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Sample records for hra human reliability

  1. Space Mission Human Reliability Analysis (HRA) Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The purpose of this project is to extend current ground-based Human Reliability Analysis (HRA) techniques to a long-duration, space-based tool to more effectively...

  2. Space Mission Human Reliability Analysis (HRA) Project

    Science.gov (United States)

    Boyer, Roger

    2014-01-01

    The purpose of the Space Mission Human Reliability Analysis (HRA) Project is to extend current ground-based HRA risk prediction techniques to a long-duration, space-based tool. Ground-based HRA methodology has been shown to be a reasonable tool for short-duration space missions, such as Space Shuttle and lunar fly-bys. However, longer-duration deep-space missions, such as asteroid and Mars missions, will require the crew to be in space for as long as 400 to 900 day missions with periods of extended autonomy and self-sufficiency. Current indications show higher risk due to fatigue, physiological effects due to extended low gravity environments, and others, may impact HRA predictions. For this project, Safety & Mission Assurance (S&MA) will work with Human Health & Performance (HH&P) to establish what is currently used to assess human reliabiilty for human space programs, identify human performance factors that may be sensitive to long duration space flight, collect available historical data, and update current tools to account for performance shaping factors believed to be important to such missions. This effort will also contribute data to the Human Performance Data Repository and influence the Space Human Factors Engineering research risks and gaps (part of the HRP Program). An accurate risk predictor mitigates Loss of Crew (LOC) and Loss of Mission (LOM).The end result will be an updated HRA model that can effectively predict risk on long-duration missions.

  3. A perspective on Human Reliability Analysis (HRA) and studies on the application of HRA to Indian Pressurised Heavy Water Reactors

    International Nuclear Information System (INIS)

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

    2000-05-01

    Probabilistic studies of risks show that the human factor contributes significantly to overall risk. The potential for and mechanisms of human error to affect plant risk and safety is evaluated by Human Reliability Analysis (HRA). HRA has quantitative and qualitative aspects, both of which are useful for Human Factors Engineering (HFE) which aims at designing operator interfaces that will minimise operator error and provide for error detection and recovery capability. HRA has therefore to be conducted as an integrated activity in support of PSA and HFE design. The objectives of HRA therefore, are to assure that potential effects on plant safety and reliability are analysed and that human actions that are important to plant risk are identified so that they can be addressed in both PSA and plant design. This report is in two parts. The first part presents a comprehensive overview of HRA. It attempts to provide an understanding of how human failures are incorporated into PSA models and how HRA is performed. The focus is on the HRA process, frameworks, techniques and models. The second part begins with a discussion on the application of HRA to IPHWRs and then continues with the presentation of three specific HRA case studies. This work was carried out by the working group on HRA constituted by AERB. Part of the work was done under the aegis of the IAEA Coordinated Research Programme (CRP) on collection and classification of human reliability data and use in PSA - Research contract No. 8239/RB. (author)

  4. Study on Performance Shaping Factors (PSFs) Quantification Method in Human Reliability Analysis (HRA)

    International Nuclear Information System (INIS)

    Kim, Ar Ryum; Jang, Inseok Jang; Seong, Poong Hyun; Park, Jinkyun; Kim, Jong Hyun

    2015-01-01

    The purpose of HRA implementation is 1) to achieve the human factor engineering (HFE) design goal of providing operator interfaces that will minimize personnel errors and 2) to conduct an integrated activity to support probabilistic risk assessment (PRA). For these purposes, 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. In performing HRA, such conditions that influence human performances have been represented via several context factors called performance shaping factors (PSFs). 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. Most HRA methods evaluate the weightings of PSFs by expert judgment and explicit guidance for evaluating the weighting is not provided. It has been widely known that the performance of the human operator is one of the critical factors to determine the safe operation of NPPs. HRA methods have been developed to identify the possibility and mechanism of human errors. In performing HRA methods, the effect of PSFs which may increase or decrease human error should be investigated. However, the effect of PSFs were estimated by expert judgment so far. Accordingly, in order to estimate the effect of PSFs objectively, the quantitative framework to estimate PSFs by using PSF profiles is introduced in this paper

  5. Development of A Standard Method for Human Reliability Analysis (HRA) of Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kang, Dae Il; Jung, Won Dea; 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 and ANS PSA standard to ensure PSA quality, the standard HRA method was developed to meet the ASME and ANS 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

  6. Development of A Standard Method for Human Reliability Analysis (HRA) of Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Dae Il; Jung, Won Dea; 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 and ANS PSA standard to ensure PSA quality, the standard HRA method was developed to meet the ASME and ANS 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.

  7. Bridging Human Reliability Analysis and Psychology, Part 2: A Cognitive Framework to Support HRA

    Energy Technology Data Exchange (ETDEWEB)

    April M. Whaley; Stacey M. L. Hendrickson; Ronald L. Boring; Jing Xing

    2012-06-01

    This is the second of two papers that discuss the literature review conducted as part of the U.S. Nuclear Regulatory Commission (NRC) effort to develop a hybrid human reliability analysis (HRA) method in response to Staff Requirements Memorandum (SRM) SRM-M061020. This review was conducted with the goal of strengthening the technical basis within psychology, cognitive science and human factors for the hybrid HRA method being proposed. An overview of the literature review approach and high-level structure is provided in the first paper, whereas this paper presents the results of the review. The psychological literature review encompassed research spanning the entirety of human cognition and performance, and consequently produced an extensive list of psychological processes, mechanisms, and factors that contribute to human performance. To make sense of this large amount of information, the results of the literature review were organized into a cognitive framework that identifies causes of failure of macrocognition in humans, and connects those proximate causes to psychological mechanisms and performance influencing factors (PIFs) that can lead to the failure. This cognitive framework can serve as a tool to inform HRA. Beyond this, however, the cognitive framework has the potential to also support addressing human performance issues identified in Human Factors applications.

  8. Human Reliability Analysis. Applicability of the HRA-concept in maintenance shutdown

    International Nuclear Information System (INIS)

    Obenius, Aino

    2007-08-01

    Probabilistic Safety Analysis (PSA) is performed for Swedish nuclear power plants in order to make predictions and improvements of system safety. The analysis of the Three Mile Island and Chernobyl accidents contributed to broaden the approach to nuclear power plant safety. A system perspective focusing on the interaction between aspects of Man, Technology and Organization (MTO) emerged in addition to the development of Human Factors knowledge. To take the human influence on the technical system into consideration when performing PSAs, a Human Reliability Analysis (HRA) is performed. PSA is performed for different stages and plant operating states, and the current state of Swedish analyses is Low power and Shutdown (LPSD), also called Shutdown PSA (SPSA). The purpose of this master's thesis is to describe methods and basic models used when analysing human reliability for the LPSD state. The following questions are at issue: 1. How can the LPSD state be characterised and defined? 2. What is important to take into consideration when performing a LPSD HRA? 3. How can human behaviour be modelled for a LPSD risk analysis? 4. According to available empirical material, how are the questions above treated in performed analysis of human operation during LPSD? 5. How does the result of the questions above affect the way methods for analysis of LPSD could and/or should be developed? The procedure of this project has mainly consisted of literature studies of available theory for modelling of human behaviour and risk analysis of the LPSD state. This study regards analysis of planned outages when maintenance, fuel change, tests and inspections are performed. The outage period is characterised by planned maintenance activities performed in rotating 3-shifts, around the clock, as well as many of the persons performing work tasks on the plant being external contractors. The working conditions are characterised by stress due to heat, radiation and physically demanding or monotonous

  9. Human Reliability Analysis. Applicability of the HRA-concept in maintenance shutdown; Analys av maensklig tillfoerlitlighet. HRA-begreppets tillaempbarhet vid revisionsavstaellning

    Energy Technology Data Exchange (ETDEWEB)

    Obenius, Aino (MTO Psykologi AB, Stockholm (SE))

    2007-08-15

    Probabilistic Safety Analysis (PSA) is performed for Swedish nuclear power plants in order to make predictions and improvements of system safety. The analysis of the Three Mile Island and Chernobyl accidents contributed to broaden the approach to nuclear power plant safety. A system perspective focusing on the interaction between aspects of Man, Technology and Organization (MTO) emerged in addition to the development of Human Factors knowledge. To take the human influence on the technical system into consideration when performing PSAs, a Human Reliability Analysis (HRA) is performed. PSA is performed for different stages and plant operating states, and the current state of Swedish analyses is Low power and Shutdown (LPSD), also called Shutdown PSA (SPSA). The purpose of this master's thesis is to describe methods and basic models used when analysing human reliability for the LPSD state. The following questions are at issue: 1. How can the LPSD state be characterised and defined? 2. What is important to take into consideration when performing a LPSD HRA? 3. How can human behaviour be modelled for a LPSD risk analysis? 4. According to available empirical material, how are the questions above treated in performed analysis of human operation during LPSD? 5. How does the result of the questions above affect the way methods for analysis of LPSD could and/or should be developed? The procedure of this project has mainly consisted of literature studies of available theory for modelling of human behaviour and risk analysis of the LPSD state. This study regards analysis of planned outages when maintenance, fuel change, tests and inspections are performed. The outage period is characterised by planned maintenance activities performed in rotating 3-shifts, around the clock, as well as many of the persons performing work tasks on the plant being external contractors. The working conditions are characterised by stress due to heat, radiation and physically demanding or

  10. 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).

  11. 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

  12. Development of Human Performance Analysis and Advanced HRA Methodology

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Won Dea; Park, Jin Kyun; Kim, Jae Whan; Kim, Seong Whan; Kim, Man Cheol; Ha, Je Joo

    2007-06-15

    The purpose of this project is to build a systematic framework that can evaluate the effect of human factors related problems on the safety of nuclear power plants (NPPs) as well as develop a technology that can be used to enhance human performance. The research goal of this project is twofold: (1) the development of a human performance database and a framework to enhance human performance, and (2) the analysis of human error with constructing technical basis for human reliability analysis. There are three kinds of main results of this study. The first result is the development of a human performance database, called OPERA-I/II (Operator Performance and Reliability Analysis, Part I and Part II). In addition, a standard communication protocol was developed based on OPERA to reduce human error caused from communication error in the phase of event diagnosis. Task complexity (TACOM) measure and the methodology of optimizing diagnosis procedures were also finalized during this research phase. The second main result is the development of a software, K-HRA, which is to support the standard HRA method. Finally, an advanced HRA method named as AGAPE-ET was developed by combining methods MDTA (misdiagnosis tree analysis technique) and K-HRA, which can be used to analyze EOC (errors of commission) and EOO (errors of ommission). These research results, such as OPERA-I/II, TACOM, a standard communication protocol, K-HRA and AGAPE-ET methods will be used to improve the quality of HRA and to enhance human performance in nuclear power plants.

  13. Development of Human Performance Analysis and Advanced HRA Methodology

    International Nuclear Information System (INIS)

    Jung, Won Dea; Park, Jin Kyun; Kim, Jae Whan; Kim, Seong Whan; Kim, Man Cheol; Ha, Je Joo

    2007-06-01

    The purpose of this project is to build a systematic framework that can evaluate the effect of human factors related problems on the safety of nuclear power plants (NPPs) as well as develop a technology that can be used to enhance human performance. The research goal of this project is twofold: (1) the development of a human performance database and a framework to enhance human performance, and (2) the analysis of human error with constructing technical basis for human reliability analysis. There are three kinds of main results of this study. The first result is the development of a human performance database, called OPERA-I/II (Operator Performance and Reliability Analysis, Part I and Part II). In addition, a standard communication protocol was developed based on OPERA to reduce human error caused from communication error in the phase of event diagnosis. Task complexity (TACOM) measure and the methodology of optimizing diagnosis procedures were also finalized during this research phase. The second main result is the development of a software, K-HRA, which is to support the standard HRA method. Finally, an advanced HRA method named as AGAPE-ET was developed by combining methods MDTA (misdiagnosis tree analysis technique) and K-HRA, which can be used to analyze EOC (errors of commission) and EOO (errors of ommission). These research results, such as OPERA-I/II, TACOM, a standard communication protocol, K-HRA and AGAPE-ET methods will be used to improve the quality of HRA and to enhance human performance in nuclear power plants

  14. 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

  15. 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

  16. 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

  17. 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

  18. SCHEME (Soft Control Human error Evaluation MEthod) for advanced MCR HRA

    International Nuclear Information System (INIS)

    Jang, Inseok; Jung, Wondea; Seong, Poong Hyun

    2015-01-01

    The Technique for Human Error Rate Prediction (THERP), Korean Human Reliability Analysis (K-HRA), Human Error Assessment and Reduction Technique (HEART), A Technique for Human Event Analysis (ATHEANA), Cognitive Reliability and Error Analysis Method (CREAM), and Simplified Plant Analysis Risk Human Reliability Assessment (SPAR-H) in relation to NPP maintenance and operation. Most of these methods were developed considering the conventional type of Main Control Rooms (MCRs). They are still used for HRA in advanced MCRs even though the operating environment of advanced MCRs in NPPs has been considerably changed by the adoption of new human-system interfaces such as computer-based soft controls. Among the many features in advanced MCRs, soft controls are an important feature because the operation action in NPP advanced MCRs is performed by soft controls. Consequently, those conventional methods may not sufficiently consider the features of soft control execution human errors. To this end, a new framework of a HRA method for evaluating soft control execution human error is suggested by performing the soft control task analysis and the literature reviews regarding widely accepted human error taxonomies. In this study, the framework of a HRA method for evaluating soft control execution human error in advanced MCRs is developed. First, the factors which HRA method in advanced MCRs should encompass are derived based on the literature review, and soft control task analysis. Based on the derived factors, execution HRA framework in advanced MCRs is developed mainly focusing on the features of soft control. Moreover, since most current HRA database deal with operation in conventional type of MCRs and are not explicitly designed to deal with digital HSI, HRA database are developed under lab scale simulation

  19. 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.

  20. 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

  1. 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.

  2. 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

  3. 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

  4. 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

  5. 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)

  6. Consideration on HRA implementation during LPSD operation

    International Nuclear Information System (INIS)

    Kim, Ar Ryum; Kim, Jae Whan; Jang, In Seok; Seong, Poong Hyun

    2014-01-01

    During low power and shutdown operation, it has been pointed out that the important of human actions is significantly increased. Because automatic control may be disabled, most control room annunciation titles indicate alarm status, and procedures are insufficient, human operators plays a much larger role during outages and in lower power state. In this regards, in order to reduce human errors and secure nuclear power plant safety, it is necessary to identify and estimate human errors during LPSD operations. However, many researchers have argued that there is no comprehensive LPSD human reliability analysis (HRA) method so far. In this study, we reviewed and implemented the existing HRA methods during LPSD operations: Korean standard HRA (K-HRA), Simplified plant analysis risk HRA (SPAR-H), Caused based decision tree (CBDT), and human cognitive reliability/ operator reliability experiments (HCR/ORE). We reviewed HRA methods based on several reports related to the unique aspects of human operator during LPSD operation and HRA requirements. In addition, the existing HRA methods were implemented during various plant operating states (POS) of LPSD operation. Loss of shutdown cooling system (SCS) was selected as initiating event. As a result of review and implementation of HRA methods, we derived the some limitations of the existing HRA methods and related procedure during loss of SCS event

  7. HRA qualitative analysis in a nuclear power plant

    International Nuclear Information System (INIS)

    Dai Licao; Zhang Li; Huang Shudong

    2004-01-01

    Human reliability analysis (HRA) is a very important part of probability safety assessment (PSA) in a nuclear power plant. Qualitative analysis is the basis and starting point of HRA. The purpose, the principle, the method and the procedure of qualitative HRA are introduced. SGTR, a pressurized nuclear power plant as an example, is used to illustrate it. (authors)

  8. Bridging the gap between HRA research and HRA practice: A Bayesian network version of SPAR-H

    International Nuclear Information System (INIS)

    Groth, Katrina M.; Swiler, Laura P.

    2013-01-01

    The shortcomings of Human Reliability Analysis (HRA) have been a topic of discussion for over two decades. Repeated attempts to address these limitations have resulted in over 50 HRA methods, and the HRA research community continues to develop new methods. However, there remains a gap between the methods developed by HRA researchers and those actually used by HRA practitioners. Bayesian Networks (BNs) have become an increasingly popular part of the risk and reliability analysis framework over the past decade. BNs provide a framework for addressing many of the shortcomings of HRA from a researcher perspective and from a practitioner perspective. Several research groups have developed advanced HRA methods based on BNs, but none of these methods has been adopted by HRA practitioners in the U.S. nuclear power industry or at the U.S. Nuclear Regulatory Commission. In this paper we bridge the gap between HRA research and HRA practice by building a BN version of the widely used SPAR-H method. We demonstrate how the SPAR-H BN can be used by HRA practitioners, and we also demonstrate how it can be modified to incorporate data and information from research to advance HRA practice. The SPAR-H BN can be used as a starting point for translating HRA research efforts and advances in scientific understanding into real, timely benefits for HRA practitioners

  9. 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.

  10. 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.

  11. 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

  12. 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)

  13. 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

  14. 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

  15. 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.

  16. Design of Qualitative HRA Database Structure

    International Nuclear Information System (INIS)

    Kim, Seunghwan; Kim, Yochan; Choi, Sun Yeong; Park, Jinkyun; Jung, Wondea

    2015-01-01

    HRA DB is to collect and store the data in a database form to manage and maintain them from the perspective of human reliability analysis. All information on the human errors taken by operators in the power plant should be systematically collected and documented in its management. KAERI is developing the simulator-based HRA data handbook. In this study, the information required to store and manage the data necessary to perform an HRA as to store the HRA data to be stored in the handbook is identified and summarized. Especially this study is to summarize the collection and classification of qualitative data as the raw data to organize the data required to draw the HEP and its DB process. Qualitative HRA DB is a storehouse of all sub-information needed to receive the human error probability for Pasa. In this study, the requirements for structural design and implementation of qualitative HRA DB must be implemented for HRA DB were summarized. The follow-up study of the quantitative HRA DB implementation should be followed to draw the substantial HEP

  17. 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.)

  18. 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.

  19. Accident Sequence Evaluation Program: Human reliability analysis procedure

    Energy Technology Data Exchange (ETDEWEB)

    Swain, A.D.

    1987-02-01

    This document presents a shortened version of the procedure, models, and data for human reliability analysis (HRA) which are presented in the Handbook of Human Reliability Analysis With emphasis on Nuclear Power Plant Applications (NUREG/CR-1278, August 1983). This shortened version was prepared and tried out as part of the Accident Sequence Evaluation Program (ASEP) funded by the US Nuclear Regulatory Commission and managed by Sandia National Laboratories. The intent of this new HRA procedure, called the ''ASEP HRA Procedure,'' is to enable systems analysts, with minimal support from experts in human reliability analysis, to make estimates of human error probabilities and other human performance characteristics which are sufficiently accurate for many probabilistic risk assessments. The ASEP HRA Procedure consists of a Pre-Accident Screening HRA, a Pre-Accident Nominal HRA, a Post-Accident Screening HRA, and a Post-Accident Nominal HRA. The procedure in this document includes changes made after tryout and evaluation of the procedure in four nuclear power plants by four different systems analysts and related personnel, including human reliability specialists. The changes consist of some additional explanatory material (including examples), and more detailed definitions of some of the terms. 42 refs.

  20. Accident Sequence Evaluation Program: Human reliability analysis procedure

    International Nuclear Information System (INIS)

    Swain, A.D.

    1987-02-01

    This document presents a shortened version of the procedure, models, and data for human reliability analysis (HRA) which are presented in the Handbook of Human Reliability Analysis With emphasis on Nuclear Power Plant Applications (NUREG/CR-1278, August 1983). This shortened version was prepared and tried out as part of the Accident Sequence Evaluation Program (ASEP) funded by the US Nuclear Regulatory Commission and managed by Sandia National Laboratories. The intent of this new HRA procedure, called the ''ASEP HRA Procedure,'' is to enable systems analysts, with minimal support from experts in human reliability analysis, to make estimates of human error probabilities and other human performance characteristics which are sufficiently accurate for many probabilistic risk assessments. The ASEP HRA Procedure consists of a Pre-Accident Screening HRA, a Pre-Accident Nominal HRA, a Post-Accident Screening HRA, and a Post-Accident Nominal HRA. The procedure in this document includes changes made after tryout and evaluation of the procedure in four nuclear power plants by four different systems analysts and related personnel, including human reliability specialists. The changes consist of some additional explanatory material (including examples), and more detailed definitions of some of the terms. 42 refs

  1. 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

  2. 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)

  3. 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)

  4. 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.

  5. 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

  6. 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.

  7. 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.

  8. 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...

  9. Human reliability

    International Nuclear Information System (INIS)

    Embrey, D.E.

    1987-01-01

    Concepts and techniques of human reliability have been developed and are used mostly in probabilistic risk assessment. For this, the major application of human reliability assessment has been to identify the human errors which have a significant effect on the overall safety of the system and to quantify the probability of their occurrence. Some of the major issues within human reliability studies are reviewed and it is shown how these are applied to the assessment of human failures in systems. This is done under the following headings; models of human performance used in human reliability assessment, the nature of human error, classification of errors in man-machine systems, practical aspects, human reliability modelling in complex situations, quantification and examination of human reliability, judgement based approaches, holistic techniques and decision analytic approaches. (UK)

  10. 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.

  11. EXAM-HRA. A comparison between HRA approaches in Nordic countries and in Germany

    International Nuclear Information System (INIS)

    Becker, G.; Hussels, U.; Schubert, B.; Maennistoe, Ilkka

    2012-01-01

    A joint project called ''EXAM-HRA'' is performed by German, Swedish and Finnish nuclear power plants. The working group presently consists of: - Kent Bladh, RAB. - Anders Karlsson, FKA. - Guenter Becker and Marina Proske, RISA. - Ilkka Maennistoeand Jan-Erik Holmberg, VTT. - Lasse Tunturivuori, TVO. - Christian Bjursten Carlsson and Anders Olsson, Scandpower. - Gunnar Johanson and Lisa Fritzson, ES-konsult. The overall project objective is to provide guidance for a ''state of the art'' human reliability analysis (HRA) for purposes of probabilistic safety assessment (PSA) to ensure that plant specific properties are properly taken into consideration in the analysis /1/. This will also provide means to improve plant features based on HRA and PSA results as well as providing means to improve Nordic and German HRA application for PSA purposes. In previous phase 1, a method of comparing existing HRA analysis has been developed and applied to tasks, which have been modelled for various German and Nordic PSA. In phase 2, three plant visits have been performed in Brunsbuettel (Germany), Forsmark (Sweden), and Olkiluoto (Finland) to compare actions in more detail, and to perform additional analysis using specific variants of HRA methods. Although both German and Nordic partners strongly rely on THERP method, the variants of this HRA method turned out to differ in some aspects. Swedish and Finnish analyses focus on the diagnosis part of the action. They are based on the time correlation of Swain /2/, but they make use of additional performance shaping factors normally not used in Germany. German analyses have a less complex model for diagnosis, but a more thorough investigation of the various action steps, especially considering various types of recovery. This will be demonstrated by an example, which will show the different analyses and their influence on results. (orig.)

  12. 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

  13. 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)

  14. 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)

  15. 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

  16. Human reliability

    International Nuclear Information System (INIS)

    Bubb, H.

    1992-01-01

    This book resulted from the activity of Task Force 4.2 - 'Human Reliability'. This group was established on February 27th, 1986, at the plenary meeting of the Technical Reliability Committee of VDI, within the framework of the joint committee of VDI on industrial systems technology - GIS. It is composed of representatives of industry, representatives of research institutes, of technical control boards and universities, whose job it is to study how man fits into the technical side of the world of work and to optimize this interaction. In a total of 17 sessions, information from the part of ergonomy dealing with human reliability in using technical systems at work was exchanged, and different methods for its evaluation were examined and analyzed. The outcome of this work was systematized and compiled in this book. (orig.) [de

  17. 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.)

  18. 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

  19. 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

  20. 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.

  1. 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)

  2. 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)

  3. 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)

  4. 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

  5. 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)

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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

  11. 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

  12. 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)

  13. 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)

  14. 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

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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

  20. 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

  1. 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.

  2. 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

  3. 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

  4. 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

  5. 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.

  6. 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)

  7. 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.

  8. 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

  9. 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

  10. 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)

  11. 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.

  12. 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.

  13. 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)

  14. 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.

  15. 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

  16. 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)

  17. 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.

  18. 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

  19. 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.

  20. 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.)

  1. 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

  2. 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.

  3. 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.)

  4. 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.

  5. 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

  6. 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)

  7. 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.

  8. 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

  9. 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)

  10. 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

  11. 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)

  12. 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).

  13. 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

  14. 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

  15. 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

  16. 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.

  17. 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.

  18. 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

  19. 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

  20. 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.

  1. 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

  2. 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.

  3. 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)

  4. 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).

  5. Human factor reliability program

    International Nuclear Information System (INIS)

    Knoblochova, L.

