Information System (INIS)

Nitoi, M.; Farcasiu, M.; Apostol, M.

2016-01-01

The human reliability assessment tools are not currently capable to model adequately the human ability to adapt, to innovate and to manage under extreme situations. The paper presents the results obtained by ICN PSA team in the frame of FP7 Advanced Safety Assessment Methodologies: extended PSA (ASAMPSA_E) project regarding the investigation of conducting HRA in human-made hazards. The paper proposes to use a 4-steps methodology for the assessment of human interactions in the external events (Definition and modelling of human interactions; Quantification of human failure events; Recovery analysis; Review). The most relevant factors with respect to HRA for man-made hazards (response execution complexity; existence of procedures with respect to the scenario in question; time available for action; timing of cues; accessibility of equipment; harsh environmental conditions) are presented and discussed thoroughly. The challenges identified in relation to man-made hazards HRA are highlighted. (authors)

Swimming pool reactor reliability and safety analysis

International Nuclear Information System (INIS)

Li Zhaohuan

1997-01-01

A reliability and safety analysis of Swimming Pool Reactor in China Institute of Atomic Energy is done by use of event/fault tree technique. The paper briefly describes the analysis model, analysis code and main results. Meanwhile it also describes the impact of unassigned operation status on safety, the estimation of effectiveness of defense tactics in maintenance against common cause failure, the effectiveness of recovering actions on the system reliability, the comparison of occurrence frequencies of the core damage by use of generic and specific data

Probabilistic risk assessment course documentation. Volume 3. System reliability and analysis techniques, Session A - reliability

International Nuclear Information System (INIS)

Lofgren, E.V.

1985-08-01

This course in System Reliability and Analysis Techniques focuses on the quantitative estimation of reliability at the systems level. Various methods are reviewed, but the structure provided by the fault tree method is used as the basis for system reliability estimates. The principles of fault tree analysis are briefly reviewed. Contributors to system unreliability and unavailability are reviewed, models are given for quantitative evaluation, and the requirements for both generic and plant-specific data are discussed. Also covered are issues of quantifying component faults that relate to the systems context in which the components are embedded. All reliability terms are carefully defined. 44 figs., 22 tabs

Application of Metric-based Software Reliability Analysis to Example Software

International Nuclear Information System (INIS)

Kim, Man Cheol; Smidts, Carol

2008-07-01

The software reliability of TELLERFAST ATM software is analyzed by using two metric-based software reliability analysis methods, a state transition diagram-based method and a test coverage-based method. The procedures for the software reliability analysis by using the two methods and the analysis results are provided in this report. It is found that the two methods have a relation of complementary cooperation, and therefore further researches on combining the two methods to reflect the benefit of the complementary cooperative effect to the software reliability analysis are recommended

Fatigue Reliability Analysis of a Mono-Tower Platform

DEFF Research Database (Denmark)

Kirkegaard, Poul Henning; Sørensen, John Dalsgaard; Brincker, Rune

1991-01-01

In this paper, a fatigue reliability analysis of a Mono-tower platform is presented. The failure mode, fatigue failure in the butt welds, is investigated with two different models. The one with the fatigue strength expressed through SN relations, the other with the fatigue strength expressed thro...... of the natural period, damping ratio, current, stress spectrum and parameters describing the fatigue strength. Further, soil damping is shown to be significant for the Mono-tower.......In this paper, a fatigue reliability analysis of a Mono-tower platform is presented. The failure mode, fatigue failure in the butt welds, is investigated with two different models. The one with the fatigue strength expressed through SN relations, the other with the fatigue strength expressed...... through linear-elastic fracture mechanics (LEFM). In determining the cumulative fatigue damage, Palmgren-Miner's rule is applied. Element reliability, as well as systems reliability, is estimated using first-order reliability methods (FORM). The sensitivity of the systems reliability to various parameters...

Hybrid instrument applied to human reliability study in event of loss of external electric power in a nuclear power plant; Instrumento hibrido aplicado ao estudo da confiabilidade humana em evento de perda de energia eletrica externa em usina nuclear

Energy Technology Data Exchange (ETDEWEB)

Martins, Eduardo Ferraz

2015-04-01

The study projects in highly complex installations involves robust modeling, supported by conceptual and mathematical tools, to carry out systematic research and structured the different risk scenarios that can lead to unwanted events from occurring equipment failures or human errors. In the context of classical modeling, the Probabilistic Safety Analysis (PSA) seeks to provide qualitative and quantitative information about the project particularity and their operational facilities, including the identification of factors or scenarios that contribute to the risk and consequent comparison options for increasing safety. In this context, the aim of the thesis is to develop a hybrid instrument (CPP-HI) innovative, from the integrated modeling techniques of Failure Mode and Effect Analysis (FMEA), concepts of Human Reliability Analysis and Probabilistic Composition of Preferences (PCP). In support of modeling and validation of the CPP-HI, a simulation was performed on a triggering event 'Loss of External Electric Power' - PEEE, in a Nuclear Power plant. The results were simulated in a virtual environment (sensitivity analysis) and are robust to the study of Human Reliability Analysis (HRA) in the context of the PSA. (author)

Prime implicants in dynamic reliability analysis

International Nuclear Information System (INIS)

Tyrväinen, Tero

2016-01-01

This paper develops an improved definition of a prime implicant for the needs of dynamic reliability analysis. Reliability analyses often aim to identify minimal cut sets or prime implicants, which are minimal conditions that cause an undesired top event, such as a system's failure. Dynamic reliability analysis methods take the time-dependent behaviour of a system into account. This means that the state of a component can change in the analysed time frame and prime implicants can include the failure of a component at different time points. There can also be dynamic constraints on a component's behaviour. For example, a component can be non-repairable in the given time frame. If a non-repairable component needs to be failed at a certain time point to cause the top event, we consider that the condition that it is failed at the latest possible time point is minimal, and the condition in which it fails earlier non-minimal. The traditional definition of a prime implicant does not account for this type of time-related minimality. In this paper, a new definition is introduced and illustrated using a dynamic flowgraph methodology model. - Highlights: • A new definition of a prime implicant is developed for dynamic reliability analysis. • The new definition takes time-related minimality into account. • The new definition is needed in dynamic flowgraph methodology. • Results can be represented by a smaller number of prime implicants.

An Empirical Investigation into Causality of Unsafe Act and Recovery during EOP Simulation

Energy Technology Data Exchange (ETDEWEB)

Choi, Sun Yeong; Jung, Won Dea [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-08-15

A data collection worksheet and guideline to collect HRA (Human Reliability Analysis) data with simulator data sources were developed for the HRA data handbook project by KAERI. Using the data worksheet, simulator data were collected and analyzed for an HRA qualitative database. The purpose of this paper is to define the causalities of operators' UAs (Unsafe Acts) ending in an inappropriate component manipulation and recovery during an EOP (Emergency Operating Procedure) operation, and to show some results for the causality from a case study. The reason we suggest the causality of an UA is because an inappropriate manipulation during an EOP operation can be resulted by the causality among operators in an MCR (Main Control Room). Therefore, a 'causality' data field was inserted into the data worksheet to identify the real initiator, and related operators for an inappropriate component manipulation. With this 'causality' data field, an HRA analyzer can establish who caused an UA (or a recovery) and who was involved in the process. They can also calculate the HEP (Human Error Probability) grouped by the initiator if they are interested in the HEP by the initiator.

