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

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

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.

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.

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

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)

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.

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)

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

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

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

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

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

Applicability of HRA to support advanced MMI design review

International Nuclear Information System (INIS)

Kim, Inn Seock

2000-01-01

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

Human Reliability Analysis. 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

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.

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

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

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.

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

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.

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

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)

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

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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.

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

Energy Technology Data Exchange (ETDEWEB)

Jacques Hugo; David I Gertman

2012-06-01

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

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

International Nuclear Information System (INIS)

Hugo, Jacques; Gertman, David I.

2012-01-01

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

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)

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)

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

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...

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

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.

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)

"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...

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.)

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

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

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...

Integrated sequence analysis. Final report

International Nuclear Information System (INIS)

Andersson, K.; Pyy, P.

1998-02-01

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

Integrated sequence analysis. Final report

Energy Technology Data Exchange (ETDEWEB)

Andersson, K.; Pyy, P

1998-02-01

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

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

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

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)

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

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

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

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)

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

International Nuclear Information System (INIS)

1996-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-12-31

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

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)

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.

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)

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

International Nuclear Information System (INIS)

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

1995-08-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-08-01

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

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.

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.

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

International Nuclear Information System (INIS)

Alvarenga, Marco Antonio Bayout; Fonseca, Renato Alves

2009-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-07-01

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

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

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)

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

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.

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.

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.

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

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)

A human error analysis methodology, AGAPE-ET, for emergency tasks in nuclear power plants and its application

Energy Technology Data Exchange (ETDEWEB)

Kim, Jae Whan; Jung, Won Dea [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-03-01

This report presents a procedurised human reliability analysis (HRA) methodology, AGAPE-ET (A Guidance And Procedure for Human Error Analysis for Emergency Tasks), for both qualitative error analysis and quantification of human error probability (HEP) of emergency tasks in nuclear power plants. The AGAPE-ET is based on the simplified cognitive model. By each cognitive function, error causes or error-likely situations have been identified considering the characteristics of the performance of each cognitive function and influencing mechanism of PIFs on the cognitive function. Then, error analysis items have been determined from the identified error causes or error-likely situations to help the analysts cue or guide overall human error analysis. A human error analysis procedure based on the error analysis items is organised. The basic scheme for the quantification of HEP consists in the multiplication of the BHEP assigned by the error analysis item and the weight from the influencing factors decision tree (IFDT) constituted by cognitive function. The method can be characterised by the structured identification of the weak points of the task required to perform and the efficient analysis process that the analysts have only to carry out with the necessary cognitive functions. The report also presents the the application of AFAPE-ET to 31 nuclear emergency tasks and its results. 42 refs., 7 figs., 36 tabs. (Author)

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

International Nuclear Information System (INIS)

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

2000-01-01

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

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

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.

Development of a systematic methodology to select hazard analysis techniques for nuclear facilities

Energy Technology Data Exchange (ETDEWEB)

Vasconcelos, Vanderley de; Reis, Sergio Carneiro dos; Costa, Antonio Carlos Lopes da [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)]. E-mails: vasconv@cdtn.br; reissc@cdtn.br; aclc@cdtn.br; Jordao, Elizabete [Universidade Estadual de Campinas (UNICAMP), SP (Brazil). Faculdade de Engenharia Quimica]. E-mail: bete@feq.unicamp.br

2008-07-01

In order to comply with licensing requirements of regulatory bodies risk assessments of nuclear facilities should be carried out. In Brazil, such assessments are part of the Safety Analysis Reports, required by CNEN (Brazilian Nuclear Energy Commission), and of the Risk Analysis Studies, required by the competent environmental bodies. A risk assessment generally includes the identification of the hazards and accident sequences that can occur, as well as the estimation of the frequencies and effects of these unwanted events on the plant, people, and environment. The hazard identification and analysis are also particularly important when implementing an Integrated Safety, Health, and Environment Management System following ISO 14001, BS 8800 and OHSAS 18001 standards. Among the myriad of tools that help the process of hazard analysis can be highlighted: CCA (Cause- Consequence Analysis); CL (Checklist Analysis); ETA (Event Tree Analysis); FMEA (Failure Mode and Effects Analysis); FMECA (Failure Mode, Effects and Criticality Analysis); FTA (Fault Tree Analysis); HAZOP (Hazard and Operability Study); HRA (Human Reliability Analysis); Pareto Analysis; PHA (Preliminary Hazard Analysis); RR (Relative Ranking); SR (Safety Review); WI (What-If); and WI/CL (What-If/Checklist Analysis). The choice of a particular technique or a combination of techniques depends on many factors like motivation of the analysis, available data, complexity of the process being analyzed, expertise available on hazard analysis, and initial perception of the involved risks. This paper presents a systematic methodology to select the most suitable set of tools to conduct the hazard analysis, taking into account the mentioned involved factors. Considering that non-reactor nuclear facilities are, to a large extent, chemical processing plants, the developed approach can also be applied to analysis of chemical and petrochemical plants. The selected hazard analysis techniques can support cost

Development of a systematic methodology to select hazard analysis techniques for nuclear facilities

International Nuclear Information System (INIS)

Vasconcelos, Vanderley de; Reis, Sergio Carneiro dos; Costa, Antonio Carlos Lopes da; Jordao, Elizabete

2008-01-01

In order to comply with licensing requirements of regulatory bodies risk assessments of nuclear facilities should be carried out. In Brazil, such assessments are part of the Safety Analysis Reports, required by CNEN (Brazilian Nuclear Energy Commission), and of the Risk Analysis Studies, required by the competent environmental bodies. A risk assessment generally includes the identification of the hazards and accident sequences that can occur, as well as the estimation of the frequencies and effects of these unwanted events on the plant, people, and environment. The hazard identification and analysis are also particularly important when implementing an Integrated Safety, Health, and Environment Management System following ISO 14001, BS 8800 and OHSAS 18001 standards. Among the myriad of tools that help the process of hazard analysis can be highlighted: CCA (Cause- Consequence Analysis); CL (Checklist Analysis); ETA (Event Tree Analysis); FMEA (Failure Mode and Effects Analysis); FMECA (Failure Mode, Effects and Criticality Analysis); FTA (Fault Tree Analysis); HAZOP (Hazard and Operability Study); HRA (Human Reliability Analysis); Pareto Analysis; PHA (Preliminary Hazard Analysis); RR (Relative Ranking); SR (Safety Review); WI (What-If); and WI/CL (What-If/Checklist Analysis). The choice of a particular technique or a combination of techniques depends on many factors like motivation of the analysis, available data, complexity of the process being analyzed, expertise available on hazard analysis, and initial perception of the involved risks. This paper presents a systematic methodology to select the most suitable set of tools to conduct the hazard analysis, taking into account the mentioned involved factors. Considering that non-reactor nuclear facilities are, to a large extent, chemical processing plants, the developed approach can also be applied to analysis of chemical and petrochemical plants. The selected hazard analysis techniques can support cost

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

International Nuclear Information System (INIS)

Mercurio, D.

2011-01-01

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

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

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.

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)

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.

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

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

International Nuclear Information System (INIS)

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

2005-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2005-12-15

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

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

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

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.

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

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.

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

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

International Nuclear Information System (INIS)

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

2010-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-06-01

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

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.

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.

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

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

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.)

Performance Shaping Factors Assessments and Application to PHWR Outages

Energy Technology Data Exchange (ETDEWEB)

Lee, Seung Woo

2007-02-15

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

Performance Shaping Factors Assessments and Application to PHWR Outages

International Nuclear Information System (INIS)

Lee, Seung Woo

2007-02-01

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

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)

Safety analysis methodology for OPR 1000

International Nuclear Information System (INIS)

Hwang-Yong, Jun

2005-01-01

Full text: Korea Electric Power Research Institute (KEPRI) has been developing inhouse safety analysis methodology based on the delicate codes available to KEPRI to overcome the problems arising from currently used vendor oriented methodologies. For the Loss of Coolant Accident (LOCA) analysis, the KREM (KEPRI Realistic Evaluation Methodology) has been developed based on the RELAP-5 code. The methodology was approved for the Westinghouse 3-loop plants by the Korean regulatory organization and the project to extent the methodology to the Optimized Power Reactor 1000 (OPR1000) has been ongoing since 2001. Also, for the Non-LOCA analysis, the KNAP (Korea Non-LOCA Analysis Package) has been developed using the UNICORN-TM code system. To demonstrate the feasibility of these codes systems and methodologies, some typical cases of the design basis accidents mentioned in the final safety analysis report (FSAR) were analyzed. (author)

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

Regional Shelter Analysis Methodology

Energy Technology Data Exchange (ETDEWEB)

Dillon, Michael B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dennison, Deborah [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kane, Jave [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Walker, Hoyt [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Miller, Paul [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2015-08-01

The fallout from a nuclear explosion has the potential to injure or kill 100,000 or more people through exposure to external gamma (fallout) radiation. Existing buildings can reduce radiation exposure by placing material between fallout particles and exposed people. Lawrence Livermore National Laboratory was tasked with developing an operationally feasible methodology that could improve fallout casualty estimates. The methodology, called a Regional Shelter Analysis, combines the fallout protection that existing buildings provide civilian populations with the distribution of people in various locations. The Regional Shelter Analysis method allows the consideration of (a) multiple building types and locations within buildings, (b) country specific estimates, (c) population posture (e.g., unwarned vs. minimally warned), and (d) the time of day (e.g., night vs. day). The protection estimates can be combined with fallout predictions (or measurements) to (a) provide a more accurate assessment of exposure and injury and (b) evaluate the effectiveness of various casualty mitigation strategies. This report describes the Regional Shelter Analysis methodology, highlights key operational aspects (including demonstrating that the methodology is compatible with current tools), illustrates how to implement the methodology, and provides suggestions for future work.

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)

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).

Preliminary safety analysis methodology for the SMART

Energy Technology Data Exchange (ETDEWEB)

Bae, Kyoo Hwan; Chung, Y. J.; Kim, H. C.; Sim, S. K.; Lee, W. J.; Chung, B. D.; Song, J. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-03-01

This technical report was prepared for a preliminary safety analysis methodology of the 330MWt SMART (System-integrated Modular Advanced ReacTor) which has been developed by Korea Atomic Energy Research Institute (KAERI) and funded by the Ministry of Science and Technology (MOST) since July 1996. This preliminary safety analysis methodology has been used to identify an envelope for the safety of the SMART conceptual design. As the SMART design evolves, further validated final safety analysis methodology will be developed. Current licensing safety analysis methodology of the Westinghouse and KSNPP PWRs operating and under development in Korea as well as the Russian licensing safety analysis methodology for the integral reactors have been reviewed and compared to develop the preliminary SMART safety analysis methodology. SMART design characteristics and safety systems have been reviewed against licensing practices of the PWRs operating or KNGR (Korean Next Generation Reactor) under construction in Korea. Detailed safety analysis methodology has been developed for the potential SMART limiting events of main steam line break, main feedwater pipe break, loss of reactor coolant flow, CEA withdrawal, primary to secondary pipe break and the small break loss of coolant accident. SMART preliminary safety analysis methodology will be further developed and validated in parallel with the safety analysis codes as the SMART design further evolves. Validated safety analysis methodology will be submitted to MOST as a Topical Report for a review of the SMART licensing safety analysis methodology. Thus, it is recommended for the nuclear regulatory authority to establish regulatory guides and criteria for the integral reactor. 22 refs., 18 figs., 16 tabs. (Author)

METHODOLOGICAL ELEMENTS OF SITUATIONAL ANALYSIS

Directory of Open Access Journals (Sweden)

Tetyana KOVALCHUK

2016-07-01

Full Text Available The article deals with the investigation of theoretical and methodological principles of situational analysis. The necessity of situational analysis is proved in modern conditions. The notion “situational analysis” is determined. We have concluded that situational analysis is a continuous system study which purpose is to identify dangerous situation signs, to evaluate comprehensively such signs influenced by a system of objective and subjective factors, to search for motivated targeted actions used to eliminate adverse effects of the exposure of the system to the situation now and in the future and to develop the managerial actions needed to bring the system back to norm. It is developed a methodological approach to the situational analysis, its goal is substantiated, proved the expediency of diagnostic, evaluative and searching functions in the process of situational analysis. The basic methodological elements of the situational analysis are grounded. The substantiation of the principal methodological elements of system analysis will enable the analyst to develop adaptive methods able to take into account the peculiar features of a unique object which is a situation that has emerged in a complex system, to diagnose such situation and subject it to system and in-depth analysis, to identify risks opportunities, to make timely management decisions as required by a particular period.

Causal Meta-Analysis : Methodology and Applications

NARCIS (Netherlands)

Bax, L.J.

2009-01-01

Meta-analysis is a statistical method to summarize research data from multiple studies in a quantitative manner. This dissertation addresses a number of methodological topics in causal meta-analysis and reports the development and validation of meta-analysis software. In the first (methodological)

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

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

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.

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-07-01

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

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

International Nuclear Information System (INIS)

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

2007-01-01

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

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)

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

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)

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

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)

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.

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)

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

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

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)

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.

A Human Error Analysis with Physiological Signals during Utilizing Digital Devices

Energy Technology Data Exchange (ETDEWEB)

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

2011-10-15

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

Update of Part 61 Impacts Analysis Methodology. Methodology report. Volume 1

International Nuclear Information System (INIS)

Oztunali, O.I.; Roles, G.W.

1986-01-01

Under contract to the US Nuclear Regulatory Commission, the Envirosphere Company has expanded and updated the impacts analysis methodology used during the development of the 10 CFR Part 61 rule to allow improved consideration of the costs and impacts of treatment and disposal of low-level waste that is close to or exceeds Class C concentrations. The modifications described in this report principally include: (1) an update of the low-level radioactive waste source term, (2) consideration of additional alternative disposal technologies, (3) expansion of the methodology used to calculate disposal costs, (4) consideration of an additional exposure pathway involving direct human contact with disposed waste due to a hypothetical drilling scenario, and (5) use of updated health physics analysis procedures (ICRP-30). Volume 1 of this report describes the calculational algorithms of the updated analysis methodology

Update of Part 61 Impacts Analysis Methodology. Methodology report. Volume 1

Energy Technology Data Exchange (ETDEWEB)

Oztunali, O.I.; Roles, G.W.

1986-01-01

Under contract to the US Nuclear Regulatory Commission, the Envirosphere Company has expanded and updated the impacts analysis methodology used during the development of the 10 CFR Part 61 rule to allow improved consideration of the costs and impacts of treatment and disposal of low-level waste that is close to or exceeds Class C concentrations. The modifications described in this report principally include: (1) an update of the low-level radioactive waste source term, (2) consideration of additional alternative disposal technologies, (3) expansion of the methodology used to calculate disposal costs, (4) consideration of an additional exposure pathway involving direct human contact with disposed waste due to a hypothetical drilling scenario, and (5) use of updated health physics analysis procedures (ICRP-30). Volume 1 of this report describes the calculational algorithms of the updated analysis methodology.

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.

Development of analysis methodology on turbulent thermal stripping

Energy Technology Data Exchange (ETDEWEB)

Yoo, Geun Jong; Jeon, Won Dae; Han, Jin Woo; Gu, Byong Kook [Changwon National University, Changwon(Korea)

2001-03-01

For developing analysis methodology, important governing factors of thermal stripping phenomena are identified as geometric configuration and flow characteristics such as velocity. Along these factors, performance of turbulence models in existing analysis methodology are evaluated against experimental data. Status of DNS application is also accessed based on literature. Evaluation results are reflected in setting up the new analysis methodology. From the evaluation of existing analysis methodology, Full Reynolds Stress model is identified as best one among other turbulence models. And LES is found to be able to provide time dependent turbulence values. Further improvements in near-wall region and temperature variance equation are required for FRS and implementation of new sub-grid scale models is also required for LES. Through these improvements, new reliable analysis methodology for thermal stripping can be developed. 30 refs., 26 figs., 6 tabs. (Author)

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.

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

International Nuclear Information System (INIS)

Oxstrand, J.

2010-12-01

The main goal of this Nordic Nuclear Safety Research Council (NKS) project is to produce guidance for how to use human reliability analysis (HRA) to strengthen overall safety. The project consists of two substudies: The Nordic Point of View - A User Needs Analysis, and The American Point of View - Insights of How the US Nuclear Industry Works with HRA. The purpose of the Nordic Point of View study was a user needs analysis that aimed to survey current HRA practices in the Nordic nuclear industry, with the main focus being to connect HRA to system design. In this study, 26 Nordic (Swedish and Finnish) nuclear power plant specialists with research, practitioner, and regulatory expertise in HRA, PRA, HSI, and human performance were interviewed. This study was completed in 2009. This study concludes that HRA is an important tool when dealing with human factors in control room design or modernizations. The Nordic Point of View study showed areas where the use of HRA in the Nordic nuclear industry could be improved. To gain more knowledge about how these improvements could be made, and what improvements to focus on, the second study was conducted. The second study is focused on the American nuclear industry, which has many more years of experience with risk assessment and human reliability than the Nordic nuclear industry. Interviews were conducted to collect information to help the author understand the similarities and differences between the American and the Nordic nuclear industries, and to find data regarding the findings from the first study. The main focus of this report is to identify potential HRA improvements based on the data collected in the American Point of View survey. (Author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-12-15

The main goal of this Nordic Nuclear Safety Research Council (NKS) project is to produce guidance for how to use human reliability analysis (HRA) to strengthen overall safety. The project consists of two substudies: The Nordic Point of View - A User Needs Analysis, and The American Point of View - Insights of How the US Nuclear Industry Works with HRA. The purpose of the Nordic Point of View study was a user needs analysis that aimed to survey current HRA practices in the Nordic nuclear industry, with the main focus being to connect HRA to system design. In this study, 26 Nordic (Swedish and Finnish) nuclear power plant specialists with research, practitioner, and regulatory expertise in HRA, PRA, HSI, and human performance were interviewed. This study was completed in 2009. This study concludes that HRA is an important tool when dealing with human factors in control room design or modernizations. The Nordic Point of View study showed areas where the use of HRA in the Nordic nuclear industry could be improved. To gain more knowledge about how these improvements could be made, and what improvements to focus on, the second study was conducted. The second study is focused on the American nuclear industry, which has many more years of experience with risk assessment and human reliability than the Nordic nuclear industry. Interviews were conducted to collect information to help the author understand the similarities and differences between the American and the Nordic nuclear industries, and to find data regarding the findings from the first study. The main focus of this report is to identify potential HRA improvements based on the data collected in the American Point of View survey. (Author)

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)

Issues in benchmarking human reliability analysis methods: A literature review

International Nuclear Information System (INIS)

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

2010-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-04-01

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

SMART performance analysis methodology

International Nuclear Information System (INIS)

Lim, H. S.; Kim, H. C.; Lee, D. J.

2001-04-01

To ensure the required and desired operation over the plant lifetime, the performance analysis for the SMART NSSS design is done by means of the specified analysis methodologies for the performance related design basis events(PRDBE). The PRDBE is an occurrence(event) that shall be accommodated in the design of the plant and whose consequence would be no more severe than normal service effects of the plant equipment. The performance analysis methodology which systematizes the methods and procedures to analyze the PRDBEs is as follows. Based on the operation mode suitable to the characteristics of the SMART NSSS, the corresponding PRDBEs and allowable range of process parameters for these events are deduced. With the developed control logic for each operation mode, the system thermalhydraulics are analyzed for the chosen PRDBEs using the system analysis code. Particularly, because of different system characteristics of SMART from the existing commercial nuclear power plants, the operation mode, PRDBEs, control logic, and analysis code should be consistent with the SMART design. This report presents the categories of the PRDBEs chosen based on each operation mode and the transition among these and the acceptance criteria for each PRDBE. It also includes the analysis methods and procedures for each PRDBE and the concept of the control logic for each operation mode. Therefore this report in which the overall details for SMART performance analysis are specified based on the current SMART design, would be utilized as a guide for the detailed performance analysis

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.)