    2017-01-01

    The human factor's reliability program was at Slovenske elektrarne, a.s. (SE) nuclear power plants. introduced as one of the components Initiatives of Excellent Performance in 2011. The initiative's goal was to increase the reliability of both people and facilities, in response to 3 major areas of improvement - Need for improvement of the results, Troubleshooting support, Supporting the achievement of the company's goals. The human agent's reliability program is in practice included: - Tools to prevent human error; - Managerial observation and coaching; - Human factor analysis; -Quick information about the event with a human agent; -Human reliability timeline and performance indicators; - Basic, periodic and extraordinary training in human factor reliability(authors)

  6. 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)

  7. 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)

  8. 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

  9. 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.)

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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.

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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

  20. Human Reliability Program Overview

    Energy Technology Data Exchange (ETDEWEB)

    Bodin, Michael

    2012-09-25

    This presentation covers the high points of the Human Reliability Program, including certification/decertification, critical positions, due process, organizational structure, program components, personnel security, an overview of the US DOE reliability program, retirees and academia, and security program integration.

  1. 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

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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

  7. 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

  8. 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)

  9. 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)

  10. 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.

  11. 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)

  12. 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.

  13. 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...

  14. 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

  15. 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

  16. 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.

  17. 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

  18. 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.

  19. 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.

  20. Reliability analysis and operator modelling

    International Nuclear Information System (INIS)

    Hollnagel, Erik

    1996-01-01

    The paper considers the state of operator modelling in reliability analysis. Operator models are needed in reliability analysis because operators are needed in process control systems. HRA methods must therefore be able to account both for human performance variability and for the dynamics of the interaction. A selected set of first generation HRA approaches is briefly described in terms of the operator model they use, their classification principle, and the actual method they propose. In addition, two examples of second generation methods are also considered. It is concluded that first generation HRA methods generally have very simplistic operator models, either referring to the time-reliability relationship or to elementary information processing concepts. It is argued that second generation HRA methods must recognise that cognition is embedded in a context, and be able to account for that in the way human reliability is analysed and assessed

  1. 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

  2. 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.

  3. 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

  4. 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)

  5. 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.

  6. 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.

  7. 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

  8. 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

  9. 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.

  10. 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.

  11. 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

  12. Human Reliability Program Workshop

    Energy Technology Data Exchange (ETDEWEB)

    Landers, John; Rogers, Erin; Gerke, Gretchen

    2014-05-18

    A Human Reliability Program (HRP) is designed to protect national security as well as worker and public safety by continuously evaluating the reliability of those who have access to sensitive materials, facilities, and programs. Some elements of a site HRP include systematic (1) supervisory reviews, (2) medical and psychological assessments, (3) management evaluations, (4) personnel security reviews, and (4) training of HRP staff and critical positions. Over the years of implementing an HRP, the Department of Energy (DOE) has faced various challenges and overcome obstacles. During this 4-day activity, participants will examine programs that mitigate threats to nuclear security and the insider threat to include HRP, Nuclear Security Culture (NSC) Enhancement, and Employee Assistance Programs. The focus will be to develop an understanding of the need for a systematic HRP and to discuss challenges and best practices associated with mitigating the insider threat.

  13. 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.

  14. 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)

  15. 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.

  16. 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

  17. 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.

  18. 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

  19. 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

  20. 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)

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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)

  6. 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

  7. 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)

  8. 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

  9. 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)

  10. 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)

  11. 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.

  12. 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.

  13. 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

  14. 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

  15. 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.

  16. 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)

  17. 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

  18. 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

  19. 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.

  20. 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)

  1. 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)

  2. 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.

  3. 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

  4. 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

  5. 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

  6. 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

  7. Human factors reliability Benchmark exercise

    International Nuclear Information System (INIS)

    Poucet, A.

    1989-06-01

    The Joint Research Centre of the European Commission has organized a Human Factors Reliability Benchmark Exercise (HF-RBE) with the aim of assessing the state of the art in human reliability modelling and assessment. Fifteen teams from eleven countries, representing industry, utilities, licensing organisations and research institutes, participated in the HF-RBE. The HF-RBE was organized around two study cases: (1) analysis of routine functional Test and Maintenance (T and M) procedures: with the aim of assessing the probability of test induced failures, the probability of failures to remain unrevealed and the potential to initiate transients because of errors performed in the test; (2) analysis of human actions during an operational transient: with the aim of assessing the probability that the operators will correctly diagnose the malfunctions and take proper corrective action. This report contains the final summary reports produced by the participants in the exercise

  8. "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...

  9. 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

  10. 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

  11. 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

  12. 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.

  13. Human factors reliability benchmark exercise

    International Nuclear Information System (INIS)

    Poucet, A.

    1989-08-01

    The Joint Research Centre of the European Commission has organised a Human Factors Reliability Benchmark Exercise (HF-RBE) with the aim of assessing the state of the art in human reliability modelling and assessment. Fifteen teams from eleven countries, representing industry, utilities, licensing organisations and research institutes, participated in the HF-RBE. The HF-RBE was organised around two study cases: (1) analysis of routine functional Test and Maintenance (TPM) procedures: with the aim of assessing the probability of test induced failures, the probability of failures to remain unrevealed and the potential to initiate transients because of errors performed in the test; (2) analysis of human actions during an operational transient: with the aim of assessing the probability that the operators will correctly diagnose the malfunctions and take proper corrective action. This report summarises the contributions received from the participants and analyses these contributions on a comparative basis. The aim of this analysis was to compare the procedures, modelling techniques and quantification methods used, to obtain insight in the causes and magnitude of the variability observed in the results, to try to identify preferred human reliability assessment approaches and to get an understanding of the current state of the art in the field identifying the limitations that are still inherent to the different approaches

  14. 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)

  15. 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

  16. 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)

  17. 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

  18. 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.

  19. 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.

  20. 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.

  1. 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

  2. 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.

  3. 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.

  4. Human reliability. Is probabilistic human reliability assessment possible?

    International Nuclear Information System (INIS)

    Mosneron Dupin, F.

    1996-01-01

    The possibility of carrying out Probabilistic Human Reliability Assessments (PHRA) is often doubted. Basing ourselves on the experience Electricite de France (EDF) has acquired in Probabilistic Safety Assessments for nuclear power plants, we show why the uncertainty of PHRA is very high. We then specify the limits of generic data and models for PHRA: very important factors are often poorly taken into account. To account for them, you need to have proper understanding of the actual context in which operators work. This demands surveys on the field (power plant and simulator) all of which must be carried out with behaviours science skills. The idea of estimating the probabilities of operator failure must not be abandoned, but probabilities must be given less importance, for they are only approximate indications. The qualitative aspects of PHRA should be given greater value (analysis process and qualitative insights). That is why the description (illustrated by case histories) of the main mechanisms of human behaviour, and of their manifestations in the nuclear power plant context (in terms of habits, attitudes, and informal methods and organization in particular) should be an important part of PHRA handbooks. These handbooks should also insist more on methods for gathering information on the actual context of the work of operators. Under these conditions, the PHRA should be possible and even desirable as a process for systematic analysis and assessment of human intervention. (author). 24 refs, 2 figs, 1 tab

  5. Human reliability. Is probabilistic human reliability assessment possible?

    Energy Technology Data Exchange (ETDEWEB)

    Mosneron Dupin, F

    1997-12-31

    The possibility of carrying out Probabilistic Human Reliability Assessments (PHRA) is often doubted. Basing ourselves on the experience Electricite de France (EDF) has acquired in Probabilistic Safety Assessments for nuclear power plants, we show why the uncertainty of PHRA is very high. We then specify the limits of generic data and models for PHRA: very important factors are often poorly taken into account. To account for them, you need to have proper understanding of the actual context in which operators work. This demands surveys on the field (power plant and simulator) all of which must be carried out with behaviours science skills. The idea of estimating the probabilities of operator failure must not be abandoned, but probabilities must be given less importance, for they are only approximate indications. The qualitative aspects of PHRA should be given greater value (analysis process and qualitative insights). That is why the description (illustrated by case histories) of the main mechanisms of human behaviour, and of their manifestations in the nuclear power plant context (in terms of habits, attitudes, and informal methods and organization in particular) should be an important part of PHRA handbooks. These handbooks should also insist more on methods for gathering information on the actual context of the work of operators. Under these conditions, the PHRA should be possible and even desirable as a process for systematic analysis and assessment of human intervention. (author). 24 refs, 2 figs, 1 tab.

  6. 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

  7. Human reliability in complex systems: an overview

    International Nuclear Information System (INIS)

    Embrey, D.E.

    1976-07-01

    A detailed analysis is presented of the main conceptual background underlying the areas of human reliability and human error. The concept of error is examined and generalized to that of human reliability, and some of the practical and methodological difficulties of reconciling the different standpoints of the human factors specialist and the engineer discussed. Following a survey of general reviews available on human reliability, quantitative techniques for prediction of human reliability are considered. An in-depth critical analysis of the various quantitative methods is then presented, together with the data bank requirements for human reliability prediction. Reliability considerations in process control and nuclear plant, and also areas of design, maintenance, testing and emergency situations are discussed. The effects of stress on human reliability are analysed and methods of minimizing these effects discussed. Finally, a summary is presented and proposals for further research are set out. (author)

  8. 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

  9. 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

  10. 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)

  11. 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

  12. 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

  13. 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

  14. Human reliability analysis—Taxonomy and praxes of human entropy boundary conditions for marine and offshore applications

    International Nuclear Information System (INIS)

    El-Ladan, S.B.; Turan, O.

    2012-01-01

    This is the first stage towards the development of a human reliability model called human entropy (HENT). The paper presents qualitative and quantitative taxonomies and praxes of performance shaping factors (PSF) for Marine and Offshore operations. Three structured and guided expert elicitation methods were used in this study. The experts interrogated accident reports and databases from which the generic root causes of failures/accidents in operations are determined. The elicitations led to the development of 9 qualitative and quantitative human influencing factors, which are called Human Entropy Boundary Conditions (HEBC). Further explications of the 9 HEBC gave birth to 137 quantifiable explanatory variables, which are called hypothetical constructs (HyC). The HyCs are used to identify potential risks due to shrinkages in safety standards. Human entropy is a detour from traditional human error and was used as a result of tripartite human failure modes; error, local rationality and extraneous acts, all of which signify disorderliness and are seemingly inevitable in maritime operations. The praxes and scaling of HEBC was developed as guidance towards a practical oriented HRA and provide inputs for measuring human disorderliness in maritime operations.

  15. 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)

  16. 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)

  17. 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.

  18. 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)

  19. 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

  20. 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

  1. 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.

  2. 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.

  3. 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

  4. Assessment of human decision reliability - a case study

    International Nuclear Information System (INIS)

    Pyy, P

    1998-01-01

    In his discussion of this case study, the author indicates that human beings are not merely machines who use rules. Thus, more focus needs to be put on studying decision making situations and their contexts. Decision theory (both normative and descriptive) and contextual psychological approaches may offer tools to cope with operator decision making. Further an ideal decision space needs to be defined for operators. The case study specifically addressed a loss of feedwater scenario and the various operator decisions that were involved in that scenario. It was concluded from this particular study that there are significant differences in the crew decision behaviours that are not explained by process variables. Through use of evidence from simulator tests with expert judgement, an approach to estimate probabilities has been developed. The modelling approach presented in this discussion is an extension of current HRA paradigms, but a natural one since all human beings make decisions

  5. 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.

  6. '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

  7. A taxonomy for human reliability analysis

    International Nuclear Information System (INIS)

    Beattie, J.D.; Iwasa-Madge, K.M.

    1984-01-01

    A human interaction taxonomy (classification scheme) was developed to facilitate human reliability analysis in a probabilistic safety evaluation of a nuclear power plant, being performed at Ontario Hydro. A human interaction occurs, by definition, when operators or maintainers manipulate, or respond to indication from, a plant component or system. The taxonomy aids the fault tree analyst by acting as a heuristic device. It helps define the range and type of human errors to be identified in the construction of fault trees, while keeping the identification by different analysts consistent. It decreases the workload associated with preliminary quantification of the large number of identified interactions by including a category called 'simple interactions'. Fault tree analysts quantify these according to a procedure developed by a team of human reliability specialists. The interactions which do not fit into this category are called 'complex' and are quantified by the human reliability team. The taxonomy is currently being used in fault tree construction in a probabilistic safety evaluation. As far as can be determined at this early stage, the potential benefits of consistency and completeness in identifying human interactions and streamlining the initial quantification are being realized

  8. 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

  9. An approach for assessing human decision reliability

    International Nuclear Information System (INIS)

    Pyy, P.

    2000-01-01

    This paper presents a method to study human reliability in decision situations related to nuclear power plant disturbances. Decisions often play a significant role in handling of emergency situations. The method may be applied to probabilistic safety assessments (PSAs) in cases where decision making is an important dimension of an accident sequence. Such situations are frequent e.g. in accident management. In this paper, a modelling approach for decision reliability studies is first proposed. Then, a case study with two decision situations with relatively different characteristics is presented. Qualitative and quantitative findings of the study are discussed. In very simple decision cases with time pressure, time reliability correlation proved out to be a feasible reliability modelling method. In all other decision situations, more advanced probabilistic decision models have to be used. Finally, decision probability assessment by using simulator run results and expert judgement is presented

  10. Operational human performance reliability assessment (OHPRA)

    International Nuclear Information System (INIS)

    Haas, P.M.; Swanson, P.J.; Connelly, E.M.

    1993-01-01

    Operational Human Performance Reliability Assessment (OHPRA) is an approach for assessing human performance that is being developed in response to demands from modern process industries for practical and effective tools to assess and improve human performance, and therefore overall system performance and safety. The single most distinguishing feature of the approach is that is defines human performance in open-quotes operationalclose quotes terms. OHPRA is focused not on generation of human error probabilities, but on practical analysis of human performance to aid management in (1) identifying open-quotes fixableclose quotes problems and (2) providing input on the importance and nature of potential improvements. Development of the model in progress uses a unique approach for eliciting expert strategies for assessing performance. A PC-based model incorporating this expertise is planned. A preliminary version of the approach has already been used successfully to identify practical human performance problems in reactor and chemical process plant operations

  11. 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.

  12. Human factors reliability benchmark exercise: a review

    International Nuclear Information System (INIS)

    Humphreys, P.

    1990-01-01

    The Human Factors Reliability Benchmark Exercise has addressed the issues of identification, analysis, representation and quantification of Human Error in order to identify the strengths and weaknesses of available techniques. Using a German PWR nuclear powerplant as the basis for the studies, fifteen teams undertook evaluations of a routine functional Test and Maintenance procedure plus an analysis of human actions during an operational transient. The techniques employed by the teams are discussed and reviewed on a comparative basis. The qualitative assessments performed by each team compare well, but at the quantification stage there is much less agreement. (author)

  13. 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.

  14. 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

  15. Benchmark of systematic human action reliability procedure

    International Nuclear Information System (INIS)

    Spurgin, A.J.; Hannaman, G.W.; Moieni, P.

    1986-01-01

    Probabilistic risk assessment (PRA) methodology has emerged as one of the most promising tools for assessing the impact of human interactions on plant safety and understanding the importance of the man/machine interface. Human interactions were considered to be one of the key elements in the quantification of accident sequences in a PRA. The approach to quantification of human interactions in past PRAs has not been very systematic. The Electric Power Research Institute sponsored the development of SHARP to aid analysts in developing a systematic approach for the evaluation and quantification of human interactions in a PRA. The SHARP process has been extensively peer reviewed and has been adopted by the Institute of Electrical and Electronics Engineers as the basis of a draft guide for the industry. By carrying out a benchmark process, in which SHARP is an essential ingredient, however, it appears possible to assess the strengths and weaknesses of SHARP to aid human reliability analysts in carrying out human reliability analysis as part of a PRA

  16. 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

  17. 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

  18. Human reliability in probabilistic safety assessments

    International Nuclear Information System (INIS)

    Nunez Mendez, J.

    1989-01-01

    Nowadays a growing interest in medioambiental aspects is detected in our country. It implies an assessment of the risk involved in the industrial processess and installations in order to determine if those are into the acceptable limits. In these safety assessments, among which PSA (Probabilistic Safety Assessments), can be pointed out the role played by the human being in the system is one of the more relevant subjects. (This relevance has been demostrated in the accidents happenned). However in Spain there aren't manuals specifically dedicated to asses the human contribution to risk in the frame of PSAs. This report aims to improve this situation providing: a) a theoretical background to help the reader in the understanding of the nature of the human error, b) a guide to carry out a Human Reliability Analysis and c) a selected overwiev of the techniques and methodologies currently applied in this area. (Author)

  19. Human Reliability in Probabilistic Safety Assessments

    International Nuclear Information System (INIS)

    Nunez Mendez, J.

    1989-01-01

    Nowadays a growing interest in environmental aspects is detected in our country. It implies an assessment of the risk involved in the industrial processes and installations in order to determine if those are into the acceptable limits. In these safety assessments, among which PSA (Probabilistic Safety Assessments), can be pointed out the role played by the human being in the system is one of the more relevant subjects (This relevance has been demonstrated in the accidents happened) . However, in Spain there aren't manuals specifically dedicated to asses the human contribution to risk in the frame of PSAs. This report aims to improve this situation providing: a) a theoretical background to help the reader in the understanding of the nature of the human error, b) a quid to carry out a Human Reliability Analysis and c) a selected overview of the techniques and methodologies currently applied in this area. (Author) 20 refs

  20. Human reliability analysis of dependent events

    International Nuclear Information System (INIS)

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

    1977-01-01

    In the human reliability analysis in WASH-1400, the continuous variable of degree of interaction among human events was approximated by selecting four points on this continuum to represent the entire continuum. The four points selected were identified as zero coupling (i.e., zero dependence), complete coupling (i.e., complete dependence), and two intermediate points--loose coupling (a moderate level of dependence) and tight coupling (a high level of dependence). The paper expands the WASH-1400 treatment of common mode failure due to the interaction of human activities. Mathematical expressions for the above four levels of dependence are derived for parallel and series systems. The psychological meaning of each level of dependence is illustrated by examples, with probability tree diagrams to illustrate the use of conditional probabilities resulting from the interaction of human actions in nuclear power plant tasks

  1. Advancing Usability Evaluation through Human Reliability Analysis

    International Nuclear Information System (INIS)

    Ronald L. Boring; David I. Gertman

    2005-01-01

    This paper introduces a novel augmentation to the current heuristic usability evaluation methodology. The SPAR-H human reliability analysis method was developed for categorizing human performance in nuclear power plants. Despite the specialized use of SPAR-H for safety critical scenarios, the method also holds promise for use in commercial off-the-shelf software usability evaluations. The SPAR-H method shares task analysis underpinnings with human-computer interaction, and it can be easily adapted to incorporate usability heuristics as performance shaping factors. By assigning probabilistic modifiers to heuristics, it is possible to arrive at the usability error probability (UEP). This UEP is not a literal probability of error but nonetheless provides a quantitative basis to heuristic evaluation. When combined with a consequence matrix for usability errors, this method affords ready prioritization of usability issues

  2. Human factors considerations for reliability and safety

    International Nuclear Information System (INIS)

    Carnino, A.

    1985-01-01

    Human factors in many industries have become an important issue, since the last few years. They should be considered during the whole life time of a plant: design, fabrication and construction, licensing, operation. Improvements have been performed in the field of man-machine interface such as procedures, control room lay-out, operator aids, training. In order to meet the needs of reliability and probabilistic risk studies, quantification of human errors has been developed but needs still improvements in the field of cognitive behaviour, diagnosis and representation errors. Data banks to support these quantifications are still in a development stage. This applies to nuclear power plants and several examples are given to illustrate the above ideas. In conclusion, human factors field is in a very quickly evolving process but the tendency is still to adapt the man to the machines whilst the reverse would be desirable

  3. Human reliability assessors guide: an overview

    International Nuclear Information System (INIS)

    Humphreys, P.

    1988-01-01

    The Human Reliability Assessors Guide 1 provides a review of techniques currently available for the quantification of Human Error Probabilities. The Guide has two main objectives. The first is to provide a clear and comprehensive description of eight major techniques which can be used to assess human reliability. This is supplemented by case studies taken from practical applications of each technique to industrial problems. The second objective is to provide practical guidelines for the selection of techniques. The selection process is aided by reference to a set of criteria against which each of the eight techniques have been evaluated. Utilising the criteria and critiques, a selection method is presented. This is designed to assist the potential user in choosing the technique, or combination of techniques, most suited to answering the users requirements. For each of the eight selected techniques, a summary of the origins of the technique is provided, together with a method description, detailed case studies, abstracted case studies and supporting references. (author)

  4. Human reliability program: Components and effects

    International Nuclear Information System (INIS)

    Baley-Downes, S.

    1986-01-01

    The term ''Human Reliability Program'' (HRP) is defined as a series of selective controls which are implemented and integrated to identify the ''insider threat'' from current and prospective employees who are dishonest, disloyal and unreliable. The HRP, although not a prediction of human behaviour, is an excellent tool for decision making and should compliment security and improve employee quality. The HRP consists of several component applications such as management evaluation; appropriate background investigative requirements; occupational health examination and laboratory testing; drug/alcohol screening; psychological testing and interviews; polygraph examination; job related aberrant behaviour recognition; on-going education and training; document control; drug/alcohol rehabilitation; periodic HRP audit; and implementation of an onsite central clearing house. The components and effects of HRP are discussed in further detail in this paper

  5. 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

  6. Human Reliability Data Bank: evaluation results

    International Nuclear Information System (INIS)

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

    1985-01-01

    The US Nuclear Regulatory Commission (NRC), Sandia National Laboratories (SNL), and General Physics Corporation are conducting a research program to determine the practicality, acceptability, and usefulness of a Human Reliability Data Bank for nuclear power industry probabilistic risk assessment (PRA). As part of this program, a survey was conducted of existing human reliability data banks from other industries, and a detailed concept of a Data Bank for the nuclear industry was developed. Subsequently, a detailed specification for implementing the Data Bank was developed. An evaluation of this specification was conducted and is described in this report. The evaluation tested data treatment, storage, and retrieval using the Data Bank structure, as modified from NUREG/CR-2744, and detailed procedures for data processing and retrieval, developed prior to this evaluation and documented in the test specification. The evaluation consisted of an Operability Demonstration and Evaluation of the data processing procedures, a Data Retrieval Demonstration and Evaluation, a Retrospective Analysis that included a survey of organizations currently operating data banks for the nuclear power industry, and an Internal Analysis of the current Data Bank System

  7. 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)

  8. 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

  9. 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)

  10. 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

  11. [Study of the relationship between human quality and reliability].

    Science.gov (United States)

    Long, S; Wang, C; Wang, L i; Yuan, J; Liu, H; Jiao, X

    1997-02-01

    To clarify the relationship between human quality and reliability, 1925 experiments in 20 subjects were carried out to study the relationship between disposition character, digital memory, graphic memory, multi-reaction time and education level and simulated aircraft operation. Meanwhile, effects of task difficulty and enviromental factor on human reliability were also studied. The results showed that human quality can be predicted and evaluated through experimental methods. The better the human quality, the higher the human reliability.

  12. Human reliability data bank: feasibility study

    International Nuclear Information System (INIS)

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

    1984-01-01

    The US Nuclear Regulatory Commission and Sandia National Laboratories have been developing a plan for a human reliability data bank since August 1981. This research is in response to the data needs of the nuclear power industry's probabilistic risk assessment community. The three phases of the program are to: (A) develop the data bank concept, (B) develop an implementation plan and conduct a feasibility study, and (C) assist a sponsor in implementing the data bank. The program is now in Phase B. This paper describes the methods used in the feasibility study. Decisions to be made in the future regarding full-scale implementation will be based, in part, on the outcome of this study. 3 references, 2 figures

  13. 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

  14. 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)

  15. Survey of methods used to asses human reliability in the human factors reliability benchmark exercise

    International Nuclear Information System (INIS)

    Poucet, A.

    1988-01-01

    The Joint Research Centre of the European Commission has organised a Human Factors Reliability Benchmark Exercise (HF-RBE) with the aim to assess the state-of-the-art in human reliability modelling and assessment. Fifteen teams from eleven countries, representing industry, utilities, licensing organisations and research institutes, participate in the HF-RBE, which is organised around two study cases: (1) analysis of routine functional test and maintenance procedures, with the aim to assess the probability of test-induced failures, the probability of failures to remain unrevealed, and the potential to initiate transients because of errors performed in the test; and (2) analysis of human actions during an operational transient, with the aim to assess the probability that the operators will correctly diagnose the malfunctions and take proper corrective action. The paper briefly reports how the HF-RBE was structured and gives an overview of the methods that have been used for predicting human reliability in both study cases. The experience in applying these methods is discussed and the results obtained are compared. (author)

  16. The development of a nuclear chemical plant human reliability management approach: HRMS and JHEDI

    International Nuclear Information System (INIS)

    Kirwan, Barry

    1997-01-01

    In the late 1980's, amidst the qualitative and quantitative validation of certain Human Reliability Assessment (HRA) techniques, there was a desire for a new technique specifically for a nuclear reprocessing plant being designed. The technique was to have the following attributes: it should be data-based rather than involving pure expert judgement; it was to be flexible, so that it would allow both relatively rapid screening and more detailed assessment; and it was to have sensitivity analysis possibilities, so that Human Factors design-related parameters, albeit at a gross level, could be brought into the risk assessment equation. The techniques and literature were surveyed, and it was decided that no one technique fulfilled these requirements, and so a new approach was developed. Two techniques were devised, the Human Reliability Management System (HRMS), and the Justification of Human Error Data Information (JHEDI) technique, the latter being essentially a quicker screening version of the former. Both techniques carry out task analysis, error analysis, and Performance Shaping Factor-based quantification, but JHEDI involves less detailed assessment than HRMS. Additionally, HRMS can be utilised to determine error reduction mechanisms, based on the way the Performance Shaping Factors are contributing to the assessed error probabilities. Both techniques are fully computerised and assessments are highly documentable and auditable, which was seen as a useful feature both by the company developing the techniques, and by the regulatory authorities assessing the final output risk assessments into which these two techniques fed data. This paper focuses in particular on the quantification process used by these techniques. The quantification approach for both techniques was principally one of extrapolation from real data to the desired Human Error Probability (HEP), based on a comparison between Performance Shaping Factor (PSF) profiles for the real, and the to

  17. 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)

  18. 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.