Framework for estimating response time data to conduct a seismic human reliability analysis - its feasibility

International Nuclear Information System (INIS)

Park, Jinkyun; Kin, Yochan; Jung, Wondea; Jang, Seung Cheol

2014-01-01

This is because the PSA has been used for several decades as the representative tool to evaluate the safety of NPPs. To this end, it is inevitable to evaluate human error probabilities (HEPs) in conducting important tasks being considered in the PSA framework (i.e., HFEs; human failure events), which are able to significantly affect the safety of NPPs. In addition, it should be emphasized that the provision of a realistic human performance data is an important precondition for calculating HEPs under a seismic condition. Unfortunately, it seems that HRA methods being currently used for calculating HEPs under a seismic event do not properly consider the performance variation of human operators. For this reason, in this paper, a framework to estimate response time data that are critical for calculating HEPs is suggested with respect to a seismic intensity. This paper suggested a systematic framework for estimating response time data that would be one of the most critical for calculating HEPs. Although extensive review of existing literatures is indispensable for identifying response times of human operators who have to conduct a series of tasks prescribed in procedures based on a couple of wrong indications, it is highly expected that response time data for seismic HRA can be properly secured through revisiting response time data collected from diverse situations without concerning a seismic event

One Size Does Not Fit All: Human Failure Event Decomposition and Task Analysis

Energy Technology Data Exchange (ETDEWEB)

Ronald Laurids Boring, PhD

2014-09-01

In the probabilistic safety assessments (PSAs) used in the nuclear industry, human failure events (HFEs) are determined as a subset of hardware failures, namely those hardware failures that could be triggered or exacerbated by human action or inaction. This approach is top-down, starting with hardware faults and deducing human contributions to those faults. Elsewhere, more traditionally human factors driven approaches would tend to look at opportunities for human errors first in a task analysis and then identify which of those errors is risk significant. The intersection of top-down and bottom-up approaches to defining HFEs has not been carefully studied. Ideally, both approaches should arrive at the same set of HFEs. This question remains central as human reliability analysis (HRA) methods are generalized to new domains like oil and gas. The HFEs used in nuclear PSAs tend to be top-down—defined as a subset of the PSA—whereas the HFEs used in petroleum quantitative risk assessments (QRAs) are more likely to be bottom-up—derived from a task analysis conducted by human factors experts. The marriage of these approaches is necessary in order to ensure that HRA methods developed for top-down HFEs are also sufficient for bottom-up applications. In this paper, I first review top-down and bottom-up approaches for defining HFEs and then present a seven-step guideline to ensure a task analysis completed as part of human error identification decomposes to a level suitable for use as HFEs. This guideline illustrates an effective way to bridge the bottom-up approach with top-down requirements.

Reliability Analysis of Elasto-Plastic Structures

DEFF Research Database (Denmark)

Thoft-Christensen, Palle; Sørensen, John Dalsgaard

1984-01-01

. Failure of this type of system is defined either as formation of a mechanism or by failure of a prescribed number of elements. In the first case failure is independent of the order in which the elements fail, but this is not so by the second definition. The reliability analysis consists of two parts...... are described and the two definitions of failure can be used by the first formulation, but only the failure definition based on formation of a mechanism by the second formulation. The second part of the reliability analysis is an estimate of the failure probability for the structure on the basis...

Bearing Procurement Analysis Method by Total Cost of Ownership Analysis and Reliability Prediction

Science.gov (United States)

Trusaji, Wildan; Akbar, Muhammad; Sukoyo; Irianto, Dradjad

2018-03-01

In making bearing procurement analysis, price and its reliability must be considered as decision criteria, since price determines the direct cost as acquisition cost and reliability of bearing determine the indirect cost such as maintenance cost. Despite the indirect cost is hard to identify and measured, it has high contribution to overall cost that will be incurred. So, the indirect cost of reliability must be considered when making bearing procurement analysis. This paper tries to explain bearing evaluation method with the total cost of ownership analysis to consider price and maintenance cost as decision criteria. Furthermore, since there is a lack of failure data when bearing evaluation phase is conducted, reliability prediction method is used to predict bearing reliability from its dynamic load rating parameter. With this method, bearing with a higher price but has higher reliability is preferable for long-term planning. But for short-term planning the cheaper one but has lower reliability is preferable. This contextuality can give rise to conflict between stakeholders. Thus, the planning horizon needs to be agreed by all stakeholder before making a procurement decision.

Reliability Analysis Techniques for Communication Networks in Nuclear Power Plant

International Nuclear Information System (INIS)

Lim, T. J.; Jang, S. C.; Kang, H. G.; Kim, M. C.; Eom, H. S.; Lee, H. J.

2006-09-01

The objectives of this project is to investigate and study existing reliability analysis techniques for communication networks in order to develop reliability analysis models for nuclear power plant's safety-critical networks. It is necessary to make a comprehensive survey of current methodologies for communication network reliability. Major outputs of this study are design characteristics of safety-critical communication networks, efficient algorithms for quantifying reliability of communication networks, and preliminary models for assessing reliability of safety-critical communication networks

Reliability analysis of grid connected small wind turbine power electronics

International Nuclear Information System (INIS)

Arifujjaman, Md.; Iqbal, M.T.; Quaicoe, J.E.

2009-01-01

Grid connection of small permanent magnet generator (PMG) based wind turbines requires a power conditioning system comprising a bridge rectifier, a dc-dc converter and a grid-tie inverter. This work presents a reliability analysis and an identification of the least reliable component of the power conditioning system of such grid connection arrangements. Reliability of the configuration is analyzed for the worst case scenario of maximum conversion losses at a particular wind speed. The analysis reveals that the reliability of the power conditioning system of such PMG based wind turbines is fairly low and it reduces to 84% of initial value within one year. The investigation is further enhanced by identifying the least reliable component within the power conditioning system and found that the inverter has the dominant effect on the system reliability, while the dc-dc converter has the least significant effect. The reliability analysis demonstrates that a permanent magnet generator based wind energy conversion system is not the best option from the point of view of power conditioning system reliability. The analysis also reveals that new research is required to determine a robust power electronics configuration for small wind turbine conversion systems.

Analysis of operating reliability of WWER-1000 unit

International Nuclear Information System (INIS)

Bortlik, J.

1985-01-01

The nuclear power unit was divided into 33 technological units. Input data for reliability analysis were surveys of operating results obtained from the IAEA information system and certain indexes of the reliability of technological equipment determined using the Bayes formula. The missing reliability data for technological equipment were used from the basic variant. The fault tree of the WWER-1000 unit was determined for the peak event defined as the impossibility of reaching 100%, 75% and 50% of rated power. The period was observed of the nuclear power plant operation with reduced output owing to defect and the respective time needed for a repair of the equipment. The calculation of the availability of the WWER-1000 unit was made for different variant situations. Certain indexes of the operating reliability of the WWER-1000 unit which are the result of a detailed reliability analysis are tabulated for selected variants. (E.S.)

Reliability analysis and assessment of structural systems

International Nuclear Information System (INIS)

Yao, J.T.P.; Anderson, C.A.

1977-01-01

The study of structural reliability deals with the probability of having satisfactory performance of the structure under consideration within any specific time period. To pursue this study, it is necessary to apply available knowledge and methodology in structural analysis (including dynamics) and design, behavior of materials and structures, experimental mechanics, and the theory of probability and statistics. In addition, various severe loading phenomena such as strong motion earthquakes and wind storms are important considerations. For three decades now, much work has been done on reliability analysis of structures, and during this past decade, certain so-called 'Level I' reliability-based design codes have been proposed and are in various stages of implementation. These contributions will be critically reviewed and summarized in this paper. Because of the undesirable consequences resulting from the failure of nuclear structures, it is important and desirable to consider the structural reliability in the analysis and design of these structures. Moreover, after these nuclear structures are constructed, it is desirable for engineers to be able to assess the structural reliability periodically as well as immediately following the occurrence of severe loading conditions such as a strong-motion earthquake. During this past decade, increasing use has been made of techniques of system identification in structural engineering. On the basis of non-destructive test results, various methods have been developed to obtain an adequate mathematical model (such as the equations of motion with more realistic parameters) to represent the structural system

Safety and reliability analysis based on nonprobabilistic methods

International Nuclear Information System (INIS)

Kozin, I.O.; Petersen, K.E.

1996-01-01

Imprecise probabilities, being developed during the last two decades, offer a considerably more general theory having many advantages which make it very promising for reliability and safety analysis. The objective of the paper is to argue that imprecise probabilities are more appropriate tool for reliability and safety analysis, that they allow to model the behavior of nuclear industry objects more comprehensively and give a possibility to solve some problems unsolved in the framework of conventional approach. Furthermore, some specific examples are given from which we can see the usefulness of the tool for solving some reliability tasks

Qualitative and quantitative methods for human factor analysis and assessment in NPP. Investigations and results

International Nuclear Information System (INIS)

Hristova, R.; Kalchev, B.; Atanasov, D.