Exploring Participatory Methodologies in Organizational Discourse Analysis

DEFF Research Database (Denmark)

Plotnikof, Mie

2014-01-01

Recent debates in the field of organizational discourse analysis stress contrasts in approaches as single-level vs. multi-level, critical vs. participatory, discursive vs. material methods. They raise methodological issues of combining such to embrace multimodality in order to enable new contribu......Recent debates in the field of organizational discourse analysis stress contrasts in approaches as single-level vs. multi-level, critical vs. participatory, discursive vs. material methods. They raise methodological issues of combining such to embrace multimodality in order to enable new...... contributions. As regards conceptual efforts are made but further exploration of methodological combinations and their practical implications are called for. This paper argues 1) to combine methodologies by approaching this as scholarly subjectification processes, and 2) to perform combinations in both...

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)

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

Energy Technology Data Exchange (ETDEWEB)

Reer, B

2007-06-15

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

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

International Nuclear Information System (INIS)

Reer, B.

2007-06-01

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

A technique for human error analysis (ATHEANA)

International Nuclear Information System (INIS)

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

1996-05-01

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

A technique for human error analysis (ATHEANA)

Energy Technology Data Exchange (ETDEWEB)

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

1996-05-01

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

Human error probability quantification using fuzzy methodology in nuclear plants; Aplicacao da metodologia fuzzy na quantificacao da probabilidade de erro humano em instalacoes nucleares

Energy Technology Data Exchange (ETDEWEB)

Nascimento, Claudio Souza do

2010-07-01

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)

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)

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

Constructive Analysis : A Study in Epistemological Methodology

DEFF Research Database (Denmark)

Ahlström, Kristoffer

, and develops a framework for a kind of analysis that is more in keeping with recent psychological research on categorization. Finally, it is shown that this kind of analysis can be applied to the concept of justification in a manner that furthers the epistemological goal of providing intellectual guidance.......The present study is concerned the viability of the primary method in contemporary philosophy, i.e., conceptual analysis. Starting out by tracing the roots of this methodology to Platonic philosophy, the study questions whether such a methodology makes sense when divorced from Platonic philosophy...

Development of Advanced Non-LOCA Analysis Methodology for Licensing

International Nuclear Information System (INIS)

Jang, Chansu; Um, Kilsup; Choi, Jaedon

2008-01-01

KNF is developing a new design methodology on the Non-LOCA analysis for the licensing purpose. The code chosen is the best-estimate transient analysis code RETRAN and the OPR1000 is aimed as a target plant. For this purpose, KNF prepared a simple nodal scheme appropriate to the licensing analyses and developed the designer-friendly analysis tool ASSIST (Automatic Steady-State Initialization and Safety analysis Tool). To check the validity of the newly developed methodology, the single CEA withdrawal and the locked rotor accidents are analyzed by using a new methodology and are compared with current design results. Comparison results show a good agreement and it is concluded that the new design methodology can be applied to the licensing calculations for OPR1000 Non-LOCA

Using a Realist Research Methodology in Policy Analysis

Science.gov (United States)

Lourie, Megan; Rata, Elizabeth

2017-01-01

The article describes the usefulness of a realist methodology in linking sociological theory to empirically obtained data through the development of a methodological device. Three layers of analysis were integrated: 1. the findings from a case study about Maori language education in New Zealand; 2. the identification and analysis of contradictions…

Methodology for Validating Building Energy Analysis Simulations

Energy Technology Data Exchange (ETDEWEB)

Judkoff, R.; Wortman, D.; O' Doherty, B.; Burch, J.

2008-04-01

The objective of this report was to develop a validation methodology for building energy analysis simulations, collect high-quality, unambiguous empirical data for validation, and apply the validation methodology to the DOE-2.1, BLAST-2MRT, BLAST-3.0, DEROB-3, DEROB-4, and SUNCAT 2.4 computer programs. This report covers background information, literature survey, validation methodology, comparative studies, analytical verification, empirical validation, comparative evaluation of codes, and conclusions.

Disposal Criticality Analysis Methodology Topical Report

International Nuclear Information System (INIS)

Horton, D.G.

1998-01-01

The fundamental objective of this topical report is to present the planned risk-informed disposal criticality analysis methodology to the NRC to seek acceptance that the principles of the methodology and the planned approach to validating the methodology are sound. The design parameters and environmental assumptions within which the waste forms will reside are currently not fully established and will vary with the detailed waste package design, engineered barrier design, repository design, and repository layout. Therefore, it is not practical to present the full validation of the methodology in this report, though a limited validation over a parameter range potentially applicable to the repository is presented for approval. If the NRC accepts the methodology as described in this section, the methodology will be fully validated for repository design applications to which it will be applied in the License Application and its references. For certain fuel types (e.g., intact naval fuel), a ny processes, criteria, codes or methods different from the ones presented in this report will be described in separate addenda. These addenda will employ the principles of the methodology described in this report as a foundation. Departures from the specifics of the methodology presented in this report will be described in the addenda

Disposal Criticality Analysis Methodology Topical Report

International Nuclear Information System (INIS)

D.G. Horton

1998-01-01

The fundamental objective of this topical report is to present the planned risk-informed disposal criticality analysis methodology to the NRC to seek acceptance that the principles of the methodology and the planned approach to validating the methodology are sound. The design parameters and environmental assumptions within which the waste forms will reside are currently not fully established and will vary with the detailed waste package design, engineered barrier design, repository design, and repository layout. Therefore, it is not practical to present the full validation of the methodology in this report, though a limited validation over a parameter range potentially applicable to the repository is presented for approval. If the NRC accepts the methodology as described in this section, the methodology will be fully validated for repository design applications to which it will be applied in the License Application and its references. For certain fuel types (e.g., intact naval fuel), any processes, criteria, codes or methods different from the ones presented in this report will be described in separate addenda. These addenda will employ the principles of the methodology described in this report as a foundation. Departures from the specifics of the methodology presented in this report will be described in the addenda

A study on the operator's errors of commission (EOC) in accident scenarios of nuclear power plants: methodology development and application

International Nuclear Information System (INIS)

Kim, Jae Whan; Jung, Won Dea; Park, Jin Kyun; Kang, Da Il

2003-04-01

As the concern on the operator's inappropriate interventions, the so-called Errors Of Commission (EOCs), that can exacerbate the plant safety has been raised, much of interest in the identification and analysis of EOC events from the risk assessment perspective has been increased. Also, one of the items in need of improvement for the conventional PSA and HRA that consider only the system-demanding human actions is the inclusion of the operator's EOC events into the PSA model. In this study, we propose a methodology for identifying and analysing human errors of commission that might be occurring from the failures in situation assessment and decision making during accident progressions given an initiating event. In order to achieve this goal, the following research items have been performed: Firstly, we analysed the error causes or situations contributed to the occurrence of EOCs in several incidents/accidents of nuclear power plants. Secondly, limitations of the advanced HRAs in treating EOCs were reviewed, and a requirement for a new methodology for analysing EOCs was established. Thirdly, based on these accomplishments a methodology for identifying and analysing EOC events inducible from the failures in situation assessment and decision making was proposed and applied to all the accident sequences of YGN 3 and 4 NPP which resulted in the identification of about 10 EOC situations

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.

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)

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-11-01

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

CONTENT ANALYSIS, DISCOURSE ANALYSIS, AND CONVERSATION ANALYSIS: PRELIMINARY STUDY ON CONCEPTUAL AND THEORETICAL METHODOLOGICAL DIFFERENCES

Directory of Open Access Journals (Sweden)

Anderson Tiago Peixoto Gonçalves

2016-08-01

Full Text Available This theoretical essay aims to reflect on three models of text interpretation used in qualitative research, which is often confused in its concepts and methodologies (Content Analysis, Discourse Analysis, and Conversation Analysis. After the presentation of the concepts, the essay proposes a preliminary discussion on conceptual and theoretical methodological differences perceived between them. A review of the literature was performed to support the conceptual and theoretical methodological discussion. It could be verified that the models have differences related to the type of strategy used in the treatment of texts, the type of approach, and the appropriate theoretical position.

Probabilistic methodology for turbine missile risk analysis

International Nuclear Information System (INIS)

Twisdale, L.A.; Dunn, W.L.; Frank, R.A.

1984-01-01

A methodology has been developed for estimation of the probabilities of turbine-generated missile damage to nuclear power plant structures and systems. Mathematical models of the missile generation, transport, and impact events have been developed and sequenced to form an integrated turbine missile simulation methodology. Probabilistic Monte Carlo techniques are used to estimate the plant impact and damage probabilities. The methodology has been coded in the TURMIS computer code to facilitate numerical analysis and plant-specific turbine missile probability assessments. Sensitivity analyses have been performed on both the individual models and the integrated methodology, and probabilities have been estimated for a hypothetical nuclear power plant case study. (orig.)

A study on the operator's errors of commission (EOC) in accident scenarios of nuclear power plants: methodology development and application

Energy Technology Data Exchange (ETDEWEB)

Kim, Jae Whan; Jung, Won Dea; Park, Jin Kyun; Kang, Da Il

2003-04-01

As the concern on the operator's inappropriate interventions, the so-called Errors Of Commission (EOCs), that can exacerbate the plant safety has been raised, much of interest in the identification and analysis of EOC events from the risk assessment perspective has been increased. Also, one of the items in need of improvement for the conventional PSA and HRA that consider only the system-demanding human actions is the inclusion of the operator's EOC events into the PSA model. In this study, we propose a methodology for identifying and analysing human errors of commission that might be occurring from the failures in situation assessment and decision making during accident progressions given an initiating event. In order to achieve this goal, the following research items have been performed: Firstly, we analysed the error causes or situations contributed to the occurrence of EOCs in several incidents/accidents of nuclear power plants. Secondly, limitations of the advanced HRAs in treating EOCs were reviewed, and a requirement for a new methodology for analysing EOCs was established. Thirdly, based on these accomplishments a methodology for identifying and analysing EOC events inducible from the failures in situation assessment and decision making was proposed and applied to all the accident sequences of YGN 3 and 4 NPP which resulted in the identification of about 10 EOC situations.

Improved Methodology of MSLB M/E Release Analysis for OPR1000

International Nuclear Information System (INIS)

Park, Seok Jeong; Kim, Cheol Woo; Seo, Jong Tae

2006-01-01

A new mass and energy (M/E) release analysis methodology for the equipment environmental qualification (EEQ) on loss-of-coolant accident (LOCA) has been recently developed and adopted on small break LOCA EEQ. The new methodology for the M/E release analysis is extended to the M/E release analysis for the containment design for large break LOCA and the main steam line break (MSLB) accident, and named KIMERA (KOPEC Improved Mass and Energy Release Analysis) methodology. The computer code systems used in this methodology is RELAP5K/CONTEMPT4 (or RELAP5-ME) which couples RELAP5/MOD3.1/K with enhanced M/E model and LOCA long term model, and CONTEMPT4/ MOD5. This KIMERA methodology is applied to the MSLB M/E release analysis to evaluate the validation of KIMERA methodology for MSLB in containment design. The results are compared with the OPR 1000 FSAR

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

Formation of the methodological matrix of the strategic analysis of the enterprise

Directory of Open Access Journals (Sweden)

N.H. Vygovskaya

2018-04-01

Full Text Available The article is devoted to the study of the methodological matrix of the strategic analysis of the enterprise. The aim of this article is to analyze the influence of methodological changes in the 20th century on the methodology of strategic analysis; critical assessment and generalization of scientific approaches to its methods. Evaluation of scientific works on analysis made it possible to identify such problems in the methodology of strategic analysis as the lack of consideration of the features of strategic analysis in the formation of its methods, which often leads to confusion of methods of financial (economic, thrifty analysis; failure to use the fact that the strategic analysis contains, besides the methods of analyzing the internal and external environment, the methods of forecast analysis aimed at forming the strategy for the development of the enterprise; identification of the concepts «image», «reception», «method» of analysis; multidirectionality and indistinctness of signs of classification of methods of strategic analysis; blind copying of foreign methods of application of techniques and methods of strategic analysis without taking into account the specifics of domestic economic conditions. The expediency of using the system approach in forming the methodological design of strategic analysis is proved, which will allow to combine the methodology as a science of methods (a broad approach to the methods of strategic analysis with methodology as a set of applied methods and methods of analysis (narrow approach to methodology. The use of the system approach allowed to distinguish three levels of the methodology of strategic analysis. The first and second levels of methodology correspond to the level of science, the third level – the practice. When developing the third level of special methods of strategic analysis, an approach is applied that differentiates them depending on the stages of strategic analysis (methods of the stage

Taipower's transient analysis methodology for pressurized water reactors

International Nuclear Information System (INIS)

Huang, Pinghue

1998-01-01

The methodology presented in this paper is a part of the 'Taipower's Reload Design and Transient Analysis Methodologies for Light Water Reactors' developed by the Taiwan Power Company (TPC) and the Institute of Nuclear Energy Research. This methodology utilizes four computer codes developed or sponsored by Electric Power Research institute: system transient analysis code RETRAN-02, core thermal-hydraulic analysis code COBRAIIIC, three-dimensional spatial kinetics code ARROTTA, and fuel rod evaluation code FREY. Each of the computer codes was extensively validated. Analysis methods and modeling techniques were conservatively established for each application using a systematic evaluation with the assistance of sensitivity studies. The qualification results and analysis methods were documented in detail in TPC topical reports. The topical reports for COBRAIIIC, ARROTTA. and FREY have been reviewed and approved by the Atomic Energy Council (ABC). TPC 's in-house transient methodology have been successfully applied to provide valuable support for many operational issues and plant improvements for TPC's Maanshan Units I and 2. Major applications include the removal of the resistance temperature detector bypass system, the relaxation of the hot-full-power moderator temperature coefficient design criteria imposed by the ROCAEC due to a concern on Anticipated Transient Without Scram, the reduction of boron injection tank concentration and the elimination of the heat tracing, and the reduction of' reactor coolant system flow. (author)

Compliance strategy for statistically based neutron overpower protection safety analysis methodology

International Nuclear Information System (INIS)

Holliday, E.; Phan, B.; Nainer, O.

2009-01-01

The methodology employed in the safety analysis of the slow Loss of Regulation (LOR) event in the OPG and Bruce Power CANDU reactors, referred to as Neutron Overpower Protection (NOP) analysis, is a statistically based methodology. Further enhancement to this methodology includes the use of Extreme Value Statistics (EVS) for the explicit treatment of aleatory and epistemic uncertainties, and probabilistic weighting of the initial core states. A key aspect of this enhanced NOP methodology is to demonstrate adherence, or compliance, with the analysis basis. This paper outlines a compliance strategy capable of accounting for the statistical nature of the enhanced NOP methodology. (author)

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.

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

A Global Sensitivity Analysis Methodology for Multi-physics Applications

Energy Technology Data Exchange (ETDEWEB)

Tong, C H; Graziani, F R

2007-02-02

Experiments are conducted to draw inferences about an entire ensemble based on a selected number of observations. This applies to both physical experiments as well as computer experiments, the latter of which are performed by running the simulation models at different input configurations and analyzing the output responses. Computer experiments are instrumental in enabling model analyses such as uncertainty quantification and sensitivity analysis. This report focuses on a global sensitivity analysis methodology that relies on a divide-and-conquer strategy and uses intelligent computer experiments. The objective is to assess qualitatively and/or quantitatively how the variabilities of simulation output responses can be accounted for by input variabilities. We address global sensitivity analysis in three aspects: methodology, sampling/analysis strategies, and an implementation framework. The methodology consists of three major steps: (1) construct credible input ranges; (2) perform a parameter screening study; and (3) perform a quantitative sensitivity analysis on a reduced set of parameters. Once identified, research effort should be directed to the most sensitive parameters to reduce their uncertainty bounds. This process is repeated with tightened uncertainty bounds for the sensitive parameters until the output uncertainties become acceptable. To accommodate the needs of multi-physics application, this methodology should be recursively applied to individual physics modules. The methodology is also distinguished by an efficient technique for computing parameter interactions. Details for each step will be given using simple examples. Numerical results on large scale multi-physics applications will be available in another report. Computational techniques targeted for this methodology have been implemented in a software package called PSUADE.

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.

Simplified methodology for Angra 1 containment analysis

International Nuclear Information System (INIS)

Neves Conti, T. das; Souza, A.L. de; Sabundjian, G.

1991-08-01

A simplified methodology of analysis was developed to simulate a Large Break Loss of Coolant Accident in the Angra 1 Nuclear Power Station. Using the RELAP5/MOD1, RELAP4/MOD5 and CONTEMPT-LT Codes, the time variation of pressure and temperature in the containment was analysed. The obtained data was compared with the Angra 1 Final Safety Analysis Report, and too those calculated by a Detailed Model. The results obtained by this new methodology such as the small computational time of simulation, were satisfactory when getting the preliminary evaluation of the Angra 1 global parameters. (author)

The methodology of semantic analysis for extracting physical effects

Science.gov (United States)

Fomenkova, M. A.; Kamaev, V. A.; Korobkin, D. M.; Fomenkov, S. A.

2017-01-01

The paper represents new methodology of semantic analysis for physical effects extracting. This methodology is based on the Tuzov ontology that formally describes the Russian language. In this paper, semantic patterns were described to extract structural physical information in the form of physical effects. A new algorithm of text analysis was described.

Development of seismic risk analysis methodologies at JAERI

International Nuclear Information System (INIS)

Tanaka, T.; Abe, K.; Ebisawa, K.; Oikawa, T.

1988-01-01

The usefulness of probabilistic safety assessment (PSA) is recognized worldwidely for balanced design and regulation of nuclear power plants. In Japan, the Japan Atomic Energy Research Institute (JAERI) has been engaged in developing methodologies necessary for carrying out PSA. The research and development program was started in 1980. In those days the effort was only for internal initiator PSA. In 1985 the program was expanded so as to include external event analysis. Although this expanded program is to cover various external initiators, the current effort is dedicated for seismic risk analysis. There are three levels of seismic PSA, similarly to internal initiator PSA: Level 1: Evaluation of core damage frequency, Level 2: Evaluation of radioactive release frequency and source terms, and Level 3: Evaluation of environmental consequence. In the JAERI's program, only the methodologies for level 1 seismic PSA are under development. The methodology development for seismic risk analysis is divided into two phases. The Phase I study is to establish a whole set of simple methodologies based on currently available data. In the Phase II, Sensitivity study will be carried out to identify the parameters whose uncertainty may result in lage uncertainty in seismic risk, and For such parameters, the methodology will be upgraded. Now the Phase I study has almost been completed. In this report, outlines of the study and some of its outcomes are described

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

Directory of Open Access Journals (Sweden)

Aiman Tulaimat

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

Disposal criticality analysis methodology for fissile waste forms

International Nuclear Information System (INIS)

Davis, J.W.; Gottlieb, P.

1998-03-01

A general methodology has been developed to evaluate the criticality potential of the wide range of waste forms planned for geologic disposal. The range of waste forms include commercial spent fuel, high level waste, DOE spent fuel (including highly enriched), MOX using weapons grade plutonium, and immobilized plutonium. The disposal of these waste forms will be in a container with sufficiently thick corrosion resistant barriers to prevent water penetration for up to 10,000 years. The criticality control for DOE spent fuel is primarily provided by neutron absorber material incorporated into the basket holding the individual assemblies. For the immobilized plutonium, the neutron absorber material is incorporated into the waste form itself. The disposal criticality analysis methodology includes the analysis of geochemical and physical processes that can breach the waste package and affect the waste forms within. The basic purpose of the methodology is to guide the criticality control features of the waste package design, and to demonstrate that the final design meets the criticality control licensing requirements. The methodology can also be extended to the analysis of criticality consequences (primarily increased radionuclide inventory), which will support the total performance assessment for the respository

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.

Go-flow: a reliability analysis methodology applicable to piping system

International Nuclear Information System (INIS)

Matsuoka, T.; Kobayashi, M.