  19. 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...

  20. 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

  1. Ergonomics design and operator training as contributors to human reliability

    International Nuclear Information System (INIS)

    Jackson, A.R.G.; Madden, V.J.; Umbers, I.G.; Williams, J.C.

    1987-01-01

    The safe operation of nuclear reactors depends not only on good physical safety engineering but on the human operators as well. The Central Electricity Generating Board's approach to human reliability includes the following aspects: ergonomics design (task analysis and the development of man-machine interfaces), analysis of human reliability, operational feedback, staff training and assessment, maintenance management, research programmes and management. This paper describes how these combine to achieve the highest practicable level of human reliability, not only for the Sizewell-B pressurized water reactor, but also for the Board's gas-cooled reactors. Examples are used to illustrate the topics considered. (UK)

  2. Human Error Probabilites (HEPs) for generic tasks and Performance Shaping Factors (PSFs) selected for railway operations

    DEFF Research Database (Denmark)

    Thommesen, Jacob; Andersen, Henning Boje

    This report describes an HRA (Human Reliability Assessment) of six generic tasks and four Perfor-mance Shaping Factors (PSFs) targeted at railway operations commissioned by Banedanmark. The selection and characterization of generic tasks and PSFs are elaborated by DTU Management in close...

  3. 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

  4. Incorporating process mining into human reliability analysis

    NARCIS (Netherlands)

    Kelly, D.L.

    2011-01-01

    It is well established that the human contribution to the risk of operation of complex technological systems is significant, with typical estimates lying in the range of 60-85%. Human errors have been a contributor to many significant catastrophic technological accidents. Examples are 1) the

  5. Review of some aspects of human reliability quantification

    International Nuclear Information System (INIS)

    Lydell, B.O.Y.; Spurgin, A.J.; Hannaman, G.W.; Lukic, Y.D.

    1986-01-01

    An area in systems reliability considered to be weak, is the characterization and quantification of the role of the operations and maintenance staff in combatting accidents. Several R and D programs are underway to improve the modeling of human interactions and some progress has been made. This paper describes a specific aspect of human reliability analysis which is referred to as modeling of cognitive processes. In particular, the basis for the so- called Human Cognitive Reliability (HCR) model is described and the focus is on its validation and on its benefits and limitations

  6. Modeling cognition dynamics and its application to human reliability analysis

    International Nuclear Information System (INIS)

    Mosleh, A.; Smidts, C.; Shen, S.H.

    1996-01-01

    For the past two decades, a number of approaches have been proposed for the identification and estimation of the likelihood of human errors, particularly for use in the risk and reliability studies of nuclear power plants. Despite the wide-spread use of the most popular among these methods, their fundamental weaknesses are widely recognized, and the treatment of human reliability has been considered as one of the soft spots of risk studies of large technological systems. To alleviate the situation, new efforts have focused on the development of human reliability models based on a more fundamental understanding of operator response and its cognitive aspects

  7. 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

  8. Basic research on human reliability in nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Li; Deng Zhiliang

    1996-10-01

    Human reliability in nuclear power plants is one of key factors in nuclear safety and economic operation. According to cognitive science, behaviour theory and ergonomic and on the bases of human cognitive behaviour characteristics, performance shaping factors, human error mechanisms and organization management, the project systematically studied the human reliability in nuclear power plant systems, established the basic theory and methods for analyzing human factor accidents and suggested feasible approaches and countermeasures for precaution against human factor accidents and improving human reliability. The achievement has been applied in operation departments, management departments and scientific research institutions of nuclear power, and has produced guiding significance and practical value to design, operation and management in nuclear power plants. (11 refs.)

  9. 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

  10. A HUMAN RELIABILITY-CENTERED APPROACH TO THE DEVELOPMENT OF JOB AIDS FOR REVIEWERS OF MEDICAL DEVICES THAT USE RADIOLOGICAL BYPRODUCT MATERIALS.

    Energy Technology Data Exchange (ETDEWEB)

    COOPER, S.E.; BROWN, W.S.; WREATHALL, J.

    2005-02-02

    The U.S. Nuclear Regulatory Commission (NRC) is engaged in an initiative to risk-inform the regulation of byproduct materials. Operating experience indicates that human actions play a dominant role in most of the activities involving byproduct materials, which are radioactive materials other than those used in nuclear power plants or in weapons production, primarily for medical or industrial purposes. The overall risk of these activities is strongly influenced by human performance. Hence, an improved understanding of human error, its causes and contexts, and human reliability analysis (HRA) is important in risk-informing the regulation of these activities. The development of the human performance job aids was undertaken by stages, with frequent interaction with the prospective users. First, potentially risk significant human actions were identified based on reviews of available risk studies for byproduct material applications and of descriptions of events for byproduct materials applications that involved potentially significant human actions. Applications from the medical and the industrial domains were sampled. Next, the specific needs of the expected users of the human performance-related capabilities were determined. To do this, NRC headquarters and region staff were interviewed to identify the types of activities (e.g., license reviews, inspections, event assessments) that need HRA support and the form in which such support might best be offered. Because the range of byproduct uses regulated by NRC is so broad, it was decided that initial development of knowledge and tools would be undertaken in the context of a specific use of byproduct material, which was selected in consultation with NRC staff. Based on needs of NRC staff and the human performance related characteristics of the context chosen, knowledge resources were then compiled to support consideration of human performance issues related to the regulation of byproduct materials. Finally, with

  11. A HUMAN RELIABILITY-CENTERED APPROACH TO THE DEVELOPMENT OF JOB AIDS FOR REVIEWERS OF MEDICAL DEVICES THAT USE RADIOLOGICAL BYPRODUCT MATERIALS

    International Nuclear Information System (INIS)

    COOPER, S.E.; BROWN, W.S.; WREATHALL, J.

    2005-01-01

    The U.S. Nuclear Regulatory Commission (NRC) is engaged in an initiative to risk-inform the regulation of byproduct materials. Operating experience indicates that human actions play a dominant role in most of the activities involving byproduct materials, which are radioactive materials other than those used in nuclear power plants or in weapons production, primarily for medical or industrial purposes. The overall risk of these activities is strongly influenced by human performance. Hence, an improved understanding of human error, its causes and contexts, and human reliability analysis (HRA) is important in risk-informing the regulation of these activities. The development of the human performance job aids was undertaken by stages, with frequent interaction with the prospective users. First, potentially risk significant human actions were identified based on reviews of available risk studies for byproduct material applications and of descriptions of events for byproduct materials applications that involved potentially significant human actions. Applications from the medical and the industrial domains were sampled. Next, the specific needs of the expected users of the human performance-related capabilities were determined. To do this, NRC headquarters and region staff were interviewed to identify the types of activities (e.g., license reviews, inspections, event assessments) that need HRA support and the form in which such support might best be offered. Because the range of byproduct uses regulated by NRC is so broad, it was decided that initial development of knowledge and tools would be undertaken in the context of a specific use of byproduct material, which was selected in consultation with NRC staff. Based on needs of NRC staff and the human performance related characteristics of the context chosen, knowledge resources were then compiled to support consideration of human performance issues related to the regulation of byproduct materials. Finally, with

  12. 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

  13. Application of DFM in human reliability analysis

    International Nuclear Information System (INIS)

    Yu Shaojie; Zhao Jun; Tong Jiejuan

    2011-01-01

    Combining with ATHEANA, the possible to identify EFCs and UAs using DFM is studied; and then Steam Generator Tube Rupture (SGTR) accident is modeled and solved. Through inductive analysis, 26 Prime Implicants (PIs) are obtained and the meaning of results is interpreted; and one of PIs is similar to the accident scenario of human failure event in one nuclear power plant. Finally, this paper discusses the methods of quantifying PIs, analysis of Error of commission (EOC) and so on. (authors)

  14. PROVIDING RELIABILITY OF HUMAN RESOURCES IN PRODUCTION MANAGEMENT PROCESS

    Directory of Open Access Journals (Sweden)

    Anna MAZUR

    2014-07-01

    Full Text Available People are the most valuable asset of an organization and the results of a company mostly depends on them. The human factor can also be a weak link in the company and cause of the high risk for many of the processes. Reliability of the human factor in the process of the manufacturing process will depend on many factors. The authors include aspects of human error, safety culture, knowledge, communication skills, teamwork and leadership role in the developed model of reliability of human resources in the management of the production process. Based on the case study and the results of research and observation of the author present risk areas defined in a specific manufacturing process and the results of evaluation of the reliability of human resources in the process.

  15. MODELING HUMAN RELIABILITY ANALYSIS USING MIDAS

    Energy Technology Data Exchange (ETDEWEB)

    Ronald L. Boring; Donald D. Dudenhoeffer; Bruce P. Hallbert; Brian F. Gore

    2006-05-01

    This paper summarizes an emerging collaboration between Idaho National Laboratory and NASA Ames Research Center regarding the utilization of high-fidelity MIDAS simulations for modeling control room crew performance at nuclear power plants. The key envisioned uses for MIDAS-based control room simulations are: (i) the estimation of human error with novel control room equipment and configurations, (ii) the investigative determination of risk significance in recreating past event scenarios involving control room operating crews, and (iii) the certification of novel staffing levels in control rooms. It is proposed that MIDAS serves as a key component for the effective modeling of risk in next generation control rooms.

  16. Collection of offshore human error probability data

    International Nuclear Information System (INIS)

    Basra, Gurpreet; Kirwan, Barry

    1998-01-01

    Accidents such as Piper Alpha have increased concern about the effects of human errors in complex systems. Such accidents can in theory be predicted and prevented by risk assessment, and in particular human reliability assessment (HRA), but HRA ideally requires qualitative and quantitative human error data. A research initiative at the University of Birmingham led to the development of CORE-DATA, a Computerised Human Error Data Base. This system currently contains a reasonably large number of human error data points, collected from a variety of mainly nuclear-power related sources. This article outlines a recent offshore data collection study, concerned with collecting lifeboat evacuation data. Data collection methods are outlined and a selection of human error probabilities generated as a result of the study are provided. These data give insights into the type of errors and human failure rates that could be utilised to support offshore risk analyses

  17. 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

  18. Predicting risk and human reliability: a new approach

    International Nuclear Information System (INIS)

    Duffey, R.; Ha, T.-S.

    2009-01-01

    Learning from experience describes human reliability and skill acquisition, and the resulting theory has been validated by comparison against millions of outcome data from multiple industries and technologies worldwide. The resulting predictions were used to benchmark the classic first generation human reliability methods adopted in probabilistic risk assessments. The learning rate, probabilities and response times are also consistent with the existing psychological models for human learning and error correction. The new approach also implies a finite lower bound probability that is not predicted by empirical statistical distributions that ignore the known and fundamental learning effects. (author)

  19. The dependence of human reliability upon task information content

    International Nuclear Information System (INIS)

    Hermanson, E.M.; Golay, M.W.

    1994-09-01

    The role of human error in safety mishaps is an important factor in system design. As systems become increasingly complex the capacity of the human to deal with the added complexity is diminished. It is therefore crucial to understand the relationship between system complexity and human reliability so that systems may be built in such a way as to minimize human error. One way of understanding this relationship is to quantify system complexity and then measure the human reaction in response to situations of varying complexity. The quantification of system complexity may be performed by determining the information content present in the tasks that the human must execute. The purpose of this work is therefore to build and perform a consistent experiment which will determine the extent to which human reliability depends upon task information content. Two main conclusions may be drawn from this work. The first is that human reliability depends upon task information content. Specifically, as the information content contained in a task increases, the capacity of a human to deal successfully with the task decreases monotonically. Here the definition of total success is the ability to complete the task at hand fully and correctly. Furthermore, there exists a value of information content below which a human can deal with the task successfully, but above which the success of an individual decreases monotonically with increasing information. These ideas should be generalizable to any model where system complexity can be clearly and consistently defined

  20. Human factors reliability benchmark exercise, report of the SRD participation

    International Nuclear Information System (INIS)

    Waters, Trevor

    1988-01-01

    Within the scope of the Human Factors Reliability Benchmark Exercise, organised by the Joint Research Centre, Ispra, Italy, the Safety and Reliability Directorate (SRD) team has performed analysis of human factors in two different activities - a routine test and a non-routine operational transient. For both activities, an 'FMEA-like' task, potential errors, and the factors which affect performance. For analysis of the non-routine activity, which involved a significant amount of cognitive processing, such as diagnosis and decision making, a new approach for qualitative analysis has been developed. Modelling has been performed using both event trees and fault trees and examples are provided. Human error probabilities were estimated using the methods Absolute Probability Judgement (APJ), Human Cognitive Reliability Method (HCR), Human Error and Assessment and Reduction Technique (HEART), Success-Likelihood Index Method (SLIM), Technica Empiriza Stima Eurori Operatori (TESEO), and Technique for Human Error Rate Prediction (THERP). A discussion is provided of the lessons learnt in the course of the exercise and unresolved difficulties in the assessment of human reliability. (author)

  1. Some areas of reliability technique which have been neglected to some extent - maintainability - human reliability - mechanical reliability - repairable systems

    International Nuclear Information System (INIS)

    Akersten, P.A.

    1985-01-01

    The present thesis consists of four papers, three of which are of a expositary nature and one more theoretical. The first two papers have a natural coupling to the man-machine interface. The first paper is devoted to the concept of maintainability and the role of man as maintenance technician. The second paper discusses aspects of human reliability, mainly studying man as operator. However, maintenance tasks can be studied in the same manner. The third paper concerns reliability prediction for mechanical components. This is an area of vital importance for the reliability practitioner, who needs realistic and easy-to-use mathematical models for different failure modes. The fourth paper discusses mathematical models for repairable systems, especially the problem of testing whether a constant event intensity model is adequate or not. (author)

  2. System ergonomics as an approach to improve human reliability

    International Nuclear Information System (INIS)

    Bubb, H.

    1988-01-01

    The application of system technics on ergonomical problems is called system ergonomics. This enables improvements of human reliability by design measures. The precondition for this is the knowledge of how information processing is performed by man and machine. By a separate consideration of sensory processing, cognitive processing, and motory processing it is possible to have a more exact idea of the system element 'man'. The system element 'machine' is well described by differential equations which allow an ergonomical assessment of the manouverability. The knowledge of information processing of man and machine enables a task analysis. This makes appear on one hand the human boundaries depending on the different properties of the task and on the other hand suitable ergonomical solution proposals which improve the reliability of the total system. It is a disadvantage, however, that the change of human reliability by such measures may not be quoted numerically at the moment. (orig.)

  3. 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)

  4. 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.

  5. 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)

  6. Human reliability impact on in-service inspection

    International Nuclear Information System (INIS)

    Spanner, J.C. Sr.

    1986-01-01

    This paper describes a study conducted to identify, characterize, and evaluate the human reliability aspects of ultrasonic testing/inservice inspection (UT/ISI). Recent measurements of UT/ISI system effectiveness have revealed wide variations in performance; suggesting that insufficient emphasis is being placed on the human reliability aspects of nondestructive examination. It appears that NDE performance can be improved through application of the human factors principles relating to the task, training, procedure, environmental, and individual difference variables. These variables are collectively referred to as performance-shaping factors. A man-machine systems model was developed to describe the UT/ISI process using functional task descriptors. The relative operating characteristic (ROC) analysis method, which is derived from signal detection theory, offers unique attributes for analyzing NDT performance. The results of a limited human factors evaluation conducted in conjunction with a mini-round robin test are also described

  7. Applications of Human Performance Reliability Evaluation Concepts and Demonstration Guidelines

    Science.gov (United States)

    1977-03-15

    ship stops dead in the water and the AN/SQS-26 operator recommends a new heading (000°). At T + 14 minutes, the target ship begins a hard turn to...Various Simulated Conditions 82 9 Hunan Reliability for Each Simulated Operator (Baseline Run) 83 10 Human and Equipment Availabilit / under

  8. Human performance for the success of equipment reliability programs

    International Nuclear Information System (INIS)

    Woodcock, J.

    2007-01-01

    Human performance is a critical element of programs directed at equipment reliability. Reliable equipment performance requires broad support from all levels of plant management and throughout all plant departments. Experience at both nuclear power plants and fuel manufacturing plants shows that human performance must be addressed during all phases of program implementation from the beginning through the establishment of a living, on-going process. At the beginning, certain organizational and management actions during the initiation of the program set the stage for successful adoption by station personnel, leading to more rapid benefits. For the long term, equipment reliability is a living process needed throughout the lifetime of a station, a program which must be motivated and measured. Sustained acceptance and participation by the plant personnel is a requirement, and culture is a key ingredient. This paper will provide an overview of key human performance issues to be considered, using the application of the INPO AP-913 Equipment Reliability Guideline as a basis and gives some best practices for training, communicating and implementing programs. The very last part includes ways to tell if the program is effective

  9. DOE Human Reliability Program Removals Report 2004-2006

    International Nuclear Information System (INIS)

    Center for Human Reliability Studies

    2007-01-01

    This report presents results of the comprehensive data analysis and assessment of all U.S. Department of Energy (DOE) and National Nuclear Security Administration (NNSA) facilities that have positions requiring workers to be certified in the Human Reliability Program (HRP). Those facilities include: Albuquerque, Amarillo, DOE Headquarters, Hanford, Idaho, Nevada, Oak Ridge, Oakland, and Savannah River. The HRP was established to ensure, through continuous review and evaluation, the reliability of individuals who have access to the DOE's most sensitive facilities, materials, and information

  10. SHARP1: A revised systematic human action reliability procedure

    International Nuclear Information System (INIS)

    Wakefield, D.J.; Parry, G.W.; Hannaman, G.W.; Spurgin, A.J.

    1990-12-01

    Individual plant examinations (IPE) are being performed by utilities to evaluate plant-specific vulnerabilities to severe accidents. A major tool in performing an IPE is a probabilistic risk assessment (PRA). The importance of human interactions in determining the plant response in past PRAs is well documented. The modeling and quantification of the probabilities of human interactions have been the subjects of considerable research by the Electric Power Research Institute (EPRI). A revised framework, SHARP1, for incorporating human interactions into PRA is summarized in this report. SHARP1 emphasizes that the process stages are iterative and directed at specific goals rather than being performed sequentially in a stepwise procedure. This expanded summary provides the reader with a flavor of the full report content. Excerpts from the full report are presented, following the same outline as the full report. In the full report, the interface of the human reliability analysis with the plant logic model development in a PRA is given special attention. In addition to describing a methodology framework, the report also discusses the types of human interactions to be evaluated, and how to formulate a project team to perform the human reliability analysis. A concise description and comparative evaluation of the selected existing methods of quantification of human error are also presented. Four case studies are also provided to illustrate the SHARP1 process

  11. A model for assessing human cognitive reliability in PRA studies

    International Nuclear Information System (INIS)

    Hannaman, G.W.; Spurgin, A.J.; Lukic, Y.

    1985-01-01

    This paper summarizes the status of a research project sponsored by EPRI as part of the Probabilistic Risk Assessment (PRA) technology improvement program and conducted by NUS Corporation to develop a model of Human Cognitive Reliability (HCR). The model was synthesized from features identified in a review of existing models. The model development was based on the hypothesis that the key factors affecting crew response times are separable. The inputs to the model consist of key parameters the values of which can be determined by PRA analysts for each accident situation being assessed. The output is a set of curves which represent the probability of control room crew non-response as a function of time for different conditions affecting their performance. The non-response probability is then a contributor to the overall non-success of operating crews to achieve a functional objective identified in the PRA study. Simulator data and some small scale tests were utilized to illustrate the calibration of interim HCR model coefficients for different types of cognitive processing since the data were sparse. The model can potentially help PRA analysts make human reliability assessments more explicit. The model incorporates concepts from psychological models of human cognitive behavior, information from current collections of human reliability data sources and crew response time data from simulator training exercises

  12. A comparative evaluation of five human reliability assessment techniques

    International Nuclear Information System (INIS)

    Kirwan, B.

    1988-01-01

    A field experiment was undertaken to evaluate the accuracy, usefulness, and resources requirements of five human reliability quantification techniques (Techniques for Human Error Rate Prediction (THERP); Paired Comparisons, Human Error Assessment and Reduction Technique (HEART), Success Liklihood Index Method (SLIM)-Multi Attribute Utility Decomposition (MAUD), and Absolute Probability Judgement). This was achieved by assessing technique predictions against a set of known human error probabilities, and by comparing their predictions on a set of five realistic Probabilisitc Risk Assessment (PRA) human error. On a combined measure of accuracy THERP and Absolute Probability Judgement performed best, whilst HEART showed indications of accuracy and was lower in resources usage than other techniques. HEART and THERP both appear to benefit from using trained assessors in order to obtain the best results. SLIM and Paired Comparisons require further research on achieving a robust calibration relationship between their scale values and absolute probabilities. (author)

  13. 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)

  14. Human reliability: an evaluation of its understanding and prediction

    International Nuclear Information System (INIS)

    Joksimovich, V.

    1987-01-01

    This paper presents a viewpoint on the state-of-the-art in human reliability. The bases for this viewpoint are, by and large, research projects conducted by the NUS for the Electric Power Research Institute (EPRI) primarily with the objective of further enhancing the credibility of PRA methodology. The presentation is divided into the following key sections: Background and Overview, Methodology and Data Base with emphasis on the simulator data base

  15. Human event observations in the individual plant examinations

    Energy Technology Data Exchange (ETDEWEB)

    Forester, J. [Sandia National Labs., Albuquerque, NM (United States)

    1995-04-01

    A major objective of the Individual Plant Examination Insights Program is to identify the important determinants of core damage frequency for the different reactor and containment types and plant designs as indicated in the IPEs. The human reliability analysis is a critical component of the probabilistic risk assessments 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. 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. Another finding was that while there were relatively wide variations in the calculated human error probabilities for similar events across plants, there was no evidence for any systematic variation as a function of the HRA methods used in the analyses. Much of the variability in HEP values can be explained by differences in plant characteristics and sequence-specific factors. Details of these results and other findings are discussed.

  16. Human event observations in the individual plant examinations

    International Nuclear Information System (INIS)

    Forester, J.

    1995-01-01

    A major objective of the Individual Plant Examination Insights Program is to identify the important determinants of core damage frequency for the different reactor and containment types and plant designs as indicated in the IPEs. The human reliability analysis is a critical component of the probabilistic risk assessments 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. 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. Another finding was that while there were relatively wide variations in the calculated human error probabilities for similar events across plants, there was no evidence for any systematic variation as a function of the HRA methods used in the analyses. Much of the variability in HEP values can be explained by differences in plant characteristics and sequence-specific factors. Details of these results and other findings are discussed

  17. Some developments in human reliability analysis approaches and tools

    Energy Technology Data Exchange (ETDEWEB)

    Hannaman, G W; Worledge, D H

    1988-01-01

    Since human actions have been recognized as an important contributor to safety of operating plants in most industries, research has been performed to better understand and account for the way operators interact during accidents through the control room and equipment interface. This paper describes the integration of a series of research projects sponsored by the Electric Power Research Institute to strengthen the methods for performing the human reliability analysis portion of the probabilistic safety studies. It focuses on the analytical framework used to guide the analysis, the development of the models for quantifying time-dependent actions, and simulator experiments used to validate the models.

  18. Reliability

    OpenAIRE

    Condon, David; Revelle, William

    2017-01-01

    Separating the signal in a test from the irrelevant noise is a challenge for all measurement. Low test reliability limits test validity, attenuates important relationships, and can lead to regression artifacts. Multiple approaches to the assessment and improvement of reliability are discussed. The advantages and disadvantages of several different approaches to reliability are considered. Practical advice on how to assess reliability using open source software is provided.

  19. Modeling human intention formation for human reliability assessment

    International Nuclear Information System (INIS)

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

    1988-01-01

    This paper describes a dynamic simulation capability for modeling how people form intentions to act in nuclear power plant emergency situations. This modeling tool, Cognitive Environment Simulation or CES, was developed based on techniques from artificial intelligence. It simulates the cognitive processes that determine situation assessment and intention formation. It can be used to investigate analytically what situations and factors lead to intention failures, what actions follow from intention failures (e.g. errors of omission, errors of commission, common mode errors), the ability to recover from errors or additional machine failures, and the effects of changes in the NPP person machine system. One application of the CES modeling environment is to enhance the measurement of the human contribution to risk in probabilistic risk assessment studies. (author)

  20. Human performance and reliability studies on nuclear power plant

    International Nuclear Information System (INIS)

    Miyaoka, S.