2005-01-01

We consider here two basic groups of methods for analysis and assessment of the human factor in the NPP area and give some results from performed analyses as well. The human factor is the human interaction with the design equipment, with the working environment and takes into account the human capabilities and limits. In the frame of the qualitative methods for analysis of the human factor are considered concepts and structural methods for classifying of the information, connected with the human factor. Emphasize is given to the HPES method for human factor analysis in NPP. Methods for quantitative assessment of the human reliability are considered. These methods allow assigning of probabilities to the elements of the already structured information about human performance. This part includes overview of classical methods for human reliability assessment (HRA, THERP), and methods taking into account specific information about human capabilities and limits and about the man-machine interface (CHR, HEART, ATHEANA). Quantitative and qualitative results concerning human factor influence in the initiating events occurrences in the Kozloduy NPP are presented. (authors)

System Reliability Analysis Considering Correlation of Performances

Energy Technology Data Exchange (ETDEWEB)

Kim, Saekyeol; Lee, Tae Hee [Hanyang Univ., Seoul (Korea, Republic of); Lim, Woochul [Mando Corporation, Seongnam (Korea, Republic of)

2017-04-15

Reliability analysis of a mechanical system has been developed in order to consider the uncertainties in the product design that may occur from the tolerance of design variables, uncertainties of noise, environmental factors, and material properties. In most of the previous studies, the reliability was calculated independently for each performance of the system. However, the conventional methods cannot consider the correlation between the performances of the system that may lead to a difference between the reliability of the entire system and the reliability of the individual performance. In this paper, the joint probability density function (PDF) of the performances is modeled using a copula which takes into account the correlation between performances of the system. The system reliability is proposed as the integral of joint PDF of performances and is compared with the individual reliability of each performance by mathematical examples and two-bar truss example.

System Reliability Analysis Considering Correlation of Performances

International Nuclear Information System (INIS)

Kim, Saekyeol; Lee, Tae Hee; Lim, Woochul

2017-01-01

Reliability analysis of a mechanical system has been developed in order to consider the uncertainties in the product design that may occur from the tolerance of design variables, uncertainties of noise, environmental factors, and material properties. In most of the previous studies, the reliability was calculated independently for each performance of the system. However, the conventional methods cannot consider the correlation between the performances of the system that may lead to a difference between the reliability of the entire system and the reliability of the individual performance. In this paper, the joint probability density function (PDF) of the performances is modeled using a copula which takes into account the correlation between performances of the system. The system reliability is proposed as the integral of joint PDF of performances and is compared with the individual reliability of each performance by mathematical examples and two-bar truss example.

Reliability analysis of digital based I and C system

Energy Technology Data Exchange (ETDEWEB)

Kang, I. S.; Cho, B. S.; Choi, M. J. [KOPEC, Yongin (Korea, Republic of)

1999-10-01

Rapidly, digital technology is being widely applied in replacing analog component installed in existing plant and designing new nuclear power plant for control and monitoring system in Korea as well as in foreign countries. Even though many merits of digital technology, it is being faced with a new problem of reliability assurance. The studies for solving this problem are being performed vigorously in foreign countries. The reliability of KNGR Engineered Safety Features Component Control System (ESF-CCS), digital based I and C system, was analyzed to verify fulfillment of the ALWR EPRI-URD requirement for reliability analysis and eliminate hazards in design applied new technology. The qualitative analysis using FMEA and quantitative analysis using reliability block diagram were performed. The results of analyses are shown in this paper.

Development of RBDGG Solver and Its Application to System Reliability Analysis

International Nuclear Information System (INIS)

Kim, Man Cheol

2010-01-01

For the purpose of making system reliability analysis easier and more intuitive, RBDGG (Reliability Block diagram with General Gates) methodology was introduced as an extension of the conventional reliability block diagram. The advantage of the RBDGG methodology is that the structure of a RBDGG model is very similar to the actual structure of the analyzed system, and therefore the modeling of a system for system reliability and unavailability analysis becomes very intuitive and easy. The main idea of the development of the RBDGG methodology is similar with that of the development of the RGGG (Reliability Graph with General Gates) methodology, which is an extension of a conventional reliability graph. The newly proposed methodology is now implemented into a software tool, RBDGG Solver. RBDGG Solver was developed as a WIN32 console application. RBDGG Solver receives information on the failure modes and failure probabilities of each component in the system, along with the connection structure and connection logics among the components in the system. Based on the received information, RBDGG Solver automatically generates a system reliability analysis model for the system, and then provides the analysis results. In this paper, application of RBDGG Solver to the reliability analysis of an example system, and verification of the calculation results are provided for the purpose of demonstrating how RBDGG Solver is used for system reliability analysis

Advances in methods and applications of reliability and safety analysis

International Nuclear Information System (INIS)

Fieandt, J.; Hossi, H.; Laakso, K.; Lyytikaeinen, A.; Niemelae, I.; Pulkkinen, U.; Pulli, T.

1986-01-01

The know-how of the reliability and safety design and analysis techniques of Vtt has been established over several years in analyzing the reliability in the Finnish nuclear power plants Loviisa and Olkiluoto. This experience has been later on applied and developed to be used in the process industry, conventional power industry, automation and electronics. VTT develops and transfers methods and tools for reliability and safety analysis to the private and public sectors. The technology transfer takes place in joint development projects with potential users. Several computer-aided methods, such as RELVEC for reliability modelling and analysis, have been developed. The tool developed are today used by major Finnish companies in the fields of automation, nuclear power, shipbuilding and electronics. Development of computer-aided and other methods needed in analysis of operating experience, reliability or safety is further going on in a number of research and development projects

Mathematical Methods in Survival Analysis, Reliability and Quality of Life

CERN Document Server

Huber, Catherine; Mesbah, Mounir

2008-01-01

Reliability and survival analysis are important applications of stochastic mathematics (probability, statistics and stochastic processes) that are usually covered separately in spite of the similarity of the involved mathematical theory. This title aims to redress this situation: it includes 21 chapters divided into four parts: Survival analysis, Reliability, Quality of life, and Related topics. Many of these chapters were presented at the European Seminar on Mathematical Methods for Survival Analysis, Reliability and Quality of Life in 2006.

Reliability analysis - systematic approach based on limited data

International Nuclear Information System (INIS)

Bourne, A.J.

1975-11-01

The initial approaches required for reliability analysis are outlined. These approaches highlight the system boundaries, examine the conditions under which the system is required to operate, and define the overall performance requirements. The discussion is illustrated by a simple example of an automatic protective system for a nuclear reactor. It is then shown how the initial approach leads to a method of defining the system, establishing performance parameters of interest and determining the general form of reliability models to be used. The overall system model and the availability of reliability data at the system level are next examined. An iterative process is then described whereby the reliability model and data requirements are systematically refined at progressively lower hierarchic levels of the system. At each stage, the approach is illustrated with examples from the protective system previously described. The main advantages of the approach put forward are the systematic process of analysis, the concentration of assessment effort in the critical areas and the maximum use of limited reliability data. (author)

A human reliability analysis (HRA) method for identifying and assessing the error of commission (EOC) from a diagnosis failure

Energy Technology Data Exchange (ETDEWEB)

Kim, Jae Whan; Jung, Won Dea; Park, Jin Yun; Kang, Dae Il

2005-01-01

The study deals with a method for systematically identifying and assessing the EOC events that might be caused from a diagnosis failure or misdiagnosis of the expected events in accident scenarios of nuclear power plants. The method for EOC identification and assessment consists of three steps: analysis of the potential for a diagnosis failure (or misdiagnosis), identification of the EOC events from the diagnosis failure, quantitative assessment of the identified EOC events. As a tool for analysing a diagnosis failure, the MisDiagnosis Tree Analysis (MDTA) technique is proposed with the taxonomy of misdiagnosis causes. Also, the guidance on the identification of EOC events and the classification system and data are given for quantitiative assessment. As an applicaton of the proposed method, the EOCs identification and assessment for Younggwang 3 and 4 plants and their impact on the plant risk were performed. As the result, six events or event sequences were considered for diagnosis failures and about 20 new Human Failure Events (HFEs) involving EOCs were identified. According to the assessment of the risk impact of the identified HFEs, they increase the CDF by 11.4 % of the current CDF value, which corresponds to 10.2 % of the new CDF. The small loss of coolant accident (SLOCA) turned out to be a major contributor to the increase of CDF resulting in 9.2 % increaseof the current CDF.