1985-01-01

Since the completion of the Reactor Safety Study, the use of probabilistic risk assessment technique has been becoming more widespread in the nuclear community. Several analytical methods are used for the reliability analysis of nuclear power plants. The GO methodology is one of these methods. Using the GO methodology, the authors performed a reliability analysis of the emergency decay heat removal system of the nuclear ship Mutsu, in order to examine its applicability to piping systems. By this analysis, the authors have found out some disadvantages of the GO methodology. In the GO methodology, the signal is on-to-off or off-to-on signal, therefore the GO finds out the time point at which the state of a system changes, and can not treat a system which state changes as off-on-off. Several computer runs are required to obtain the time dependent failure probability of a system. In order to overcome these disadvantages, the authors propose a new analytical methodology: GO-FLOW. In GO-FLOW, the modeling method (chart) and the calculation procedure are similar to those in the GO methodology, but the meaning of signal and time point, and the definitions of operators are essentially different. In the paper, the GO-FLOW methodology is explained and two examples of the analysis by GO-FLOW are given

A Review of Citation Analysis Methodologies for Collection Management

Science.gov (United States)

Hoffmann, Kristin; Doucette, Lise

2012-01-01

While there is a considerable body of literature that presents the results of citation analysis studies, most researchers do not provide enough detail in their methodology to reproduce the study, nor do they provide rationale for methodological decisions. In this paper, we review the methodologies used in 34 recent articles that present a…

Methodology of Credit Analysis Development

Directory of Open Access Journals (Sweden)

Slađana Neogradi

2017-12-01

Full Text Available The subject of research presented in this paper refers to the definition of methodology for the development of credit analysis in companies and its application in lending operations in the Republic of Serbia. With the developing credit market, there is a growing need for a well-developed risk and loss prevention system. In the introduction the process of bank analysis of the loan applicant is presented in order to minimize and manage the credit risk. By examining the subject matter, the process of processing the credit application is described, the procedure of analyzing the financial statements in order to get an insight into the borrower's creditworthiness. In the second part of the paper, the theoretical and methodological framework is presented applied in the concrete company. In the third part, models are presented which banks should use to protect against exposure to risks, i.e. their goal is to reduce losses on loan operations in our country, as well as to adjust to market conditions in an optimal way.

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-03-15

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

A Research Roadmap for Computation-Based Human Reliability Analysis

Energy Technology Data Exchange (ETDEWEB)

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

2015-08-01

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

A Research Roadmap for Computation-Based Human Reliability Analysis

International Nuclear Information System (INIS)

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

2015-01-01

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

The analysis of RWAP(Rod Withdrawal at Power) using the KEPRI methodology

International Nuclear Information System (INIS)

Yang, C. K.; Kim, Y. H.

2001-01-01

KEPRI developed new methodology which was based on RASP(Reactor Analysis Support Package). In this paper, The analysis of RWAP(Rod Withdrawal at Power) accident which can result in reactivity and power distribution anomaly was performed using the KEPRI methodology. The calculation describes RWAP transient and documents the analysis, including the computer code modeling assumptions and input parameters used in the analysis. To validity for the new methodology, the result of calculation was compared with FSAR. As compared with FSAR, result of the calculation using the KEPRI Methodology is similar to FSAR's. And result of the sensitivity of postulated parameters were similar to the existing methodology

Comparative analysis of proliferation resistance assessment methodologies

International Nuclear Information System (INIS)

Takaki, Naoyuki; Kikuchi, Masahiro; Inoue, Naoko; Osabe, Takeshi

2005-01-01

Comparative analysis of the methodologies was performed based on the discussions in the international workshop on 'Assessment Methodology of Proliferation Resistance for Future Nuclear Energy Systems' held in Tokyo, on March 2005. Through the workshop and succeeding considerations, it is clarified that the proliferation resistance assessment methodologies are affected by the broader nuclear options being pursued and also by the political situations of the state. Even the definition of proliferation resistance, despite the commonality of fundamental issues, derives from perceived threat and implementation circumstances inherent to the larger programs. Deep recognitions of the 'difference' among communities would help us to make further essential and progressed discussion with harmonization. (author)

Safety analysis and evaluation methodology for fusion systems

International Nuclear Information System (INIS)

Fujii-e, Y.; Kozawa, Y.; Namba, C.

1987-03-01

Fusion systems which are under development as future energy systems have reached a stage that the break even is expected to be realized in the near future. It is desirable to demonstrate that fusion systems are well acceptable to the societal environment. There are three crucial viewpoints to measure the acceptability, that is, technological feasibility, economy and safety. These three points have close interrelation. The safety problem is more important since three large scale tokamaks, JET, TFTR and JT-60, start experiment, and tritium will be introduced into some of them as the fusion fuel. It is desirable to establish a methodology to resolve the safety-related issues in harmony with the technological evolution. The promising fusion system toward reactors is not yet settled. This study has the objective to develop and adequate methodology which promotes the safety design of general fusion systems and to present a basis for proposing the R and D themes and establishing the data base. A framework of the methodology, the understanding and modeling of fusion systems, the principle of ensuring safety, the safety analysis based on the function and the application of the methodology are discussed. As the result of this study, the methodology for the safety analysis and evaluation of fusion systems was developed. New idea and approach were presented in the course of the methodology development. (Kako, I.)

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.

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

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

Nondestructive assay methodologies in nuclear forensics analysis

International Nuclear Information System (INIS)

Tomar, B.S.

2016-01-01

In the present chapter, the nondestructive assay (NDA) methodologies used for analysis of nuclear materials as a part of nuclear forensic investigation have been described. These NDA methodologies are based on (i) measurement of passive gamma and neutrons emitted by the radioisotopes present in the nuclear materials, (ii) measurement of gamma rays and neutrons emitted after the active interrogation of the nuclear materials with a source of X-rays, gamma rays or neutrons

Methodology for Mode Selection in Corridor Analysis of Freight Transportation

OpenAIRE

Kanafani, Adib

1984-01-01

The purpose of tins report is to outline a methodology for the analysis of mode selection in freight transportation. This methodology is intended to partake of transportation corridor analysts, a component of demand analysis that is part of a national transportation process. The methodological framework presented here provides a basis on which specific models and calculation procedures might be developed. It also provides a basis for the development of a data management system suitable for co...

Severe accident analysis methodology in support of accident management

International Nuclear Information System (INIS)

Boesmans, B.; Auglaire, M.; Snoeck, J.

1997-01-01

The author addresses the implementation at BELGATOM of a generic severe accident analysis methodology, which is intended to support strategic decisions and to provide quantitative information in support of severe accident management. The analysis methodology is based on a combination of severe accident code calculations, generic phenomenological information (experimental evidence from various test facilities regarding issues beyond present code capabilities) and detailed plant-specific technical information

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-08-15

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Supplement to the Disposal Criticality Analysis Methodology

International Nuclear Information System (INIS)

Thomas, D.A.

1999-01-01

The methodology for evaluating criticality potential for high-level radioactive waste and spent nuclear fuel after the repository is sealed and permanently closed is described in the Disposal Criticality Analysis Methodology Topical Report (DOE 1998b). The topical report provides a process for validating various models that are contained in the methodology and states that validation will be performed to support License Application. The Supplement to the Disposal Criticality Analysis Methodology provides a summary of data and analyses that will be used for validating these models and will be included in the model validation reports. The supplement also summarizes the process that will be followed in developing the model validation reports. These reports will satisfy commitments made in the topical report, and thus support the use of the methodology for Site Recommendation and License Application. It is concluded that this report meets the objective of presenting additional information along with references that support the methodology presented in the topical report and can be used both in validation reports and in answering request for additional information received from the Nuclear Regulatory Commission concerning the topical report. The data and analyses summarized in this report and presented in the references are not sufficient to complete a validation report. However, this information will provide a basis for several of the validation reports. Data from several references in this report have been identified with TBV-1349. Release of the TBV governing this data is required prior to its use in quality affecting activities and for use in analyses affecting procurement, construction, or fabrication. Subsequent to the initiation of TBV-1349, DOE issued a concurrence letter (Mellington 1999) approving the request to identify information taken from the references specified in Section 1.4 as accepted data

Risk analysis methodologies for the transportation of radioactive materials

International Nuclear Information System (INIS)

Geffen, C.A.

1983-05-01

Different methodologies have evolved for consideration of each of the many steps required in performing a transportation risk analysis. Although there are techniques that attempt to consider the entire scope of the analysis in depth, most applications of risk assessment to the transportation of nuclear fuel cycle materials develop specific methodologies for only one or two parts of the analysis. The remaining steps are simplified for the analyst by narrowing the scope of the effort (such as evaluating risks for only one material, or a particular set of accident scenarios, or movement over a specific route); performing a qualitative rather than a quantitative analysis (probabilities may be simply ranked as high, medium or low, for instance); or assuming some generic, conservative conditions for potential release fractions and consequences. This paper presents a discussion of the history and present state-of-the-art of transportation risk analysis methodologies. Many reports in this area were reviewed as background for this presentation. The literature review, while not exhaustive, did result in a complete representation of the major methods used today in transportation risk analysis. These methodologies primarily include the use of severity categories based on historical accident data, the analysis of specifically assumed accident sequences for the transportation activity of interest, and the use of fault or event tree analysis. Although the focus of this work has generally been on potential impacts to public groups, some effort has been expended in the estimation of risks to occupational groups in transportation activities

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)

Analysis of Alternatives for Risk Assessment Methodologies and Tools

Energy Technology Data Exchange (ETDEWEB)

Nachtigal, Noel M. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). System Analytics; Fruetel, Julia A. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Systems Research and Analysis; Gleason, Nathaniel J. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Systems Research and Analysis; Helms, Jovana [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Systems Research and Analysis; Imbro, Dennis Raymond [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Systems Research and Analysis; Sumner, Matthew C. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Systems Research and Analysis

2013-10-01

The purpose of this document is to provide a basic overview and understanding of risk assessment methodologies and tools from the literature and to assess the suitability of these methodologies and tools for cyber risk assessment. Sandia National Laboratories (SNL) performed this review in support of risk modeling activities performed for the Stakeholder Engagement and Cyber Infrastructure Resilience (SECIR) division of the Department of Homeland Security (DHS) Office of Cybersecurity and Communications (CS&C). The set of methodologies and tools covered in this document is not intended to be exhaustive; instead, it focuses on those that are commonly used in the risk assessment community. The classification of methodologies and tools was performed by a group of analysts with experience in risk analysis and cybersecurity, and the resulting analysis of alternatives has been tailored to address the needs of a cyber risk assessment.

Disposal criticality analysis methodology's principal isotope burnup credit

International Nuclear Information System (INIS)

Doering, T.W.; Thomas, D.A.

2001-01-01

This paper presents the burnup credit aspects of the United States Department of Energy Yucca Mountain Project's methodology for performing criticality analyses for commercial light-water-reactor fuel. The disposal burnup credit methodology uses a 'principal isotope' model, which takes credit for the reduced reactivity associated with the build-up of the primary principal actinides and fission products in irradiated fuel. Burnup credit is important to the disposal criticality analysis methodology and to the design of commercial fuel waste packages. The burnup credit methodology developed for disposal of irradiated commercial nuclear fuel can also be applied to storage and transportation of irradiated commercial nuclear fuel. For all applications a series of loading curves are developed using a best estimate methodology and depending on the application, an additional administrative safety margin may be applied. The burnup credit methodology better represents the 'true' reactivity of the irradiated fuel configuration, and hence the real safety margin, than do evaluations using the 'fresh fuel' assumption. (author)

Two methodologies for optical analysis of contaminated engine lubricants

International Nuclear Information System (INIS)

Aghayan, Hamid; Yang, Jun; Bordatchev, Evgueni

2012-01-01

The performance, efficiency and lifetime of modern combustion engines significantly depend on the quality of the engine lubricants. However, contaminants, such as gasoline, moisture, coolant and wear particles, reduce the life of engine mechanical components and lubricant quality. Therefore, direct and indirect measurements of engine lubricant properties, such as physical-mechanical, electro-magnetic, chemical and optical properties, are intensively utilized in engine condition monitoring systems and sensors developed within the last decade. Such sensors for the measurement of engine lubricant properties can be used to detect a functional limit of the in-use lubricant, increase drain interval and reduce the environmental impact. This paper proposes two new methodologies for the quantitative and qualitative analysis of the presence of contaminants in the engine lubricants. The methodologies are based on optical analysis of the distortion effect when an object image is obtained through a thin random optical medium (e.g. engine lubricant). The novelty of the proposed methodologies is in the introduction of an object with a known periodic shape behind a thin film of the contaminated lubricant. In this case, an acquired image represents a combined lubricant–object optical appearance, where an a priori known periodic structure of the object is distorted by a contaminated lubricant. In the object shape-based optical analysis, several parameters of an acquired optical image, such as the gray scale intensity of lubricant and object, shape width at object and lubricant levels, object relative intensity and width non-uniformity coefficient are newly proposed. Variations in the contaminant concentration and use of different contaminants lead to the changes of these parameters measured on-line. In the statistical optical analysis methodology, statistical auto- and cross-characteristics (e.g. auto- and cross-correlation functions, auto- and cross-spectrums, transfer function

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)

Update of Part 61 impacts analysis methodology

International Nuclear Information System (INIS)

Oztunali, O.I.; Roles, G.W.

1986-01-01

The US Nuclear Regulatory Commission is expanding the impacts analysis methodology used during the development of the 10 CFR Part 61 rule to allow improved consideration of costs and impacts of disposal of waste that exceeds Class C concentrations. The project includes updating the computer codes that comprise the methodology, reviewing and updating data assumptions on waste streams and disposal technologies, and calculation of costs for small as well as large disposal facilities. This paper outlines work done to date on this project

Update of Part 61 impacts analysis methodology

International Nuclear Information System (INIS)

Oztunali, O.I.; Roles, G.W.; US Nuclear Regulatory Commission, Washington, DC 20555)

1985-01-01

The US Nuclear Regulatory Commission is expanding the impacts analysis methodology used during the development of the 10 CFR Part 61 regulation to allow improved consideration of costs and impacts of disposal of waste that exceeds Class C concentrations. The project includes updating the computer codes that comprise the methodology, reviewing and updating data assumptions on waste streams and disposal technologies, and calculation of costs for small as well as large disposal facilities. This paper outlines work done to date on this project

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.

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

Internal fire analysis screening methodology for the Salem Nuclear Generating Station

International Nuclear Information System (INIS)

Eide, S.; Bertucio, R.; Quilici, M.; Bearden, R.

1989-01-01

This paper reports on an internal fire analysis screening methodology that has been utilized for the Salem Nuclear Generating Station (SNGS) Probabilistic Risk Assessment (PRA). The methodology was first developed and applied in the Brunswick Steam Electric Plant (BSEP) PRA. The SNGS application includes several improvements and extensions to the original methodology. The SNGS approach differs significantly from traditional fire analysis methodologies by providing a much more detailed treatment of transient combustibles. This level of detail results in a model which is more usable for assisting in the management of fire risk at the plant

Simplified methodology for analysis of Angra-1 containing

International Nuclear Information System (INIS)

Neves Conti, T. das; Souza, A.L. de; Sabundjian, G.

1988-01-01

A simplified methodology of analysis was developed to simulate a Large Break Loss of Coolant Accident in the Angra 1 Nuclear Power Station. Using the RELAP5/MOD1, RELAP4/MOD5 and CONTEMPT-LT Codes, the time the variation of pressure and temperature in the containment was analysed. The obtained data was compared with the Angra 1 Final Safety Analysis Report, and too those calculated by a Detailed Model. The results obtained by this new methodology such as the small computational time of simulation, were satisfactory when getting the preliminar avaliation of the Angra 1 global parameters. (author) [pt

The development of a safety analysis methodology for the optimized power reactor 1000

International Nuclear Information System (INIS)

Hwang-Yong, Jun; Yo-Han, Kim

2005-01-01

Korea Electric Power Research Institute (KEPRI) has been developing inhouse safety analysis methodology based on the delicate codes available to KEPRI to overcome the problems arising from currently used vendor oriented methodologies. For the Loss of Coolant Accident (LOCA) analysis, the KREM (KEPRI Realistic Evaluation Methodology) has been developed based on the RELAP-5 code. The methodology was approved for the Westinghouse 3-loop plants by the Korean regulatory organization and the project to extent the methodology to the Optimized Power Reactor 1000 (OPR1000) has been ongoing since 2001. Also, for the Non-LOCA analysis, the KNAP (Korea Non-LOCA Analysis Package) has been developed using the UNICORN-TM code system. To demonstrate the feasibility of these codes systems and methodologies, some typical cases of the design basis accidents mentioned in the final safety analysis report (FSAR) were analyzed. (author)

A study on safety analysis methodology in spent fuel dry storage facility

Energy Technology Data Exchange (ETDEWEB)

Che, M. S.; Ryu, J. H.; Kang, K. M.; Cho, N. C.; Kim, M. S. [Hanyang Univ., Seoul (Korea, Republic of)

2004-02-15

Collection and review of the domestic and foreign technology related to spent fuel dry storage facility. Analysis of a reference system. Establishment of a framework for criticality safety analysis. Review of accident analysis methodology. Establishment of accident scenarios. Establishment of scenario analysis methodology.

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

International Nuclear Information System (INIS)

Cepis, M.

2007-01-01

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

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

International Nuclear Information System (INIS)

Cepin, M.

2008-01-01

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

Methodology for flood risk analysis for nuclear power plants

International Nuclear Information System (INIS)

Wagner, D.P.; Casada, M.L.; Fussell, J.B.

1984-01-01

The methodology for flood risk analysis described here addresses the effects of a flood on nuclear power plant safety systems. Combining the results of this method with the probability of a flood allows the effects of flooding to be included in a probabilistic risk assessment. The five-step methodology includes accident sequence screening to focus the detailed analysis efforts on the accident sequences that are significantly affected by a flood event. The quantitative results include the flood's contribution to system failure probability, accident sequence occurrence frequency and consequence category occurrence frequency. The analysis can be added to existing risk assessments without a significant loss in efficiency. The results of two example applications show the usefulness of the methodology. Both examples rely on the Reactor Safety Study for the required risk assessment inputs and present changes in the Reactor Safety Study results as a function of flood probability

Clean Energy Manufacturing Analysis Center Benchmark Report: Framework and Methodologies

Energy Technology Data Exchange (ETDEWEB)

Sandor, Debra [National Renewable Energy Lab. (NREL), Golden, CO (United States); Chung, Donald [National Renewable Energy Lab. (NREL), Golden, CO (United States); Keyser, David [National Renewable Energy Lab. (NREL), Golden, CO (United States); Mann, Margaret [National Renewable Energy Lab. (NREL), Golden, CO (United States); Engel-Cox, Jill [National Renewable Energy Lab. (NREL), Golden, CO (United States)

2017-05-23

This report documents the CEMAC methodologies for developing and reporting annual global clean energy manufacturing benchmarks. The report reviews previously published manufacturing benchmark reports and foundational data, establishes a framework for benchmarking clean energy technologies, describes the CEMAC benchmark analysis methodologies, and describes the application of the methodologies to the manufacturing of four specific clean energy technologies.

Scenario aggregation and analysis via Mean-Shift Methodology

International Nuclear Information System (INIS)

Mandelli, D.; Yilmaz, A.; Metzroth, K.; Aldemir, T.; Denning, R.

2010-01-01

A new generation of dynamic methodologies is being developed for nuclear reactor probabilistic risk assessment (PRA) which explicitly account for the time element in modeling the probabilistic system evolution and use numerical simulation tools to account for possible dependencies between failure events. The dynamic event tree (DET) approach is one of these methodologies. One challenge with dynamic PRA methodologies is the large amount of data they produce which may be difficult to analyze without appropriate software tools. The concept of 'data mining' is well known in the computer science community and several methodologies have been developed in order to extract useful information from a dataset with a large number of records. Using the dataset generated by the DET analysis of the reactor vessel auxiliary cooling system (RVACS) of an ABR-1000 for an aircraft crash recovery scenario and the Mean-Shift Methodology for data mining, it is shown how clusters of transients with common characteristics can be identified and classified. (authors)

Proceedings of the twenty-fourth water reactor safety information meeting. Volume 3: PRA and HRA; Probabilistic seismic hazard assessment and seismic siting criteria

Energy Technology Data Exchange (ETDEWEB)

Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

1997-02-01

This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: PRA and HRA and probabilistic seismic hazard assessment and seismic siting criteria. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

Proceedings of the twenty-fourth water reactor safety information meeting. Volume 3: PRA and HRA; Probabilistic seismic hazard assessment and seismic siting criteria

International Nuclear Information System (INIS)

Monteleone, S.