    1988-01-01

    The TMI accident in USA, the Chernobyl accident in USSR and other major accidents overseas have shown that it is necessary to investigate and research human factor problems related to operation, maintenance and others in order to increase the safety and reliability of nuclear power plants. Although a variety of countermeasures have been devised, the accidents and failures due to human factors still occur. So far, the problems related to human factors have not been fundamantally and systematically investigated. Also the data base related to this problem has not been developed. Therefore, the government and electric utility industry began the research on the prevention of the accidents caused by human errors. The basic research is carried out by the government, and the applied research is done by electric utility industry. The Central Research Institute of Electric Power Industry established the Human Factors Research Center on July 1, 1987. The research program in the Human Factors Research Center is divided into the basic research to clarity fundamental human characteristics, the systematic research to apply this information and the analytical research on human error experience. These research activities are reported. (Kako, I.)

  1. Human Reliability Assessment and Human Performance Evaluation: Research and Analysis Activities at the U.S. NRC

    International Nuclear Information System (INIS)

    Ramey-Smith, A.M.

    1998-01-01

    The author indicates the themes of the six programs identified by the US NRC mission on human performance and human reliability activities. They aim at developing the technical basis to support human performance, at developing and updating a model of human performance and human reliability, at fostering national and international dialogue and cooperation efforts on human performance evaluation, at conducting operating events analysis and database development, and at providing support to human performance and human reliability inspection

  2. Human reliability analysis in Loviisa probabilistic safety analysis

    International Nuclear Information System (INIS)

    Illman, L.; Isaksson, J.; Makkonen, L.; Vaurio, J.K.; Vuorio, U.

    1986-01-01

    The human reliability analysis in the Loviisa PSA project is carried out for three major groups of errors in human actions: (A) errors made before an initiating event, (B) errors that initiate a transient and (C) errors made during transients. Recovery possibilities are also included in each group. The methods used or planned for each group are described. A simplified THERP approach is used for group A, with emphasis on test and maintenance error recovery aspects and dependencies between redundancies. For group B, task analyses and human factors assessments are made for startup, shutdown and operational transients, with emphasis on potential common cause initiators. For group C, both misdiagnosis and slow decision making are analyzed, as well as errors made in carrying out necessary or backup actions. New or advanced features of the methodology are described

  3. Meeting Human Reliability Requirements through Human Factors Design, Testing, and Modeling

    Energy Technology Data Exchange (ETDEWEB)

    R. L. Boring

    2007-06-01

    In the design of novel systems, it is important for the human factors engineer to work in parallel with the human reliability analyst to arrive at the safest achievable design that meets design team safety goals and certification or regulatory requirements. This paper introduces the System Development Safety Triptych, a checklist of considerations for the interplay of human factors and human reliability through design, testing, and modeling in product development. This paper also explores three phases of safe system development, corresponding to the conception, design, and implementation of a system.

  4. Demonstration Integrated Knowledge-Based System for Estimating Human Error Probabilities

    Energy Technology Data Exchange (ETDEWEB)

    Auflick, Jack L.

    1999-04-21

    Human Reliability Analysis (HRA) is currently comprised of at least 40 different methods that are used to analyze, predict, and evaluate human performance in probabilistic terms. Systematic HRAs allow analysts to examine human-machine relationships, identify error-likely situations, and provide estimates of relative frequencies for human errors on critical tasks, highlighting the most beneficial areas for system improvements. Unfortunately, each of HRA's methods has a different philosophical approach, thereby producing estimates of human error probabilities (HEPs) that area better or worse match to the error likely situation of interest. Poor selection of methodology, or the improper application of techniques can produce invalid HEP estimates, where that erroneous estimation of potential human failure could have potentially severe consequences in terms of the estimated occurrence of injury, death, and/or property damage.

  5. Concept development of the human reliability data bank

    International Nuclear Information System (INIS)

    Miller, D.P.

    1984-01-01

    The US Nuclear Regulatory Commission and Sandia National Laboratories initiated a three-phased research program in 1981 to develop a plan for a human reliability data bank. This research initiative was in response to the data needs of the nuclear power industry's probabilistic risk assessment community. The three phases are: (1) develop the data bank concept; (2) develop an implementation plan and conduct a feasibility test; and (3) assist the sponsor in implementing the data bank. This paper briefly describes some of the results of the work performed during Phase A and outlines the program elements schedules for Phase B

  6. Launch and Assembly Reliability Analysis for Human Space Exploration Missions

    Science.gov (United States)

    Cates, Grant; Gelito, Justin; Stromgren, Chel; Cirillo, William; Goodliff, Kandyce

    2012-01-01

    NASA's future human space exploration strategy includes single and multi-launch missions to various destinations including cis-lunar space, near Earth objects such as asteroids, and ultimately Mars. Each campaign is being defined by Design Reference Missions (DRMs). Many of these missions are complex, requiring multiple launches and assembly of vehicles in orbit. Certain missions also have constrained departure windows to the destination. These factors raise concerns regarding the reliability of launching and assembling all required elements in time to support planned departure. This paper describes an integrated methodology for analyzing launch and assembly reliability in any single DRM or set of DRMs starting with flight hardware manufacturing and ending with final departure to the destination. A discrete event simulation is built for each DRM that includes the pertinent risk factors including, but not limited to: manufacturing completion; ground transportation; ground processing; launch countdown; ascent; rendezvous and docking, assembly, and orbital operations leading up to trans-destination-injection. Each reliability factor can be selectively activated or deactivated so that the most critical risk factors can be identified. This enables NASA to prioritize mitigation actions so as to improve mission success.

  7. 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)

  8. 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.

  9. 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

  10. Human reliability assessment on the basis of operating experience

    International Nuclear Information System (INIS)

    Straeter, O.

    1997-01-01

    For development of methodology, available models for qualitative assessment of human errors (e.g. by Swain, Hacker, Rasmussen) and a variety of known systematic approaches for quantitiative assessment of inadequate human action (e.g. THERP, ASEP, HCR, SLIM) were taken as a basis to establish a job specification, which in turn was used for developing a method for acquisition, characterisation and evaluation of errors. This method encompasses the two processes of event analysis and event evaluation: The first step comprises analysis of events by analysis of information describing the conditions and scenarios of relevance to the inadequate human action examined. In addition to the description of process sequences, information is taken into account on possible conditions that may bring about failure. As an assessment of human reliability requires manifold approaches for evaluation, a connectionistic procedure was developed for evaluation of the compilation of events based on a debate about various approaches from the domain of artificial intelligence (AI). This procedure yields both qualitative and quantitative information through a homogenous approach. (orig./GL) [de

  11. 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)

  12. Process evaluation of the human reliability data bank

    International Nuclear Information System (INIS)

    Miller, D.P.; Comer, K.

    1985-01-01

    The US Nuclear Regulatory Commission and Sandia National Laboratories have been developing a plan for a human reliability data bank since August 1981. This research is in response to the data need of the nuclear power industry's probabilistic risk assessment community. The three phases of the program are to: (a) develop the data bank concept, (b) develop an implementation plan and conduct a process evaluation, and (c) assist a sponsor in implementing the data bank. The program is now in Phase B. This paper describes the methods used and the results of the process evaluation. Decisions to be made in the future regarding full-scale implementation will be based, in part, on the outcome of this study

  13. Process evaluation of the human reliability data bank

    International Nuclear Information System (INIS)

    Miller, D.P.; Comer, K.

    1984-01-01

    The US Nuclear Regulatory Commission and Sandia National Laboratories have been developing a plan for a human reliability data bank since August 1981. This research is in response to the data needs of the nuclear power industry's probabilistic risk assessment community. The three phases of the program are to: (A) develop the data bank concept, (B) develop an implementation plan and conduct a process evaluation, and (C) assist a sponsor in implementing the data bank. The program is now in Phase B. This paper describes the methods used and the results of the process evaluation. Decisions to be made in the future regarding full-scale implementation will be based in part on the outcome of this study

  14. Simulator training and human factor reliability in Kozloduy NPP, Bulgaria

    International Nuclear Information System (INIS)

    Stoychev, Kosta

    2007-01-01

    This is a PowerPoint presentation. Situated in North Bulgaria, in the vicinity of the town of Kozloduy, near the Danube River bank, there is the Bulgarian Kozloduy Nuclear Power plant operating four WWER-440 and two WWER-1000 units. Units 1 and 2 were commissioned in July, 1974 and November, 1975, respectively. These were shut down at the end of 2003. Units 3 and 4 were commissioned in December, 1980 and May, 1982. They were shut down at the end of 2006 as a precondition for Bulgaria's accession to the European Union. The 1000 MW units 5 and 6 of Kozloduy NPP were commissioned in September, 1988 and December, 1993, respectively. Large-scale modernization have been implemented and now the units meet all international safety standards. The paper describes the multifunctional simulator Kozloduy NPP for the operational staff training. The training stages are as follows: - Preparatory; -Theoretical studies; - Training at the Training Centre by means of technical devices; - Preparation and sitting for an exam before a Kozloduy NPP expert commission; - Simulator training ; - Preparation to obtain a permit for a license, corresponding to the position to begin work at the NPP; - Exams before the Nuclear Regulatory Agency (NRA) and licensing; - Shadow training at the working place; - Permission for unaided operation. The following positions are addressed by the simulator training: - Chief Plant Supervisor; - Shift Unit Supervisor; - Senior Reactor Operator; - Simulator Instructor; - Controller physicist; -Senior Turbine Operator; - Senior Operator of Turbine Feedwater Pumps of Kozloduy NPP. Improving of training method led to a reduction of number of significant events while worldwide practice proves that improvement of engineering resulted in an increase in the percentage of events, related to human factor. Analysis of human reliability in 2005 and 2006 in cooperation with representatives from Great Britain and the Technical University in Sofia were worked on the DTI NSP B

  15. Sample Size Bounding and Context Ranking as Approaches to the Human Error Quantification Problem

    Energy Technology Data Exchange (ETDEWEB)

    Reer, B

    2004-03-01

    The paper describes a technique denoted as Sub-Sample-Size Bounding (SSSB), which is useable for the statistical derivation of context-specific probabilities from data available in existing reports on operating experience. Applications to human reliability analysis (HRA) are emphasised in the presentation of this technique. Exemplified by a sample of 180 abnormal event sequences, the manner in which SSSB can provide viable input for the quantification of errors of commission (EOCs) are outlined. (author)

  16. Sample Size Bounding and Context Ranking as Approaches to the Human Error Quantification Problem

    International Nuclear Information System (INIS)

    Reer, B.

    2004-01-01

    The paper describes a technique denoted as Sub-Sample-Size Bounding (SSSB), which is useable for the statistical derivation of context-specific probabilities from data available in existing reports on operating experience. Applications to human reliability analysis (HRA) are emphasised in the presentation of this technique. Exemplified by a sample of 180 abnormal event sequences, the manner in which SSSB can provide viable input for the quantification of errors of commission (EOCs) are outlined. (author)

  17. 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.)

  18. Generating human reliability estimates using expert judgment. Volume 2. Appendices

    International Nuclear Information System (INIS)

    Comer, M.K.; Seaver, D.A.; Stillwell, W.G.; Gaddy, C.D.

    1984-11-01

    The US Nuclear Regulatory Commission is conducting a research program to determine the practicality, acceptability, and usefulness of several different methods for obtaining human reliability data and estimates that can be used in nuclear power plant probabilistic risk assessments (PRA). One method, investigated as part of this overall research program, uses expert judgment to generate human error probability (HEP) estimates and associated uncertainty bounds. The project described in this document evaluated two techniques for using expert judgment: paired comparisons and direct numerical estimation. Volume 2 provides detailed procedures for using the techniques, detailed descriptions of the analyses performed to evaluate the techniques, and HEP estimates generated as part of this project. The results of the evaluation indicate that techniques using expert judgment should be given strong consideration for use in developing HEP estimates. Judgments were shown to be consistent and to provide HEP estimates with a good degree of convergent validity. Of the two techniques tested, direct numerical estimation appears to be preferable in terms of ease of application and quality of results

  19. 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.

  20. Condition-based Human Reliability Assessment for digitalized control room

    International Nuclear Information System (INIS)

    Kang, H. G.; Jang, S. C.; Eom, H. S.; Ha, J. J.

    2005-04-01

    In safety-critical systems, the generation failure of an actuation signal is caused by the concurrent failures of the automated systems and an operator action. These two sources of safety signals are complicatedly correlated. The failures of sensors or automated systems will cause a lack of necessary information for a human operator and result in error-forcing contexts such as the loss of corresponding alarms and indications. In the conventional analysis, the Human Error Probabilities (HEP) are estimated based on the assumption of 'normal condition of indications and alarms'. In order to construct a more realistic signal-generation failure model, we have to consider more complicated conditions in a more realistic manner. In this study, we performed two kinds of investigation for addressing this issue. We performed the analytic calculations for estimating the effect of sensors failures on the system unavailability and plant risk. For the single-parameter safety signals, the analysis result reveals that the quantification of the HEP should be performed by focusing on the 'no alarm from the automatic system and corresponding indications unavailable' situation. This study also proposes a Condition-Based Human Reliability Assessment (CBHRA) method in order to address these complicated conditions in a practical way. We apply the CBHRA method to the manual actuation of the safety features such as a reactor trip and auxiliary feedwater actuation in Korean Standard Nuclear Power Plants. In the case of conventional single HEP method, it is very hard to consider the multiple HE conditions. The merit of CBHRA is clearly shown in the application to the AFAS generation where no dominating HE condition exits. In this case, even if the HE conditions are carefully investigated, the single HEP method cannot accommodate the multiple conditions in a fault tree. On the other hand, the application result of the reactor trip in SLOCA shows that if there is a dominating condition, the use

  1. 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

  2. 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

  3. The Concept of Human Error and the Design of Reliable Human-Machine Systems

    DEFF Research Database (Denmark)

    Rasmussen, Jens

    1995-01-01

    The concept of human error is unreliable as a basis for design of reliable human-machine systems. Humans are basically highly adaptive and 'errors' are closely related to the process of adaptation and learning. Therefore, reliability of system operation depends on an interface that is not designed...... so as to support a pre-conceived operating procedure, but, instead, makes visible the deep, functional structure of the system together with the boundaries of acceptable operation in away that allows operators to 'touch' the boundaries and to learn to cope with the effects of errors in a reversible...... way. The concepts behind such 'ecological' interfaces are discussed, an it is argued that a 'typology' of visualization concepts is a pressing research need....

  4. 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)

  5. Human reliability and human factors in complex organizations: epistemological and critical analysis - practical avenues to action

    International Nuclear Information System (INIS)

    Llory, A.

    1991-08-01

    This article starts out with comment on the existence of persistent problems inherent to probabilistic safety assessments (PSA). It first surveys existing American documents on the subject which make a certain number of criticisms on human reliability analyses, e.g. limitations due to the scant quantities of data available, lack of a basic theoretical model, non-reproducibility of analyses, etc. The article therefore examines and criticizes the epistemological bases of these analyses. One of the fundamental points stressed is that human reliability analyses do not take account of all the special features of the work situation which result in human error (so as to draw up statistical data from a sufficiently representative number of cases), and consequently lose all notion of the 'relationships' between human errors and the different aspects of the working environment. The other key points of criticism concern the collective nature of work which is not taken into account, and the frequent confusion between what operatives actually do and their formally prescribed job-tasks. The article proposes aspects to be given thought in order to overcome these difficulties, e.g. quantitative assessment of the social environment within a company, non-linear model for assessment of the accident rate, analysis of stress levels in staff on off-shore platforms. The method approaches used in these three studies are of the same type, and could be transposed to human-reliability problems. The article then goes into greater depth on thinking aimed at developing a 'positive' view of the human factor (and not just a 'negative' one, i.e. centred on human errors and organizational malfunctions), applying investigation methods developed in the occupational human sciences (occupational psychodynamics, ergonomics, occupational sociology). The importance of operatives working as actors of a team is stressed

  6. The Measure of Human Error: Direct and Indirect Performance Shaping Factors

    Energy Technology Data Exchange (ETDEWEB)

    Ronald L. Boring; Candice D. Griffith; Jeffrey C. Joe

    2007-08-01

    The goal of performance shaping factors (PSFs) is to provide measures to account for human performance. PSFs fall into two categories—direct and indirect measures of human performance. While some PSFs such as “time to complete a task” are directly measurable, other PSFs, such as “fitness for duty,” can only be measured indirectly through other measures and PSFs, such as through fatigue measures. This paper explores the role of direct and indirect measures in human reliability analysis (HRA) and the implications that measurement theory has on analyses and applications using PSFs. The paper concludes with suggestions for maximizing the reliability and validity of PSFs.

  7. 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

  8. Human Reliability and the Current Dilemma in Human-Machine Interface Design Strategies

    International Nuclear Information System (INIS)

    Passalacqua, Roberto; Yamada, Fumiaki

    2002-01-01

    Since human error dominates the probability of failures of still-existing human-requiring systems (as the Monju reactor), the human-machine interface needs to be improved. Several rationales may lead to the conclusion that 'humans' should limit themselves to monitor the 'machine'. For example, this is the trend in the aviation industry: newest aircrafts are designed to be able to return to a safe state by the use of control systems, which do not need human intervention. Thus, the dilemma whether we really need operators (for example in the nuclear industry) might arise. However, social-technical approaches in recent human error analyses are pointing out the so-called 'organizational errors' and the importance of a human-machine interface harmonization. Typically plant's operators are a 'redundant' safety system with a much lower reliability (than the machine): organizational factors and harmonization requirements suggest designing the human-machine interface in a way that allows improvement of operator's reliability. In addition, taxonomy studies of accident databases have also proved that operators' training should promote processes of decision-making. This is accomplished in the latest trends of PSA technology by introducing the concept of a 'Safety Monitor' that is a computer-based tool that uses a level 1 PSA model of the plant. Operators and maintenance schedulers of the Monju FBR will be able to perform real-time estimations of the plant risk level. The main benefits are risk awareness and improvements in decision-making by operators. Also scheduled maintenance can be approached in a more rational (safe and economic) way. (authors)

  9. Human event observations in the individual plant examinations

    International Nuclear Information System (INIS)

    Lois, E.; Forester, J.

    1994-01-01

    A main objective of the Nuclear Regulatory Commission's (NRC) Individual Plant Examination (IPE) Insights Program is to document significant safety insights relative to 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 NRCs 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 results from the IPEs

  10. 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.

  11. Human reliability data, human error and accident models--illustration through the Three Mile Island accident analysis

    International Nuclear Information System (INIS)

    Le Bot, Pierre

    2004-01-01

    Our first objective is to provide a panorama of Human Reliability data used in EDF's Safety Probabilistic Studies, and then, since these concepts are at the heart of Human Reliability and its methods, to go over the notion of human error and the understanding of accidents. We are not sure today that it is actually possible to provide in this field a foolproof and productive theoretical framework. Consequently, the aim of this article is to suggest potential paths of action and to provide information on EDF's progress along those paths which enables us to produce the most potentially useful Human Reliability analyses while taking into account current knowledge in Human Sciences. The second part of this article illustrates our point of view as EDF researchers through the analysis of the most famous civil nuclear accident, the Three Mile Island unit accident in 1979. Analysis of this accident allowed us to validate our positions regarding the need to move, in the case of an accident, from the concept of human error to that of systemic failure in the operation of systems such as a nuclear power plant. These concepts rely heavily on the notion of distributed cognition and we will explain how we applied it. These concepts were implemented in the MERMOS Human Reliability Probabilistic Assessment methods used in the latest EDF Probabilistic Human Reliability Assessment. Besides the fact that it is not very productive to focus exclusively on individual psychological error, the design of the MERMOS method and its implementation have confirmed two things: the significance of qualitative data collection for Human Reliability, and the central role held by Human Reliability experts in building knowledge about emergency operation, which in effect consists of Human Reliability data collection. The latest conclusion derived from the implementation of MERMOS is that, considering the difficulty in building 'generic' Human Reliability data in the field we are involved in, the best

  12. Team performance modeling for HRA in dynamic situations

    International Nuclear Information System (INIS)

    Shu Yufei; Furuta, Kazuo; Kondo, Shunsuke

    2002-01-01

    This paper proposes a team behavior network model that can simulate and analyze response of an operator team to an incident in a dynamic and context-sensitive situation. The model is composed of four sub-models, which describe the context of team performance. They are task model, event model, team model and human-machine interface model. Each operator demonstrates aspects of his/her specific cognitive behavior and interacts with other operators and the environment in order to deal with an incident. Individual human factors, which determine the basis of communication and interaction between individuals, and cognitive process of an operator, such as information acquisition, state-recognition, decision-making and action execution during development of an event scenario are modeled. A case of feed and bleed operation in pressurized water reactor under an emergency situation was studied and the result was compared with an experiment to check the validity of the proposed model

  13. A computational method for probabilistic safety assessment of I and C systems and human operators in nuclear power plants

    International Nuclear Information System (INIS)

    Kim, Man Cheol; Seong, Poong Hyun

    2006-01-01

    To make probabilistic safety assessment (PSA) more realistic, the improvements of human reliability analysis (HRA) are essential. But, current HRA methods have many limitations including the lack of considerations on the interdependency between instrumentation and control (I and C) systems and human operators, and lack of theoretical basis for situation assessment of human operators. To overcome these limitations, we propose a new method for the quantitative safety assessment of I and C systems and human operators. The proposed method is developed based on the computational models for the knowledge-driven monitoring and the situation assessment of human operators, with the consideration of the interdependency between I and C systems and human operators. The application of the proposed method to an example situation demonstrates that the quantitative description by the proposed method for a probable scenario well matches with the qualitative description of the scenario. It is also demonstrated that the proposed method can probabilistically consider all possible scenarios and the proposed method can be used to quantitatively evaluate the effects of various context factor on the safety of nuclear power plants. In our opinion, the proposed method can be used as the basis for the development of advanced HRA methods

  14. 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.

  15. Human Reliability and the Cost of Doing Business

    Science.gov (United States)

    DeMott, D. L.

    2014-01-01

    Human error cannot be defined unambiguously in advance of it happening, it often becomes an error after the fact. The same action can result in a tragic accident for one situation or a heroic action given a more favorable outcome. People often forget that we employ humans in business and industry for the flexibility and capability to change when needed. In complex systems, operations are driven by their specifications of the system and the system structure. People provide the flexibility to make it work. Human error has been reported as being responsible for 60%-80% of failures, accidents and incidents in high-risk industries. We don't have to accept that all human errors are inevitable. Through the use of some basic techniques, many potential human error events can be addressed. There are actions that can be taken to reduce the risk of human error.

  16. A priori and a posteriori approaches in human reliability

    International Nuclear Information System (INIS)

    Griffon-Fouco, M.; Gagnolet, P.

    1981-09-01

    The French atomic energy commission (CEA) and the French supplier in electric power (EDF) have joint studies on human factors in nuclear safety. This paper deals with these studies which are a combination of two approaches: - An a posteriori approach so as to know the rate of human errors and their causes: an analysis of incident data banks and an analysis of human errors on simulator are presented. - An a priori approach so as to know the potential factors of human errors: an analysis of the control rooms design and an analysis of the writing of procedures are presented. The possibility to take into account these two approaches to prevent and quantify human errors is discussed

  17. Human reliability in high dose rate afterloading radiotherapy based on FMECA

    International Nuclear Information System (INIS)

    Deng Jun; Fan Yaohua; Yue Baorong; Wei Kedao; Ren Fuli

    2012-01-01

    Objective: To put forward reasonable and feasible recommendations against the procedure with relative high risk during the high dose rate (HDR) afterloading radiotherapy, so as to enhance its clinical application safety, through studying the human reliability in the process of carrying out the HDR afterloading radiotherapy. Methods: Basic data were collected by on-site investigation and process analysis as well as expert evaluation. Failure mode, effect and criticality analysis (FMECA) employed to study the human reliability in the execution of HDR afterloading radiotherapy. Results: The FMECA model of human reliability for HDR afterloading radiotherapy was established, through which 25 procedures with relative high risk index were found,accounting for 14.1% of total 177 procedures. Conclusions: FMECA method in human reliability study for HDR afterloading radiotherapy is feasible. The countermeasures are put forward to reduce the human error, so as to provide important basis for enhancing clinical application safety of HDR afterloading radiotherapy. (authors)

  18. Sensitivity evaluation of human factors for reliability of the containment spray system

    International Nuclear Information System (INIS)

    Tsujimura, Yasuhiro; Suzuki, Eiji

    1988-01-01

    Evaluation of the human reliability is one of the most difficult problems that deal with the safety and reliability of large systems, especially of the Engineered Safety Features (ESF) of the nuclear power plant. Influences of human factors on the reliability of the Containment Spray System in the ESF were estimated by using the FTA method in this paper. As a result, the adequacy of the system structure and the effects of human factors on variations of the design of the system structure were explained. (author)

  19. Human error and the associated recovery probabilities for soft control being used in the advanced MCRs of NPPs

    International Nuclear Information System (INIS)

    Jang, Inseok; Jung, Wondea; Seong, Poong Hyun

    2016-01-01

    Highlights: • The operation environment of MCRs in NPPs has changed by adopting digital HSIs. • Most current HRA databases are not explicitly designed to deal with digital HSI. • Empirical analysis for new HRA DB under an advanced MCR mockup are carried. • It is expected that the results can be used for advanced MCR HRA. - 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 studies were focused on considering the conventional Main Control Room (MCR) environment. However, the operating environment of MCRs in NPPs has changed with the adoption of new human-system interfaces (HSI) largely based on up-to-date digital 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 because operating actions in advanced MCRs are performed by soft control. Due to the difference in interfaces between soft control and hardwired conventional controls, different HEP should be used in the HRA for advanced MCRs. Unfortunately, most current HRA databases deal with operations in conventional MCRs and are not explicitly designed to deal with digital Human System Interface (HSI). For this reason, empirical human error and the associated error recovery probabilities were collected from the mockup of an advanced MCR equipped with soft controls. To this end, small-scaled experiments are conducted with 48 graduated students in the department of nuclear engineering in Korea Advanced Institute of Science and Technology (KAIST) are participated, and accident scenarios are designed with respect to the typical Design Basis Accidents (DBAs) in NPPs, such as Steam Generator Tube Rupture

  20. Expanding the human-machine interface model to address the effect of leadership and management on performance

    International Nuclear Information System (INIS)

    Briant, V.S.; Childress, J.R.; Hannaman, G.W.