Bayesian Inference for NASA Probabilistic Risk and Reliability Analysis

Science.gov (United States)

Dezfuli, Homayoon; Kelly, Dana; Smith, Curtis; Vedros, Kurt; Galyean, William

2009-01-01

This document, Bayesian Inference for NASA Probabilistic Risk and Reliability Analysis, is intended to provide guidelines for the collection and evaluation of risk and reliability-related data. It is aimed at scientists and engineers familiar with risk and reliability methods and provides a hands-on approach to the investigation and application of a variety of risk and reliability data assessment methods, tools, and techniques. This document provides both: A broad perspective on data analysis collection and evaluation issues. A narrow focus on the methods to implement a comprehensive information repository. The topics addressed herein cover the fundamentals of how data and information are to be used in risk and reliability analysis models and their potential role in decision making. Understanding these topics is essential to attaining a risk informed decision making environment that is being sought by NASA requirements and procedures such as 8000.4 (Agency Risk Management Procedural Requirements), NPR 8705.05 (Probabilistic Risk Assessment Procedures for NASA Programs and Projects), and the System Safety requirements of NPR 8715.3 (NASA General Safety Program Requirements).

Digital Processor Module Reliability Analysis of Nuclear Power Plant

International Nuclear Information System (INIS)

Lee, Sang Yong; Jung, Jae Hyun; Kim, Jae Ho; Kim, Sung Hun

2005-01-01

The system used in plant, military equipment, satellite, etc. consists of many electronic parts as control module, which requires relatively high reliability than other commercial electronic products. Specially, Nuclear power plant related to the radiation safety requires high safety and reliability, so most parts apply to Military-Standard level. Reliability prediction method provides the rational basis of system designs and also provides the safety significance of system operations. Thus various reliability prediction tools have been developed in recent decades, among of them, the MI-HDBK-217 method has been widely used as a powerful tool for the prediction. In this work, It is explained that reliability analysis work for Digital Processor Module (DPM, control module of SMART) is performed by Parts Stress Method based on MIL-HDBK-217F NOTICE2. We are using the Relex 7.6 of Relex software corporation, because reliability analysis process requires enormous part libraries and data for failure rate calculation

Preliminary Analysis of LORAN-C System Reliability for Civil Aviation.

Science.gov (United States)

1981-09-01

overviev of the analysis technique. Section 3 describes the computerized LORAN-C coverage model which is used extensively in the reliability analysis...Xth Plenary Assembly, Geneva, 1963, published by International Telecomunications Union. S. Braff, R., Computer program to calculate a Karkov Chain Reliability Model, unpublished york, MITRE Corporation. A-1 I.° , 44J Ili *Y 0E 00 ...F i8 1110 Prelim inary Analysis of Program Engineering & LORAN’C System ReliabilityMaintenance Service i ~Washington. D.C.

Reliability analysis of digital I and C systems at KAERI

International Nuclear Information System (INIS)

Kim, Man Cheol

2013-01-01

This paper provides an overview of the ongoing research activities on a reliability analysis of digital instrumentation and control (I and C) systems of nuclear power plants (NPPs) performed by the Korea Atomic Energy Research Institute (KAERI). The research activities include the development of a new safety-critical software reliability analysis method by integrating the advantages of existing software reliability analysis methods, a fault coverage estimation method based on fault injection experiments, and a new human reliability analysis method for computer-based main control rooms (MCRs) based on human performance data from the APR-1400 full-scope simulator. The research results are expected to be used to address various issues such as the licensing issues related to digital I and C probabilistic safety assessment (PSA) for advanced digital-based NPPs. (author)

Reliability analysis of stiff versus flexible piping

International Nuclear Information System (INIS)

Lu, S.C.

1985-01-01

The overall objective of this research project is to develop a technical basis for flexible piping designs which will improve piping reliability and minimize the use of pipe supports, snubbers, and pipe whip restraints. The current study was conducted to establish the necessary groundwork based on the piping reliability analysis. A confirmatory piping reliability assessment indicated that removing rigid supports and snubbers tends to either improve or affect very little the piping reliability. The authors then investigated a couple of changes to be implemented in Regulatory Guide (RG) 1.61 and RG 1.122 aimed at more flexible piping design. They concluded that these changes substantially reduce calculated piping responses and allow piping redesigns with significant reduction in number of supports and snubbers without violating ASME code requirements. Furthermore, the more flexible piping redesigns are capable of exhibiting reliability levels equal to or higher than the original stiffer design. An investigation of the malfunction of pipe whip restraints confirmed that the malfunction introduced higher thermal stresses and tended to reduce the overall piping reliability. Finally, support and component reliabilities were evaluated based on available fragility data. Results indicated that the support reliability usually exhibits a moderate decrease as the piping flexibility increases. Most on-line pumps and valves showed an insignificant reduction in reliability for a more flexible piping design

Reliability analysis for Atucha II reactor protection system signals

International Nuclear Information System (INIS)

Roca, Jose Luis

1996-01-01

Atucha II is a 745 MW Argentine Power Nuclear Reactor constructed by ENACE SA, Nuclear Argentine Company for Electrical Power Generation and SIEMENS AG KWU, Erlangen, Germany. A preliminary modular logic analysis of RPS (Reactor Protection System) signals was performed by means of the well known Swedish professional risk and reliability software named Risk-Spectrum taking as a basis a reference signal coded as JR17ER003 which command the two moderator loops valves. From the reliability and behavior knowledge for this reference signal follows an estimation of the reliability for the other 97 RPS signals. Because the preliminary character of this analysis Main Important Measures are not performed at this stage. Reliability is by the statistic value named unavailability predicted. The scope of this analysis is restricted from the measurement elements to the RPS buffer outputs. In the present context only one redundancy is analyzed so in the Instrumentation and Control area there no CCF (Common Cause Failures) present for signals. Finally those unavailability values could be introduced in the failure domain for the posterior complete Atucha II reliability analysis which includes all mechanical and electromechanical features. Also an estimation of the spurious frequency of RPS signals defined as faulty by no trip is performed

Reliability analysis for Atucha II reactor protection system signals

International Nuclear Information System (INIS)

Roca, Jose L.

2000-01-01

Atucha II is a 745 MW Argentine power nuclear reactor constructed by Nuclear Argentine Company for Electric Power Generation S.A. (ENACE S.A.) and SIEMENS AG KWU, Erlangen, Germany. A preliminary modular logic analysis of RPS (Reactor Protection System) signals was performed by means of the well known Swedish professional risk and reliability software named Risk-Spectrum taking as a basis a reference signal coded as JR17ER003 which command the two moderator loops valves. From the reliability and behavior knowledge for this reference signal follows an estimation of the reliability for the other 97 RPS signals. Because the preliminary character of this analysis Main Important Measures are not performed at this stage. Reliability is by the statistic value named unavailability predicted. The scope of this analysis is restricted from the measurement elements to the RPS buffer outputs. In the present context only one redundancy is analyzed so in the Instrumentation and Control area there no CCF (Common Cause Failures) present for signals. Finally those unavailability values could be introduced in the failure domain for the posterior complete Atucha II reliability analysis which includes all mechanical and electromechanical features. Also an estimation of the spurious frequency of RPS signals defined as faulty by no trip is performed. (author)

Comparing cultural profiles of MCR operators with those of non-MCR operators working in domestic Nuclear Power Plants

International Nuclear Information System (INIS)

Park, Jinkyun; Jung, Wondea

2015-01-01

Operation experience of complex socio-technical systems such as Nuclear Power Plants (NPPs) shows that most significant events are attributable to human error. Thus, various kinds of Human Reliability Analysis (HRA) techniques were used to manage human error for safety-critical tasks. However, a lack of available HRA data is a critical issue in conducting an HRA. For this reason, many researchers have tried to provide HRA data extracted from simulated emergency conditions. Unfortunately, it is still doubtful to directly use these HRA data because different operational cultures may result in different human performances even under a similar task context. From this concern, previous studies claimed that Main Control Room (MCR) operators working in different NPPs share very similar cultural profile. In order to confirm this claim, the culture profiles of operating personnel (i.e., non-MCR and MCR operators) working in domestic NPPs are compared. As a result, although some discrepancies are observed, it is positive to say that operating personnel of NPPs share similar cultural profiles to some extent. This result can be regarded as the first step to provide technical underpinnings that are helpful for understanding human performance data collected from different countries. - Highlights: • National and organizational culture of MCR and non-MCR operators are compared. • Organizational culture profiles are almost identical. • National culture profiles are not quite different overall. • Operating personnel in Nuclear Power Plants seems to share similar cultural profiles

Interactive reliability analysis project. FY 80 progress report

International Nuclear Information System (INIS)

Rasmuson, D.M.; Shepherd, J.C.