1997-02-01

This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: PRA and HRA and probabilistic seismic hazard assessment and seismic siting criteria. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

GO-FLOW methodology. Basic concept and integrated analysis framework for its applications

International Nuclear Information System (INIS)

Matsuoka, Takeshi

2010-01-01

GO-FLOW methodology is a success oriented system analysis technique, and is capable of evaluating a large system with complex operational sequences. Recently an integrated analysis framework of the GO-FLOW has been developed for the safety evaluation of elevator systems by the Ministry of Land, Infrastructure, Transport and Tourism, Japanese Government. This paper describes (a) an Overview of the GO-FLOW methodology, (b) Procedure of treating a phased mission problem, (c) Common cause failure analysis, (d) Uncertainty analysis, and (e) Integrated analysis framework. The GO-FLOW methodology is a valuable and useful tool for system reliability analysis and has a wide range of applications. (author)

An Evaluation Methodology for Protocol Analysis Systems

Science.gov (United States)

2007-03-01

Main Memory Requirement NS: Needham-Schroeder NSL: Needham-Schroeder-Lowe OCaml : Objective Caml POSIX: Portable Operating System...methodology is needed. A. PROTOCOL ANALYSIS FIELD As with any field, there is a specialized language used within the protocol analysis community. Figure...ProVerif requires that Objective Caml ( OCaml ) be installed on the system, OCaml version 3.09.3 was installed. C. WINDOWS CONFIGURATION OS

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-03-15

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

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

International Nuclear Information System (INIS)

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

2007-03-01

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

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...

Nuclear methodology development for clinical analysis

International Nuclear Information System (INIS)

Oliveira, Laura Cristina de

2003-01-01

In the present work the viability of using the neutron activation analysis to perform urine and blood clinical analysis was checked. The aim of this study is to investigate the biological behavior of animals that has been fed with chow doped by natural uranium for a long period. Aiming at time and cost reduction, the absolute method was applied to determine element concentration on biological samples. The quantitative results of urine sediment using NAA were compared with the conventional clinical analysis and the results were compatible. This methodology was also used on bone and body organs such as liver and muscles to help the interpretation of possible anomalies. (author)

Failure mode effect analysis and fault tree analysis as a combined methodology in risk management

Science.gov (United States)

Wessiani, N. A.; Yoshio, F.

2018-04-01

There have been many studies reported the implementation of Failure Mode Effect Analysis (FMEA) and Fault Tree Analysis (FTA) as a method in risk management. However, most of the studies usually only choose one of these two methods in their risk management methodology. On the other side, combining these two methods will reduce the drawbacks of each methods when implemented separately. This paper aims to combine the methodology of FMEA and FTA in assessing risk. A case study in the metal company will illustrate how this methodology can be implemented. In the case study, this combined methodology will assess the internal risks that occur in the production process. Further, those internal risks should be mitigated based on their level of risks.

Development and application of a deterministic-realistic hybrid methodology for LOCA licensing analysis

International Nuclear Information System (INIS)

Liang, Thomas K.S.; Chou, Ling-Yao; Zhang, Zhongwei; Hsueh, Hsiang-Yu; Lee, Min

2011-01-01

Highlights: → A new LOCA licensing methodology (DRHM, deterministic-realistic hybrid methodology) was developed. → DRHM involves conservative Appendix K physical models and statistical treatment of plant status uncertainties. → DRHM can generate 50-100 K PCT margin as compared to a traditional Appendix K methodology. - Abstract: It is well recognized that a realistic LOCA analysis with uncertainty quantification can generate greater safety margin as compared with classical conservative LOCA analysis using Appendix K evaluation models. The associated margin can be more than 200 K. To quantify uncertainty in BELOCA analysis, generally there are two kinds of uncertainties required to be identified and quantified, which involve model uncertainties and plant status uncertainties. Particularly, it will take huge effort to systematically quantify individual model uncertainty of a best estimate LOCA code, such as RELAP5 and TRAC. Instead of applying a full ranged BELOCA methodology to cover both model and plant status uncertainties, a deterministic-realistic hybrid methodology (DRHM) was developed to support LOCA licensing analysis. Regarding the DRHM methodology, Appendix K deterministic evaluation models are adopted to ensure model conservatism, while CSAU methodology is applied to quantify the effect of plant status uncertainty on PCT calculation. Generally, DRHM methodology can generate about 80-100 K margin on PCT as compared to Appendix K bounding state LOCA analysis.

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

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

Energy Technology Data Exchange (ETDEWEB)

Swain, A D; Guttmann, H E

1983-08-01

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

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

International Nuclear Information System (INIS)

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

1983-08-01

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

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

Establishing Equivalence: Methodological Progress in Group-Matching Design and Analysis

Science.gov (United States)

Kover, Sara T.; Atwood, Amy K.

2013-01-01

This methodological review draws attention to the challenges faced by intellectual and developmental disabilities researchers in the appropriate design and analysis of group comparison studies. We provide a brief overview of matching methodologies in the field, emphasizing group-matching designs used in behavioral research on cognition and…

LOFT uncertainty-analysis methodology

International Nuclear Information System (INIS)

Lassahn, G.D.

1983-01-01

The methodology used for uncertainty analyses of measurements in the Loss-of-Fluid Test (LOFT) nuclear-reactor-safety research program is described and compared with other methodologies established for performing uncertainty analyses

LOFT uncertainty-analysis methodology

International Nuclear Information System (INIS)

Lassahn, G.D.

1983-01-01

The methodology used for uncertainty analyses of measurements in the Loss-of-Fluid Test (LOFT) nuclear reactor safety research program is described and compared with other methodologies established for performing uncertainty analyses

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.)

Proposed methodology for completion of scenario analysis for the Basalt Waste Isolation Project

International Nuclear Information System (INIS)

Roberds, W.J.; Plum, R.J.; Visca, P.J.

1984-11-01

This report presents the methodology to complete an assessment of postclosure performance, considering all credible scenarios, including the nominal case, for a proposed repository for high-level nuclear waste at the Hanford Site, Washington State. The methodology consists of defensible techniques for identifying and screening scenarios, and for then assessing the risks associated with each. The results of the scenario analysis are used to comprehensively determine system performance and/or risk for evaluation of compliance with postclosure performance criteria (10 CFR 60 and 40 CFR 191). In addition to describing the proposed methodology, this report reviews available methodologies for scenario analysis, discusses pertinent performance assessment and uncertainty concepts, advises how to implement the methodology (including the organizational requirements and a description of tasks) and recommends how to use the methodology in guiding future site characterization, analysis, and engineered subsystem design work. 36 refs., 24 figs., 1 tab

Fire risk analysis for nuclear power plants: Methodological developments and applications

International Nuclear Information System (INIS)

Kazarians, M.; Apostolakis, G.; Siv, N.O.

1985-01-01

A methodology to quantify the risk from fires in nuclear power plants is described. This methodology combines engineering judgment, statistical evidence, fire phenomenology, and plant system analysis. It can be divided into two major parts: (1) fire scenario identification and quantification, and (2) analysis of the impact on plant safety. This article primarily concentrates on the first part. Statistical analysis of fire occurrence data is used to establish the likelihood of ignition. The temporal behaviors of the two competing phenomena, fire propagation and fire detection and suppression, are studied and their characteristic times are compared. Severity measures are used to further specialize the frequency of the fire scenario. The methodology is applied to a switchgear room of a nuclear power plant

Methodology for dimensional variation analysis of ITER integrated systems

International Nuclear Information System (INIS)

Fuentes, F. Javier; Trouvé, Vincent; Cordier, Jean-Jacques; Reich, Jens

2016-01-01

Highlights: • Tokamak dimensional management methodology, based on 3D variation analysis, is presented. • Dimensional Variation Model implementation workflow is described. • Methodology phases are described in detail. The application of this methodology to the tolerance analysis of ITER Vacuum Vessel is presented. • Dimensional studies are a valuable tool for the assessment of Tokamak PCR (Project Change Requests), DR (Deviation Requests) and NCR (Non-Conformance Reports). - Abstract: The ITER machine consists of a large number of complex systems highly integrated, with critical functional requirements and reduced design clearances to minimize the impact in cost and performances. Tolerances and assembly accuracies in critical areas could have a serious impact in the final performances, compromising the machine assembly and plasma operation. The management of tolerances allocated to part manufacture and assembly processes, as well as the control of potential deviations and early mitigation of non-compliances with the technical requirements, is a critical activity on the project life cycle. A 3D tolerance simulation analysis of ITER Tokamak machine has been developed based on 3DCS dedicated software. This integrated dimensional variation model is representative of Tokamak manufacturing functional tolerances and assembly processes, predicting accurate values for the amount of variation on critical areas. This paper describes the detailed methodology to implement and update the Tokamak Dimensional Variation Model. The model is managed at system level. The methodology phases are illustrated by its application to the Vacuum Vessel (VV), considering the status of maturity of VV dimensional variation model. The following topics are described in this paper: • Model description and constraints. • Model implementation workflow. • Management of input and output data. • Statistical analysis and risk assessment. The management of the integration studies based on

Methodology for dimensional variation analysis of ITER integrated systems

Energy Technology Data Exchange (ETDEWEB)

Fuentes, F. Javier, E-mail: FranciscoJavier.Fuentes@iter.org [ITER Organization, Route de Vinon-sur-Verdon—CS 90046, 13067 St Paul-lez-Durance (France); Trouvé, Vincent [Assystem Engineering & Operation Services, rue J-M Jacquard CS 60117, 84120 Pertuis (France); Cordier, Jean-Jacques; Reich, Jens [ITER Organization, Route de Vinon-sur-Verdon—CS 90046, 13067 St Paul-lez-Durance (France)

2016-11-01

Highlights: • Tokamak dimensional management methodology, based on 3D variation analysis, is presented. • Dimensional Variation Model implementation workflow is described. • Methodology phases are described in detail. The application of this methodology to the tolerance analysis of ITER Vacuum Vessel is presented. • Dimensional studies are a valuable tool for the assessment of Tokamak PCR (Project Change Requests), DR (Deviation Requests) and NCR (Non-Conformance Reports). - Abstract: The ITER machine consists of a large number of complex systems highly integrated, with critical functional requirements and reduced design clearances to minimize the impact in cost and performances. Tolerances and assembly accuracies in critical areas could have a serious impact in the final performances, compromising the machine assembly and plasma operation. The management of tolerances allocated to part manufacture and assembly processes, as well as the control of potential deviations and early mitigation of non-compliances with the technical requirements, is a critical activity on the project life cycle. A 3D tolerance simulation analysis of ITER Tokamak machine has been developed based on 3DCS dedicated software. This integrated dimensional variation model is representative of Tokamak manufacturing functional tolerances and assembly processes, predicting accurate values for the amount of variation on critical areas. This paper describes the detailed methodology to implement and update the Tokamak Dimensional Variation Model. The model is managed at system level. The methodology phases are illustrated by its application to the Vacuum Vessel (VV), considering the status of maturity of VV dimensional variation model. The following topics are described in this paper: • Model description and constraints. • Model implementation workflow. • Management of input and output data. • Statistical analysis and risk assessment. The management of the integration studies based on

A methodology for strain-based fatigue reliability analysis

International Nuclear Information System (INIS)

Zhao, Y.X.

2000-01-01

A significant scatter of the cyclic stress-strain (CSS) responses should be noted for a nuclear reactor material, 1Cr18Ni9Ti pipe-weld metal. Existence of the scatter implies that a random cyclic strain applied history will be introduced under any of the loading modes even a deterministic loading history. A non-conservative evaluation might be given in the practice without considering the scatter. A methodology for strain-based fatigue reliability analysis, which has taken into account the scatter, is developed. The responses are approximately modeled by probability-based CSS curves of Ramberg-Osgood relation. The strain-life data are modeled, similarly, by probability-based strain-life curves of Coffin-Manson law. The reliability assessment is constructed by considering interference of the random fatigue strain applied and capacity histories. Probability density functions of the applied and capacity histories are analytically given. The methodology could be conveniently extrapolated to the case of deterministic CSS relation as the existent methods did. Non-conservative evaluation of the deterministic CSS relation and availability of present methodology have been indicated by an analysis of the material test results

MMOSA – A new approach of the human and organizational factor analysis in PSA

International Nuclear Information System (INIS)

Farcasiu, M.; Prisecaru, I.

2014-01-01

The results of many Probabilistic Safety Assessment (PSA) studies show a very significant contribution of human errors to nuclear installations failure. This paper is intended to analyze both the human performance importance in PSA studies and the elements that influence it. Starting from Man–Machine–Organization System (MMOS) concept a new approach (MMOSA) was developed to allow an explicit incorporation of the human and organizational factor in PSA studies. This method uses old techniques from Human Reliability Analysis (HRA) methods (THERP, SPAR-H) and new techniques to analyze human performance. The main novelty included in MMOSA is the identification of the machine–organization interfaces (maintenance, modification and aging management plan and state of man–machine interface) and the human performance evaluation based on them. A detailed result of the Human Performance Analysis (HPA) using the MMOSA methodology can identify any serious deficiencies of human performance which can usually be corrected through the improvement of the related MMOS interfaces. - Highlights: • MMOSA allows the incorporation of the human and organizational factor in PSA. • The method uses old techniques and new techniques to analyze human performance. • The main novelty is the identification of the machine–organization interfaces. • The MMOSA methodology identifies any serious deficiencies which can be corrected

Requirements Analysis in the Value Methodology

Energy Technology Data Exchange (ETDEWEB)

Conner, Alison Marie

2001-05-01

The Value Methodology (VM) study brings together a multidisciplinary team of people who own the problem and have the expertise to identify and solve it. With the varied backgrounds and experiences the team brings to the study, come different perspectives on the problem and the requirements of the project. A requirements analysis step can be added to the Information and Function Analysis Phases of a VM study to validate whether the functions being performed are required, either regulatory or customer prescribed. This paper will provide insight to the level of rigor applied to a requirements analysis step and give some examples of tools and techniques utilized to ease the management of the requirements and functions those requirements support for highly complex problems.

Attack Methodology Analysis: Emerging Trends in Computer-Based Attack Methodologies and Their Applicability to Control System Networks

Energy Technology Data Exchange (ETDEWEB)

Bri Rolston

2005-06-01

Threat characterization is a key component in evaluating the threat faced by control systems. Without a thorough understanding of the threat faced by critical infrastructure networks, adequate resources cannot be allocated or directed effectively to the defense of these systems. Traditional methods of threat analysis focus on identifying the capabilities and motivations of a specific attacker, assessing the value the adversary would place on targeted systems, and deploying defenses according to the threat posed by the potential adversary. Too many effective exploits and tools exist and are easily accessible to anyone with access to an Internet connection, minimal technical skills, and a significantly reduced motivational threshold to be able to narrow the field of potential adversaries effectively. Understanding how hackers evaluate new IT security research and incorporate significant new ideas into their own tools provides a means of anticipating how IT systems are most likely to be attacked in the future. This research, Attack Methodology Analysis (AMA), could supply pertinent information on how to detect and stop new types of attacks. Since the exploit methodologies and attack vectors developed in the general Information Technology (IT) arena can be converted for use against control system environments, assessing areas in which cutting edge exploit development and remediation techniques are occurring can provide significance intelligence for control system network exploitation, defense, and a means of assessing threat without identifying specific capabilities of individual opponents. Attack Methodology Analysis begins with the study of what exploit technology and attack methodologies are being developed in the Information Technology (IT) security research community within the black and white hat community. Once a solid understanding of the cutting edge security research is established, emerging trends in attack methodology can be identified and the gap between

Physical data generation methodology for return-to-power steam line break analysis

Energy Technology Data Exchange (ETDEWEB)

Zee, Sung Kyun; Lee, Chung Chan; Lee, Chang Kue [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-02-01

Current methodology to generate physics data for steamline break accident analysis of CE-type nuclear plant such as Yonggwang Unit 3 is valid only if the core reactivity does not reach the criticality after shutdown. Therefore, the methodology requires tremendous amount of net scram worth, specially at the end of the cycle when moderator temperature coefficient is most negative. Therefore, we need a new methodology to obtain reasonably conservation physics data, when the reactor returns to power condition. Current methodology used ROCS which include only closed channel model. But it is well known that the closed channel model estimates the core reactivity too much negative if core flow rate is low. Therefore, a conservative methodology is presented which utilizes open channel 3D HERMITE model. Current methodology uses ROCS which include only closed channel model. But it is well known that the closed channel model estimates the core reactivity too much negative if core flow rate is low. Therefore, a conservative methodology is presented which utilizes open channel 3D HERMITE model. Return-to-power reactivity credit is produced to assist the reactivity table generated by closed channel model. Other data includes hot channel axial power shape, peaking factor and maximum quality for DNBR analysis. It also includes pin census for radiological consequence analysis. 48 figs., 22 tabs., 18 refs. (Author) .new.

Interaction between core analysis methodology and nuclear design: some PWR examples

International Nuclear Information System (INIS)

Rothleder, B.M.; Eich, W.J.

1982-01-01

The interaction between core analysis methodology and nuclear design is exemplified by PSEUDAX, a major improvement related to the Advanced Recycle methodology program (ARMP) computer code system, still undergoing development by the Electric Power Research Institute. The mechanism of this interaction is explored by relating several specific nulcear design changes to the demands placed by these changes on the ARMP system, and by examining the meeting of these demands, first within the standard ARMP methodology and then through augmentation of the standard methodology by development of PSEUDAX

Prototype application of best estimate and uncertainty safety analysis methodology to large LOCA analysis

International Nuclear Information System (INIS)

Luxat, J.C.; Huget, R.G.

2001-01-01

Development of a methodology to perform best estimate and uncertainty nuclear safety analysis has been underway at Ontario Power Generation for the past two and one half years. A key driver for the methodology development, and one of the major challenges faced, is the need to re-establish demonstrated safety margins that have progressively been undermined through excessive and compounding conservatism in deterministic analyses. The major focus of the prototyping applications was to quantify the safety margins that exist at the probable range of high power operating conditions, rather than the highly improbable operating states associated with Limit of the Envelope (LOE) assumptions. In LOE, all parameters of significance to the consequences of a postulated accident are assumed to simultaneously deviate to their limiting values. Another equally important objective of the prototyping was to demonstrate the feasibility of conducting safety analysis as an incremental analysis activity, as opposed to a major re-analysis activity. The prototype analysis solely employed prior analyses of Bruce B large break LOCA events - no new computer simulations were undertaken. This is a significant and novel feature of the prototyping work. This methodology framework has been applied to a postulated large break LOCA in a Bruce generating unit on a prototype basis. This paper presents results of the application. (author)

Extracting and Converting Quantitative Data into Human Error Probabilities

Energy Technology Data Exchange (ETDEWEB)

Tuan Q. Tran; Ronald L. Boring; Jeffrey C. Joe; Candice D. Griffith

2007-08-01

This paper discusses a proposed method using a combination of advanced statistical approaches (e.g., meta-analysis, regression, structural equation modeling) that will not only convert different empirical results into a common metric for scaling individual PSFs effects, but will also examine the complex interrelationships among PSFs. Furthermore, the paper discusses how the derived statistical estimates (i.e., effect sizes) can be mapped onto a HRA method (e.g. SPAR-H) to generate HEPs that can then be use in probabilistic risk assessment (PRA). The paper concludes with a discussion of the benefits of using academic literature in assisting HRA analysts in generating sound HEPs and HRA developers in validating current HRA models and formulating new HRA models.