    1988-01-01

    The US nuclear industry now focuses on improving plant performance in measurable areas such as availability, safety, and operations. Risk assessment and pioneering work in human reliability analysis (HRA) have provided methods to identify and prioritize numerous design improvements. Improvements such as control room design, training, and procedures have contributed positively to plant performance. Human performance is increasingly recognized as a fundamental contributor to safe, economic, and reliable operation. Industry leaders suggest that improved leadership and management are keys to enhanced plant performance. This paper identifies several critical aspects of individual and group behavior that, if managed, could significantly contribute to improved performance. Some existing tools for measuring performance are cited

  1. Visualizing team characteristics from the HRA perspective-its feasibility

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jinkyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In this study, as a method to visualize the nature of the team characteristics, the applicability of Hofstede's culture model is investigated. In this study, as a promising solution for visualizing the team characteristics, the applicability of Hofstede's culture model is briefly investigated. As a result, it is expected that the use of Hofstede's culture model allows us to have another viewpoint that is helpful for understanding the relationship between the team characteristics and the associated team performance. Actually, the following excerpts advocate the importance of a cultural aspect in understanding the responses of operating personnel working in NPPs: 'However, the Fukushima Daiichi NPP accident identified significant and new human, organisational and cultural challenges that also need to be addressed. The accident revealed the importance of applying existing concepts and technical knowledge throughout the decision-making processes for design, operation and accident management.' (p. 54) 'Nuclear industry culture has been described as culture of control where organisations and workers emphasise that risks are in control and they do not appreciate the inherent uncertainties. This has also been recognised by some of the Fukushima accident investigations. In reality many activities in the nuclear industry, for example maintenance, design work, construction of new plants and emergency management involve dealing with unforeseen situations and performing underspecified work tasks. In order to smoothly cope with these the culture of the organisation need to support flexibility and adaptability to some degree.' (p. 19) Therefore, although this study is still premature, it is reasonable to say that this study is a good starting point to scrutinize the nature of the team characteristics in a systematic manner, at least to some extent.

  2. An Impact of Thermodynamic Processes in Human Bodies on Performance Reliability of Individuals

    Directory of Open Access Journals (Sweden)

    Smalko Zbigniew

    2015-01-01

    Full Text Available The article presents the problem of the influence of thermodynamic factors on human fallibility in different zones of thermal discomfort. Describes the processes of energy in the human body. Been given a formal description of the energy balance of the human body thermoregulation. Pointed to human reactions to temperature changes of internal and external environment, including reactions associated with exercise. The methodology to estimate and determine the reliability of indicators of human basal acting in different zones of thermal discomfort. The significant effect of thermodynamic factors on the reliability and security ofperson.

  3. Comparative reliability of cheiloscopy and palatoscopy in human identification

    Directory of Open Access Journals (Sweden)

    Sharma Preeti

    2009-01-01

    Full Text Available Background: Establishing a person′s identity in postmortem scenarios can be a very difficult process. Dental records, fingerprint and DNA comparisons are probably the most common techniques used in this context, allowing fast and reliable identification processes. However, under certain circumstances they cannot always be used; sometimes it is necessary to apply different and less known techniques. In forensic identification, lip prints and palatal rugae patterns can lead us to important information and help in a person′s identification. This study aims to ascertain the use of lip prints and palatal rugae pattern in identification and sex differentiation. Materials and Methods: A total of 100 subjects, 50 males and 50 females were selected from among the students of Subharti Dental College, Meerut. The materials used to record lip prints were lipstick, bond paper, cellophane tape, a brush for applying the lipstick, and a magnifying lens. To study palatal rugae, alginate impressions were taken and the dental casts analyzed for their various patterns. Results: Statistical analysis (applying Z-test for proportion showed significant difference for type I, I′, IV and V lip patterns (P < 0.05 in males and females, while no significant difference was observed for the same in the palatal rugae patterns (P > 0.05. Conclusion: This study not only showed that palatal rugae and lip prints are unique to an individual, but also that lip prints is more reliable for recognition of the sex of an individual.

  4. The Answer Is Blowing in the Wind. Investment in Training from a Human Resource Accounting Perspective.

    Science.gov (United States)

    Johanson, Ulf

    1998-01-01

    Presents components of human resource accounting (HRA)--description of human resource costs, estimation of return on investment, estimation of human resource values. Reviews research on the influence of HRA on decision making, concluding that a number of factors inhibit its effective use. (SK)

  5. Analysis Testing of Sociocultural Factors Influence on Human Reliability within Sociotechnical Systems: The Algerian Oil Companies

    Directory of Open Access Journals (Sweden)

    Abdelbaki Laidoune

    2016-09-01

    Conclusion: The explored sociocultural factors influence the human reliability both in qualitative and quantitative manners. The proposed model shows how reliability can be enhanced by some measures such as experience feedback based on, for example, safety improvements, training, and information. With that is added the continuous systems improvements to improve sociocultural reality and to reduce negative behaviors.

  6. Application of expert elicitation techniques in human reliability, assessment

    International Nuclear Information System (INIS)

    Sanyasi Rao, V.V.S.; Saraf, R.K.; Ghosh, A.K.; Kushwaha, H.S.

    2006-01-01

    Expert elicitation techniques are being used, in the area of technological forecasting, in estimating data needed for analysis when it is either difficult to arrive at the data by experimental means or when it is quite involved to plan and conduct the experiment. In this study, expert elicitation techniques are applied to the evaluation of the frequencies of the various accident sequences that can result from the initiating event (IE) 'High Pressure Process Water (HPPW) system failure' in typical Indian Pressurised Heavy Water Reactor (IPHWR) of the older generation. The Operating Procedure under Emergency Conditions (OPEC) for this IE involves human actions according to a pre-defined procedure. The Human Error Probabilities for all these human actions are obtained using expert elicitation techniques. These techniques aim at eliciting the opinion of the experts in the area of interest with regard to the issue in question. The uncertainty is analysed by employing the measure of dissonance and the most probable range of human error probabilities are arrived at by maximizing this measure. These values are combined using the same procedures mentioned above to yield a distribution representing the uncertainty associated with the predictions. (author)

  7. On modeling human reliability in space flights - Redundancy and recovery operations

    Science.gov (United States)

    Aarset, M.; Wright, J. F.

    The reliability of humans is of paramount importance to the safety of space flight systems. This paper describes why 'back-up' operators might not be the best solution, and in some cases, might even degrade system reliability. The problem associated with human redundancy calls for special treatment in reliability analyses. The concept of Standby Redundancy is adopted, and psychological and mathematical models are introduced to improve the way such problems can be estimated and handled. In the past, human reliability has practically been neglected in most reliability analyses, and, when included, the humans have been modeled as a component and treated numerically the way technical components are. This approach is not wrong in itself, but it may lead to systematic errors if too simple analogies from the technical domain are used in the modeling of human behavior. In this paper redundancy in a man-machine system will be addressed. It will be shown how simplification from the technical domain, when applied to human components of a system, may give non-conservative estimates of system reliability.

  8. An Analysis and Quantification Method of Human Errors of Soft Controls in Advanced MCRs

    International Nuclear Information System (INIS)

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

    2011-01-01

    In this work, a method was proposed for quantifying human errors that may occur during operation executions using soft control. Soft controls of advanced main control rooms (MCRs) have totally different features from conventional controls, and thus they may have different human error modes and occurrence probabilities. It is important to define the human error modes and to quantify the error probability for evaluating the reliability of the system and preventing errors. This work suggests a modified K-HRA method for quantifying error probability

  9. Research on cognitive reliability model for main control room considering human factors in nuclear power plants

    International Nuclear Information System (INIS)

    Jiang Jianjun; Zhang Li; Wang Yiqun; Zhang Kun; Peng Yuyuan; Zhou Cheng

    2012-01-01

    Facing the shortcomings of the traditional cognitive factors and cognitive model, this paper presents a Bayesian networks cognitive reliability model by taking the main control room as a reference background and human factors as the key points. The model mainly analyzes the cognitive reliability affected by the human factors, and for the cognitive node and influence factors corresponding to cognitive node, a series of methods and function formulas to compute the node cognitive reliability is proposed. The model and corresponding methods can be applied to the evaluation of cognitive process for the nuclear power plant operators and have a certain significance for the prevention of safety accidents in nuclear power plants. (authors)

  10. Cognitive modelling: a basic complement of human reliability analysis

    International Nuclear Information System (INIS)

    Bersini, U.; Cacciabue, P.C.; Mancini, G.

    1988-01-01

    In this paper the issues identified in modelling humans and machines are discussed in the perspective of the consideration of human errors managing complex plants during incidental as well as normal conditions. The dichotomy between the use of a cognitive versus a behaviouristic model approach is discussed and the complementarity aspects rather than the differences of the two methods are identified. A cognitive model based on a hierarchical goal-oriented approach and driven by fuzzy logic methodology is presented as the counterpart to the 'classical' THERP methodology for studying human errors. Such a cognitive model is discussed at length and its fundamental components, i.e. the High Level Decision Making and the Low Level Decision Making models, are reviewed. Finally, the inadequacy of the 'classical' THERP methodology to deal with cognitive errors is discussed on the basis of a simple test case. For the same case the cognitive model is then applied showing the flexibility and adequacy of the model to dynamic configuration with time-dependent failures of components and with consequent need for changing of strategy during the transient itself. (author)

  11. 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

  12. 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

  13. Advanced control rooms and crew performance issues: Implications for human reliability

    International Nuclear Information System (INIS)

    O'Hara, J.M.; Hall, R.E.

    1991-01-01

    Recent trends in advanced control room (ACR) design are considered with respect to their impact on human performance. It is concluded that potentially negative influences exist, however, a variety of factors make it difficult to model, analyze, and quantify these effects for human reliability analyses (HRAs)

  14. Development of slim-maud: a multi-attribute utility approach to human reliability evaluation

    International Nuclear Information System (INIS)

    Embrey, D.E.

    1984-01-01

    This paper describes further work on the Success Likelihood Index Methodology (SLIM), a procedure for quantitatively evaluating human reliability in nuclear power plants and other systems. SLIM was originally developed by Human Reliability Associates during an earlier contract with Brookhaven National Laboratory (BNL). A further development of SLIM, SLIM-MAUD (Multi-Attribute Utility Decomposition) is also described. This is an extension of the original approach using an interactive, computer-based system. All of the work described in this report was supported by the Human Factors and Safeguards Branch of the US Nuclear Regulatory Commission

  15. Role of frameworks, models, data, and judgment in human reliability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Hannaman, G W

    1986-05-01

    Many advancements in the methods for treating human interactions in PRA studies have occurred in the last decade. These advancements appear to increase the capability of PRAs to extend beyond just the assessment of the human's importance to safety. However, variations in the application of these advanced models, data, and judgements in recent PRAs make quantitative comparisons among studies extremely difficult. This uncertainty in the analysis diminishes the usefulness of the PRA study for upgrading procedures, enhancing traning, simulator design, technical specification guidance, and for aid in designing the man-machine interface. Hence, there is a need for a framework to guide analysts in incorporating human interactions into the PRA systems analyses so that future users of a PRA study will have a clear understanding of the approaches, models, data, and assumptions which were employed in the initial study. This paper describes the role of the systematic human action reliability procedure (SHARP) in providing a road map through the complex terrain of human reliability that promises to improve the reproducibility of such analysis in the areas of selecting the models, data, representations, and assumptions. Also described is the role that a human cognitive reliability model can have in collecting data from simulators and helping analysts assign human reliability parameters in a PRA study. Use of these systematic approaches to perform or upgrade existing PRAs promises to make PRA studies more useful as risk management tools.

  16. Human reliability in non-destructive inspections of nuclear power plant components: modeling and analysis

    International Nuclear Information System (INIS)

    Vasconcelos, Vanderley de; Soares, Wellington Antonio; Marques, Raíssa Oliveira; Silva Júnior, Silvério Ferreira da; Raso, Amanda Laureano

    2017-01-01

    Non-destructive inspection (NDI) is one of the key elements in ensuring quality of engineering systems and their safe use. NDI is a very complex task, during which the inspectors have to rely on their sensory, perceptual, cognitive, and motor skills. It requires high vigilance once it is often carried out on large components, over a long period of time, and in hostile environments and restriction of workplace. A successful NDI requires careful planning, choice of appropriate NDI methods and inspection procedures, as well as qualified and trained inspection personnel. A failure of NDI to detect critical defects in safety-related components of nuclear power plants, for instance, may lead to catastrophic consequences for workers, public and environment. Therefore, ensuring that NDI methods are reliable and capable of detecting all critical defects is of utmost importance. Despite increased use of automation in NDI, human inspectors, and thus human factors, still play an important role in NDI reliability. Human reliability is the probability of humans conducting specific tasks with satisfactory performance. Many techniques are suitable for modeling and analyzing human reliability in NDI of nuclear power plant components. Among these can be highlighted Failure Modes and Effects Analysis (FMEA) and THERP (Technique for Human Error Rate Prediction). The application of these techniques is illustrated in an example of qualitative and quantitative studies to improve typical NDI of pipe segments of a core cooling system of a nuclear power plant, through acting on human factors issues. (author)

  17. Human reliability in non-destructive inspections of nuclear power plant components: modeling and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Vasconcelos, Vanderley de; Soares, Wellington Antonio; Marques, Raíssa Oliveira; Silva Júnior, Silvério Ferreira da; Raso, Amanda Laureano, E-mail: vasconv@cdtn.br, E-mail: soaresw@cdtn.br, E-mail: raissaomarques@gmail.com, E-mail: silvasf@cdtn.br, E-mail: amandaraso@hotmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    Non-destructive inspection (NDI) is one of the key elements in ensuring quality of engineering systems and their safe use. NDI is a very complex task, during which the inspectors have to rely on their sensory, perceptual, cognitive, and motor skills. It requires high vigilance once it is often carried out on large components, over a long period of time, and in hostile environments and restriction of workplace. A successful NDI requires careful planning, choice of appropriate NDI methods and inspection procedures, as well as qualified and trained inspection personnel. A failure of NDI to detect critical defects in safety-related components of nuclear power plants, for instance, may lead to catastrophic consequences for workers, public and environment. Therefore, ensuring that NDI methods are reliable and capable of detecting all critical defects is of utmost importance. Despite increased use of automation in NDI, human inspectors, and thus human factors, still play an important role in NDI reliability. Human reliability is the probability of humans conducting specific tasks with satisfactory performance. Many techniques are suitable for modeling and analyzing human reliability in NDI of nuclear power plant components. Among these can be highlighted Failure Modes and Effects Analysis (FMEA) and THERP (Technique for Human Error Rate Prediction). The application of these techniques is illustrated in an example of qualitative and quantitative studies to improve typical NDI of pipe segments of a core cooling system of a nuclear power plant, through acting on human factors issues. (author)

  18. A human reliability analysis of the University of New Mexico's AGN- 201M nuclear research reactor

    International Nuclear Information System (INIS)

    Brumburgh, G.P.; Heger, A.S.

    1992-01-01

    During 1990--1991, a probabilistic risk assessment was conducted on the University of New Mexico's AGN-201M nuclear research reactor to address the risk and consequence of a maximum hypothetical release accident. The assessment indicated a potential for consequential human error to precipitate Chis scenario. Subsequently, a human reliability analysis was performed to evaluate the significance of human interaction on the reactor's safety systems. This paper presents the results of that investigation

  19. reliability reliability

    African Journals Online (AJOL)

    eobe

    Corresponding author, Tel: +234-703. RELIABILITY .... V , , given by the code of practice. However, checks must .... an optimization procedure over the failure domain F corresponding .... of Concrete Members based on Utility Theory,. Technical ...

  20. The human factor in operation and maintenance of complex high-reliability systems

    International Nuclear Information System (INIS)

    Ryan, T.G.

    1989-01-01

    Human factors issues in probabilistic risk assessment (PRAs) of complex high-reliability systems are addressed. These PRAs influence system operation and technical support programs such as maintainability, test, and surveillance. Using the U.S. commercial nuclear power industry as the setting, the paper addresses the manner in which PRAs currently treat human performance, the state of quantification methods and source data for analyzing human performance, and the role of human factors specialist in the analysis. The paper concludes with a presentation of TALENT, an emerging concept for fully integrating broad-based human factors expertise into the PRA process, is presented. 47 refs

  1. 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

  2. An Evidential Reasoning-Based CREAM to Human Reliability Analysis in Maritime Accident Process.

    Science.gov (United States)

    Wu, Bing; Yan, Xinping; Wang, Yang; Soares, C Guedes

    2017-10-01

    This article proposes a modified cognitive reliability and error analysis method (CREAM) for estimating the human error probability in the maritime accident process on the basis of an evidential reasoning approach. This modified CREAM is developed to precisely quantify the linguistic variables of the common performance conditions and to overcome the problem of ignoring the uncertainty caused by incomplete information in the existing CREAM models. Moreover, this article views maritime accident development from the sequential perspective, where a scenario- and barrier-based framework is proposed to describe the maritime accident process. This evidential reasoning-based CREAM approach together with the proposed accident development framework are applied to human reliability analysis of a ship capsizing accident. It will facilitate subjective human reliability analysis in different engineering systems where uncertainty exists in practice. © 2017 Society for Risk Analysis.

  3. The application of cognitive models to the evaluation and prediction of human reliability

    International Nuclear Information System (INIS)

    Embrey, D.E.; Reason, J.T.

    1986-01-01

    The first section of the paper provides a brief overview of a number of important principles relevant to human reliability modeling that have emerged from cognitive models, and presents a synthesis of these approaches in the form of a Generic Error Modeling System (GEMS). The next section illustrates the application of GEMS to some well known nuclear power plant (NPP) incidents in which human error was a major contributor. The way in which design recommendations can emerge from analyses of this type is illustrated. The third section describes the use of cognitive models in the classification of human errors for prediction and data collection purposes. The final section addresses the predictive modeling of human error as part of human reliability assessment in Probabilistic Risk Assessment

  4. The INEL Human Reliability Program: The first two years of experience

    International Nuclear Information System (INIS)

    Minner, D.E.

    1986-01-01

    This paper provides a review of the design, implementation, and operation of the INEL Human Reliability Program from January 1984 through June of 1986. Human Reliability Programs are defined in terms of the ''insider threat'' to security of nuclear facilities. The design of HRP's are discussed with special attention given the special challenge of the disgruntled employee. Each component of an HRP is reviewed noting pitfalls and opportunities with each: drug testing of applicants and incumbents, psychological evaluation by management, security clearance procedures and administration including the use of an Employee Review Board to recommend action prior to final management decision

  5. Cognitive human reliability analysis for an assessment of the safety significance of complex transients

    International Nuclear Information System (INIS)

    Amico, P.J.; Hsu, C.J.; Youngblood, R.W.; Fitzpatrick, R.G.

    1989-01-01

    This paper reports that as part of a probabilistic assessment of the safety significance of complex transients at certain PWR power plants, it was necessary to perform a cognitive human reliability analysis. To increase the confidence in the results, it was desirable to make use of actual observations of operator response which were available for the assessment. An approach was developed which incorporated these observations into the human cognitive reliability (HCR) modeling approach. The results obtained provided additional insights over what would have been found using other approaches. These insights were supported by the observations, and it is suggested that this approach be considered for use in future probabilistic safety assessments

  6. Latest scientific and technological knowledge of human-reliability quantification - December 1991

    International Nuclear Information System (INIS)

    Berg, H.P.; Schott, H.

    1992-02-01

    Again an again real incidents and accidents show that human factors may seriously affect the safety of plants. This is also true for, e.g. nuclear facilities. The major methods which are used to quantify the reliability of humans are described. These methods are applied in the framework of German and international risk analyses. Since in probabilistic safety analyses data bases are of great importance of the, however, naturally very difficult quantitative evaluation of human errors, the study also discusses the present limits to the treatment of human misbehavior in safety analyses. (orig.) [de

  7. Neuropsychological Aspects Observed in a Nuclear Plant Simulator and its Relation with Human Reliability Analysis

    International Nuclear Information System (INIS)

    Prado, E.A.P. do; Martins, M.; Pinheiro, A.; Silveira, J.

    2016-01-01

    This paper will discuss preliminary results of an evaluation methodology for the analysis and quantification of errors in manual (human) operation by training cognitive parameters and skill levels in the complex control system operation using Neuropsychophysiology and Neuro feedback equipment. The research was conducted using a game (nuclear power plant simulator) that simulates concepts of operation of a nuclear plant with a split sample evaluating aspects of learning and knowledge in the nuclear area. Operators were monitored using biomarkers (ECG, EEG, GSR, face detection and eye tracking) and the results were analyzed by Statistical multivariate techniques. An important component in the evaluation of complex systems is the human reliability during operation. Human reliability refers to the probability of the human element perform the tasks scheduled during the defined period for system operation when tested under specified environmental conditions, and additionally not to take any action detrimental to system operation.

  8. Techniques and applications of the human reliability analysis in nuclear facilities

    International Nuclear Information System (INIS)

    Pinto, Fausto C.

    1995-01-01

    The analysis and prediction of the man-machine interaction are the objectives of human reliability analysis. In this work is presented in a manner that could be used by experts in the field of Probabilistic Safety Assessment, considering primarily the aspects of human errors. The Technique of Human Error Rate Prediction (THERP) is used in large scale to obtain data on human error. Applications of this technique are presented, as well as aspects of the state-of-art and of research and development of this particular field of work, where the construction of a reliable data bank is considered essential. In this work is also developed an application of the THERP for the TRIGA Mark 1 IPR R-1 Reactor of the Centro de Desenvolvimento de Tecnologia Nuclear, Brazilian research institute of nuclear technology. The results indicate that some changes must be made in the emergency procedures of the reactor, in order to achieve a higher level of safety

  9. Lessons learned from the PISC III study of the influence on human factors on inspection reliability

    International Nuclear Information System (INIS)

    Murgatroyd, R.A.; Worrall, G.M.; Crutzen, S.

    1995-01-01

    Results from the PISC II Programme suggested that differences existed between manual inspectors in terms of their skills, knowledge and working practices which could exert a significant influence on the reliability on an inspection. Therefore, a programme of work on human reliability studies was initiated in the PISC III Programme as Action 7, with the objectives of studying and identifying causes of variability in inspection activities, and identifying some of the factors influencing the reliability of inspection in industrial conditions. It was foreseen that the information from Action 7 would aid in the development of methods for reducing the incidence of human error in inspection activities. This paper gives a brief summary of the programme and describes the lessons learned as a result of the work. A considerably more detailed description of the work is available as a PISC report. 3 refs, 3 figs

  10. EVALUATION OF HUMAN RELIABILITY IN SELECTED ACTIVITIES IN THE RAILWAY INDUSTRY

    Directory of Open Access Journals (Sweden)

    Erika SUJOVÁ

    2016-07-01

    Full Text Available The article focuses on evaluation of human reliability in the human – machine system in the railway industry. Based on a survey of a train dispatcher and of selected activities, we have identified risk factors affecting the dispatcher‘s work and the evaluated risk level of their influence on the reliability and safety of preformed activities. The research took place at the authors‘ work place between 2012-2013. A survey method was used. With its help, authors were able to identify selected work activities of train dispatcher’s risk factors that affect his/her work and the evaluated seriousness of its in-fluence on the reliability and safety of performed activities. Amongst the most important finding fall expressions of un-clear and complicated internal regulations and work processes, a feeling of being overworked, fear for one’s safety at small, insufficiently protected stations.