1981-03-01

This report summarizes the progress to date in the interactive reliability analysis project. Purpose is to develop and demonstrate a reliability and safety technique that can be incorporated early in the design process. Details are illustrated in a simple example of a reactor safety system

78 FR 45447 - Revisions to Modeling, Data, and Analysis Reliability Standard

Science.gov (United States)

2013-07-29

...; Order No. 782] Revisions to Modeling, Data, and Analysis Reliability Standard AGENCY: Federal Energy... Analysis (MOD) Reliability Standard MOD- 028-2, submitted to the Commission for approval by the North... Organization. The Commission finds that the proposed Reliability Standard represents an improvement over the...

State of the art report on aging reliability analysis

International Nuclear Information System (INIS)

Choi, Sun Yeong; Yang, Joon Eon; Han, Sang Hoon; Ha, Jae Joo

2002-03-01

The goal of this report is to describe the state of the art on aging analysis methods to calculate the effects of component aging quantitatively. In this report, we described some aging analysis methods which calculate the increase of Core Damage Frequency (CDF) due to aging by including the influence of aging into PSA. We also described several research topics required for aging analysis for components of domestic NPPs. We have described a statistical model and reliability physics model which calculate the effect of aging quantitatively by using PSA method. It is expected that the practical use of the reliability-physics model will be increased though the process with the reliability-physics model is more complicated than statistical model

Reliability of the Emergency Severity Index: Meta-analysis

Directory of Open Access Journals (Sweden)

Amir Mirhaghi

2015-01-01

Full Text Available Objectives: Although triage systems based on the Emergency Severity Index (ESI have many advantages in terms of simplicity and clarity, previous research has questioned their reliability in practice. Therefore, the aim of this meta-analysis was to determine the reliability of ESI triage scales. Methods: This metaanalysis was performed in March 2014. Electronic research databases were searched and articles conforming to the Guidelines for Reporting Reliability and Agreement Studies were selected. Two researchers independently examined selected abstracts. Data were extracted in the following categories: version of scale (latest/older, participants (adult/paediatric, raters (nurse, physician or expert, method of reliability (intra/inter-rater, reliability statistics (weighted/unweighted kappa and the origin and publication year of the study. The effect size was obtained by the Z-transformation of reliability coefficients. Data were pooled with random-effects models and a meta-regression was performed based on the method of moments estimator. Results: A total of 19 studies from six countries were included in the analysis. The pooled coefficient for the ESI triage scales was substantial at 0.791 (95% confidence interval: 0.787‒0.795. Agreement was higher with the latest and adult versions of the scale and among expert raters, compared to agreement with older and paediatric versions of the scales and with other groups of raters, respectively. Conclusion: ESI triage scales showed an acceptable level of overall reliability. However, ESI scales require more development in order to see full agreement from all rater groups. Further studies concentrating on other aspects of reliability assessment are needed.

Reliability analysis in interdependent smart grid systems

Science.gov (United States)

Peng, Hao; Kan, Zhe; Zhao, Dandan; Han, Jianmin; Lu, Jianfeng; Hu, Zhaolong

2018-06-01

Complex network theory is a useful way to study many real complex systems. In this paper, a reliability analysis model based on complex network theory is introduced in interdependent smart grid systems. In this paper, we focus on understanding the structure of smart grid systems and studying the underlying network model, their interactions, and relationships and how cascading failures occur in the interdependent smart grid systems. We propose a practical model for interdependent smart grid systems using complex theory. Besides, based on percolation theory, we also study the effect of cascading failures effect and reveal detailed mathematical analysis of failure propagation in such systems. We analyze the reliability of our proposed model caused by random attacks or failures by calculating the size of giant functioning components in interdependent smart grid systems. Our simulation results also show that there exists a threshold for the proportion of faulty nodes, beyond which the smart grid systems collapse. Also we determine the critical values for different system parameters. In this way, the reliability analysis model based on complex network theory can be effectively utilized for anti-attack and protection purposes in interdependent smart grid systems.

Adjoint sensitivity analysis of dynamic reliability models based on Markov chains - II: Application to IFMIF reliability assessment

Energy Technology Data Exchange (ETDEWEB)

Cacuci, D. G. [Commiss Energy Atom, Direct Energy Nucl, Saclay, (France); Cacuci, D. G.; Balan, I. [Univ Karlsruhe, Inst Nucl Technol and Reactor Safetly, Karlsruhe, (Germany); Ionescu-Bujor, M. [Forschungszentrum Karlsruhe, Fus Program, D-76021 Karlsruhe, (Germany)

2008-07-01

In Part II of this work, the adjoint sensitivity analysis procedure developed in Part I is applied to perform sensitivity analysis of several dynamic reliability models of systems of increasing complexity, culminating with the consideration of the International Fusion Materials Irradiation Facility (IFMIF) accelerator system. Section II presents the main steps of a procedure for the automated generation of Markov chains for reliability analysis, including the abstraction of the physical system, construction of the Markov chain, and the generation and solution of the ensuing set of differential equations; all of these steps have been implemented in a stand-alone computer code system called QUEFT/MARKOMAG-S/MCADJSEN. This code system has been applied to sensitivity analysis of dynamic reliability measures for a paradigm '2-out-of-3' system comprising five components and also to a comprehensive dynamic reliability analysis of the IFMIF accelerator system facilities for the average availability and, respectively, the system's availability at the final mission time. The QUEFT/MARKOMAG-S/MCADJSEN has been used to efficiently compute sensitivities to 186 failure and repair rates characterizing components and subsystems of the first-level fault tree of the IFMIF accelerator system. (authors)

Adjoint sensitivity analysis of dynamic reliability models based on Markov chains - II: Application to IFMIF reliability assessment

International Nuclear Information System (INIS)

Cacuci, D. G.; Cacuci, D. G.; Balan, I.; Ionescu-Bujor, M.

2008-01-01

In Part II of this work, the adjoint sensitivity analysis procedure developed in Part I is applied to perform sensitivity analysis of several dynamic reliability models of systems of increasing complexity, culminating with the consideration of the International Fusion Materials Irradiation Facility (IFMIF) accelerator system. Section II presents the main steps of a procedure for the automated generation of Markov chains for reliability analysis, including the abstraction of the physical system, construction of the Markov chain, and the generation and solution of the ensuing set of differential equations; all of these steps have been implemented in a stand-alone computer code system called QUEFT/MARKOMAG-S/MCADJSEN. This code system has been applied to sensitivity analysis of dynamic reliability measures for a paradigm '2-out-of-3' system comprising five components and also to a comprehensive dynamic reliability analysis of the IFMIF accelerator system facilities for the average availability and, respectively, the system's availability at the final mission time. The QUEFT/MARKOMAG-S/MCADJSEN has been used to efficiently compute sensitivities to 186 failure and repair rates characterizing components and subsystems of the first-level fault tree of the IFMIF accelerator system. (authors)

A reliability simulation language for reliability analysis

International Nuclear Information System (INIS)

Deans, N.D.; Miller, A.J.; Mann, D.P.