SINGULAR SPECTRUM ANALYSIS: METHODOLOGY AND APPLICATION TO ECONOMICS DATA

Institute of Scientific and Technical Information of China (English)

Hossein HASSANI; Anatoly ZHIGLJAVSKY

2009-01-01

This paper describes the methodology of singular spectrum analysis (SSA) and demonstrate that it is a powerful method of time series analysis and forecasting, particulary for economic time series. The authors consider the application of SSA to the analysis and forecasting of the Iranian national accounts data as provided by the Central Bank of the Islamic Republic of lran.

Methodological developments and applications of neutron activation analysis

International Nuclear Information System (INIS)

Kucera, J.

2007-01-01

The paper reviews the author's experience acquired and achievements made in methodological developments of neutron activation analysis (NAA) of mostly biological materials. These involve epithermal neutron activation analysis, radiochemical neutron activation analysis using both single- and multi-element separation procedures, use of various counting modes, and the development and use of the self-verification principle. The role of NAA in the detection of analytical errors is discussed and examples of applications of the procedures developed are given. (author)

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-06-01

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

Reliability analysis for power supply system in a reprocessing facility based on GO methodology

International Nuclear Information System (INIS)

Wang Renze

2014-01-01

GO methodology was applied to analyze the reliability of power supply system in a typical reprocessing facility. Based on the fact that tie breakers are set in the system, tie breaker operator was defined. Then GO methodology modeling and quantitative analysis were performed sequently, minimal cut sets and average unavailability of the system were obtained. Parallel analysis between GO methodology and fault tree methodology was also performed. The results showed that setup of tie breakers was rational and necessary and that the modeling was much easier and the chart was much more succinct for GO methodology parallel with fault tree methodology to analyze the reliability of the power supply system. (author)

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-09-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

International Nuclear Information System (INIS)

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

2009-01-01

Since the Reactor Safety Study in the early 1970's, human reliability analysis (HRA) has been evolving towards a better ability to account for the factors and conditions that can lead humans to take unsafe actions and thereby provide better estimates of the likelihood of human error for probabilistic risk assessments (PRAs). The purpose of this paper is to provide an overview of recent reviews of operational events and advances in the behavioral sciences that have impacted the evolution of HRA methods and contributed to improvements. The paper discusses the importance of human errors in complex human-technical systems, examines why humans contribute to accidents and unsafe conditions, and discusses how lessons learned over the years have changed the perspective and approach for modeling human behavior in PRAs of complicated domains such as nuclear power plants. It is argued that it has become increasingly more important to understand and model the more cognitive aspects of human performance and to address the broader range of factors that have been shown to influence human performance in complex domains. The paper concludes by addressing the current ability of HRA to adequately predict human failure events and their likelihood

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-01-01

Since the Reactor Safety Study in the early 1970's, human reliability analysis (HRA) has been evolving towards a better ability to account for the factors and conditions that can lead humans to take unsafe actions and thereby provide better estimates of the likelihood of human error for probabilistic risk assessments (PRAs). The purpose of this paper is to provide an overview of recent reviews of operational events and advances in the behavioral sciences that have impacted the evolution of HRA methods and contributed to improvements. The paper discusses the importance of human errors in complex human-technical systems, examines why humans contribute to accidents and unsafe conditions, and discusses how lessons learned over the years have changed the perspective and approach for modeling human behavior in PRAs of complicated domains such as nuclear power plants. It is argued that it has become increasingly more important to understand and model the more cognitive aspects of human performance and to address the broader range of factors that have been shown to influence human performance in complex domains. The paper concludes by addressing the current ability of HRA to adequately predict human failure events and their likelihood.

RAMS (Risk Analysis - Modular System) methodology

Energy Technology Data Exchange (ETDEWEB)

Stenner, R.D.; Strenge, D.L.; Buck, J.W. [and others

1996-10-01

The Risk Analysis - Modular System (RAMS) was developed to serve as a broad scope risk analysis tool for the Risk Assessment of the Hanford Mission (RAHM) studies. The RAHM element provides risk analysis support for Hanford Strategic Analysis and Mission Planning activities. The RAHM also provides risk analysis support for the Hanford 10-Year Plan development activities. The RAMS tool draws from a collection of specifically designed databases and modular risk analysis methodologies and models. RAMS is a flexible modular system that can be focused on targeted risk analysis needs. It is specifically designed to address risks associated with overall strategy, technical alternative, and `what if` questions regarding the Hanford cleanup mission. RAMS is set up to address both near-term and long-term risk issues. Consistency is very important for any comparative risk analysis, and RAMS is designed to efficiently and consistently compare risks and produce risk reduction estimates. There is a wide range of output information that can be generated by RAMS. These outputs can be detailed by individual contaminants, waste forms, transport pathways, exposure scenarios, individuals, populations, etc. However, they can also be in rolled-up form to support high-level strategy decisions.

Theoretical and methodological approaches in discourse analysis.

Science.gov (United States)

Stevenson, Chris

2004-01-01

Discourse analysis (DA) embodies two main approaches: Foucauldian DA and radical social constructionist DA. Both are underpinned by social constructionism to a lesser or greater extent. Social constructionism has contested areas in relation to power, embodiment, and materialism, although Foucauldian DA does focus on the issue of power Embodiment and materialism may be especially relevant for researchers of nursing where the physical body is prominent. However, the contested nature of social constructionism allows a fusion of theoretical and methodological approaches tailored to a specific research interest. In this paper, Chris Stevenson suggests a framework for working out and declaring the DA approach to be taken in relation to a research area, as well as to aid anticipating methodological critique. Method, validity, reliability and scholarship are discussed from within a discourse analytic frame of reference.

Theoretical and methodological approaches in discourse analysis.

Science.gov (United States)

Stevenson, Chris

2004-10-01

Discourse analysis (DA) embodies two main approaches: Foucauldian DA and radical social constructionist DA. Both are underpinned by social constructionism to a lesser or greater extent. Social constructionism has contested areas in relation to power, embodiment, and materialism, although Foucauldian DA does focus on the issue of power. Embodiment and materialism may be especially relevant for researchers of nursing where the physical body is prominent. However, the contested nature of social constructionism allows a fusion of theoretical and methodological approaches tailored to a specific research interest. In this paper, Chris Stevenson suggests a frame- work for working out and declaring the DA approach to be taken in relation to a research area, as well as to aid anticipating methodological critique. Method, validity, reliability and scholarship are discussed from within a discourse analytic frame of reference.

Stakeholder analysis methodologies resource book

Energy Technology Data Exchange (ETDEWEB)

Babiuch, W.M.; Farhar, B.C.

1994-03-01

Stakeholder analysis allows analysts to identify how parties might be affected by government projects. This process involves identifying the likely impacts of a proposed action and stakeholder groups affected by that action. Additionally, the process involves assessing how these groups might be affected and suggesting measures to mitigate any adverse effects. Evidence suggests that the efficiency and effectiveness of government actions can be increased and adverse social impacts mitigated when officials understand how a proposed action might affect stakeholders. This report discusses how to conduct useful stakeholder analyses for government officials making decisions on energy-efficiency and renewable-energy technologies and their commercialization. It discusses methodological issues that may affect the validity and reliability of findings, including sampling, generalizability, validity, ``uncooperative`` stakeholder groups, using social indicators, and the effect of government regulations. The Appendix contains resource directories and a list of specialists in stakeholder analysis and involvement.

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.)

Development of the GO-FLOW reliability analysis methodology for nuclear reactor system

International Nuclear Information System (INIS)

Matsuoka, Takeshi; Kobayashi, Michiyuki

1994-01-01

Probabilistic Safety Assessment (PSA) is important in the safety analysis of technological systems and processes, such as, nuclear plants, chemical and petroleum facilities, aerospace systems. Event trees and fault trees are the basic analytical tools that have been most frequently used for PSAs. Several system analysis methods can be used in addition to, or in support of, the event- and fault-tree analysis. The need for more advanced methods of system reliability analysis has grown with the increased complexity of engineered systems. The Ship Research Institute has been developing a new reliability analysis methodology, GO-FLOW, which is a success-oriented system analysis technique, and is capable of evaluating a large system with complex operational sequences. The research has been supported by the special research fund for Nuclear Technology, Science and Technology Agency, from 1989 to 1994. This paper describes the concept of the Probabilistic Safety Assessment (PSA), an overview of various system analysis techniques, an overview of the GO-FLOW methodology, the GO-FLOW analysis support system, procedure of treating a phased mission problem, a function of common cause failure analysis, a function of uncertainty analysis, a function of common cause failure analysis with uncertainty, and printing out system of the results of GO-FLOW analysis in the form of figure or table. Above functions are explained by analyzing sample systems, such as PWR AFWS, BWR ECCS. In the appendices, the structure of the GO-FLOW analysis programs and the meaning of the main variables defined in the GO-FLOW programs are described. The GO-FLOW methodology is a valuable and useful tool for system reliability analysis, and has a wide range of applications. With the development of the total system of the GO-FLOW, this methodology has became a powerful tool in a living PSA. (author) 54 refs

Methodology for object-oriented real-time systems analysis and design: Software engineering

Science.gov (United States)

Schoeffler, James D.

1991-01-01

Successful application of software engineering methodologies requires an integrated analysis and design life-cycle in which the various phases flow smoothly 'seamlessly' from analysis through design to implementation. Furthermore, different analysis methodologies often lead to different structuring of the system so that the transition from analysis to design may be awkward depending on the design methodology to be used. This is especially important when object-oriented programming is to be used for implementation when the original specification and perhaps high-level design is non-object oriented. Two approaches to real-time systems analysis which can lead to an object-oriented design are contrasted: (1) modeling the system using structured analysis with real-time extensions which emphasizes data and control flows followed by the abstraction of objects where the operations or methods of the objects correspond to processes in the data flow diagrams and then design in terms of these objects; and (2) modeling the system from the beginning as a set of naturally occurring concurrent entities (objects) each having its own time-behavior defined by a set of states and state-transition rules and seamlessly transforming the analysis models into high-level design models. A new concept of a 'real-time systems-analysis object' is introduced and becomes the basic building block of a series of seamlessly-connected models which progress from the object-oriented real-time systems analysis and design system analysis logical models through the physical architectural models and the high-level design stages. The methodology is appropriate to the overall specification including hardware and software modules. In software modules, the systems analysis objects are transformed into software objects.

Vulnerability and Risk Analysis Program: Overview of Assessment Methodology

National Research Council Canada - National Science Library

2001-01-01

.... Over the last three years, a team of national laboratory experts, working in partnership with the energy industry, has successfully applied the methodology as part of OCIP's Vulnerability and Risk Analysis Program (VRAP...

Environmental impact statement analysis: dose methodology

International Nuclear Information System (INIS)

Mueller, M.A.; Strenge, D.L.; Napier, B.A.

1981-01-01

Standardized sections and methodologies are being developed for use in environmental impact statements (EIS) for activities to be conducted on the Hanford Reservation. Five areas for standardization have been identified: routine operations dose methodologies, accident dose methodology, Hanford Site description, health effects methodology, and socioeconomic environment for Hanford waste management activities

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

A methodology for radiological accidents analysis in industrial gamma radiography

International Nuclear Information System (INIS)

Silva, F.C.A. da.

1990-01-01

A critical review of 34 published severe radiological accidents in industrial gamma radiography, that happened in 15 countries, from 1960 to 1988, was performed. The most frequent causes, consequences and dose estimation methods were analysed, aiming to stablish better procedures of radiation safety and accidents analysis. The objective of this work is to elaborate a radiological accidents analysis methodology in industrial gamma radiography. The suggested methodology will enable professionals to determine the true causes of the event and to estimate the dose with a good certainty. The technical analytical tree, recommended by International Atomic Energy Agency to perform radiation protection and nuclear safety programs, was adopted in the elaboration of the suggested methodology. The viability of the use of the Electron Gamma Shower 4 Computer Code System to calculate the absorbed dose in radiological accidents in industrial gamma radiography, mainly at sup(192)Ir radioactive source handling situations was also studied. (author)

Development of Audit Calculation Methodology for RIA Safety Analysis

Energy Technology Data Exchange (ETDEWEB)

Lee, Joosuk; Kim, Gwanyoung; Woo, Swengwoong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2015-05-15

The interim criteria contain more stringent limits than previous ones. For example, pellet-to-cladding mechanical interaction(PCMI) was introduced as a new failure criteria. And both short-term (e.g. fuel-to coolant interaction, rod burst) and long-term(e.g., fuel rod ballooning, flow blockage) phenomena should be addressed for core coolability assurance. For dose calculations, transient-induced fission gas release has to be accounted additionally. Traditionally, the approved RIA analysis methodologies for licensing application are developed based on conservative approach. But newly introduced safety criteria tend to reduce the margins to the criteria. Thereby, licensees are trying to improve the margins by utilizing a less conservative approach. In this situation, to cope with this trend, a new audit calculation methodology needs to be developed. In this paper, the new methodology, which is currently under developing in KINS, was introduced. For the development of audit calculation methodology of RIA safety analysis based on the realistic evaluation approach, preliminary calculation by utilizing the best estimate code has been done on the initial core of APR1400. Followings are main conclusions. - With the assumption of single full-strength control rod ejection in HZP condition, rod failure due to PCMI is not predicted. - And coolability can be assured in view of entalphy and fuel melting. - But, rod failure due to DNBR is expected, and there is possibility of fuel failure at the rated power conditions also.

The assessment of cervical sensory motor control: a systematic review focusing on measuring methods and their clinimetric characteristics.

Science.gov (United States)

Michiels, Sarah; De Hertogh, Willem; Truijen, Steven; November, Danny; Wuyts, Floris; Van de Heyning, Paul

2013-05-01

Cervical sensorimotor control (CSMC) becomes increasingly important in the assessment and treatment of patients with neck pain. This review aims to compare commonly used CSMC measuring methods in terms of required tasks, measuring device and clinimetric properties. A systematic review of two databases, followed by methodological quality assessment (CBO guidelines). The methodological quality of 34 included articles was generally good (five to seven out of eight), the inter-rater agreement was excellent (κw=0.966, pvirtual reality test, a continuous linear movement technique (CLMT) and an object following non-linear movement technique (NLMT) (The Fly™). Test-retest reliability was fair to excellent (ICC 0.35-0.87) for the HRA-to-NHP, very bad to excellent (ICC 0.01-0.90) for the HRA-to-target, fair to good (ICC 0.25-0.77) for the virtual reality test and moderate to excellent (ICC: 0.60-0.86) for The Fly™. The reliability of the CLMT was not documented. The HRA-to-NHP, The Fly™ and the CLMT can discriminate between patients with neck complaints and controls (discriminant validity). Currently, only The Fly™ can discriminate between different patient populations (post-traumatic and non-traumatic neck pain). The sensitivity, specificity and responsiveness of the methods have to be assessed in future research. The dynamic method The Fly™ appears to be more reliable than the HRA-to-NHP and is able to discriminate between different patient populations. The diagnostic potential is to be confirmed in future research. Copyright © 2012 Elsevier B.V. All rights reserved.

Opening Remarks of the Acquisition Path Analysis Methodology Session

International Nuclear Information System (INIS)

Renis, T.

2015-01-01

An overview of the recent development work that has been done on acquisition path analysis, implementation of the methodologies within the Department of Safeguards, lessons learned and future areas for development will be provided. (author)

An economic analysis methodology for project evaluation and programming.

Science.gov (United States)

2013-08-01

Economic analysis is a critical component of a comprehensive project or program evaluation methodology that considers all key : quantitative and qualitative impacts of highway investments. It allows highway agencies to identify, quantify, and value t...

Eco-efficiency analysis methodology on the example of the chosen polyolefins production

OpenAIRE

K. Czaplicka-Kolarz; D. Burchart-Korol; P. Krawczyk

2010-01-01

the chosen polyolefins production. The article presents also main tools of eco-efficiency analysis: Life Cycle Assessment (LCA) and Net Present Value (NPV).Design/methodology/approach: On the basis of LCA and NPV of high density polyethylene (HDPE) and low density polyethylene (LDPE) production, eco-efficiency analysis is conducted.Findings: In this article environmental and economic performance of the chosen polyolefins production was presented. The basis phases of eco-efficiency methodology...

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.

Development of a Long Term Cooling Analysis Methodology Using Rappel

International Nuclear Information System (INIS)

Lee, S. I.; Jeong, J. H.; Ban, C. H.; Oh, S. J.

2012-01-01

Since the revision of the 10CFR50.46 in 1988, which allowed BE (Best-Estimate) method in analyzing the safety performance of a nuclear power plant, safety analysis methodologies have been changed continuously from conservative EM (Evaluation Model) approaches to BE ones. In this context, LSC (Long-Term core Cooling) methodologies have been reviewed by the regulatory bodies of USA and Korea. Some non-conservatism and improperness of the old methodology have been identified, and as a result, USNRC suspended the approval of CENPD-254-P-A which is the old LSC methodology for CE-designed NPPs. Regulatory bodies requested to remove the non-conservatisms and to reflect system transient behaviors in all the LSC methodologies used. In the present study, a new LSC methodology using RELAP5 is developed. RELAP5 and a newly developed code, BACON (Boric Acid Concentration Of Nuclear power plant) are used to calculate the transient behavior of the system and the boric acid concentration, respectively. Full range of break spectrum is considered and the applicability is confirmed through plant demonstration calculations. The result shows a good comparison with the old-fashioned ones, therefore, the methodology could be applied with no significant changes of current LSC plans

THEORETICAL AND METHODOLOGICAL PRINCIPLES OF THE STRATEGIC FINANCIAL ANALYSIS OF CAPITAL

Directory of Open Access Journals (Sweden)

Olha KHUDYK

2016-07-01

Full Text Available The article is devoted to the theoretical and methodological principles of strategic financial analysis of capital. The necessity of strategic financial analysis of capital as a methodological basis for study strategies is proved in modern conditions of a high level of dynamism, uncertainty and risk. The methodological elements of the strategic indicators, the factors, the methods of study, the subjects of analysis, the sources of incoming and outgoing information are justified in the system of financial management, allowing to improve its theoretical foundations. It is proved that the strategic financial analysis of capital is a continuous process, carried out in an appropriate sequence at each stage of capital circulation. The system of indexes is substantiated, based on the needs of the strategic financial analysis. The classification of factors determining the size and structure of company’s capital is grounded. The economic nature of capital of the company is clarified. We consider that capital is a stock of economic resources in the form of cash, tangible and intangible assets accumulated by savings, which is used by its owner as a factor of production and investment resource in the economic process in order to obtain profit, to ensure the growth of owners’ prosperity and to achieve social effect.

EUROCONTROL-Systemic Occurrence Analysis Methodology (SOAM)-A 'Reason'-based organisational methodology for analysing incidents and accidents

International Nuclear Information System (INIS)

Licu, Tony; Cioran, Florin; Hayward, Brent; Lowe, Andrew

2007-01-01

The Safety Occurrence Analysis Methodology (SOAM) developed for EUROCONTROL is an accident investigation methodology based on the Reason Model of organisational accidents. The purpose of a SOAM is to broaden the focus of an investigation from human involvement issues, also known as 'active failures of operational personnel' under Reason's original model, to include analysis of the latent conditions deeper within the organisation that set the context for the event. Such an approach is consistent with the tenets of Just Culture in which people are encouraged to provide full and open information about how incidents occurred, and are not penalised for errors. A truly systemic approach is not simply a means of transferring responsibility for a safety occurrence from front-line employees to senior managers. A consistent philosophy must be applied, where the investigation process seeks to correct deficiencies wherever they may be found, without attempting to apportion blame or liability

Design and analysis of sustainable computer mouse using design for disassembly methodology

Science.gov (United States)

Roni Sahroni, Taufik; Fitri Sukarman, Ahmad; Agung Mahardini, Karunia

2017-12-01

This paper presents the design and analysis of computer mouse using Design for Disassembly methodology. Basically, the existing computer mouse model consist a number of unnecessary part that cause the assembly and disassembly time in production. The objective of this project is to design a new computer mouse based on Design for Disassembly (DFD) methodology. The main methodology of this paper was proposed from sketch generation, concept selection, and concept scoring. Based on the design screening, design concept B was selected for further analysis. New design of computer mouse is proposed using fastening system. Furthermore, three materials of ABS, Polycarbonate, and PE high density were prepared to determine the environmental impact category. Sustainable analysis was conducted using software SolidWorks. As a result, PE High Density gives the lowers amount in the environmental category with great maximum stress value.