  11. Systamatic approach to integration of a human reliability analysis into a NPP probabalistic risk assessment

    International Nuclear Information System (INIS)

    Fragola, J.R.

    1984-01-01

    This chapter describes the human reliability analysis tasks which were employed in the evaluation of the overall probability of an internal flood sequence and its consequences in terms of disabling vulnerable risk significant equipment. Topics considered include the problem familiarization process, the identification and classification of key human interactions, a human interaction review of potential initiators, a maintenance and operations review, human interaction identification, quantification model selection, the definition of operator-induced sequences, the quantification of specific human interactions, skill- and rule-based interactions, knowledge-based interactions, and the incorporation of human interaction-related events into the event tree structure. It is concluded that an integrated approach to the analysis of human interaction within the context of a Probabilistic Risk Assessment (PRA) is feasible

  12. Guidelines for system modeling: pre-accident human errors, rev.0

    International Nuclear Information System (INIS)

    Kang, Dae Il; Jung, W. D.; Lee, Y. H.; Hwang, M. J.; Yang, J. E.

    2004-01-01

    The evaluation results of Human Reliability Analysis (HRA) of pre-accident human errors in the probabilistic safety assessment (PSA) for the Korea Standard Nuclear Power Plant (KSNP) using the ASME PRA standard show that more than 50% of 10 items to be improved are related to the identification and screening analysis for them. Thus, we developed a guideline for modeling pre-accident human errors for the system analyst to resolve some items to be improved for them. The developed guideline consists of modeling criteria for the pre-accident human errors (identification, qualitative screening, and common restoration errors) and detailed guidelines for pre-accident human errors relating to testing, maintenance, and calibration works of nuclear power plants (NPPs). The system analyst use the developed guideline and he or she applies it to the system which he or she takes care of. The HRA analyst review the application results of the system analyst. We applied the developed guideline to the auxiliary feed water system of the KSNP to show the usefulness of it. The application results of the developed guideline show that more than 50% of the items to be improved for pre-accident human errors of auxiliary feed water system are resolved. The guideline for modeling pre-accident human errors developed in this study can be used for other NPPs as well as the KSNP. It is expected that both use of the detailed procedure, to be developed in the future, for the quantification of pre-accident human errors and the guideline developed in this study will greatly enhance the PSA quality in the HRA of pre-accident human errors

  13. Guidelines for system modeling: pre-accident human errors, rev.0

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Dae Il; Jung, W. D.; Lee, Y. H.; Hwang, M. J.; Yang, J. E

    2004-01-01

    The evaluation results of Human Reliability Analysis (HRA) of pre-accident human errors in the probabilistic safety assessment (PSA) for the Korea Standard Nuclear Power Plant (KSNP) using the ASME PRA standard show that more than 50% of 10 items to be improved are related to the identification and screening analysis for them. Thus, we developed a guideline for modeling pre-accident human errors for the system analyst to resolve some items to be improved for them. The developed guideline consists of modeling criteria for the pre-accident human errors (identification, qualitative screening, and common restoration errors) and detailed guidelines for pre-accident human errors relating to testing, maintenance, and calibration works of nuclear power plants (NPPs). The system analyst use the developed guideline and he or she applies it to the system which he or she takes care of. The HRA analyst review the application results of the system analyst. We applied the developed guideline to the auxiliary feed water system of the KSNP to show the usefulness of it. The application results of the developed guideline show that more than 50% of the items to be improved for pre-accident human errors of auxiliary feed water system are resolved. The guideline for modeling pre-accident human errors developed in this study can be used for other NPPs as well as the KSNP. It is expected that both use of the detailed procedure, to be developed in the future, for the quantification of pre-accident human errors and the guideline developed in this study will greatly enhance the PSA quality in the HRA of pre-accident human errors.

  14. Human reliability analysis for probabilistic safety assessments - review of methods and issues

    International Nuclear Information System (INIS)

    Srinivas, G.; Guptan, Rajee; Malhotra, P.K.; Ghadge, S.G.; Chandra, Umesh

    2011-01-01

    It is well known that the two major events in World Nuclear Power Plant Operating history, namely the Three Mile Island and Chernobyl, were Human failure events. Subsequent to these two events, several significant changes have been incorporated in Plant Design, Control Room Design and Operator Training to reduce the possibility of Human errors during plant transients. Still, human error contribution to Risk in Nuclear Power Plant operations has been a topic of continued attention for research, development and analysis. Probabilistic Safety Assessments attempt to capture all potential human errors with a scientifically computed failure probability, through Human Reliability Analysis. Several methods are followed by different countries to quantify the Human error probability. This paper reviews the various popular methods being followed, critically examines them with reference to their criticisms and brings out issues for future research. (author)

  15. Analysis Testing of Sociocultural Factors Influence on Human Reliability within Sociotechnical Systems: The Algerian Oil Companies.

    Science.gov (United States)

    Laidoune, Abdelbaki; Rahal Gharbi, Med El Hadi

    2016-09-01

    The influence of sociocultural factors on human reliability within an open sociotechnical systems is highlighted. The design of such systems is enhanced by experience feedback. The study was focused on a survey related to the observation of working cases, and by processing of incident/accident statistics and semistructured interviews in the qualitative part. In order to consolidate the study approach, we considered a schedule for the purpose of standard statistical measurements. We tried to be unbiased by supporting an exhaustive list of all worker categories including age, sex, educational level, prescribed task, accountability level, etc. The survey was reinforced by a schedule distributed to 300 workers belonging to two oil companies. This schedule comprises 30 items related to six main factors that influence human reliability. Qualitative observations and schedule data processing had shown that the sociocultural factors can negatively and positively influence operator behaviors. The explored sociocultural factors influence the human reliability both in qualitative and quantitative manners. The proposed model shows how reliability can be enhanced by some measures such as experience feedback based on, for example, safety improvements, training, and information. With that is added the continuous systems improvements to improve sociocultural reality and to reduce negative behaviors.

  16. Safety, reliability, risk management and human factors: an integrated engineering approach applied to nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    Vasconcelos, Vanderley de; Silva, Eliane Magalhaes Pereira da; Costa, Antonio Carlos Lopes da; Reis, Sergio Carneiro dos [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)], e-mail: vasconv@cdtn.br, e-mail: silvaem@cdtn.br, e-mail: aclc@cdtn.br, e-mail: reissc@cdtn.br

    2009-07-01

    Nuclear energy has an important engineering legacy to share with the conventional industry. Much of the development of the tools related to safety, reliability, risk management, and human factors are associated with nuclear plant processes, mainly because the public concern about nuclear power generation. Despite the close association between these subjects, there are some important different approaches. The reliability engineering approach uses several techniques to minimize the component failures that cause the failure of the complex systems. These techniques include, for instance, redundancy, diversity, standby sparing, safety factors, and reliability centered maintenance. On the other hand system safety is primarily concerned with hazard management, that is, the identification, evaluation and control of hazards. Rather than just look at failure rates or engineering strengths, system safety would examine the interactions among system components. The events that cause accidents may be complex combinations of component failures, faulty maintenance, design errors, human actions, or actuation of instrumentation and control. Then, system safety deals with a broader spectrum of risk management, including: ergonomics, legal requirements, quality control, public acceptance, political considerations, and many other non-technical influences. Taking care of these subjects individually can compromise the completeness of the analysis and the measures associated with both risk reduction, and safety and reliability increasing. Analyzing together the engineering systems and controls of a nuclear facility, their management systems and operational procedures, and the human factors engineering, many benefits can be realized. This paper proposes an integration of these issues based on the application of systems theory. (author)

  17. Safety, reliability, risk management and human factors: an integrated engineering approach applied to nuclear facilities

    International Nuclear Information System (INIS)

    Vasconcelos, Vanderley de; Silva, Eliane Magalhaes Pereira da; Costa, Antonio Carlos Lopes da; Reis, Sergio Carneiro dos

    2009-01-01

    Nuclear energy has an important engineering legacy to share with the conventional industry. Much of the development of the tools related to safety, reliability, risk management, and human factors are associated with nuclear plant processes, mainly because the public concern about nuclear power generation. Despite the close association between these subjects, there are some important different approaches. The reliability engineering approach uses several techniques to minimize the component failures that cause the failure of the complex systems. These techniques include, for instance, redundancy, diversity, standby sparing, safety factors, and reliability centered maintenance. On the other hand system safety is primarily concerned with hazard management, that is, the identification, evaluation and control of hazards. Rather than just look at failure rates or engineering strengths, system safety would examine the interactions among system components. The events that cause accidents may be complex combinations of component failures, faulty maintenance, design errors, human actions, or actuation of instrumentation and control. Then, system safety deals with a broader spectrum of risk management, including: ergonomics, legal requirements, quality control, public acceptance, political considerations, and many other non-technical influences. Taking care of these subjects individually can compromise the completeness of the analysis and the measures associated with both risk reduction, and safety and reliability increasing. Analyzing together the engineering systems and controls of a nuclear facility, their management systems and operational procedures, and the human factors engineering, many benefits can be realized. This paper proposes an integration of these issues based on the application of systems theory. (author)

  18. 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.

  19. 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

  20. Review of the human reliability analysis performed for Empire State Electric Energy Research Corporation

    International Nuclear Information System (INIS)

    Swart, D.; Banz, I.

    1985-01-01

    The Empire State Electric Energy Research Corporation (ESEERCO) commissioned Westinghouse to conduct a human reliability analysis to identify and quantify human error probabilities associated with operator actions for four specific events which may occur in light water reactors: loss of coolant accident, steam generator tube rupture, steam/feed line break, and stuck open pressurizer spray valve. Human Error Probabilities (HEPs) derived from Swain's Technique for Human Error Rate Prediction (THERP) were compared to data obtained from simulator exercises. A correlation was found between the HEPs derived from Swain and the results of the simulator data. The results of this study provide a unique insight into human factors analysis. The HEPs obtained from such probabilistic studies can be used to prioritize scenarios for operator training situations, and thus improve the correlation between simulator exercises and real control room experiences

  1. Balancing human and technical reliability in the design of advanced nuclear reactors

    International Nuclear Information System (INIS)

    Papin, Bernard

    2011-01-01

    Highlights: ► Human factors exigencies are often overseen during the early design phases of NPP. ► Optimization of reactors safety is only based on technical reliability considerations. ► The search for more technical reliability often leads to more system complexity. ► System complexity is a major contributor to the operator's poor performance. ► Our method enables to assess plant complexity and it's impact on human performance. - Abstract: The strong influence of human factors (HF) on the safety of nuclear facilities is nowadays recognised and the designers are now enforced to consider HF requirements in the design of new facilities. Yet, this consideration of human factors requirements is still more or less restricted to the latest phases of the projects, essentially for the design of human-system interfaces (HSI's) and control rooms, although the design options influencing at most the human performance in operation are indeed fixed during the very early phases of the new reactors projects. The main reason of this late consideration of HF is that there exist few methods and models for anticipating the influence of fundamental design options on the future performance of operation teams. This paper describes a set of new tools permitting (i) determination of the impact of the fundamental process design options on the future activity of the operation teams and (ii) assessment of the influence of these operational constraints on teams performance. These tools are intended to guide the design of future 4th generation (GEN4) reactors, within the frame of a global risk-informed design approach, considering technical and human reliability exigencies in a balanced way.

  2. Improving human reliability through better nuclear power plant system design. Progress report

    International Nuclear Information System (INIS)

    Golay, M.W.

    1995-01-01

    The project on open-quotes Development of a Theory of the Dependence of Human Reliability upon System Designs as a Means of Improving Nuclear Power Plant Performanceclose quotes has been undertaken in order to address the important problem of human error in advanced nuclear power plant designs. Most of the creativity in formulating such concepts has focused upon improving the mechanical reliability of safety related plant systems. However, the lack of a mature theory has retarded similar progress in reducing the likely frequencies of human errors. The main design mechanism used to address this class of concerns has been to reduce or eliminate the human role in plant operations and accident response. The plan of work being pursued in this project is to perform a set of experiments involving human subject who are required to operate, diagnose and respond to changes in computer-simulated systems, relevant to those encountered in nuclear power plants. In the tests the systems are made to differ in complexity in a systematic manner. The computer program used to present the problems to be solved also records the response of the operator as it unfolds. Ultimately this computer is also to be used in compiling the results of the project. The work of this project is focused upon nuclear power plant applications. However, the persuasiveness of human errors in using all sorts of electromechanical machines gives it a much greater potential importance. Because of this we are attempting to pursue our work in a fashion permitting broad generalizations

  3. Optimal design methods for a digital human-computer interface based on human reliability in a nuclear power plant

    International Nuclear Information System (INIS)

    Jiang, Jianjun; Zhang, Li; Xie, Tian; Wu, Daqing; Li, Min; Wang, Yiqun; Peng, Yuyuan; Peng, Jie; Zhang, Mengjia; Li, Peiyao; Ma, Congmin; Wu, Xing

    2017-01-01

    Highlights: • A complete optimization process is established for digital human-computer interfaces of Npps. • A quick convergence search method is proposed. • The authors propose an affinity error probability mapping function to test human reliability. - Abstract: This is the second in a series of papers describing the optimal design method for a digital human-computer interface of nuclear power plant (Npp) from three different points based on human reliability. The purpose of this series is to explore different optimization methods from varying perspectives. This present paper mainly discusses the optimal design method for quantity of components of the same factor. In monitoring process, quantity of components has brought heavy burden to operators, thus, human errors are easily triggered. To solve the problem, the authors propose an optimization process, a quick convergence search method and an affinity error probability mapping function. Two balanceable parameter values of the affinity error probability function are obtained by experiments. The experimental results show that the affinity error probability mapping function about human-computer interface has very good sensitivity and stability, and that quick convergence search method for fuzzy segments divided by component quantity has better performance than general algorithm.

  4. Evidential analytic hierarchy process dependence assessment methodology in human reliability analysis

    International Nuclear Information System (INIS)

    Chen, Lu Yuan; Zhou, Xinyi; Xiao, Fuyuan; Deng, Yong; Mahadevan, Sankaran

    2017-01-01

    In human reliability analysis, dependence assessment is an important issue in risky large complex systems, such as operation of a nuclear power plant. Many existing methods depend on an expert's judgment, which contributes to the subjectivity and restrictions of results. Recently, a computational method, based on the Dempster-Shafer evidence theory and analytic hierarchy process, has been proposed to handle the dependence in human reliability analysis. The model can deal with uncertainty in an analyst's judgment and reduce the subjectivity in the evaluation process. However, the computation is heavy and complicated to some degree. The most important issue is that the existing method is in a positive aspect, which may cause an underestimation of the risk. In this study, a new evidential analytic hierarchy process dependence assessment methodology, based on the improvement of existing methods, has been proposed, which is expected to be easier and more effective

  5. Evidential Analytic Hierarchy Process Dependence Assessment Methodology in Human Reliability Analysis

    Directory of Open Access Journals (Sweden)

    Luyuan Chen

    2017-02-01

    Full Text Available In human reliability analysis, dependence assessment is an important issue in risky large complex systems, such as operation of a nuclear power plant. Many existing methods depend on an expert's judgment, which contributes to the subjectivity and restrictions of results. Recently, a computational method, based on the Dempster–Shafer evidence theory and analytic hierarchy process, has been proposed to handle the dependence in human reliability analysis. The model can deal with uncertainty in an analyst's judgment and reduce the subjectivity in the evaluation process. However, the computation is heavy and complicated to some degree. The most important issue is that the existing method is in a positive aspect, which may cause an underestimation of the risk. In this study, a new evidential analytic hierarchy process dependence assessment methodology, based on the improvement of existing methods, has been proposed, which is expected to be easier and more effective.

  6. Evidential analytic hierarchy process dependence assessment methodology in human reliability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Lu Yuan; Zhou, Xinyi; Xiao, Fuyuan; Deng, Yong [School of Computer and Information Science, Southwest University, Chongqing (China); Mahadevan, Sankaran [School of Engineering, Vanderbilt University, Nashville (United States)

    2017-02-15

    In human reliability analysis, dependence assessment is an important issue in risky large complex systems, such as operation of a nuclear power plant. Many existing methods depend on an expert's judgment, which contributes to the subjectivity and restrictions of results. Recently, a computational method, based on the Dempster-Shafer evidence theory and analytic hierarchy process, has been proposed to handle the dependence in human reliability analysis. The model can deal with uncertainty in an analyst's judgment and reduce the subjectivity in the evaluation process. However, the computation is heavy and complicated to some degree. The most important issue is that the existing method is in a positive aspect, which may cause an underestimation of the risk. In this study, a new evidential analytic hierarchy process dependence assessment methodology, based on the improvement of existing methods, has been proposed, which is expected to be easier and more effective.

  7. The NUCLARR databank: Human reliability and hardware failure data for the nuclear power industry

    International Nuclear Information System (INIS)

    Reece, W.J.

    1993-01-01

    Under the sponsorship of the US Nuclear Regulatory Commission (NRC), the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) was developed to provide human reliability and hardware failure data to analysts in the nuclear power industry. This IBM-compatible databank is contained on a set of floppy diskettes which include data files and a menu-driven system for locating, reviewing, sorting, and retrieving the data. NUCLARR contains over 2500 individual data records, drawn from more, than 60 sources. The system is upgraded annually, to include additional human error and hardware component failure data and programming enhancements (i.e., increased user-friendliness). NUCLARR is available from the NRC through project staff at the INEL

  8. A human error probability estimate methodology based on fuzzy inference and expert judgment on nuclear plants

    International Nuclear Information System (INIS)

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

    2009-01-01

    Recent studies point human error as an important factor for many industrial and nuclear accidents: Three Mile Island (1979), Bhopal (1984), Chernobyl and Challenger (1986) are classical examples. Human contribution to these accidents may be better understood and analyzed by using Human Reliability Analysis (HRA), which has being taken as an essential part on Probabilistic Safety Analysis (PSA) of nuclear plants. Both HRA and PSA depend on Human Error Probability (HEP) for a quantitative analysis. These probabilities are extremely affected by the Performance Shaping Factors (PSF), which has a direct effect on human behavior and thus shape HEP according with specific environment conditions and personal individual characteristics which are responsible for these actions. This PSF dependence raises a great problem on data availability as turn these scarcely existent database too much generic or too much specific. Besides this, most of nuclear plants do not keep historical records of human error occurrences. Therefore, in order to overcome this occasional data shortage, a methodology based on Fuzzy Inference and expert judgment was employed in this paper in order to determine human error occurrence probabilities and to evaluate PSF's on performed actions by operators in a nuclear power plant (IEA-R1 nuclear reactor). Obtained HEP values were compared with reference tabled data used on current literature in order to show method coherence and valid approach. This comparison leads to a conclusion that this work results are able to be employed both on HRA and PSA enabling efficient prospection of plant safety conditions, operational procedures and local working conditions potential improvements (author)

  9. Use of eye tracking equipment for human reliability analysis applied to complex system operations

    International Nuclear Information System (INIS)

    Pinheiro, Andre Ricardo Mendonça; Prado, Eugenio Anselmo Pessoa do; Martins, Marcelo Ramos

    2017-01-01

    This article will discuss the preliminary results of an evaluation methodology for the analysis and quantification of manual character errors (human), by monitoring cognitive parameters and skill levels in the operation of a complex control system based on parameters provided by a eye monitoring equipment (Eye Tracker). The research was conducted using a simulator (game) that plays concepts of operation of a nuclear reactor with a split sample for evaluation of aspects of learning, knowledge and standard operating within the context addressed. bridge operators were monitored using the EYE TRACKING, eliminating the presence of the analyst in the evaluation of the operation, allowing the analysis of the results by means of multivariate statistical techniques within the scope of system reliability. The experiments aim to observe state change situations such as stops and scheduled departures, incidents assumptions and common operating characteristics. Preliminary results of this research object indicate that technical and cognitive aspects can contribute to improving the reliability of the available techniques in human reliability, making them more realistic both in the context of quantitative approaches to regulatory and training purposes, as well as reduced incidence of human error. (author)

  10. Use of eye tracking equipment for human reliability analysis applied to complex system operations

    Energy Technology Data Exchange (ETDEWEB)

    Pinheiro, Andre Ricardo Mendonça; Prado, Eugenio Anselmo Pessoa do; Martins, Marcelo Ramos, E-mail: andrericardopinheiro@usp.br, E-mail: eugenio.prado@labrisco.usp.br, E-mail: mrmatins@usp.br [Universidade de Sao Paulo (LABRISCO/USP), Sao Paulo, SP (Brazil). Lab. de Análise, Avaliação e Gerenciamento de Risco

    2017-07-01

    This article will discuss the preliminary results of an evaluation methodology for the analysis and quantification of manual character errors (human), by monitoring cognitive parameters and skill levels in the operation of a complex control system based on parameters provided by a eye monitoring equipment (Eye Tracker). The research was conducted using a simulator (game) that plays concepts of operation of a nuclear reactor with a split sample for evaluation of aspects of learning, knowledge and standard operating within the context addressed. bridge operators were monitored using the EYE TRACKING, eliminating the presence of the analyst in the evaluation of the operation, allowing the analysis of the results by means of multivariate statistical techniques within the scope of system reliability. The experiments aim to observe state change situations such as stops and scheduled departures, incidents assumptions and common operating characteristics. Preliminary results of this research object indicate that technical and cognitive aspects can contribute to improving the reliability of the available techniques in human reliability, making them more realistic both in the context of quantitative approaches to regulatory and training purposes, as well as reduced incidence of human error. (author)

  11. Improving human reliability through better nuclear power plant system design: Program for advanced nuclear power studies

    International Nuclear Information System (INIS)

    Golay, M.W.

    1993-01-01

    The project on ''Development of a Theory of the Dependence of Human Reliability upon System Designs as a Means of Improving Nuclear Power Plant Performance'' was been undertaken in order to address the problem of human error in advanced nuclear power plant designs. Lack of a mature theory has retarded progress in reducing likely frequencies of human errors. Work being pursued in this project is to perform a set of experiments involving human subjects who are required to operate, diagnose and respond to changes in computer-simulated systems, relevant to those encountered in nuclear power plants, which are made to differ in complexity in a systematic manner. The computer program used to present the problems to be solved also records the response of the operator as it unfolds

  12. Human Reliability in Probabilistic Safety Assessments; Fiabilidad Humana en los Analisis Probabilisticos de Seguridad

    Energy Technology Data Exchange (ETDEWEB)

    Nunez Mendez, J

    1989-07-01

    Nowadays a growing interest in environmental aspects is detected in our country. It implies an assessment of the risk involved in the industrial processes and installations in order to determine if those are into the acceptable limits. In these safety assessments, among which PSA (Probabilistic Safety Assessments), can be pointed out the role played by the human being in the system is one of the more relevant subjects (This relevance has been demonstrated in the accidents happened) . However, in Spain there aren't manuals specifically dedicated to asses the human contribution to risk in the frame of PSAs. This report aims to improve this situation providing: a) a theoretical background to help the reader in the understanding of the nature of the human error, b) a quid to carry out a Human Reliability Analysis and c) a selected overview of the techniques and methodologies currently applied in this area. (Author) 20 refs.

  13. Human Reliability in Probabilistic Safety Assessments; Fiabilidad Humana en los Analisis Probabilisticos de Seguridad

    Energy Technology Data Exchange (ETDEWEB)

    Nunez Mendez, J.

    1989-07-01

    Nowadays a growing interest in environmental aspects is detected in our country. It implies an assessment of the risk involved in the industrial processes and installations in order to determine if those are into the acceptable limits. In these safety assessments, among which PSA (Probabilistic Safety Assessments), can be pointed out the role played by the human being in the system is one of the more relevant subjects (This relevance has been demonstrated in the accidents happened) . However, in Spain there aren't manuals specifically dedicated to asses the human contribution to risk in the frame of PSAs. This report aims to improve this situation providing: a) a theoretical background to help the reader in the understanding of the nature of the human error, b) a quid to carry out a Human Reliability Analysis and c) a selected overview of the techniques and methodologies currently applied in this area. (Author) 20 refs.

  14. Human reliability and risk management in the transportation of spent nuclear fuel

    International Nuclear Information System (INIS)

    Tuler, S.; Kasperson, R.E.; Ratick, S.

    1989-01-01

    This paper summarizes work on human factor contributions to risks from spent nuclear fuel transportation. Human participation may have significant effects on the levels and types of risks by enabling or initiating incidents and exacerbating adverse consequences. Human errors are defined to be the result of mismatches between perceived system state and actual system state. In complex transportation systems such mismatches may be distributed in time (e.g., during different stages of design, implementation, operation, maintenance) and location (e.g., human error, its identification, and its recovery may be geographically and institutionally separate). Risk management programs may decrease the probability of undesirable events or attenuate the consequences of mismatches. This paper presents a methodology to identify the scope and types of human-task mismatches and to identify potential management options for their prevention, mitigation, or recovery. A review of transportation accident databases, in conjunction with human error models, is used to develop a taxonomy of human errors during design for the pre-identification of potential mismatches or after incidents have occurred to evaluate their causes. Risk management options to improve human reliability are identified by a matrix that relates the multiple stages of a spent nuclear fuel transportation system to management options (e.g., training, data analysis, regulation). The paper concludes with examples to illustrate how the methodology may be applied. (author)

  15. 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

  16. Quantitative evaluation of the impact of human reliability in risk assessment for nuclear power plants

    International Nuclear Information System (INIS)

    Samanta, P.K.