1986-01-01

The results of work being undertaken to develop a Reliability Description Language (RDL) which will enable reliability analysts to describe complex reliability problems in a simple, clear and unambiguous way are described. Component and system features can be stated in a formal manner and subsequently used, along with control statements to form a structured program. The program can be compiled and executed on a general-purpose computer or special-purpose simulator. (DG)

Durability reliability analysis for corroding concrete structures under uncertainty

Science.gov (United States)

Zhang, Hao

2018-02-01

This paper presents a durability reliability analysis of reinforced concrete structures subject to the action of marine chloride. The focus is to provide insight into the role of epistemic uncertainties on durability reliability. The corrosion model involves a number of variables whose probabilistic characteristics cannot be fully determined due to the limited availability of supporting data. All sources of uncertainty, both aleatory and epistemic, should be included in the reliability analysis. Two methods are available to formulate the epistemic uncertainty: the imprecise probability-based method and the purely probabilistic method in which the epistemic uncertainties are modeled as random variables. The paper illustrates how the epistemic uncertainties are modeled and propagated in the two methods, and shows how epistemic uncertainties govern the durability reliability.

Structural reliability analysis based on the cokriging technique

International Nuclear Information System (INIS)

Zhao Wei; Wang Wei; Dai Hongzhe; Xue Guofeng

2010-01-01

Approximation methods are widely used in structural reliability analysis because they are simple to create and provide explicit functional relationships between the responses and variables in stead of the implicit limit state function. Recently, the kriging method which is a semi-parameter interpolation technique that can be used for deterministic optimization and structural reliability has gained popularity. However, to fully exploit the kriging method, especially in high-dimensional problems, a large number of sample points should be generated to fill the design space and this can be very expensive and even impractical in practical engineering analysis. Therefore, in this paper, a new method-the cokriging method, which is an extension of kriging, is proposed to calculate the structural reliability. cokriging approximation incorporates secondary information such as the values of the gradients of the function being approximated. This paper explores the use of the cokriging method for structural reliability problems by comparing it with the Kriging method based on some numerical examples. The results indicate that the cokriging procedure described in this work can generate approximation models to improve on the accuracy and efficiency for structural reliability problems and is a viable alternative to the kriging.

Reliability analysis of Angra I safety systems

International Nuclear Information System (INIS)

Oliveira, L.F.S. de; Soto, J.B.; Maciel, C.C.; Gibelli, S.M.O.; Fleming, P.V.; Arrieta, L.A.

1980-07-01

An extensive reliability analysis of some safety systems of Angra I, are presented. The fault tree technique, which has been successfully used in most reliability studies of nuclear safety systems performed to date is employed. Results of a quantitative determination of the unvailability of the accumulator and the containment spray injection systems are presented. These results are also compared to those reported in WASH-1400. (E.G.) [pt

Development of a taxonomy of performance influencing factors for human reliability assessment of accident management tasks and its application

International Nuclear Information System (INIS)

Kim, Jae Whan; Jung, Won Dae; Kang, Dae Il; Ha, Jae Joo

1999-06-01

In this study, a new PIF taxonomy for HRA of the tasks during emergency operation and accident management situations. We collected the existing PIF taxonomies as many as possible. Then, we analyzed the trend in the selection of PIFs, the frequency of use between PIFs in HRA methods, and the level of definition of PIFs, in order to reflect these characteristics into the development of a new PIF taxonomy. Next, we analyzed the principal task context during accident management to draw the context specific PIFs. Afterwards, we established several criteria for the selection of the appropriate PIFs for HRA under emergency operation and accident management situations. Finally, the final PIF taxonomy containing the subitems for assessing each PIF was constructed based on the results of the previous steps and the selection criteria. The final result of this study is the new PIF taxonomy for HRA of the tasks during emergency operation and accident management situations. The selected 11 PIFs in the study are as follows: training and experience, availability and quality of information, status and trend of critical parameters, status of safety system/component, time pressure, working environment features, team cooperation and communication, plant policy and safety culture. (author). 35 refs., 23 tabs

Development of a taxonomy of performance influencing factors for human reliability assessment of accident management tasks and its application

Energy Technology Data Exchange (ETDEWEB)

Kim, Jae Whan; Jung, Won Dae; Kang, Dae Il; Ha, Jae Joo

1999-06-01

In this study, a new PIF taxonomy for HRA of the tasks during emergency operation and accident management situations. We collected the existing PIF taxonomies as many as possible. Then, we analyzed the trend in the selection of PIFs, the frequency of use between PIFs in HRA methods, and the level of definition of PIFs, in order to reflect these characteristics into the development of a new PIF taxonomy. Next, we analyzed the principal task context during accident management to draw the context specific PIFs. Afterwards, we established several criteria for the selection of the appropriate PIFs for HRA under emergency operation and accident management situations. Finally, the final PIF taxonomy containing the subitems for assessing each PIF was constructed based on the results of the previous steps and the selection criteria. The final result ofthis study is the new PIF taxonomy for HRA of the tasks during emergency operation and accident management situations. The selected 11 PIFs in the study are as follows: training and experience, availability and quality of information, status and trend of critical parameters, status of safety system/component, time pressure, working environment features, team cooperation and communication, plant policy and safety culture. (author). 35 refs., 23 tabs.

Reliability analysis of RC containment structures under combined loads

International Nuclear Information System (INIS)

Hwang, H.; Reich, M.; Kagami, S.

1984-01-01

This paper discusses a reliability analysis method and load combination design criteria for reinforced concrete containment structures under combined loads. The probability based reliability analysis method is briefly described. For load combination design criteria, derivations of the load factors for accidental pressure due to a design basis accident and safe shutdown earthquake (SSE) for three target limit state probabilities are presented

Discrete event simulation versus conventional system reliability analysis approaches

DEFF Research Database (Denmark)

Kozine, Igor

2010-01-01

Discrete Event Simulation (DES) environments are rapidly developing and appear to be promising tools for building reliability and risk analysis models of safety-critical systems and human operators. If properly developed, they are an alternative to the conventional human reliability analysis models...... and systems analysis methods such as fault and event trees and Bayesian networks. As one part, the paper describes briefly the author’s experience in applying DES models to the analysis of safety-critical systems in different domains. The other part of the paper is devoted to comparing conventional approaches...

Sensitivity analysis in a structural reliability context

International Nuclear Information System (INIS)

Lemaitre, Paul

2014-01-01

This thesis' subject is sensitivity analysis in a structural reliability context. The general framework is the study of a deterministic numerical model that allows to reproduce a complex physical phenomenon. The aim of a reliability study is to estimate the failure probability of the system from the numerical model and the uncertainties of the inputs. In this context, the quantification of the impact of the uncertainty of each input parameter on the output might be of interest. This step is called sensitivity analysis. Many scientific works deal with this topic but not in the reliability scope. This thesis' aim is to test existing sensitivity analysis methods, and to propose more efficient original methods. A bibliographical step on sensitivity analysis on one hand and on the estimation of small failure probabilities on the other hand is first proposed. This step raises the need to develop appropriate techniques. Two variables ranking methods are then explored. The first one proposes to make use of binary classifiers (random forests). The second one measures the departure, at each step of a subset method, between each input original density and the density given the subset reached. A more general and original methodology reflecting the impact of the input density modification on the failure probability is then explored. The proposed methods are then applied on the CWNR case, which motivates this thesis. (author)

Reliability and risk analysis methods research plan

International Nuclear Information System (INIS)

1984-10-01

This document presents a plan for reliability and risk analysis methods research to be performed mainly by the Reactor Risk Branch (RRB), Division of Risk Analysis and Operations (DRAO), Office of Nuclear Regulatory Research. It includes those activities of other DRAO branches which are very closely related to those of the RRB. Related or interfacing programs of other divisions, offices and organizations are merely indicated. The primary use of this document is envisioned as an NRC working document, covering about a 3-year period, to foster better coordination in reliability and risk analysis methods development between the offices of Nuclear Regulatory Research and Nuclear Reactor Regulation. It will also serve as an information source for contractors and others to more clearly understand the objectives, needs, programmatic activities and interfaces together with the overall logical structure of the program

Representative Sampling for reliable data analysis

DEFF Research Database (Denmark)

Petersen, Lars; Esbensen, Kim Harry

2005-01-01

regime in order to secure the necessary reliability of: samples (which must be representative, from the primary sampling onwards), analysis (which will not mean anything outside the miniscule analytical volume without representativity ruling all mass reductions involved, also in the laboratory) and data...