Methodology, Measurement and Analysis of Flow Table Update Characteristics in Hardware OpenFlow Switches

KAUST Repository

Kuźniar, Maciej; Pereší ni, Peter; Kostić, Dejan; Canini, Marco

2018-01-01

and performance characteristics is essential for ensuring successful and safe deployments.We propose a systematic methodology for SDN switch performance analysis and devise a series of experiments based on this methodology. The methodology relies on sending a

THEORETICAL AND METHODOLOGICAL PRINCIPLES OF THE STRATEGIC FINANCIAL ANALYSIS OF CAPITAL

Directory of Open Access Journals (Sweden)

Olha KHUDYK

2016-07-01

Full Text Available The article is devoted to the theoretical and methodological principles of strategic financial analysis of capital. The necessity of strategic financial analysis of capital as a methodological basis for study strategies is proved in modern conditions of a high level of dynamism, uncertainty and risk. The methodological elements of the strategic financial analysis of capital (the object of investigation, the indicators, the factors, the methods of study, the subjects of analysis, the sources of incoming and outgoing information are justified in the system of financial management, allowing to improve its theoretical foundations. It is proved that the strategic financial analysis of capital is a continuous process, carried out in an appropriate sequence at each stage of capital circulation. The system of indexes is substantiated, based on the needs of the strategic financial analysis. The classification of factors determining the size and structure of company’s capital is grounded. The economic nature of capital of the company is clarified. We consider that capital is a stock of economic resources in the form of cash, tangible and intangible assets accumulated by savings, which is used by its owner as a factor of production and investment resource in the economic process in order to obtain profit, to ensure the growth of owners’ prosperity and to achieve social effect.

Accidental safety analysis methodology development in decommission of the nuclear facility

Energy Technology Data Exchange (ETDEWEB)

Park, G. H.; Hwang, J. H.; Jae, M. S.; Seong, J. H.; Shin, S. H.; Cheong, S. J.; Pae, J. H.; Ang, G. R.; Lee, J. U. [Seoul National Univ., Seoul (Korea, Republic of)

2002-03-15

Decontamination and Decommissioning (D and D) of a nuclear reactor cost about 20% of construction expense and production of nuclear wastes during decommissioning makes environmental issues. Decommissioning of a nuclear reactor in Korea is in a just beginning stage, lacking clear standards and regulations for decommissioning. This work accident safety analysis in decommissioning of the nuclear facility can be a solid ground for the standards and regulations. For source term analysis for Kori-1 reactor vessel, MCNP/ORIGEN calculation methodology was applied. The activity of each important nuclide in the vessel was estimated at a time after 2008, the year Kori-1 plant is supposed to be decommissioned. And a methodology for risk analysis assessment in decommissioning was developed.

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

Development of analysis methodology for hot leg break mass and energy release

International Nuclear Information System (INIS)

Song, Jin Ho; Kim, Cheol Woo; Kwon, Young Min; Kim, Sook Kwan

1995-04-01

A study for the development of an analysis methodology for hot leg break mass and energy release is performed. For the blowdown period a modified CEFLASH-4A methodology is suggested. For the post blowdown period a modified CONTRAST boil-off model is suggested. By using these computer code improved mass and energy release data are generated. Also, a RELAP5/MOD3 analysis for finally the FLOOD-3 computer code has been modified for use in the analysis of hot leg break. The results of analysis using modified FLOOD-3 are reasonable as we expected and their trends are good. 66 figs., 8 tabs. (Author) .new

Using functional analysis in archival appraisal a practical and effective alternative to traditional appraisal methodologies

CERN Document Server

Robyns, Marcus C

2014-01-01

In an age of scarcity and the challenge of electronic records, can archivists and records managers continue to rely upon traditional methodology essentially unchanged since the early 1950s? Using Functional Analysis in Archival Appraisal: A Practical and Effective Alternative to Traditional Appraisal Methodologies shows how archivists in other countries are already using functional analysis, which offers a better, more effective, and imminently more practical alternative to traditional appraisal methodologies that rely upon an analysis of the records themselves.

Reactor analysis support package (RASP). Volume 7. PWR set-point methodology. Final report

International Nuclear Information System (INIS)

Temple, S.M.; Robbins, T.R.

1986-09-01

This report provides an overview of the basis and methodology requirements for determining Pressurized Water Reactor (PWR) technical specifications related setpoints and focuses on development of the methodology for a reload core. Additionally, the report documents the implementation and typical methods of analysis used by PWR vendors during the 1970's to develop Protection System Trip Limits (or Limiting Safety System Settings) and Limiting Conditions for Operation. The descriptions of the typical setpoint methodologies are provided for Nuclear Steam Supply Systems as designed and supplied by Babcock and Wilcox, Combustion Engineering, and Westinghouse. The description of the methods of analysis includes the discussion of the computer codes used in the setpoint methodology. Next, the report addresses the treatment of calculational and measurement uncertainties based on the extent to which such information was available for each of the three types of PWR. Finally, the major features of the setpoint methodologies are compared, and the principal effects of each particular methodology on plant operation are summarized for each of the three types of PWR

Using HABIT to Establish the Chemicals Analysis Methodology for Maanshan Nuclear Power Plant

OpenAIRE

J. R. Wang; S. W. Chen; Y. Chiang; W. S. Hsu; J. H. Yang; Y. S. Tseng; C. Shih

2017-01-01

In this research, the HABIT analysis methodology was established for Maanshan nuclear power plant (NPP). The Final Safety Analysis Report (FSAR), reports, and other data were used in this study. To evaluate the control room habitability under the CO2 storage burst, the HABIT methodology was used to perform this analysis. The HABIT result was below the R.G. 1.78 failure criteria. This indicates that Maanshan NPP habitability can be maintained. Additionally, the sensitivity study of the paramet...

AGAPE-ET for human error analysis of emergency tasks and its application

International Nuclear Information System (INIS)

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

2002-01-01

The paper presents a proceduralised human reliability analysis (HRA) methodology, AGAPE-ET (A Guidance And Procedure for Human Error Analysis for Emergency Tasks), covering both qualitative error analysis and quantification of human error probability (HEP) of emergency tasks in nuclear power plants. The AGAPE-ET method is based on the simplified cognitive model. By each cognitive function, error causes or error-likely situations have been identified considering the characteristics of the performance of each cognitive function and influencing mechanism of the performance influencing factors (PIFs) on the cognitive function. Then, error analysis items have been determined from the identified error causes or error-likely situations and a human error analysis procedure based on the error analysis items is organised to help the analysts cue or guide overall human error analysis. The basic scheme for the quantification of HEP consists in the multiplication of the BHEP assigned by the error analysis item and the weight from the influencing factors decision tree (IFDT) constituted by cognitive function. The method can be characterised by the structured identification of the weak points of the task required to perform and the efficient analysis process that the analysts have only to carry out with the necessary cognitive functions. The paper also presents the application of AGAPE-ET to 31 nuclear emergency tasks and its results

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

Methodology for reactor core physics analysis - part 2

International Nuclear Information System (INIS)

Ponzoni Filho, P.; Fernandes, V.B.; Lima Bezerra, J. de; Santos, T.I.C.

1992-12-01

The computer codes used for reactor core physics analysis are described. The modifications introduced in the public codes and the technical basis for the codes developed by the FURNAS utility are justified. An evaluation of the impact of these modifications on the parameter involved in qualifying the methodology is included. (F.E.). 5 ref, 7 figs, 5 tabs

Methodologies for uncertainty analysis in the level 2 PSA and their implementation procedures

International Nuclear Information System (INIS)

Ahn, Kwang Il; Yang, Joon Eun; Kim, Dong Ha

2002-04-01

Main purpose of this report to present standardized methodologies for uncertainty analysis in the Level 2 Probabilistic Safety Assessment (PSA) and their implementation procedures, based on results obtained through a critical review of the existing methodologies for the analysis of uncertainties employed in the Level 2 PSA, especially Accident Progression Event Tree (APET). Uncertainties employed in the Level 2 PSA, quantitative expressions of overall knowledge of analysts' and experts' participating in the probabilistic quantification process of phenomenological accident progressions ranging from core melt to containment failure, their numerical values are directly related to the degree of confidence that the analyst has that a given phenomenological event or accident process will or will not occur, or analyst's subjective probabilities of occurrence. These results that are obtained from Level 2 PSA uncertainty analysis, become an essential contributor to the plant risk, in addition to the Level 1 PSA and Level 3 PSA uncertainties. Uncertainty analysis methodologies and their implementation procedures presented in this report was prepared based on the following criteria: 'uncertainty quantification process must be logical, scrutable, complete, consistent and in an appropriate level of detail, as mandated by the Level 2 PSA objectives'. For the aforementioned purpose, this report deals mainly with (1) summary of general or Level 2 PSA specific uncertainty analysis methodologies, (2) selection of phenomenological branch events for uncertainty analysis in the APET, methodology for quantification of APET uncertainty inputs and its implementation procedure, (3) statistical propagation of uncertainty inputs through APET and its implementation procedure, and (4) formal procedure for quantification of APET uncertainties and source term categories (STCs) through the Level 2 PSA quantification codes

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

Energy Technology Data Exchange (ETDEWEB)

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

2006-07-01

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

Methodology for Design and Analysis of Reactive Distillation Involving Multielement Systems

DEFF Research Database (Denmark)

Jantharasuk, Amnart; Gani, Rafiqul; Górak, Andrzej

2011-01-01

A new methodology for design and analysis of reactive distillation has been developed. In this work, the elementbased approach, coupled with a driving force diagram, has been extended and applied to the design of a reactive distillation column involving multielement (multicomponent) systems...... consisting of two components. Based on this methodology, an optimal design configuration is identified using the equivalent binary-element-driving force diagram. Two case studies of methyl acetate (MeOAc) synthesis and methyl-tert-butyl ether (MTBE) synthesis have been considered to demonstrate...... the successful applications of the methodology. Moreover, energy requirements for various column configurations corresponding to different feed locatio...

Analysis of Interbrand, BrandZ and BAV brand valuation methodologies

Directory of Open Access Journals (Sweden)

Krstić Bojan

2011-01-01

Full Text Available Brand valuation is considered for one of the most significant challenges of not only theory and practice of contemporary marketing, but other disciplines, as well. Namely, the complex nature of this issue implies the need for multidisciplinary approach and creation of methodology which goes beyond the borders of marketing as a discipline, and includes knowledge derived from accounting, finance and other areas. However, mostly one-sided approaches oriented towards determining brand value either based on research results of consumer behavior and attitudes or based on financial success of the brand are dominant in marketing and financial literature. Simultaneously with these theoretical methodologies, agencies for consultancy and marketing and other subjects have been developing their own brand valuation methods and models. Some of them can be appointed to comprehensive approach to brand valuation, which overcomes mentioned problem considering one-sided analysis of brand value. The comprehensive approach, namely, presumes brand valuation based on benefits which brand provides to both customers and enterprise that owns it, in other words - based on qualitative and quantitative measures respectively reflecting behavior and attitudes of consumers and assumed financial value of the brand, or, more precisely, brand value capitalization. According to the defined research subject, this paper is structured as follows: importance and problem of brand value are reviewed in the Introduction, and three most well-known brand valuation methodologies developed by consultancy agencies - Interbrand methodology and BrandZ and BAV models are analyzed in the next section. In the further considerations the results of comparative analysis of these methodologies are presented and implications for adequate brand valuation suggested.

Latest developments on safety analysis methodologies at the Juzbado plant

International Nuclear Information System (INIS)

Zurron-Cifuentes, Oscar; Ortiz-Trujillo, Diego; Blanco-Fernandez, Luis A.

2010-01-01

Over the last few years the Juzbado Plant has developed and implemented several analysis methodologies to cope with specific issues regarding safety management. This paper describes the three most outstanding of them, so as to say, the Integrated Safety Analysis (ISA) project, the adaptation of the MARSSIM methodology for characterization surveys of radioactive contamination spots, and the programme for the Systematic Review of the Operational Conditions of the Safety Systems (SROCSS). Several reasons motivated the decision to implement such methodologies, such as Regulator requirements, operational experience and of course, the strong commitment of ENUSA to maintain the highest standards of nuclear industry on all the safety relevant activities. In this context, since 2004 ENUSA is undertaking the ISA project, which consists on a systematic examination of plant's processes, equipment, structures and personnel activities to ensure that all relevant hazards that could result in unacceptable consequences have been adequately evaluated and the appropriate protective measures have been identified. On the other hand and within the framework of a current programme to ensure the absence of radioactive contamination spots on unintended areas, the MARSSIM methodology is being applied as a tool to conduct the radiation surveys and investigation of potentially contaminated areas. Finally, the SROCSS programme was initiated earlier this year 2009 to assess the actual operating conditions of all the systems with safety relevance, aiming to identify either potential non-conformities or areas for improvement in order to ensure their high performance after years of operation. The following paragraphs describe the key points related to these three methodologies as well as an outline of the results obtained so far. (authors)

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

Energy Technology Data Exchange (ETDEWEB)

Torralba, B.; Martinez-Arias, R.

2007-07-01

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

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

International Nuclear Information System (INIS)

Torralba, B.; Martinez-Arias, R.

2007-01-01

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

Methodology for the analysis of pollutant emissions from a city bus

International Nuclear Information System (INIS)

Armas, Octavio; Lapuerta, Magín; Mata, Carmen

2012-01-01

In this work a methodology is proposed for measurement and analysis of gaseous emissions and particle size distributions emitted by a diesel city bus during its typical operation under urban driving conditions. As test circuit, a passenger transportation line at a Spanish city was used. Different ways for data processing and representation were studied and, derived from this work, a new approach is proposed. The methodology was useful to detect the most important uncertainties arising during registration and processing of data derived from a measurement campaign devoted to determine the main pollutant emissions. A HORIBA OBS-1300 gas analyzer and a TSI engine exhaust particle spectrometer were used with 1 Hz frequency data recording. The methodology proposed allows for the comparison of results (in mean values) derived from the analysis of either complete cycles or specific categories (or sequences). The analysis by categories is demonstrated to be a robust and helpful tool to isolate the effect of the main vehicle parameters (relative fuel–air ratio and velocity) on pollutant emissions. It was shown that acceleration sequences have the highest contribution to the total emissions, whereas deceleration sequences have the least. (paper)

Methodology for the analysis of pollutant emissions from a city bus

Science.gov (United States)

Armas, Octavio; Lapuerta, Magín; Mata, Carmen

2012-04-01

In this work a methodology is proposed for measurement and analysis of gaseous emissions and particle size distributions emitted by a diesel city bus during its typical operation under urban driving conditions. As test circuit, a passenger transportation line at a Spanish city was used. Different ways for data processing and representation were studied and, derived from this work, a new approach is proposed. The methodology was useful to detect the most important uncertainties arising during registration and processing of data derived from a measurement campaign devoted to determine the main pollutant emissions. A HORIBA OBS-1300 gas analyzer and a TSI engine exhaust particle spectrometer were used with 1 Hz frequency data recording. The methodology proposed allows for the comparison of results (in mean values) derived from the analysis of either complete cycles or specific categories (or sequences). The analysis by categories is demonstrated to be a robust and helpful tool to isolate the effect of the main vehicle parameters (relative fuel-air ratio and velocity) on pollutant emissions. It was shown that acceleration sequences have the highest contribution to the total emissions, whereas deceleration sequences have the least.

Towards a Multimodal Methodology for the Analysis of Translated/Localised Games

Directory of Open Access Journals (Sweden)

Bárbara Resende Coelho

2016-12-01

Full Text Available Multimedia materials require research methodologies that are able to comprehend all of their assets. Videogames are the epitome of multimedia, joining image, sound, video, animation, graphics and text with the interactivity factor. A methodology to conduct research into translation and localisation of videogames should be able to analyse all of its assets and features. This paper sets out to develop a research methodology for games and their translations/localisations that goes beyond the collection and analysis of “screenshots” and includes as many of their assets as possible. Using the fully localised version of the game Watchdogs, this papers shows how tools and technologies allow for transcending the mere analysis of linguistic contents within multimedia materials. Using software ELAN Language Archive to analyse Portuguese-language dubbed and English-language subtitled excerpts from the videogame, it was possible to identify patterns in both linguistic and audio-visual elements, as well as to correlate them.

Application of survival analysis methodology to the quantitative analysis of LC-MS proteomics data

KAUST Repository

Tekwe, C. D.; Carroll, R. J.; Dabney, A. R.

2012-01-01

positive, skewed and often left-censored, we propose using survival methodology to carry out differential expression analysis of proteins. Various standard statistical techniques including non-parametric tests such as the Kolmogorov-Smirnov and Wilcoxon

Differing Developmental Trajectories in Heart Rate Responses to Speech Stimuli in Infants at High and Low Risk for Autism Spectrum Disorder.

Science.gov (United States)

Perdue, Katherine L; Edwards, Laura A; Tager-Flusberg, Helen; Nelson, Charles A

2017-08-01

We investigated heart rate (HR) in infants at 3, 6, 9, and 12 months of age, at high (HRA) and low (LRC) familial risk for ASD, to identify potential endophenotypes of ASD risk related to attentional responses. HR was extracted from functional near-infrared spectroscopy recordings while infants listened to speech stimuli. Longitudinal analysis revealed that HRA infants and males generally had lower baseline HR than LRC infants and females. HRA infants showed decreased HR responses to early trials over development, while LRC infants showed increased responses. These findings suggest altered developmental trajectories in physiological responses to speech stimuli over the first year of life, with HRA infants showing less social orienting over time.

Systemic design methodologies for electrical energy systems analysis, synthesis and management

CERN Document Server

Roboam, Xavier

2012-01-01

This book proposes systemic design methodologies applied to electrical energy systems, in particular analysis and system management, modeling and sizing tools. It includes 8 chapters: after an introduction to the systemic approach (history, basics & fundamental issues, index terms) for designing energy systems, this book presents two different graphical formalisms especially dedicated to multidisciplinary devices modeling, synthesis and analysis: Bond Graph and COG/EMR. Other systemic analysis approaches for quality and stability of systems, as well as for safety and robustness analysis tools are also proposed. One chapter is dedicated to energy management and another is focused on Monte Carlo algorithms for electrical systems and networks sizing. The aim of this book is to summarize design methodologies based in particular on a systemic viewpoint, by considering the system as a whole. These methods and tools are proposed by the most important French research laboratories, which have many scientific partn...

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.

Efficacy of Cladribine Tablets in high disease activity subgroups of patients with relapsing multiple sclerosis: A post hoc analysis of the CLARITY study.

Science.gov (United States)

Giovannoni, Gavin; Soelberg Sorensen, Per; Cook, Stuart; Rammohan, Kottil W; Rieckmann, Peter; Comi, Giancarlo; Dangond, Fernando; Hicking, Christine; Vermersch, Patrick

2018-04-01

In the CLARITY (CLAdRIbine Tablets treating multiple sclerosis orallY) study, Cladribine Tablets significantly improved clinical and magnetic resonance imaging (MRI) outcomes (vs placebo) in patients with relapsing-remitting multiple sclerosis. Describe two clinically relevant definitions for patients with high disease activity (HDA) at baseline of the CLARITY study (utility verified in patients receiving placebo) and assess the treatment effects of Cladribine Tablets 3.5 mg/kg compared with the overall study population. Outcomes of patients randomised to Cladribine Tablets 3.5 mg/kg or placebo were analysed for subgroups using HDA definitions based on high relapse activity (HRA; patients with ⩾2 relapses during the year prior to study entry, whether on DMD treatment or not) or HRA plus disease activity on treatment (HRA + DAT; patients with ⩾2 relapses during the year prior to study entry, whether on DMD treatment or not, PLUS patients with ⩾1 relapse during the year prior to study entry while on therapy with other DMDs and ⩾1 T1 Gd+ or ⩾9 T2 lesions). In the overall population, Cladribine Tablets 3.5 mg/kg reduced the risk of 6-month-confirmed Expanded Disability Status Scale (EDSS) worsening by 47% vs placebo. A risk reduction of 82% vs placebo was seen in both the HRA and HRA + DAT subgroups (vs 19% for non-HRA and 18% for non-HRA + DAT), indicating greater responsiveness to Cladribine Tablets 3.5 mg/kg in patients with HDA. There were consistent results for other efficacy endpoints. The safety profile in HDA patients was consistent with the overall CLARITY population. Patients with HDA showed clinical and MRI responses to Cladribine Tablets 3.5 mg/kg that were generally better than, or at least comparable with, the outcomes seen in the overall CLARITY population.