    1981-01-01

    The role of human beings in the safe operation of a nuclear power plant has been a matter of concern. This study describes methods for the quantitative description of that role and its impact on the risk from nuclear power plants. The impact of human errors was calculated by observing the changes in risk parameters, such as core melt probability, release category probabilities, accident sequence probabilities and system unavailabilities due to changes in the contribution to unavailablity of human errors, within the framework of risk assessment methodology. It was found that for operational pressurized water reactors the opportunity for reduction in core melt probability by reducing the human error rates without simultaneous reduction of hardware failures is limited, but that core melt probability would significantly increase as human error rates increased. More importantly, most of the dominant accident sequences showed a significant increase in their probabilities with an increase in human error rates. Release categories resulting in high consequences showed a much larger sensitivity to human errors than categories resulting in low consequences. A combination of structural importance and reliability importance measure was used to describe the importance of individual errors

  17. Human and organisational factors in the reliability of non-destructive testing (NOT)

    International Nuclear Information System (INIS)

    Norros, L.

    1998-01-01

    Non-destructive testing used in in-service inspections can be seen as a complicated activity system including three mutually related sub-activities: (1) definition of inspection programs and necessary resources, (2) carrying out diagnostic inspections, and (3) interpretation of the results from the view of plant safety and corrective measures. Various studies to investigate and measure the NDT performance have produced disappointing result. No clear correlations between single human factors and performance have been identified even though general agreement exists concerning the significance of human factors to the reliability of testing. Another incentive for our studies has been to test and evaluate the applicability of the international results in the Finnish circumstances. Three successive studies have thus been carried out on the human and organisational factors in non-destructive testing. (author)

  18. Post-event human decision errors: operator action tree/time reliability correlation

    Energy Technology Data Exchange (ETDEWEB)

    Hall, R E; Fragola, J; Wreathall, J

    1982-11-01

    This report documents an interim framework for the quantification of the probability of errors of decision on the part of nuclear power plant operators after the initiation of an accident. The framework can easily be incorporated into an event tree/fault tree analysis. The method presented consists of a structure called the operator action tree and a time reliability correlation which assumes the time available for making a decision to be the dominating factor in situations requiring cognitive human response. This limited approach decreases the magnitude and complexity of the decision modeling task. Specifically, in the past, some human performance models have attempted prediction by trying to emulate sequences of human actions, or by identifying and modeling the information processing approach applicable to the task. The model developed here is directed at describing the statistical performance of a representative group of hypothetical individuals responding to generalized situations.

  19. Post-event human decision errors: operator action tree/time reliability correlation

    International Nuclear Information System (INIS)

    Hall, R.E.; Fragola, J.; Wreathall, J.

    1982-11-01

    This report documents an interim framework for the quantification of the probability of errors of decision on the part of nuclear power plant operators after the initiation of an accident. The framework can easily be incorporated into an event tree/fault tree analysis. The method presented consists of a structure called the operator action tree and a time reliability correlation which assumes the time available for making a decision to be the dominating factor in situations requiring cognitive human response. This limited approach decreases the magnitude and complexity of the decision modeling task. Specifically, in the past, some human performance models have attempted prediction by trying to emulate sequences of human actions, or by identifying and modeling the information processing approach applicable to the task. The model developed here is directed at describing the statistical performance of a representative group of hypothetical individuals responding to generalized situations

  20. Notes on human factors problems in process plant reliability and safety prediction

    International Nuclear Information System (INIS)

    Rasmussen, J.; Taylor, J.R.

    1976-09-01

    The basis for plant operator reliability evaluation is described. Principles for plant design, necessary to permit reliability evaluation, are outlined. Five approaches to the plant operator reliability problem are described. Case stories, illustrating operator reliability problems, are given. (author)

  1. A human reliability based usability evaluation method for safety-critical software

    International Nuclear Information System (INIS)

    Boring, R. L.; Tran, T. Q.; Gertman, D. I.; Ragsdale, A.

    2006-01-01

    Boring and Gertman (2005) introduced a novel method that augments heuristic usability evaluation methods with that of the human reliability analysis method of SPAR-H. By assigning probabilistic modifiers to individual heuristics, it is possible to arrive at the usability error probability (UEP). Although this UEP is not a literal probability of error, it nonetheless provides a quantitative basis to heuristic evaluation. This method allows one to seamlessly prioritize and identify usability issues (i.e., a higher UEP requires more immediate fixes). However, the original version of this method required the usability evaluator to assign priority weights to the final UEP, thus allowing the priority of a usability issue to differ among usability evaluators. The purpose of this paper is to explore an alternative approach to standardize the priority weighting of the UEP in an effort to improve the method's reliability. (authors)

  2. HUMAN ERROR QUANTIFICATION USING PERFORMANCE SHAPING FACTORS IN THE SPAR-H METHOD

    Energy Technology Data Exchange (ETDEWEB)

    Harold S. Blackman; David I. Gertman; Ronald L. Boring

    2008-09-01

    This paper describes a cognitively based human reliability analysis (HRA) quantification technique for estimating the human error probabilities (HEPs) associated with operator and crew actions at nuclear power plants. The method described here, Standardized Plant Analysis Risk-Human Reliability Analysis (SPAR-H) method, was developed to aid in characterizing and quantifying human performance at nuclear power plants. The intent was to develop a defensible method that would consider all factors that may influence performance. In the SPAR-H approach, calculation of HEP rates is especially straightforward, starting with pre-defined nominal error rates for cognitive vs. action-oriented tasks, and incorporating performance shaping factor multipliers upon those nominal error rates.

  3. In-plant application of industry experience to enhance human reliability

    International Nuclear Information System (INIS)

    Hannaman, G.W.; Singh, A.

    1993-01-01

    This paper describes the way that modern data-base computer tools can enhance the ability to collect, organize, evaluate, and use industry experience. By combining the computer tools with knowledge from human reliability assessment tools, data, and frameworks, the data base can become a tool for collecting and assessing the lessons learned from past events. By integrating the data-base system with plant risk models, engineers can focus on those activities that can enhance over-all system reliability. The evaluation helps identify technology and tools to reduce human errors during operations and maintenance. Learning from both in-plant and industry experience can help enhance safety and reduce the cost of plant operations. Utility engineers currently assess events that occur in nuclear plants throughout the world for in-plant applicability. Established computer information networks, documents, bulletins, and other information sources provide a large number of event descriptions to help individual plants benefit from this industry experience. The activities for coordinating reviews of event descriptions from other plants for in-plant applications require substantial engineering time to collect, organize, evaluate, and apply. Data-base tools can help engineers efficiently handle and sort the data so that they can concentrate on understanding the importance of the event, developing cost-effective interventions, and communicating implementation plans for plant improvement. An Electric Power Research Institute human reliability project has developed a classification system with modern data-base software to help engineers efficiently process, assess, and apply information contained in the events to enhance plant operation. Plant-specific classification of industry experience provides a practical method for efficiently taking into account industry when planning maintenance activities and reviewing plant safety

  4. On action- and affectpsychology of human reliability. An access by training simulators for complex man-machine systems

    International Nuclear Information System (INIS)

    Schuette, M.

    2002-02-01

    Theoretical part and its topics: errors at the interface between man and machine; reliability analysis for man; the psychological explanation of action reliability of man (intention and control); a paradigma for human reliability (frustration and regression). Empirical part: Control room in a nuclear power plant: Influences on repeated blockages on component care in case of start-up operation; ship bridge: Frustration and regression while steering in a bight. Appendix: analysis of a social interaction.(GL)

  5. The application of two recently developed human reliability techniques to cognitive error analysis

    International Nuclear Information System (INIS)

    Gall, W.

    1990-01-01

    Cognitive error can lead to catastrophic consequences for manned systems, including those whose design renders them immune to the effects of physical slips made by operators. Four such events, pressurized water and boiling water reactor accidents which occurred recently, were analysed. The analysis identifies the factors which contributed to the errors and suggests practical strategies for error recovery or prevention. Two types of analysis were conducted: an unstructured analysis based on the analyst's knowledge of psychological theory, and a structured analysis using two recently-developed human reliability analysis techniques. In general, the structured techniques required less effort to produce results and these were comparable to those of the unstructured analysis. (author)

  6. Basic human error probabilities in advanced MCRs when using soft control

    International Nuclear Information System (INIS)

    Jang, In Seok; Seong, Poong Hyun; Kang, Hyun Gook; Lee, Seung Jun

    2012-01-01

    In a report on one of the renowned HRA methods, Technique for Human Error Rate Prediction (THERP), it is pointed out that 'The paucity of actual data on human performance continues to be a major problem for estimating HEPs and performance times in nuclear power plant (NPP) task'. However, another critical difficulty is that most current HRA databases deal with operation in conventional type of MCRs. With the adoption of new human system interfaces that are based on computer based technologies, the operation environment of MCRs in NPPs has changed. The MCRs including these digital and computer technologies, such as large display panels, computerized procedures, soft controls, and so on, are called advanced MCRs. Because of the different interfaces, different Basic Human Error Probabilities (BHEPs) should be considered in human reliability analyses (HRAs) for advanced MCRs. This study carries out an empirical analysis of human error considering soft controls. The aim of this work is not only to compile a database using the simulator for advanced MCRs but also to compare BHEPs with those of a conventional MCR database

  7. 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)

  8. Structure design of human factor data management system for Daya Bay NPP

    International Nuclear Information System (INIS)

    Zhang Li; Zhang Ning; Guo Jianbing; Huang Weigang; Zhu Minhong; Wang Jin

    2000-01-01

    Collection, analysis and quantification of human factor data are important compositions of human reliability analysis (HRA) and probabilistic risk assessment (PRA). Various human factor databases have been set up, but there are comparatively little human factor data management systems which can be uses for collection, classification, analysis, calculation and predication of the human factor data. Therefore, the human factor data management system for Daya Bay NPP is developed, with the following three modules and four databases: original data module, computing module, introduced data module, and basic database, other data source of the plant, external database and introduced database. The foundational problems about human factor data and the systemic structure and function are described. The data structure in the database is also discussed, because it is of the most importance in the system

  9. Interobserver Reliability of the Total Body Score System for Quantifying Human Decomposition.

    Science.gov (United States)

    Dabbs, Gretchen R; Connor, Melissa; Bytheway, Joan A

    2016-03-01

    Several authors have tested the accuracy of the Total Body Score (TBS) method for quantifying decomposition, but none have examined the reliability of the method as a scoring system by testing interobserver error rates. Sixteen participants used the TBS system to score 59 observation packets including photographs and written descriptions of 13 human cadavers in different stages of decomposition (postmortem interval: 2-186 days). Data analysis used a two-way random model intraclass correlation in SPSS (v. 17.0). The TBS method showed "almost perfect" agreement between observers, with average absolute correlation coefficients of 0.990 and average consistency correlation coefficients of 0.991. While the TBS method may have sources of error, scoring reliability is not one of them. Individual component scores were examined, and the influences of education and experience levels were investigated. Overall, the trunk component scores were the least concordant. Suggestions are made to improve the reliability of the TBS method. © 2016 American Academy of Forensic Sciences.

  10. Human error recovery failure probability when using soft controls in computerized control rooms

    International Nuclear Information System (INIS)

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

    2014-01-01

    Many literatures categorized recovery process into three phases; detection of problem situation, explanation of problem causes or countermeasures against problem, and end of recovery. Although the focus of recovery promotion has been on categorizing recovery phases and modeling recovery process, research related to human recovery failure probabilities has not been perform actively. On the other hand, a few study regarding recovery failure probabilities were implemented empirically. Summarizing, researches that have performed so far have several problems in terms of use in human reliability analysis (HRA). By adopting new human-system interfaces that are based on computer-based technologies, the operation environment of MCRs in NPPs has changed from conventional MCRs to advanced MCRs. Because of the different interfaces between conventional and advanced MCRs, different recovery failure probabilities should be considered in the HRA for advanced MCRs. Therefore, this study carries out an empirical analysis of human error recovery probabilities under an advanced MCR mockup called compact nuclear simulator (CNS). The aim of this work is not only to compile a recovery failure probability database using the simulator for advanced MCRs but also to collect recovery failure probability according to defined human error modes to compare that which human error mode has highest recovery failure probability. The results show that recovery failure probability regarding wrong screen selection was lowest among human error modes, which means that most of human error related to wrong screen selection can be recovered. On the other hand, recovery failure probabilities of operation selection omission and delayed operation were 1.0. These results imply that once subject omitted one task in the procedure, they have difficulties finding and recovering their errors without supervisor's assistance. Also, wrong screen selection had an effect on delayed operation. That is, wrong screen

  11. The treatment of commission errors in first generation human reliability analysis methods

    Energy Technology Data Exchange (ETDEWEB)

    Alvarengga, Marco Antonio Bayout; Fonseca, Renato Alves da, E-mail: bayout@cnen.gov.b, E-mail: rfonseca@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN) Rio de Janeiro, RJ (Brazil); Melo, Paulo Fernando Frutuoso e, E-mail: frutuoso@nuclear.ufrj.b [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear

    2011-07-01

    Human errors in human reliability analysis can be classified generically as errors of omission and commission errors. Omission errors are related to the omission of any human action that should have been performed, but does not occur. Errors of commission are those related to human actions that should not be performed, but which in fact are performed. Both involve specific types of cognitive error mechanisms, however, errors of commission are more difficult to model because they are characterized by non-anticipated actions that are performed instead of others that are omitted (omission errors) or are entered into an operational task without being part of the normal sequence of this task. The identification of actions that are not supposed to occur depends on the operational context that will influence or become easy certain unsafe actions of the operator depending on the operational performance of its parameters and variables. The survey of operational contexts and associated unsafe actions is a characteristic of second-generation models, unlike the first generation models. This paper discusses how first generation models can treat errors of commission in the steps of detection, diagnosis, decision-making and implementation, in the human information processing, particularly with the use of THERP tables of errors quantification. (author)

  12. Raman spectroscopy of human skin: looking for a quantitative algorithm to reliably estimate human age

    Science.gov (United States)

    Pezzotti, Giuseppe; Boffelli, Marco; Miyamori, Daisuke; Uemura, Takeshi; Marunaka, Yoshinori; Zhu, Wenliang; Ikegaya, Hiroshi

    2015-06-01

    The possibility of examining soft tissues by Raman spectroscopy is challenged in an attempt to probe human age for the changes in biochemical composition of skin that accompany aging. We present a proof-of-concept report for explicating the biophysical links between vibrational characteristics and the specific compositional and chemical changes associated with aging. The actual existence of such links is then phenomenologically proved. In an attempt to foster the basics for a quantitative use of Raman spectroscopy in assessing aging from human skin samples, a precise spectral deconvolution is performed as a function of donors' ages on five cadaveric samples, which emphasizes the physical significance and the morphological modifications of the Raman bands. The outputs suggest the presence of spectral markers for age identification from skin samples. Some of them appeared as authentic "biological clocks" for the apparent exactness with which they are related to age. Our spectroscopic approach yields clear compositional information of protein folding and crystallization of lipid structures, which can lead to a precise identification of age from infants to adults. Once statistically validated, these parameters might be used to link vibrational aspects at the molecular scale for practical forensic purposes.

  13. 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

  14. A structural approach to constructing perspective efficient and reliable human-computer interfaces

    International Nuclear Information System (INIS)

    Balint, L.

    1989-01-01

    The principles of human-computer interface (HCI) realizations are investigated with the aim of getting closer to a general framework and thus, to a more or less solid background of constructing perspective efficient, reliable and cost-effective human-computer interfaces. On the basis of characterizing and classifying the different HCI solutions, the fundamental problems of interface construction are pointed out especially with respect to human error occurrence possibilities. The evolution of HCI realizations is illustrated by summarizing the main properties of past, present and foreseeable future interface generations. HCI modeling is pointed out to be a crucial problem in theoretical and practical investigations. Suggestions concerning HCI structure (hierarchy and modularity), HCI functional dynamics (mapping from input to output information), minimization of human error caused system failures (error-tolerance, error-recovery and error-correcting) as well as cost-effective HCI design and realization methodology (universal and application-oriented vs. application-specific solutions) are presented. The concept of RISC-based and SCAMP-type HCI components is introduced with the aim of having a reduced interaction scheme in communication and a well defined architecture in HCI components' internal structure. HCI efficiency and reliability are dealt with, by taking into account complexity and flexibility. The application of fast computerized prototyping is also briefly investigated as an experimental device of achieving simple, parametrized, invariant HCI models. Finally, a concise outline of an approach of how to construct ideal HCI's is also suggested by emphasizing the open questions and the need of future work related to the proposals, as well. (author). 14 refs, 6 figs

  15. NKA/KRU project on operator training, control room designing and human reliability. Summary report

    International Nuclear Information System (INIS)

    1981-06-01

    A Nordic integrated project on human reliability in the conditions of new advanced technology seeks to establish: - The actual repertoire of activities and tasks performed by the operating staff of a nuclear power plant and its dependence on the present and future levels of automation. - The knowledge required for these activities and appropriate means for training plant operators and for competence evaluation and retraining in coping with the rare events. - Models of human operator performance; how do operators read information and make decisions under normal and abnormal plant conditions and how does their performance depend upon control room design. - The typical limits of human capabilities and mechanisms of human errors as they are represented in existing records of incidents and accidents in industrial plants. - The use of process computers for improved design of data presentation and operator support systems, especially for disturbance analysis and diagnosis during infrequent plant disturbance. - Development of experimental techniques to validate research results and proposals for improved man/machine interfaces and other computer-based support systems. (EG)

  16. Postretrieval new learning does not reliably induce human memory updating via reconsolidation.

    Science.gov (United States)

    Hardwicke, Tom E; Taqi, Mahdi; Shanks, David R

    2016-05-10

    Reconsolidation theory proposes that retrieval can destabilize an existing memory trace, opening a time-dependent window during which that trace is amenable to modification. Support for the theory is largely drawn from nonhuman animal studies that use invasive pharmacological or electroconvulsive interventions to disrupt a putative postretrieval restabilization ("reconsolidation") process. In human reconsolidation studies, however, it is often claimed that postretrieval new learning can be used as a means of "updating" or "rewriting" existing memory traces. This proposal warrants close scrutiny because the ability to modify information stored in the memory system has profound theoretical, clinical, and ethical implications. The present study aimed to replicate and extend a prominent 3-day motor-sequence learning study [Walker MP, Brakefield T, Hobson JA, Stickgold R (2003) Nature 425(6958):616-620] that is widely cited as a convincing demonstration of human reconsolidation. However, in four direct replication attempts (n = 64), we did not observe the critical impairment effect that has previously been taken to indicate disruption of an existing motor memory trace. In three additional conceptual replications (n = 48), we explored the broader validity of reconsolidation-updating theory by using a declarative recall task and sequences similar to phone numbers or computer passwords. Rather than inducing vulnerability to interference, memory retrieval appeared to aid the preservation of existing sequence knowledge relative to a no-retrieval control group. These findings suggest that memory retrieval followed by new learning does not reliably induce human memory updating via reconsolidation.

  17. Selection of suitable hand gestures for reliable myoelectric human computer interface.

    Science.gov (United States)

    Castro, Maria Claudia F; Arjunan, Sridhar P; Kumar, Dinesh K

    2015-04-09

    Myoelectric controlled prosthetic hand requires machine based identification of hand gestures using surface electromyogram (sEMG) recorded from the forearm muscles. This study has observed that a sub-set of the hand gestures have to be selected for an accurate automated hand gesture recognition, and reports a method to select these gestures to maximize the sensitivity and specificity. Experiments were conducted where sEMG was recorded from the muscles of the forearm while subjects performed hand gestures and then was classified off-line. The performances of ten gestures were ranked using the proposed Positive-Negative Performance Measurement Index (PNM), generated by a series of confusion matrices. When using all the ten gestures, the sensitivity and specificity was 80.0% and 97.8%. After ranking the gestures using the PNM, six gestures were selected and these gave sensitivity and specificity greater than 95% (96.5% and 99.3%); Hand open, Hand close, Little finger flexion, Ring finger flexion, Middle finger flexion and Thumb flexion. This work has shown that reliable myoelectric based human computer interface systems require careful selection of the gestures that have to be recognized and without such selection, the reliability is poor.

  18. Are We Hoping For A Bounce A Study On Resilience And Human Relations In A High Reliability Organization

    Science.gov (United States)

    2016-03-01

    negatively impact the organization’s resilience when faced with workplace stressors (Gittell, 2008, p. 26). Consequently, this reduces an organization’s...2014). Organizational resilience and the challenge for human resource management: Conceptualizations and frameworks for theory and practice. Paper...A BOUNCE? A STUDY ON RESILIENCE AND HUMAN RELATIONS IN A HIGH RELIABILITY ORGANIZATION by Robert D. Johns March 2016 Thesis Advisor

  19. MAPPS (Maintenance Personnel Performance Simulation): a computer simulation model for human reliability analysis

    International Nuclear Information System (INIS)

    Knee, H.E.; Haas, P.M.

    1985-01-01

    A computer model has been developed, sensitivity tested, and evaluated capable of generating reliable estimates of human performance measures in the nuclear power plant (NPP) maintenance context. The model, entitled MAPPS (Maintenance Personnel Performance Simulation), is of the simulation type and is task-oriented. It addresses a number of person-machine, person-environment, and person-person variables and is capable of providing the user with a rich spectrum of important performance measures including mean time for successful task performance by a maintenance team and maintenance team probability of task success. These two measures are particularly important for input to probabilistic risk assessment (PRA) studies which were the primary impetus for the development of MAPPS. The simulation nature of the model along with its generous input parameters and output variables allows its usefulness to extend beyond its input to PRA

  20. Human and organisational factors influencing the reliability of non-destructive testing. An international literary survey

    International Nuclear Information System (INIS)

    Kettunen, J.; Norros, L.

    1996-04-01

    The aim of the study is to chart human and organisational factors influencing the reliability of non-destructive testing (NDT). The emphasis will be in ultrasonic testing (UT) and in the planning and execution of in-service inspections during nuclear power plant maintenance outages. Being a literary survey this study is mainly based on the foreign and domestic research available on the topic. In consequence, the results presented in this report reflect the ideas of international research community. In addition to this, Finnish nuclear power plant operators (Imatran Voima Oy and Teollisuuden Voima Oy), independent inspection organisations and the Finnish Centre for Radiation and Nuclear Safety have provided us with valuable information on NDT theory and practice. Especially, a kind of 'big picture' of non-destructive testing has been pursued in the study. (6 figs., 2 tabs.)

  1. An empirical study on the human error recovery failure probability when using soft controls in NPP advanced MCRs

    International Nuclear Information System (INIS)

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

    2014-01-01

    Highlights: • Many researchers have tried to understand human recovery process or step. • Modeling human recovery process is not sufficient to be applied to HRA. • The operation environment of MCRs in NPPs has changed by adopting new HSIs. • Recovery failure probability in a soft control operation environment is investigated. • Recovery failure probability here would be important evidence for expert judgment. - Abstract: It is well known that probabilistic safety assessments (PSAs) today consider not just hardware failures and environmental events that can impact upon risk, but also human error contributions. Consequently, the focus on reliability and performance management has been on the prevention of human errors and failures rather than the recovery of human errors. However, the recovery of human errors is as important as the prevention of human errors and failures for the safe operation of nuclear power plants (NPPs). For this reason, many researchers have tried to find a human recovery process or step. However, modeling the human recovery process is not sufficient enough to be applied to human reliability analysis (HRA), which requires human error and recovery probabilities. In this study, therefore, human error recovery failure probabilities based on predefined human error modes were investigated by conducting experiments in the operation mockup of advanced/digital main control rooms (MCRs) in NPPs. To this end, 48 subjects majoring in nuclear engineering participated in the experiments. In the experiments, using the developed accident scenario based on tasks from the standard post trip action (SPTA), the steam generator tube rupture (SGTR), and predominant soft control tasks, which are derived from the loss of coolant accident (LOCA) and the excess steam demand event (ESDE), all error detection and recovery data based on human error modes were checked with the performance sheet and the statistical analysis of error recovery/detection was then

  2. Digital music exposure reliably induces temporary threshold shift in normal-hearing human subjects.

    Science.gov (United States)

    Le Prell, Colleen G; Dell, Shawna; Hensley, Brittany; Hall, James W; Campbell, Kathleen C M; Antonelli, Patrick J; Green, Glenn E; Miller, James M; Guire, Kenneth

    2012-01-01

    One of the challenges for evaluating new otoprotective agents for potential benefit in human populations is the availability of an established clinical paradigm with real-world relevance. These studies were explicitly designed to develop a real-world digital music exposure that reliably induces temporary threshold shift (TTS) in normal-hearing human subjects. Thirty-three subjects participated in studies that measured effects of digital music player use on hearing. Subjects selected either rock or pop music, which was then presented at 93 to 95 (n = 10), 98 to 100 (n = 11), or 100 to 102 (n = 12) dBA in-ear exposure level for a period of 4 hr. Audiograms and distortion product otoacoustic emissions (DPOAEs) were measured before and after music exposure. Postmusic tests were initiated 15 min, 1 hr 15 min, 2 hr 15 min, and 3 hr 15 min after the exposure ended. Additional tests were conducted the following day and 1 week later. Changes in thresholds after the lowest-level exposure were difficult to distinguish from test-retest variability; however, TTS was reliably detected after higher levels of sound exposure. Changes in audiometric thresholds had a "notch" configuration, with the largest changes observed at 4 kHz (mean = 6.3 ± 3.9 dB; range = 0-14 dB). Recovery was largely complete within the first 4 hr postexposure, and all subjects showed complete recovery of both thresholds and DPOAE measures when tested 1 week postexposure. These data provide insight into the variability of TTS induced by music-player use in a healthy, normal-hearing, young adult population, with music playlist, level, and duration carefully controlled. These data confirm the likelihood of temporary changes in auditory function after digital music-player use. Such data are essential for the development of a human clinical trial protocol that provides a highly powered design for evaluating novel therapeutics in human clinical trials. Care must be taken to fully inform potential subjects in

  3. Digital music exposure reliably induces temporary threshold shift (TTS) in normal hearing human subjects

    Science.gov (United States)

    Le Prell, C. G.; Dell, S.; Hensley, B.; Hall, J. W.; Campbell, K. C. M.; Antonelli, P. J.; Green, G. E.; Miller, J. M.; Guire, K.