A double-loop adaptive sampling approach for sensitivity-free dynamic reliability analysis

International Nuclear Information System (INIS)

Wang, Zequn; Wang, Pingfeng

2015-01-01

Dynamic reliability measures reliability of an engineered system considering time-variant operation condition and component deterioration. Due to high computational costs, conducting dynamic reliability analysis at an early system design stage remains challenging. This paper presents a confidence-based meta-modeling approach, referred to as double-loop adaptive sampling (DLAS), for efficient sensitivity-free dynamic reliability analysis. The DLAS builds a Gaussian process (GP) model sequentially to approximate extreme system responses over time, so that Monte Carlo simulation (MCS) can be employed directly to estimate dynamic reliability. A generic confidence measure is developed to evaluate the accuracy of dynamic reliability estimation while using the MCS approach based on developed GP models. A double-loop adaptive sampling scheme is developed to efficiently update the GP model in a sequential manner, by considering system input variables and time concurrently in two sampling loops. The model updating process using the developed sampling scheme can be terminated once the user defined confidence target is satisfied. The developed DLAS approach eliminates computationally expensive sensitivity analysis process, thus substantially improves the efficiency of dynamic reliability analysis. Three case studies are used to demonstrate the efficacy of DLAS for dynamic reliability analysis. - Highlights: • Developed a novel adaptive sampling approach for dynamic reliability analysis. • POD Developed a new metric to quantify the accuracy of dynamic reliability estimation. • Developed a new sequential sampling scheme to efficiently update surrogate models. • Three case studies were used to demonstrate the efficacy of the new approach. • Case study results showed substantially enhanced efficiency with high accuracy

DiapHRaGM: A mnemonic to describe the work of breathing in patients with respiratory failure.

Directory of Open Access Journals (Sweden)

Aiman Tulaimat

Full Text Available The assessment of the work of breathing in the definitions of respiratory failure is vague and variable.Identify a parsimonious set of signs to describe the work of breathing in hypoxemic, acutely ill patients.We examined consecutive medical ICU patients receiving oxygen with a mask, non-invasive ventilation, or T-piece. A physician inspected each patient for 10 seconds, rated the level of respiratory distress, and then examined the patient for vital signs and 17 other physical signs. We used the rating of distress as a surrogate for measuring the work of breathing, constructed three multivariate models to identify the one with the smallest number of signs and largest explained variance, and validated it with bootstrap analysis.We performed 402 observations on 240 patients. Respiratory distress was absent in 78, mild in 157, moderate in 107, and severe in 60. Respiratory rate, hypoxia, heart rate, and frequency of most signs increased as distress increased. Respiratory rate and hypoxia explained 43% of the variance in respiratory distress. Diaphoresis, gasping, and contraction of the sternomastoid explained an additional 28%. Heart rate, blood pressure, alertness, agitation, body posture, nasal flaring, audible breathing, cyanosis, tracheal tug, retractions, paradox, scalene or abdominal muscles contraction did not increase the explained variance in respiratory distress.Most of the variance is respiratory distress can be explained by five signs summarized by the mnemonic DiapHRaGM (diaphoresis, hypoxia, respiratory rate, gasping, accessory muscle. This set of signs may allow for efficient, standardized assessments of the work of breathing of hypoxic patients.

Reliability Analysis of Fatigue Fracture of Wind Turbine Drivetrain Components

DEFF Research Database (Denmark)

Berzonskis, Arvydas; Sørensen, John Dalsgaard

2016-01-01

in the volume of the casted ductile iron main shaft, on the reliability of the component. The probabilistic reliability analysis conducted is based on fracture mechanics models. Additionally, the utilization of the probabilistic reliability for operation and maintenance planning and quality control is discussed....

Reliability analysis of cluster-based ad-hoc networks

International Nuclear Information System (INIS)

Cook, Jason L.; Ramirez-Marquez, Jose Emmanuel

2008-01-01

The mobile ad-hoc wireless network (MAWN) is a new and emerging network scheme that is being employed in a variety of applications. The MAWN varies from traditional networks because it is a self-forming and dynamic network. The MAWN is free of infrastructure and, as such, only the mobile nodes comprise the network. Pairs of nodes communicate either directly or through other nodes. To do so, each node acts, in turn, as a source, destination, and relay of messages. The virtue of a MAWN is the flexibility this provides; however, the challenge for reliability analyses is also brought about by this unique feature. The variability and volatility of the MAWN configuration makes typical reliability methods (e.g. reliability block diagram) inappropriate because no single structure or configuration represents all manifestations of a MAWN. For this reason, new methods are being developed to analyze the reliability of this new networking technology. New published methods adapt to this feature by treating the configuration probabilistically or by inclusion of embedded mobility models. This paper joins both methods together and expands upon these works by modifying the problem formulation to address the reliability analysis of a cluster-based MAWN. The cluster-based MAWN is deployed in applications with constraints on networking resources such as bandwidth and energy. This paper presents the problem's formulation, a discussion of applicable reliability metrics for the MAWN, and illustration of a Monte Carlo simulation method through the analysis of several example networks

A Review: Passive System Reliability Analysis – Accomplishments and Unresolved Issues

Energy Technology Data Exchange (ETDEWEB)

Nayak, Arun Kumar, E-mail: arunths@barc.gov.in [Reactor Engineering Division, Reactor Design and Development Group, Bhabha Atomic Research Centre, Mumbai (India); Chandrakar, Amit [Homi Bhabha National Institute, Mumbai (India); Vinod, Gopika [Reactor Safety Division, Reactor Design and Development Group, Bhabha Atomic Research Centre, Mumbai (India)

2014-10-10

Reliability assessment of passive safety systems is one of the important issues, since safety of advanced nuclear reactors rely on several passive features. In this context, a few methodologies such as reliability evaluation of passive safety system (REPAS), reliability methods for passive safety functions (RMPS), and analysis of passive systems reliability (APSRA) have been developed in the past. These methodologies have been used to assess reliability of various passive safety systems. While these methodologies have certain features in common, but they differ in considering certain issues; for example, treatment of model uncertainties, deviation of geometric, and process parameters from their nominal values. This paper presents the state of the art on passive system reliability assessment methodologies, the accomplishments, and remaining issues. In this review, three critical issues pertaining to passive systems performance and reliability have been identified. The first issue is applicability of best estimate codes and model uncertainty. The best estimate codes based phenomenological simulations of natural convection passive systems could have significant amount of uncertainties, these uncertainties must be incorporated in appropriate manner in the performance and reliability analysis of such systems. The second issue is the treatment of dynamic failure characteristics of components of passive systems. REPAS, RMPS, and APSRA methodologies do not consider dynamic failures of components or process, which may have strong influence on the failure of passive systems. The influence of dynamic failure characteristics of components on system failure probability is presented with the help of a dynamic reliability methodology based on Monte Carlo simulation. The analysis of a benchmark problem of Hold-up tank shows the error in failure probability estimation by not considering the dynamism of components. It is thus suggested that dynamic reliability methodologies must be

Analysis and assessment of water treatment plant reliability

Directory of Open Access Journals (Sweden)

Szpak Dawid

2017-03-01

Full Text Available The subject of the publication is the analysis and assessment of the reliability of the surface water treatment plant (WTP. In the study the one parameter method of reliability assessment was used. Based on the flow sheet derived from the water company the reliability scheme of the analysed WTP was prepared. On the basis of the daily WTP work report the availability index Kg for the individual elements included in the WTP, was determined. Then, based on the developed reliability scheme showing the interrelationships between elements, the availability index Kg for the whole WTP was determined. The obtained value of the availability index Kg was compared with the criteria values.

Time-dependent reliability analysis of nuclear reactor operators using probabilistic network models

International Nuclear Information System (INIS)

Oka, Y.; Miyata, K.; Kodaira, H.; Murakami, S.; Kondo, S.; Togo, Y.