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.

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.

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

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

Network meta-analysis-highly attractive but more methodological research is needed

Directory of Open Access Journals (Sweden)

Singh Sonal

2011-06-01

Full Text Available Abstract Network meta-analysis, in the context of a systematic review, is a meta-analysis in which multiple treatments (that is, three or more are being compared using both direct comparisons of interventions within randomized controlled trials and indirect comparisons across trials based on a common comparator. To ensure validity of findings from network meta-analyses, the systematic review must be designed rigorously and conducted carefully. Aspects of designing and conducting a systematic review for network meta-analysis include defining the review question, specifying eligibility criteria, searching for and selecting studies, assessing risk of bias and quality of evidence, conducting a network meta-analysis, interpreting and reporting findings. This commentary summarizes the methodologic challenges and research opportunities for network meta-analysis relevant to each aspect of the systematic review process based on discussions at a network meta-analysis methodology meeting we hosted in May 2010 at the Johns Hopkins Bloomberg School of Public Health. Since this commentary reflects the discussion at that meeting, it is not intended to provide an overview of the field.

Poster - 30: Use of a Hazard-Risk Analysis for development of a new eye immobilization tool for treatment of choroidal melanoma

International Nuclear Information System (INIS)

Prooijen, Monique van; Breen, Stephen

2016-01-01

Purpose: Our treatment for choroidal melanoma utilizes the GTC frame. The patient looks at a small LED to stabilize target position. The LED is attached to a metal arm attached to the GTC frame. A camera on the arm allows therapists to monitor patient compliance. To move to mask-based immobilization we need a new LED/camera attachment mechanism. We used a Hazard-Risk Analysis (HRA) to guide the design of the new tool. Method: A pre-clinical model was built with input from therapy and machine shop personnel. It consisted of an aluminum frame placed in aluminum guide posts attached to the couch top. Further development was guided by the Department of Defense Standard Practice - System Safety hazard risk analysis technique. Results: An Orfit mask was selected because it allowed access to indexes on the couch top which assist with setup reproducibility. The first HRA table was created considering mechanical failure modes of the device. Discussions with operators and manufacturers identified other failure modes and solutions. HRA directed the design towards a safe clinical device. Conclusion: A new immobilization tool has been designed using hazard-risk analysis which resulted in an easier-to-use and safer tool compared to the initial design. The remaining risks are all low probability events and not dissimilar from those currently faced with the GTC setup. Given the gains in ease of use for therapists and patients as well as the lower costs for the hospital, we will implement this new tool.

Poster - 30: Use of a Hazard-Risk Analysis for development of a new eye immobilization tool for treatment of choroidal melanoma

Energy Technology Data Exchange (ETDEWEB)

Prooijen, Monique van; Breen, Stephen [Princess Margaret Cancer Centre, Princess Margaret Hospital-UHN (Canada)

2016-08-15

Purpose: Our treatment for choroidal melanoma utilizes the GTC frame. The patient looks at a small LED to stabilize target position. The LED is attached to a metal arm attached to the GTC frame. A camera on the arm allows therapists to monitor patient compliance. To move to mask-based immobilization we need a new LED/camera attachment mechanism. We used a Hazard-Risk Analysis (HRA) to guide the design of the new tool. Method: A pre-clinical model was built with input from therapy and machine shop personnel. It consisted of an aluminum frame placed in aluminum guide posts attached to the couch top. Further development was guided by the Department of Defense Standard Practice - System Safety hazard risk analysis technique. Results: An Orfit mask was selected because it allowed access to indexes on the couch top which assist with setup reproducibility. The first HRA table was created considering mechanical failure modes of the device. Discussions with operators and manufacturers identified other failure modes and solutions. HRA directed the design towards a safe clinical device. Conclusion: A new immobilization tool has been designed using hazard-risk analysis which resulted in an easier-to-use and safer tool compared to the initial design. The remaining risks are all low probability events and not dissimilar from those currently faced with the GTC setup. Given the gains in ease of use for therapists and patients as well as the lower costs for the hospital, we will implement this new tool.

Full cost accounting in the analysis of separated waste collection efficiency: A methodological proposal.

Science.gov (United States)

D'Onza, Giuseppe; Greco, Giulio; Allegrini, Marco

2016-02-01

Recycling implies additional costs for separated municipal solid waste (MSW) collection. The aim of the present study is to propose and implement a management tool - the full cost accounting (FCA) method - to calculate the full collection costs of different types of waste. Our analysis aims for a better understanding of the difficulties of putting FCA into practice in the MSW sector. We propose a FCA methodology that uses standard cost and actual quantities to calculate the collection costs of separate and undifferentiated waste. Our methodology allows cost efficiency analysis and benchmarking, overcoming problems related to firm-specific accounting choices, earnings management policies and purchase policies. Our methodology allows benchmarking and variance analysis that can be used to identify the causes of off-standards performance and guide managers to deploy resources more efficiently. Our methodology can be implemented by companies lacking a sophisticated management accounting system. Copyright © 2015 Elsevier Ltd. All rights reserved.

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

PIXE methodology of rare earth element analysis and its applications

International Nuclear Information System (INIS)

Ma Xinpei

1992-01-01

The Proton Induced X-ray Emission (PIXE) methodology of rare earth element (REEs) analysis is discussed, including the significance of REE analysis, the principle of PIXE applied to REE, selection of characteristic X-ray for Lanthanide series elements, deconvolution of highly over lapped PIXE spectrum and minimum detection limit (MDL) of REEs. Some practical applications are presented. And the specialities of PIXE analysis to the high pure REE chemicals are discussed. (author)

submitter Methodologies for the Statistical Analysis of Memory Response to Radiation

CERN Document Server

Bosser, Alexandre L; Tsiligiannis, Georgios; Frost, Christopher D; Zadeh, Ali; Jaatinen, Jukka; Javanainen, Arto; Puchner, Helmut; Saigne, Frederic; Virtanen, Ari; Wrobel, Frederic; Dilillo, Luigi

2016-01-01

Methodologies are proposed for in-depth statistical analysis of Single Event Upset data. The motivation for using these methodologies is to obtain precise information on the intrinsic defects and weaknesses of the tested devices, and to gain insight on their failure mechanisms, at no additional cost. The case study is a 65 nm SRAM irradiated with neutrons, protons and heavy ions. This publication is an extended version of a previous study [1].

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.

Cost analysis methodology of spent fuel storage

International Nuclear Information System (INIS)

1994-01-01

The report deals with the cost analysis of interim spent fuel storage; however, it is not intended either to give a detailed cost analysis or to compare the costs of the different options. This report provides a methodology for calculating the costs of different options for interim storage of the spent fuel produced in the reactor cores. Different technical features and storage options (dry and wet, away from reactor and at reactor) are considered and the factors affecting all options defined. The major cost categories are analysed. Then the net present value of each option is calculated and the levelized cost determined. Finally, a sensitivity analysis is conducted taking into account the uncertainty in the different cost estimates. Examples of current storage practices in some countries are included in the Appendices, with description of the most relevant technical and economic aspects. 16 figs, 14 tabs

A methodology for the data energy regional consumption consistency analysis

International Nuclear Information System (INIS)

Canavarros, Otacilio Borges; Silva, Ennio Peres da

1999-01-01

The article introduces a methodology for data energy regional consumption consistency analysis. The work was going based on recent studies accomplished by several cited authors and boarded Brazilian matrices and Brazilian energetics regional balances. The results are compared and analyzed

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)

Development of Non-LOCA Safety Analysis Methodology with RETRAN-3D and VIPRE-01/K

International Nuclear Information System (INIS)

Kim, Yo-Han; Cheong, Ae-Ju; Yang, Chang-Keun

2004-01-01

Korea Electric Power Research Institute has launched a project to develop an in-house non-loss-of-coolant-accident analysis methodology to overcome the hardships caused by the narrow analytical scopes of existing methodologies. Prior to the development, some safety analysis codes were reviewed, and RETRAN-3D and VIPRE-01 were chosen as the base codes. The codes have been modified to improve the analytical capabilities required to analyze the nuclear power plants in Korea. The methodologies of the vendors and the Electric Power Research Institute have been reviewed, and some documents of foreign utilities have been used to compensate for the insufficiencies. For the next step, a draft methodology for pressurized water reactors has been developed and modified to apply to Westinghouse-type plants in Korea. To verify the feasibility of the methodology, some events of Yonggwang Units 1 and 2 have been analyzed from the standpoints of reactor coolant system pressure and the departure from nucleate boiling ratio. The results of the analyses show trends similar to those of the Final Safety Analysis Report

Uncertainty and sensitivity analysis methodology in a level-I PSA (Probabilistic Safety Assessment)

International Nuclear Information System (INIS)

Nunez McLeod, J.E.; Rivera, S.S.

1997-01-01

This work presents a methodology for sensitivity and uncertainty analysis, applicable to a probabilistic safety assessment level I. The work contents are: correct association of distributions to parameters, importance and qualification of expert opinions, generations of samples according to sample sizes, and study of the relationships among system variables and system response. A series of statistical-mathematical techniques are recommended along the development of the analysis methodology, as well different graphical visualization for the control of the study. (author) [es

APPROPRIATE ALLOCATION OF CONTINGENCY USING RISK ANALYSIS METHODOLOGY

Directory of Open Access Journals (Sweden)

Andi Andi

2004-01-01

Full Text Available Many cost overruns in the world of construction are attributable to either unforeseen events or foreseen events for which uncertainty was not appropriately accommodated. It is argued that a significant improvement to project management performance may result from greater attention to the process of analyzing project risks. The objective of this paper is to propose a risk analysis methodology for appropriate allocation of contingency in project cost estimation. In the first step, project risks will be identified. Influence diagramming technique is employed to identify and to show how the risks affect the project cost elements and also the relationships among the risks themselves. The second step is to assess the project costs with regards to the risks under consideration. Using a linguistic approach, the degree of uncertainty of identified project risks is assessed and quantified. The problem of dependency between risks is taken into consideration during this analysis. For the final step, as the main purpose of this paper, a method for allocating appropriate contingency is presented. Two types of contingencies, i.e. project contingency and management reserve are proposed to accommodate the risks. An illustrative example is presented at the end to show the application of the methodology.

Development of the fire PSA methodology and the fire analysis computer code system

International Nuclear Information System (INIS)

Katsunori, Ogura; Tomomichi, Ito; Tsuyoshi, Uchida; Yusuke, Kasagawa

2009-01-01

Fire PSA methodology has been developed and was applied to NPPs in Japan for power operation and LPSD states. CDFs of preliminary fire PSA for power operation were the higher than that of internal events. Fire propagation analysis code system (CFAST/FDS Network) was being developed and verified thru OECD-PRISME Project. Extension of the scope for LPSD state is planned to figure out the risk level. In order to figure out the fire risk level precisely, the enhancement of the methodology is planned. Verification and validation of phenomenological fire propagation analysis code (CFAST/FDS Network) in the context of Fire PSA. Enhancement of the methodology such as an application of 'Electric Circuit Analysis' in NUREG/CR-6850 and related tests in order to quantify the hot-short effect precisely. Development of seismic-induced fire PSA method being integration of existing seismic PSA and fire PSA methods is ongoing. Fire PSA will be applied to review the validity of fire prevention and mitigation measures

Similarities and differences in philanthropic and federal support for medical research in the United States: an analysis of funding by nonprofits in 2006-2008.

Science.gov (United States)

Myers, Elizabeth R; Alciati, Marianne H; Ahlport, Kathryn N; Sung, Nancy S

2012-11-01

The medical community currently has no detailed source of information on philanthropic research funding. The authors sought to identify trends in research funding by members of the Health Research Alliance (HRA), a consortium of nonprofit funders of biomedical research, and compare findings with research support from the federal government. Thirty-two HRA members uploaded information about grants with start dates in 2006, 2007, and 2008. Data were collected about each grant, investigator, and recipient institution. Disease categorization codes were assigned by a computer process similar to that used by the National Institutes of Health (NIH). In the three years under study, HRA members awarded 9,934 grants, totaling $2,712,418,254 in research and training support. Grant funding increased by 26% between 2006 and 2008. In contrast, NIH research spending increased by only 3% over the same time. Fifty-six percent of HRA grant dollars supported research projects, whereas 30% supported career development and training. During the same period, more than two-thirds of NIH grant dollars supported research projects, although NIH invested proportionally less in career development and training (7%). The largest proportion of HRA grant dollars addressed cancer, followed by diabetes and genetics. Sixty-three percent of HRA-supported investigators were men and 36% were women; 66% of investigators were white, 32% Asian, and fewer than 2% black. These results indicate that nonprofit organizations play an important role in developing careers and advancing research in significant disease areas such as cancer and diabetes, and in basic science areas such as genetics.

Methodology for Quantitative Analysis of Large Liquid Samples with Prompt Gamma Neutron Activation Analysis using Am-Be Source

International Nuclear Information System (INIS)

Idiri, Z.; Mazrou, H.; Beddek, S.; Amokrane, A.

2009-01-01

An optimized set-up for prompt gamma neutron activation analysis (PGNAA) with Am-Be source is described and used for large liquid samples analysis. A methodology for quantitative analysis is proposed: it consists on normalizing the prompt gamma count rates with thermal neutron flux measurements carried out with He-3 detector and gamma attenuation factors calculated using MCNP-5. The relative and absolute methods are considered. This methodology is then applied to the determination of cadmium in industrial phosphoric acid. The same sample is then analyzed by inductively coupled plasma (ICP) method. Our results are in good agreement with those obtained with ICP method.

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

International Nuclear Information System (INIS)

Lee, Chang Ju

2012-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Fuzzy Clustering based Methodology for Multidimensional Data Analysis in Computational Forensic Domain

OpenAIRE

Kilian Stoffel; Paul Cotofrei; Dong Han

2012-01-01

As interdisciplinary domain requiring advanced and innovative methodologies the computational forensics domain is characterized by data being simultaneously large scaled and uncertain multidimensional and approximate. Forensic domain experts trained to discover hidden pattern from crime data are limited in their analysis without the assistance of a computational intelligence approach. In this paper a methodology and an automatic procedure based on fuzzy set theory and designed to infer precis...

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

Energy Technology Data Exchange (ETDEWEB)

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

2005-11-15

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Association of Wage With Employee Participation in Health Assessments and Biometric Screening.

Science.gov (United States)

Sherman, Bruce W; Addy, Carol

2018-02-01

To understand differences in health risk assessment (HRA) and biometric screening participation rates among benefits-enrolled employees in association with wage category. Cross-sectional analysis of employee eligibility file and health benefits (wellness and claims) data. Data from self-insured employers participating in the RightOpt private exchange (Conduent HR Services) during 2014. Active employees from 4 companies continuously enrolled in health insurance for which wage data were available. Measures included HRA and biometric screening participation rates and wage status, with employee age, sex, employer, job tenure, household income, geographic location, and health benefits deductible as a percentage of total wages serving as covariates. Employees were separated into 5 groups based on wage status. Logistic regression analysis incorporated other measures as covariates to adjust for differences between groups, with HRA and biometric screening participation rates determined as binary outcomes. Participation rates for HRA and biometric screening were 90% and 87%, respectively, in the highest wage category, decreasing to 67% and 60%, respectively, among the lowest wage category. Employee wage status is associated with significant differences in HRA and biometric participation rates. Generalizing the results generated by modest participation in these offerings to entire populations may risk misinterpretation of results based on variable participation rates across wage categories.

Methodological Analysis of Gregarious Behaviour of Agents in the Financial Markets

OpenAIRE

Solodukhin Stanislav V.

2013-01-01

The article considers methodological approaches to analysis of gregarious behaviour of agents in the financial markets and also studies foundations of the agent modelling of decision making processes with consideration of the gregarious instinct.

BWR stability analysis: methodology of the stability analysis and results of PSI for the NEA/NCR benchmark task

International Nuclear Information System (INIS)

Hennig, D.; Nechvatal, L.

1996-09-01

The report describes the PSI stability analysis methodology and the validation of this methodology based on the international OECD/NEA BWR stability benchmark task. In the frame of this work, the stability properties of some operation points of the NPP Ringhals 1 have been analysed and compared with the experimental results. (author) figs., tabs., 45 refs

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

Analysis of Human Error Types and Performance Shaping Factors in the Next Generation Main Control Room

International Nuclear Information System (INIS)

Sin, Y. C.; Jung, Y. S.; Kim, K. H.; Kim, J. H.

2008-04-01

Main control room of nuclear power plants has been computerized and digitalized in new and modernized plants, as information and digital technologies make great progresses and become mature. Survey on human factors engineering issues in advanced MCRs: Model-based approach, Literature survey-based approach. Analysis of human error types and performance shaping factors is analysis of three human errors. The results of project can be used for task analysis, evaluation of human error probabilities, and analysis of performance shaping factors in the HRA analysis

Analysis Planning Methodology: For Thesis, Joint Applied Project, & MBA Research Reports

OpenAIRE

Naegle, Brad R.

2010-01-01

Acquisition Research Handbook Series Purpose: This guide provides the graduate student researcher—you—with techniques and advice on creating an effective analysis plan, and it provides methods for focusing the data-collection effort based on that analysis plan. As a side benefit, this analysis planning methodology will help you to properly scope the research effort and will provide you with insight for changes in that effort. The information presented herein was supported b...

Analysis of market competitive structure: The new methodological approach based in the using

International Nuclear Information System (INIS)

Romero de la Fuente, J.; Yague Guillen, M. J.

2007-01-01

This paper proposes a new methodological approach to identify market competitive structure, applying usage situation concept in positioning analysis. Dimensions used by consumer to classify products are identified using Correspondence Analysis and competitive groups are formed. Results are validated with Discriminant Analysis. (Author) 23 refs

Comparative analysis as a basic research orientation: Key methodological problems

Directory of Open Access Journals (Sweden)

N P Narbut

2015-12-01

Full Text Available To date, the Sociological Laboratory of the Peoples’ Friendship University of Russia has accumulated a vast experience in the field of cross-cultural studies reflected in the publications based on the results of mass surveys conducted in Moscow, Maikop, Beijing, Guangzhou, Prague, Belgrade, and Pristina. However, these publications mainly focus on the comparisons of the empirical data rather than methodological and technical issues, that is why the aim of this article is to identify key problems of the comparative analysis in cross-cultural studies that become evident only if you conduct an empirical research yourself - from the first step of setting the problem and approving it by all the sides (countries involved to the last step of interpreting and comparing the data obtained. The authors are sure that no sociologist would ever doubt the necessity and importance of comparative analysis in the broadest sense of the word, but at the same time very few are ready to discuss its key methodological challenges and prefer to ignore them completely. We summarize problems of the comparative analysis in sociology as follows: (1 applying research techniques to the sample in another country - both in translating and adapting them to different social realities and worldview (in particular, the problematic status of standardization and qualitative approach; (2 choosing “right” respondents to question and relevant cases (cultures to study; (3 designing the research scheme, i.e. justifying the sequence of steps (what should go first - methodology or techniques; (4 accepting the procedures that are correct within one country for cross-cultural work (whether or not that is an appropriate choice.