    2012-01-01

    Objectives One of the challenges for evaluating new otoprotective agents for potential benefit in human populations is availability of an established clinical paradigm with real world relevance. These studies were explicitly designed to develop a real-world digital music exposure that reliably induces temporary threshold shift (TTS) in normal hearing human subjects. Design Thirty-three subjects participated in studies that measured effects of digital music player use on hearing. Subjects selected either rock or pop music, which was then presented at 93–95 (n=10), 98–100 (n=11), or 100–102 (n=12) dBA in-ear exposure level for a period of four hours. Audiograms and distortion product otoacoustic emissions (DPOAEs) were measured prior to and after music exposure. Post-music tests were initiated 15 min, 1 hr 15 min, 2 hr 15 min, and 3 hr 15 min after the exposure ended. Additional tests were conducted the following day and one week later. Results Changes in thresholds after the lowest level exposure were difficult to distinguish from test-retest variability; however, TTS was reliably detected after higher levels of sound exposure. Changes in audiometric thresholds had a “notch” configuration, with the largest changes observed at 4 kHz (mean=6.3±3.9dB; range=0–13 dB). Recovery was largely complete within the first 4 hours post-exposure, and all subjects showed complete recovery of both thresholds and DPOAE measures when tested 1-week post-exposure. Conclusions These data provide insight into the variability of TTS induced by music player use in a healthy, normal-hearing, young adult population, with music playlist, level, and duration carefully controlled. These data confirm the likelihood of temporary changes in auditory function following digital music player use. Such data are essential for the development of a human clinical trial protocol that provides a highly powered design for evaluating novel therapeutics in human clinical trials. Care must be

  4. Human reliability analysis during PSA at Trillo NPP: main characteristics and analysis of diagnostic errors

    International Nuclear Information System (INIS)

    Barquin, M.A.; Gomez, F.

    1998-01-01

    The design difference between Trillo NPP and other Spanish nuclear power plants (basic Westinghouse and General Electric designs) were made clear in the Human Reliability Analysis of the Probabilistic Safety Analysis (PSA) for Trillo NPP. The object of this paper is to describe the most significant characteristics of the Human Reliability Analysis carried out in the PSA, with special emphasis on the possible diagnostic errors and their consequences, based on the characteristics in the Emergency Operations Manual for Trillo NPP. - In the case of human errors before the initiating event (type 1), the existence of four redundancies in most of the plant safety systems, means that the impact of this type or error on the final results of the PSA is insignificant. However, in the case common cause errors, especially in certain calibration errors, some actions are significant in the final equation for core damage - The number of human actions that the operator has to carry out during the accidents (type 3) modelled, is relatively small in comparison with this value in other PSAs. This is basically due to the high level of automation at Rillo NPP - The Plant Operations Manual cannot be strictly considered to be a symptoms-based procedure. The operation Group must select the chapter from the Operations Manual to be followed, after having diagnosed the perturbing event, using for this purpose and Emergency and Anomaly Decision Tree (M.O.3.0.1) based on the different indications, alarms and symptoms present in the plant after the perturbing event. For this reason, it was decided to analyse the possible diagnosis errors. In the bibliography on diagnosis and commission errors available at the present time, there is no precise methodology for the analysis of this type of error and its incorporation into PSAs. The method used in the PSA for Trillo y NPP to evaluate this type of interaction, is to develop a Diagnosis Error Table, the object of which is to identify the situations in

  5. Human error mode identification for NPP main control room operations using soft controls

    International Nuclear Information System (INIS)

    Lee, Seung Jun; Kim, Jaewhan; Jang, Seung-Cheol

    2011-01-01

    The operation environment of main control rooms (MCRs) in modern nuclear power plants (NPPs) has considerably changed over the years. Advanced MCRs, which have been designed by adapting digital and computer technologies, have simpler interfaces using large display panels, computerized displays, soft controls, computerized procedure systems, and so on. The actions for the NPP operations are performed using soft controls in advanced MCRs. Soft controls have different features from conventional controls. Operators need to navigate the screens to find indicators and controls and manipulate controls using a mouse, touch screens, and so on. Due to these different interfaces, different human errors should be considered in the human reliability analysis (HRA) for advanced MCRs. In this work, human errors that could occur during operation executions using soft controls were analyzed. This work classified the human errors in soft controls into six types, and the reasons that affect the occurrence of the human errors were also analyzed. (author)

  6. Generating human reliability estimates using expert judgment. Volume 1. Main report

    International Nuclear Information System (INIS)

    Comer, M.K.; Seaver, D.A.; Stillwell, W.G.; Gaddy, C.D.

    1984-11-01

    The US Nuclear Regulatory Commission is conducting a research program to determine the practicality, acceptability, and usefulness of several different methods for obtaining human reliability data and estimates that can be used in nuclear power plant probabilistic risk assessment (PRA). One method, investigated as part of this overall research program, uses expert judgment to generate human error probability (HEP) estimates and associated uncertainty bounds. The project described in this document evaluated two techniques for using expert judgment: paired comparisons and direct numerical estimation. Volume 1 of this report provides a brief overview of the background of the project, the procedure for using psychological scaling techniques to generate HEP estimates and conclusions from evaluation of the techniques. Results of the evaluation indicate that techniques using expert judgment should be given strong consideration for use in developing HEP estimates. In addition, HEP estimates for 35 tasks related to boiling water reactors (BMRs) were obtained as part of the evaluation. These HEP estimates are also included in the report

  7. Human factors perspective on the reliability of NDT in nuclear applications

    International Nuclear Information System (INIS)

    Bertovic, Marija; Mueller, Christina; Fahlbruch, Babette

    2013-01-01

    A series of research studies have been conducted over the course of five years venturing into the fields of in-service inspections (ISI) in nuclear power plants (NPPs) and inspection of manufactured components to be used for permanent nuclear waste disposal. This paper will provide an overview of four research studies, present selected experimental results and suggest ways for optimization of the NDT process, procedures, and training. The experimental results have shown that time pressure and mental workload negatively influence the quality of the manual inspection performance. Noticeable were influences of the organization of the working schedule, communication, procedures, supervision, and demonstration task. Customized Failure Mode and Effects Analysis (FMEA) was used to identify potential human risks, arising during acquisition and evaluation of NDT data. Several preventive measures were suggested and furthermore discussed, with respect to problems that could arise from their application. Experimental results show that implementing human redundancy in critical tasks, such as defect identification, as well as using an automated aid (software) to help operators in decision making about the existence and size of defects, could lead to other kinds of problems, namely social loafing and automation bias that might affect the reliability of NDT in an undesired manner. Shifting focus from the operator, as the main source of errors, to the organization, as the underlying source, is a recommended approach to ensure safety. (orig.) [de

  8. Reliability, Resilience, and Vulnerability criteria for the evaluation of Human Health Risks

    Science.gov (United States)

    Rodak, C. M.; Silliman, S. E.; Bolster, D.

    2011-12-01

    Understanding the impact of water quality on the health of a general population is challenging due high degrees of uncertainty and variability in hydrological, toxicological and human aspects of the system. Assessment of the impact of changes in water quality of a public water supply is critical to management of that water supply. We propose the use of three different system evaluation criteria: Reliability, Resilience and Vulnerability (RRV) as a tool for assessing the impact of uncertainty in the arrival of contaminant mass through time with respect to human health risks on a variable population. These criteria were first introduced to the water resources community by Hashimoto et al (1982). Most simply one can understand these criteria as the following: Reliability is the likelihood of the system being in a state of success; Resilience is the probability that the system will return to a state of success at t+1 if it is in failure at time step t, and Vulnerability is the severity of failure, which here is defined as the maximum health risk. These concepts are applied to a theoretical example where the water quality at a water supply well varies over time: health impact is considered based on sliding, 30-year windows of exposure to water derived from the well. We apply the methodology, in terms of uncertainty in water quality deviations, to eight simulated breakthrough curves of a contaminant at the well: each curve represents equal mass of contaminant arriving at the well over a 70-year lifetime of the well, but different mass distributions over time. These curves are used to investigate the impact of uncertainty in the distribution through time of the contaminant mass at the well, as well as the initial arrival of the contaminant over the 70-year lifetime of the well. In addition to extending the health risk through time with uncertainty in mass distribution, we incorporate variability in the human population to examine the evolution of the three criteria within

  9. A task analysis-linked approach for integrating the human factor in reliability assessments of nuclear power plants

    International Nuclear Information System (INIS)

    Ryan, T.G.

    1988-01-01

    This paper describes an emerging Task Analysis-Linked Evaluation Technique (TALENT) for assessing the contributions of human error to nuclear power plant systems unreliability and risk. Techniques such as TALENT are emerging as a recognition that human error is a primary contributor to plant safety, however, it has been a peripheral consideration to data in plant reliability evaluations. TALENT also recognizes that involvement of persons with behavioral science expertise is required to support plant reliability and risk analyses. A number of state-of-knowledge human reliability analysis tools are also discussed which support the TALENT process. The core of TALENT is comprised of task, timeline and interface analysis data which provide the technology base for event and fault tree development, serve as criteria for selecting and evaluating performance shaping factors, and which provide a basis for auditing TALENT results. Finally, programs and case studies used to refine the TALENT process are described along with future research needs in the area. (author)

  10. Fuzzy sets as extension of probabilistic models for evaluating human reliability

    International Nuclear Information System (INIS)

    Przybylski, F.

    1996-11-01

    On the base of a survey of established quantification methodologies for evaluating human reliability, a new computerized methodology was developed in which a differential consideration of user uncertainties is made. In this quantification method FURTHER (FUzzy Sets Related To Human Error Rate Prediction), user uncertainties are quantified separately from model and data uncertainties. As tools fuzzy sets are applied which, however, stay hidden to the method's user. The user in the quantification process only chooses an action pattern, performance shaping factors and natural language expressions. The acknowledged method HEART (Human Error Assessment and Reduction Technique) serves as foundation of the fuzzy set approach FURTHER. By means of this method, the selection of a basic task in connection with its basic error probability, the decision how correct the basic task's selection is, the selection of a peformance shaping factor, and the decision how correct the selection and how important the performance shaping factor is, were identified as aspects of fuzzification. This fuzzification is made on the base of data collection and information from literature as well as of the estimation by competent persons. To verify the ammount of additional information to be received by the usage of fuzzy sets, a benchmark session was accomplished. In this benchmark twelve actions were assessed by five test-persons. In case of the same degree of detail in the action modelling process, the bandwidths of the interpersonal evaluations decrease in FURTHER in comparison with HEART. The uncertainties of the single results could not be reduced up to now. The benchmark sessions conducted so far showed plausible results. A further testing of the fuzzy set approach by using better confirmed fuzzy sets can only be achieved in future practical application. Adequate procedures, however, are provided. (orig.) [de

  11. Reliable generation of induced pluripotent stem cells from human lymphoblastoid cell lines.

    Science.gov (United States)

    Barrett, Robert; Ornelas, Loren; Yeager, Nicole; Mandefro, Berhan; Sahabian, Anais; Lenaeus, Lindsay; Targan, Stephan R; Svendsen, Clive N; Sareen, Dhruv

    2014-12-01

    Patient-specific induced pluripotent stem cells (iPSCs) hold great promise for many applications, including disease modeling to elucidate mechanisms involved in disease pathogenesis, drug screening, and ultimately regenerative medicine therapies. A frequently used starting source of cells for reprogramming has been dermal fibroblasts isolated from skin biopsies. However, numerous repositories containing lymphoblastoid cell lines (LCLs) generated from a wide array of patients also exist in abundance. To date, this rich bioresource has been severely underused for iPSC generation. We first attempted to create iPSCs from LCLs using two existing methods but were unsuccessful. Here we report a new and more reliable method for LCL reprogramming using episomal plasmids expressing pluripotency factors and p53 shRNA in combination with small molecules. The LCL-derived iPSCs (LCL-iPSCs) exhibited identical characteristics to fibroblast-derived iPSCs (fib-iPSCs), wherein they retained their genotype, exhibited a normal pluripotency profile, and readily differentiated into all three germ-layer cell types. As expected, they also maintained rearrangement of the heavy chain immunoglobulin locus. Importantly, we also show efficient iPSC generation from LCLs of patients with spinal muscular atrophy and inflammatory bowel disease. These LCL-iPSCs retained the disease mutation and could differentiate into neurons, spinal motor neurons, and intestinal organoids, all of which were virtually indistinguishable from differentiated cells derived from fib-iPSCs. This method for reliably deriving iPSCs from patient LCLs paves the way for using invaluable worldwide LCL repositories to generate new human iPSC lines, thus providing an enormous bioresource for disease modeling, drug discovery, and regenerative medicine applications. ©AlphaMed Press.

  12. Human error probability evaluation as part of reliability analysis of digital protection system of advanced pressurized water reactor - APR 1400

    International Nuclear Information System (INIS)

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

    2003-03-01

    A case of study on human reliability analysis has been performed as part of reliability analysis of digital protection system of the reactor automatically actuates the shutdown system of the reactor when demanded. However, the safety analysis takes credit for operator action as a diverse mean for tripping the reactor for, though a low probability, ATWS scenario. Based on the available information two cases, viz., human error in tripping the reactor and calibration error for instrumentations in protection system, have been analyzed. Wherever applicable a parametric study has also been performed

  13. Human events reference for ATHEANA (HERA) database description and preliminary user's manual

    International Nuclear Information System (INIS)

    Auflick, J.L.; Hahn, H.A.; Pond, D.J.

    1998-01-01

    The Technique for Human Error Analysis (ATHEANA) is a newly developed human reliability analysis (HRA) methodology that aims to facilitate better representation and integration of human performance into probabilistic risk assessment (PRA) modeling and quantification by analyzing risk-significant operating experience in the context of existing behavioral science models. The fundamental premise of ATHEANA is that error-forcing contexts (EFCs), which refer to combinations of equipment/material conditions and performance shaping factors (PSFs), set up or create the conditions under which unsafe actions (UAs) can occur. Because ATHEANA relies heavily on the analysis of operational events that have already occurred as a mechanism for generating creative thinking about possible EFCs, a database, called the Human Events Reference for ATHEANA (HERA), has been developed to support the methodology. This report documents the initial development efforts for HERA

  14. Human events reference for ATHEANA (HERA) database description and preliminary user`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Auflick, J.L.; Hahn, H.A.; Pond, D.J.

    1998-05-27

    The Technique for Human Error Analysis (ATHEANA) is a newly developed human reliability analysis (HRA) methodology that aims to facilitate better representation and integration of human performance into probabilistic risk assessment (PRA) modeling and quantification by analyzing risk-significant operating experience in the context of existing behavioral science models. The fundamental premise of ATHEANA is that error-forcing contexts (EFCs), which refer to combinations of equipment/material conditions and performance shaping factors (PSFs), set up or create the conditions under which unsafe actions (UAs) can occur. Because ATHEANA relies heavily on the analysis of operational events that have already occurred as a mechanism for generating creative thinking about possible EFCs, a database, called the Human Events Reference for ATHEANA (HERA), has been developed to support the methodology. This report documents the initial development efforts for HERA.

  15. Human Events Reference for ATHEANA (HERA) Database Description and Preliminary User's Manual

    Energy Technology Data Exchange (ETDEWEB)

    Auflick, J.L.

    1999-08-12

    The Technique for Human Error Analysis (ATHEANA) is a newly developed human reliability analysis (HRA) methodology that aims to facilitate better representation and integration of human performance into probabilistic risk assessment (PRA) modeling and quantification by analyzing risk-significant operating experience in the context of existing behavioral science models. The fundamental premise of ATHEANA is that error forcing contexts (EFCs), which refer to combinations of equipment/material conditions and performance shaping factors (PSFs), set up or create the conditions under which unsafe actions (UAs) can occur. Because ATHEANA relies heavily on the analysis of operational events that have already occurred as a mechanism for generating creative thinking about possible EFCs, a database (db) of analytical operational events, called the Human Events Reference for ATHEANA (HERA), has been developed to support the methodology. This report documents the initial development efforts for HERA.

  16. Human Events Reference for ATHEANA (HERA) Database Description and Preliminary User's Manual

    International Nuclear Information System (INIS)

    Auflick, J.L.

    1999-01-01

    The Technique for Human Error Analysis (ATHEANA) is a newly developed human reliability analysis (HRA) methodology that aims to facilitate better representation and integration of human performance into probabilistic risk assessment (PRA) modeling and quantification by analyzing risk-significant operating experience in the context of existing behavioral science models. The fundamental premise of ATHEANA is that error forcing contexts (EFCs), which refer to combinations of equipment/material conditions and performance shaping factors (PSFs), set up or create the conditions under which unsafe actions (UAs) can occur. Because ATHEANA relies heavily on the analysis of operational events that have already occurred as a mechanism for generating creative thinking about possible EFCs, a database (db) of analytical operational events, called the Human Events Reference for ATHEANA (HERA), has been developed to support the methodology. This report documents the initial development efforts for HERA

  17. Human Reliability Analysis for In-Tank Precipitation Alignment and Startup of Emergency Purge Ventilation Equipment. Revision 3

    International Nuclear Information System (INIS)

    Shapiro, B.J.; Britt, T.E.

    1994-10-01

    This report documents the methodology used for calculating the human error probability for establishing air based ventilation using emergency purge ventilation equipment on In-Tank Precipitation (ITP) processing tanks 48 and 49 after failure of the nitrogen purge system following a seismic event. The analyses were performed according to THERP (Technique for Human Error Rate Prediction) as described in NUREG/CR-1278-F, ''Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications.'' The calculated human error probabilities are provided as input to the Fault Tree Analysis for the ITP Nitrogen Purge System

  18. Human error probability quantification using fuzzy methodology in nuclear plants

    International Nuclear Information System (INIS)

    Nascimento, Claudio Souza do

    2010-01-01

    This work obtains Human Error Probability (HEP) estimates from operator's actions in response to emergency situations a hypothesis on Research Reactor IEA-R1 from IPEN. It was also obtained a Performance Shaping Factors (PSF) evaluation in order to classify them according to their influence level onto the operator's actions and to determine these PSF actual states over the plant. Both HEP estimation and PSF evaluation were done based on Specialists Evaluation using interviews and questionnaires. Specialists group was composed from selected IEA-R1 operators. Specialist's knowledge representation into linguistic variables and group evaluation values 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 Human Reliability Analysis (HRA). (author)

  19. Quantitative developments in the cognitive reliability and error analysis method (CREAM) for the assessment of human performance

    International Nuclear Information System (INIS)

    Marseguerra, Marzio; Zio, Enrico; Librizzi, Massimo

    2006-01-01

    The current 'second generation' approaches in human reliability analysis focus their attention on the contextual conditions under which a given action is performed rather than on the notion of inherent human error probabilities, as was done in the earlier 'first generation' techniques. Among the 'second generation' methods, this paper considers the Cognitive Reliability and Error Analysis Method (CREAM) and proposes some developments with respect to a systematic procedure for computing probabilities of action failure. The starting point for the quantification is a previously introduced fuzzy version of the CREAM paradigm which is here further extended to include uncertainty on the qualification of the conditions under which the action is performed and to account for the fact that the effects of the common performance conditions (CPCs) on performance reliability may not all be equal. By the proposed approach, the probability of action failure is estimated by rating the performance conditions in terms of their effect on the action

  20. Assessments and applications to enhance human reliability and reduce risk during less-than-full-power operations

    International Nuclear Information System (INIS)

    Hannaman, G.W.; Singh, A.

    1992-01-01

    Study of events, interviews with plant personnel, and applications of risk studies indicate that the risk of a potential accident during less-than-full-power (LTFP) operation is becoming a greater fraction of the risk as improvements are made to the full-power operations. Industry efforts have been increased to reduce risk and the cost of shutdown operations. These efforts consider the development and application of advanced tools to help utilities proactively identify issues and develop contingencies and interventions to enhance reliability and reduce risk of low-power operations at nuclear power plants. The role for human reliability assessments is to help improve utility outage planning to better achieve schedule and risk control objectives. Improvements are expected to include intervention tools to identify and reduce human error, definition of new instructional modules, and prioritization of risk reduction issues for operators. The Electric Power Research Institute is sponsoring a project to address the identification and quantification of factors that affect human reliability during LTFP operation of nuclear power plants. The results of this project are expected to promote the development of proactively applied interventions and contingencies for enhanced human reliability during shutdown operations

  1. Importance of independent and dependent human error to system reliability and plant safety

    International Nuclear Information System (INIS)

    Dach, K.

    1988-08-01

    Uncertainty analysis of the quantification of the unavailability for the emergency core cooling system was made. The reliability analysis of the low pressure injection system (LPIS) of the ECCS of WWER-440 reactor was also performed. Results of reliability analysis proved that LPIS reliability under normal conditions is sufficient and can be increased by two orders of magnitude. This increase in reliability can be achieved by means of simple changes such as securing an opening of the quick-acting fittings at LPIS discharge line. A method for analysis of systems uncertainty with periodic inspected components was elaborated and verified by performing an analysis of the medium size system. Refs, figs and tabs

  2. 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)

  3. Prevent cervical cancer by screening with reliable human papillomavirus detection and genotyping

    International Nuclear Information System (INIS)

    Ge, Shichao; Gong, Bo; Cai, Xushan; Yang, Xiaoer; Gan, Xiaowei; Tong, Xinghai; Li, Haichuan; Zhu, Meijuan; Yang, Fengyun; Zhou, Hongrong; Hong, Guofan

    2012-01-01

    The incidence of cervical cancer is expected to rise sharply in China. A reliable routine human papillomavirus (HPV) detection and genotyping test to be supplemented by the limited Papanicolaou cytology facilities is urgently needed to help identify the patients with cervical precancer for preventive interventions. To this end, we evaluated a nested polymerase chain reaction (PCR) protocol for detection of HPV L1 gene DNA in cervicovaginal cells. The PCR amplicons were genotyped by direct DNA sequencing. In parallel, split samples were subjected to a Digene HC2 HPV test which has been widely used for “cervical cancer risk” screen. Of the 1826 specimens, 1655 contained sufficient materials for analysis and 657 were truly negative. PCR/DNA sequencing showed 674 infected by a single high-risk HPV, 188 by a single low-risk HPV, and 136 by multiple HPV genotypes with up to five HPV genotypes in one specimen. In comparison, the HC2 test classified 713 specimens as infected by high-risk HPV, and 942 as negative for HPV infections. The high-risk HC2 test correctly detected 388 (57.6%) of the 674 high-risk HPV isolates in clinical specimens, mislabeled 88 (46.8%) of the 188 low-risk HPV isolates as high-risk genotypes, and classified 180 (27.4%) of the 657 “true-negative” samples as being infected by high-risk HPV. It was found to cross-react with 20 low-risk HPV genotypes. We conclude that nested PCR detection of HPV followed by short target DNA sequencing can be used for screening and genotyping to formulate a paradigm in clinical management of HPV-related disorders in a rapidly developing economy

  4. Human decomposition and the reliability of a 'Universal' model for post mortem interval estimations.

    Science.gov (United States)

    Cockle, Diane L; Bell, Lynne S

    2015-08-01

    Human decomposition is a complex biological process driven by an array of variables which are not clearly understood. The medico-legal community have long been searching for a reliable method to establish the post-mortem interval (PMI) for those whose deaths have either been hidden, or gone un-noticed. To date, attempts to develop a PMI estimation method based on the state of the body either at the scene or at autopsy have been unsuccessful. One recent study has proposed that two simple formulae, based on the level of decomposition humidity and temperature, could be used to accurately calculate the PMI for bodies outside, on or under the surface worldwide. This study attempted to validate 'Formula I' [1] (for bodies on the surface) using 42 Canadian cases with known PMIs. The results indicated that bodies exposed to warm temperatures