1987-01-01

Human factors are very important for the reliability of a nuclear power plant. Human behavior has essentially a time-dependent nature. The details of thinking and decision making processes are important for detailed analysis of human reliability. They have, however, not been well considered by the conventional methods of human reliability analysis. The present paper describes the models for the time-dependent and detailed human reliability analysis. Recovery by an operator is taken into account and two-operators models are also presented

Procedure for conducting a human-reliability analysis for nuclear power plants. Final report

International Nuclear Information System (INIS)

Bell, B.J.; Swain, A.D.

1983-05-01

This document describes in detail a procedure to be followed in conducting a human reliability analysis as part of a probabilistic risk assessment when such an analysis is performed according to the methods described in NUREG/CR-1278, Handbook for Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications. An overview of the procedure describing the major elements of a human reliability analysis is presented along with a detailed description of each element and an example of an actual analysis. An appendix consists of some sample human reliability analysis problems for further study

Root cause analysis in support of reliability enhancement of engineering components

International Nuclear Information System (INIS)

Kumar, Sachin; Mishra, Vivek; Joshi, N.S.; Varde, P.V.

2014-01-01

Reliability based methods have been widely used for the safety assessment of plant system, structures and components. These methods provide a quantitative estimation of system reliability but do not give insight into the failure mechanism. Understanding the failure mechanism is a must to avoid the recurrence of the events and enhancement of the system reliability. Root cause analysis provides a tool for gaining detailed insights into the causes of failure of component with particular attention to the identification of fault in component design, operation, surveillance, maintenance, training, procedures and policies which must be improved to prevent repetition of incidents. Root cause analysis also helps in developing Probabilistic Safety Analysis models. A probabilistic precursor study provides a complement to the root cause analysis approach in event analysis by focusing on how an event might have developed adversely. This paper discusses the root cause analysis methodologies and their application in the specific case studies for enhancement of system reliability. (author)

DATMAN: A reliability data analysis program using Bayesian updating

International Nuclear Information System (INIS)

Becker, M.; Feltus, M.A.

1996-01-01

Preventive maintenance (PM) techniques focus on the prevention of failures, in particular, system components that are important to plant functions. Reliability-centered maintenance (RCM) improves on the PM techniques by introducing a set of guidelines by which to evaluate the system functions. It also minimizes intrusive maintenance, labor, and equipment downtime without sacrificing system performance when its function is essential for plant safety. Both the PM and RCM approaches require that system reliability data be updated as more component failures and operation time are acquired. Systems reliability and the likelihood of component failures can be calculated by Bayesian statistical methods, which can update these data. The DATMAN computer code has been developed at Penn State to simplify the Bayesian analysis by performing tedious calculations needed for RCM reliability analysis. DATMAN reads data for updating, fits a distribution that best fits the data, and calculates component reliability. DATMAN provides a user-friendly interface menu that allows the user to choose from several common prior and posterior distributions, insert new failure data, and visually select the distribution that matches the data most accurately

Human event observations in the individual plant examinations

International Nuclear Information System (INIS)

Forester, J.

1995-01-01

A major objective of the Nuclear Regulatory Commission's (NRC) Individual Plant Examination (IPE) Insights Program is to identify the important determinants of core damage frequency (CDF) for the different reactor and containment types and plant designs as indicated in the IPEs. The human reliability analysis (HRA) is a critical component of the probabilistic risk assessments (PRAS) which were done for the IPES. The determination and selection of human actions for incorporation into the event and fault tree models and the quantification of their failure probabilities can have an important impact on the resulting estimates of CDF and risk. Therefore, two important goals of the NRCs IPE Insights Program are (1) to determine the extent to which human actions and their corresponding failure probabilities influenced the results of the IPEs and (2) to identify which factors played significant roles in determining the differences and similarities in the results of the HRA analyses across the different plants. To obtain the relevant information, the NRC's IPE database, which contains information on plant design, CDF, and containment performance obtained from the IPES, was used in conjunction with a systematic examination of the HRA analyses and results from the IPES. Regarding the extent to which the results of the HRA analyses were significant contributors to the plants' CDFs, examinations of several different measures indicated that while individual human actions could have important influences on CDF for particular initiators, the HRA results did not appear to be the most significant driver of plant risk (CDF). Another finding was that while there were relatively wide variations in the calculated human error probabilities (HEPs) for similar events across plants, there was no evidence for any systematic variation as a function of the HRA methods used in the analyses

The development of a reliable amateur boxing performance analysis template.

Science.gov (United States)

Thomson, Edward; Lamb, Kevin; Nicholas, Ceri

2013-01-01

The aim of this study was to devise a valid performance analysis system for the assessment of the movement characteristics associated with competitive amateur boxing and assess its reliability using analysts of varying experience of the sport and performance analysis. Key performance indicators to characterise the demands of an amateur contest (offensive, defensive and feinting) were developed and notated using a computerised notational analysis system. Data were subjected to intra- and inter-observer reliability assessment using median sign tests and calculating the proportion of agreement within predetermined limits of error. For all performance indicators, intra-observer reliability revealed non-significant differences between observations (P > 0.05) and high agreement was established (80-100%) regardless of whether exact or the reference value of ±1 was applied. Inter-observer reliability was less impressive for both analysts (amateur boxer and experienced analyst), with the proportion of agreement ranging from 33-100%. Nonetheless, there was no systematic bias between observations for any indicator (P > 0.05), and the proportion of agreement within the reference range (±1) was 100%. A reliable performance analysis template has been developed for the assessment of amateur boxing performance and is available for use by researchers, coaches and athletes to classify and quantify the movement characteristics of amateur boxing.

Reliability Worth Analysis of Distribution Systems Using Cascade Correlation Neural Networks

DEFF Research Database (Denmark)

Heidari, Alireza; Agelidis, Vassilios; Pou, Josep

2018-01-01

Reliability worth analysis is of great importance in the area of distribution network planning and operation. The reliability worth's precision can be affected greatly by the customer interruption cost model used. The choice of the cost models can change system and load point reliability indices....... In this study, a cascade correlation neural network is adopted to further develop two cost models comprising a probabilistic distribution model and an average or aggregate model. A contingency-based analytical technique is adopted to conduct the reliability worth analysis. Furthermore, the possible effects...

A new approach for reliability analysis with time-variant performance characteristics

International Nuclear Information System (INIS)

Wang, Zequn; Wang, Pingfeng

2013-01-01

Reliability represents safety level in industry practice and may variant due to time-variant operation condition and components deterioration throughout a product life-cycle. Thus, the capability to perform time-variant reliability analysis is of vital importance in practical engineering applications. This paper presents a new approach, referred to as nested extreme response surface (NERS), that can efficiently tackle time dependency issue in time-variant reliability analysis and enable to solve such problem by easily integrating with advanced time-independent tools. The key of the NERS approach is to build a nested response surface of time corresponding to the extreme value of the limit state function by employing Kriging model. To obtain the data for the Kriging model, the efficient global optimization technique is integrated with the NERS to extract the extreme time responses of the limit state function for any given system input. An adaptive response prediction and model maturation mechanism is developed based on mean square error (MSE) to concurrently improve the accuracy and computational efficiency of the proposed approach. With the nested response surface of time, the time-variant reliability analysis can be converted into the time-independent reliability analysis and existing advanced reliability analysis methods can be used. Three case studies are used to demonstrate the efficiency and accuracy of NERS approach

Methodology for reliability allocation based on fault tree analysis and dualistic contrast

Institute of Scientific and Technical Information of China (English)

TONG Lili; CAO Xuewu

2008-01-01

Reliability allocation is a difficult multi-objective optimization problem.This paper presents a methodology for reliability allocation that can be applied to determine the reliability characteristics of reactor systems or subsystems.The dualistic contrast,known as one of the most powerful tools for optimization problems,is applied to the reliability allocation model of a typical system in this article.And the fault tree analysis,deemed to be one of the effective methods of reliability analysis,is also adopted.Thus a failure rate allocation model based on the fault tree analysis and dualistic contrast is achieved.An application on the emergency diesel generator in the nuclear power plant is given to illustrate the proposed method.

Reliability analysis of protection system of advanced pressurized water reactor - APR 1400

International Nuclear Information System (INIS)

Varde, P. V.; Choi, J. G.; Lee, D. Y.; Han, J. B.

2003-04-01