PROBLEMS AND METHODOLOGY OF THE PETROLOGIC ANALYSIS OF COAL FACIES.

Science.gov (United States)

Chao, Edward C.T.

1983-01-01

This condensed synthesis gives a broad outline of the methodology of coal facies analysis, procedures for constructing sedimentation and geochemical formation curves, and micro- and macrostratigraphic analysis. The hypothetical coal bed profile has a 3-fold cycle of material characteristics. Based on studies of other similar profiles of the same coal bed, and on field studies of the sedimentary rock types and their facies interpretation, one can assume that the 3-fold subdivision is of regional significance.

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.

A Systematic Review of Brief Functional Analysis Methodology with Typically Developing Children

Science.gov (United States)

Gardner, Andrew W.; Spencer, Trina D.; Boelter, Eric W.; DuBard, Melanie; Jennett, Heather K.

2012-01-01

Brief functional analysis (BFA) is an abbreviated assessment methodology derived from traditional extended functional analysis methods. BFAs are often conducted when time constraints in clinics, schools or homes are of concern. While BFAs have been used extensively to identify the function of problem behavior for children with disabilities, their…

Predictive Ability of Pender's Health Promotion Model for Physical Activity and Exercise in People with Spinal Cord Injuries: A Hierarchical Regression Analysis

Science.gov (United States)

Keegan, John P.; Chan, Fong; Ditchman, Nicole; Chiu, Chung-Yi

2012-01-01

The main objective of this study was to validate Pender's Health Promotion Model (HPM) as a motivational model for exercise/physical activity self-management for people with spinal cord injuries (SCIs). Quantitative descriptive research design using hierarchical regression analysis (HRA) was used. A total of 126 individuals with SCI were recruited…

Trial application of a technique for human error analysis (ATHEANA)

International Nuclear Information System (INIS)

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

1996-01-01

The new method for HRA, ATHEANA, has been developed based on a study of the operating history of serious accidents and an understanding of the reasons why people make errors. Previous publications associated with the project have dealt with the theoretical framework under which errors occur and the retrospective analysis of operational events. This is the first attempt to use ATHEANA in a prospective way, to select and evaluate human errors within the PSA context

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

3-D rod ejection analysis using a conservative methodology

Energy Technology Data Exchange (ETDEWEB)

Park, Min Ho; Park, Jin Woo; Park, Guen Tae; Um, Kil Sup; Ryu, Seok Hee; Lee, Jae Il; Choi, Tong Soo [KEPCO, Daejeon (Korea, Republic of)

2016-05-15

The point kinetics model which simplifies the core phenomena and physical specifications is used for the conventional rod ejection accident analysis. The point kinetics model is convenient to assume conservative core parameters but this simplification loses large amount of safety margin. The CHASER system couples the three-dimensional core neutron kinetics code ASTRA, the sub-channel analysis code THALES and the fuel performance analysis code FROST. The validation study for the CHASER system is addressed using the NEACRP three-dimensional PWR core transient benchmark problem. A series of conservative rod ejection analyses for the APR1400 type plant is performed for both hot full power (HFP) and hot zero power (HZP) conditions to determine the most limiting cases. The conservative rod ejection analysis methodology is designed to properly consider important phenomena and physical parameters.

Methodological challenges in qualitative content analysis: A discussion paper.

Science.gov (United States)

Graneheim, Ulla H; Lindgren, Britt-Marie; Lundman, Berit

2017-09-01

This discussion paper is aimed to map content analysis in the qualitative paradigm and explore common methodological challenges. We discuss phenomenological descriptions of manifest content and hermeneutical interpretations of latent content. We demonstrate inductive, deductive, and abductive approaches to qualitative content analysis, and elaborate on the level of abstraction and degree of interpretation used in constructing categories, descriptive themes, and themes of meaning. With increased abstraction and interpretation comes an increased challenge to demonstrate the credibility and authenticity of the analysis. A key issue is to show the logic in how categories and themes are abstracted, interpreted, and connected to the aim and to each other. Qualitative content analysis is an autonomous method and can be used at varying levels of abstraction and interpretation. Copyright © 2017 Elsevier Ltd. All rights reserved.

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

A Local Approach Methodology for the Analysis of Ultimate Strength ...

African Journals Online (AJOL)

The local approach methodology in contrast to classical fracture mechanics can be used to predict the onset of tearing fracture, and the effects of geometry in tubular joints. Finite element analysis of T-joints plate geometries, and tubular joints has been done. The parameters of constraint, equivalent stress, plastic strain and ...

Human Schedule Performance, Protocol Analysis, and the "Silent Dog" Methodology

Science.gov (United States)

Cabello, Francisco; Luciano, Carmen; Gomez, Inmaculada; Barnes-Holmes, Dermot

2004-01-01

The purpose of the current experiment was to investigate the role of private verbal behavior on the operant performances of human adults, using a protocol analysis procedure with additional methodological controls (the "silent dog" method). Twelve subjects were exposed to fixed ratio 8 and differential reinforcement of low rate 3-s schedules. For…

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-07-01

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

Methodological Choices in Muscle Synergy Analysis Impact Differentiation of Physiological Characteristics Following Stroke

Directory of Open Access Journals (Sweden)

Caitlin L. Banks

2017-08-01

Full Text Available Muscle synergy analysis (MSA is a mathematical technique that reduces the dimensionality of electromyographic (EMG data. Used increasingly in biomechanics research, MSA requires methodological choices at each stage of the analysis. Differences in methodological steps affect the overall outcome, making it difficult to compare results across studies. We applied MSA to EMG data collected from individuals post-stroke identified as either responders (RES or non-responders (nRES on the basis of a critical post-treatment increase in walking speed. Importantly, no clinical or functional indicators identified differences between the cohort of RES and nRES at baseline. For this exploratory study, we selected the five highest RES and five lowest nRES available from a larger sample. Our goal was to assess how the methodological choices made before, during, and after MSA affect the ability to differentiate two groups with intrinsic physiologic differences based on MSA results. We investigated 30 variations in MSA methodology to determine which choices allowed differentiation of RES from nRES at baseline. Trial-to-trial variability in time-independent synergy vectors (SVs and time-varying neural commands (NCs were measured as a function of: (1 number of synergies computed; (2 EMG normalization method before MSA; (3 whether SVs were held constant across trials or allowed to vary during MSA; and (4 synergy analysis output normalization method after MSA. MSA methodology had a strong effect on our ability to differentiate RES from nRES at baseline. Across all 10 individuals and MSA variations, two synergies were needed to reach an average of 90% variance accounted for (VAF. Based on effect sizes, differences in SV and NC variability between groups were greatest using two synergies with SVs that varied from trial-to-trial. Differences in SV variability were clearest using unit magnitude per trial EMG normalization, while NC variability was less sensitive to EMG

Analysis of gaming community using Soft System Methodology

OpenAIRE

Hurych, Jan

2015-01-01

This diploma thesis aims to analyse virtual gaming community and it's problems in case of community belonging to EU server of the game called World of Tanks. To solve these problems, Soft System Methodology by P. Checkland, is used. The thesis includes analysis of significance of gaming communities for the gaming industry as a whole. Gaming community is then defined as a soft system. There are 3 problems analysed in the practical part of the thesis using newer version of SSM. One iteration of...

Proposal of methodology of tsunami accident sequence analysis induced by earthquake using DQFM methodology

International Nuclear Information System (INIS)

Muta, Hitoshi; Muramatsu, Ken

2017-01-01

Since the Fukushima-Daiichi nuclear power station accident, the Japanese regulatory body has improved and upgraded the regulation of nuclear power plants, and continuous effort is required to enhance risk management in the mid- to long term. Earthquakes and tsunamis are considered as the most important risks, and the establishment of probabilistic risk assessment (PRA) methodologies for these events is a major issue of current PRA. The Nuclear Regulation Authority (NRA) addressed the PRA methodology for tsunamis induced by earthquakes, which is one of the methodologies that should be enhanced step by step for the improvement and maturity of PRA techniques. The AESJ standard for the procedure of seismic PRA for nuclear power plants in 2015 provides the basic concept of the methodology; however, details of the application to the actual plant PRA model have not been sufficiently provided. This study proposes a detailed PRA methodology for tsunamis induced by earthquakes using the DQFM methodology, which contributes to improving the safety of nuclear power plants. Furthermore, this study also states the issues which need more research. (author)

Applications of a methodology for the analysis of learning trends in nuclear power plants

International Nuclear Information System (INIS)

Cho, Hang Youn; Choi, Sung Nam; Yun, Won Yong

1995-01-01

A methodology is applied to identify the learning trend related to the safety and availability of U.S. commercial nuclear power plants. The application is intended to aid in reducing likelihood of human errors. To assure that the methodology can be easily adapted to various types of classification schemes of operation data, a data bank classified by the Transient Analysis Classification and Evaluation(TRACE) scheme is selected for the methodology. The significance criteria for human-initiated events affecting the systems and for events caused by human deficiencies were used. Clustering analysis was used to identify the learning trend in multi-dimensional histograms. A computer code is developed based on the K-Means algorithm and applied to find the learning period in which error rates are monotonously decreasing with plant age

Methodology for the analysis of self-tensioned wooden structural floors

Directory of Open Access Journals (Sweden)

F. Suárez-Riestra

2017-09-01

Full Text Available It is described a self-tensioning system constituted by a force multiplying device which, attached to the supports of the ends of the structural element, is able to convert the vertical resultant from the gravitatonial actions into an effective tensioning action, through the movement that was induced by a set of rods. The self-tensioning system is able to offer a high performance, thanks to the beneficial effect of the opposite deflection generated by the tensioning, in proportion to the increasing of the gravitational action. This allows to design long-span timber ribbed floors using reduced depths. The complexity of calculation due to the non-linearity of the system can be obviated with the methodology of analysis developed in the article. In order to illustrate the advantages of the self-tensioning system and the methodology of analysis which were developed, six cases of ribbed floors have been analysed, with spans of 9, 12 and 15 m and variable using loads of 3,00 kN/m2 and 5,00 kN/m2.

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

International Nuclear Information System (INIS)

Jafarian, Reza; Sepanloo, Kamran

2006-01-01

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

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

International Nuclear Information System (INIS)

Jafarian, R.; Sepanloo, K.

2005-01-01

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

A framework for characterizing usability requirements elicitation and analysis methodologies (UREAM)

NARCIS (Netherlands)

Trienekens, J.J.M.; Kusters, R.J.; Mannaert, H.

2012-01-01

Dedicated methodologies for the elicitation and analysis of usability requirements have been proposed in literature, usually developed by usability experts. The usability of these approaches by non-expert software engineers is not obvious. In this paper, the objective is to support developers and

Methodology of a PWR containment analysis during a thermal-hydraulic accident

Energy Technology Data Exchange (ETDEWEB)

Silva, Dayane F.; Sabundjian, Gaiane; Lima, Ana Cecilia S., E-mail: dayane.silva@usp.br, E-mail: gdjian@ipen.br, E-mail: aclima@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

The aim of this work is to present the methodology of calculation to Angra 2 reactor containment during accidents of the type Loss of Coolant Accident (LOCA). This study will be possible to ensure the safety of the population of the surroundings upon the occurrence of accidents. One of the programs used to analyze containment of a nuclear plant is the CONTAIN. This computer code is an analysis tool used for predicting the physical conditions and distributions of radionuclides inside a containment building following the release of material from the primary system in a light-water reactor during an accident. The containment of the type PWR plant is a concrete building covered internally by metallic material and has limits of design pressure. The methodology of containment analysis must estimate the limits of pressure during a LOCA. The boundary conditions for the simulation are obtained from RELAP5 code. (author)

Methodology of a PWR containment analysis during a thermal-hydraulic accident

International Nuclear Information System (INIS)

Silva, Dayane F.; Sabundjian, Gaiane; Lima, Ana Cecilia S.

2015-01-01

The aim of this work is to present the methodology of calculation to Angra 2 reactor containment during accidents of the type Loss of Coolant Accident (LOCA). This study will be possible to ensure the safety of the population of the surroundings upon the occurrence of accidents. One of the programs used to analyze containment of a nuclear plant is the CONTAIN. This computer code is an analysis tool used for predicting the physical conditions and distributions of radionuclides inside a containment building following the release of material from the primary system in a light-water reactor during an accident. The containment of the type PWR plant is a concrete building covered internally by metallic material and has limits of design pressure. The methodology of containment analysis must estimate the limits of pressure during a LOCA. The boundary conditions for the simulation are obtained from RELAP5 code. (author)

Effects of methodology and analysis strategy on robustness of pestivirus phylogeny.

Science.gov (United States)

Liu, Lihong; Xia, Hongyan; Baule, Claudia; Belák, Sándor; Wahlberg, Niklas

2010-01-01

Phylogenetic analysis of pestiviruses is a useful tool for classifying novel pestiviruses and for revealing their phylogenetic relationships. In this study, robustness of pestivirus phylogenies has been compared by analyses of the 5'UTR, and complete N(pro) and E2 gene regions separately and combined, performed by four methods: neighbour-joining (NJ), maximum parsimony (MP), maximum likelihood (ML), and Bayesian inference (BI). The strategy of analysing the combined sequence dataset by BI, ML, and MP methods resulted in a single, well-supported tree topology, indicating a reliable and robust pestivirus phylogeny. By contrast, the single-gene analysis strategy resulted in 12 trees of different topologies, revealing different relationships among pestiviruses. These results indicate that the strategies and methodologies are two vital aspects affecting the robustness of the pestivirus phylogeny. The strategy and methodologies outlined in this paper may have a broader application in inferring phylogeny of other RNA viruses.

Draft Hanford Remedial Action Environmental Impact Statement and Comprehensive Land Use Plan. Volume 4 of 4

International Nuclear Information System (INIS)

1996-08-01

The US Department of Energy's (DOE) is preparing this ''Hanford Site Comprehensive Land Use Plan'' (Comprehensive Plan), Appendix M to address future land uses for the Hanford Site. The DOE has integrated this land-use planning initiative with the development of the HRA-EIS to facilitate and expedite land-use and remediation decision making, reduce time and cost of remediation, and optimize the usefulness of the planning process. The HRA-EIS is being developed to evaluate the potential environmental impacts associated with remediation, create a remedial baseline for the Environmental Restoration Program, and provide a framework for future uses at the Hanford Site. This Comprehensive Plan identifies current assets and resources related to land-use planning, and provides the analysis and recommendations for future land sues and accompanying restrictions at the Hanford Site over a 50-year period. This Comprehensive Plan relies on the analysis of environmental impacts in the HRA-EIS. The National Environmental Policy Act of 1969 (NEPA) Record of Decision (ROD) issued for the HRA-EIS will be the decision process for finalization and adoption of this Comprehensive Plan. The HRA-EIS and this Comprehensive Plan will provide a basis for remediation decisions to be identified and contained in site- and area-specific Comprehensive Environmental Response, Compensation and Liability Act of 1980 ROD

A new approach to incorporate operator actions in the simulation of accident sequences

International Nuclear Information System (INIS)

Antonio Exposito; Juan Antonio Quiroga; Javier Hortal; John-Einar Hulsund

2006-01-01

Full text of publication follows: Nowadays, simulation-based human reliability analysis (HRA) methods seem to provide a new direction for the development of advanced methodologies to study operator actions effect during accident sequences. Due to this, the Spanish Nuclear Safety Council (CSN) started a working group which has, among other objectives, to develop such simulation-based HRA methodology. As a result of its activities, a new methodology, named Integrated Safety Assessment (ISA), has been developed and is currently being incorporated into licensing activities at CSN. One of the key aspects of this approach is the incorporation of the capability to simulate operator actions, expanding the ISA methodology scopes to make HRA studies. For this reason, CSN is involved in several activities oriented to develop a new tool, which must be able to incorporate operator actions in conventional thermohydraulic (TH) simulations. One of them is the collaboration project between CSN, Halden Reactor Project (HRP) and the Department of Energy Systems (DSE) of the Polytechnic University of Madrid that started in 2003. The basic aim of the project is to develop a software tool that consists of a closed-loop plant/operator simulator, a thermal hydraulic (TH) code for simulating the plant transient and the procedures processor to give the information related with operator actions to the TH code, both coupled by a data communication system which allows the information exchange. For the plant simulation we have a plant transient simulator code (TRETA/TIZONA for PWR/BWR NPPs respectively), developed by the CSN, with PWR/BWR full scope models. The functionality of these thermalhydraulic codes has been expanded, allowing control the overall information flow between coupled codes, simulating the TH transient and determining when the operator actions must be considered. In the other hand, we have the COPMA-III code, a computerized procedure system able to manage XML operational

Human Error Analysis by Fuzzy-Set

International Nuclear Information System (INIS)

Situmorang, Johnny

1996-01-01

In conventional HRA the probability of Error is treated as a single and exact value through constructing even tree, but in this moment the Fuzzy-Set Theory is used. Fuzzy set theory treat the probability of error as a plausibility which illustrate a linguistic variable. Most parameter or variable in human engineering been defined verbal good, fairly good, worst etc. Which describe a range of any value of probability. For example this analysis is quantified the human error in calibration task, and the probability of miscalibration is very low

Development of a heat exchanger root-cause analysis methodology

International Nuclear Information System (INIS)

Jarrel, D.B.

1989-01-01

The objective of this work is to determine a generic methodology for approaching the accurate identification of the root cause of component failure. Root-cause determinations are an everyday challenge to plant personnel, but they are handled with widely differing degrees of success due to the approaches, levels of diagnostic expertise, and documentation. The criterion for success is simple: If the root cause of the failure has truly been determined and corrected, the same causal failure relationship will not be demonstrated again in the future. The approach to root-cause analysis (RCA) element definition was to first selectively choose and constrain a functionally significant component (in this case a component cooling water to service water heat exchanger) that has demonstrated prevalent failures. Then a root cause of failure analysis was performed by a systems engineer on a large number of actual failure scenarios. The analytical process used by the engineer was documented and evaluated to abstract the logic model used to arrive at the root cause. For the case of the heat exchanger, the actual root-cause diagnostic approach is described. A generic methodology for the solution of the root cause of component failure is demonstrable for this general heat exchanger sample

Methodology of the Integrated Analysis of Company's Financial Status and Its Performance Results

OpenAIRE

Mackevičius, Jonas; Valkauskas, Romualdas

2010-01-01

Information about company's financial status and its performance results is very important for the objective evaluation of company's position in the market and competitive possibilities in the future. Such information is provided in the financial statement. It is important to apply and investigate this information properly. The methodology of company's financial status and performance results integrated analysis is recommended in this article. This methodology consists of these three elements...

Analysis of Feedback processes in Online Group Interaction: a methodological model

Directory of Open Access Journals (Sweden)

Anna Espasa

2013-06-01

Full Text Available The aim of this article is to present a methodological model to analyze students' group interaction to improve their essays in online learning environments, based on asynchronous and written communication. In these environments teacher and student scaffolds for discussion are essential to promote interaction. One of these scaffolds can be the feedback. Research on feedback processes has predominantly focused on feedback design rather than on how students utilize feedback to improve learning. This methodological model fills this gap contributing to analyse the implementation of the feedback processes while students discuss collaboratively in a specific case of writing assignments. A review of different methodological models was carried out to define a framework adjusted to the analysis of the relationship of written and asynchronous group interaction, and students' activity and changes incorporated into the final text. The model proposed includes the following dimensions: 1 student participation 2 nature of student learning and 3 quality of student learning. The main contribution of this article is to present the methodological model and also to ascertain the model's operativity regarding how students incorporate such feedback into their essays.

Applying rigorous decision analysis methodology to optimization of a tertiary recovery project

International Nuclear Information System (INIS)

Wackowski, R.K.; Stevens, C.E.; Masoner, L.O.; Attanucci, V.; Larson, J.L.; Aslesen, K.S.

1992-01-01

This paper reports that the intent of this study was to rigorously look at all of the possible expansion, investment, operational, and CO 2 purchase/recompression scenarios (over 2500) to yield a strategy that would maximize net present value of the CO 2 project at the Rangely Weber Sand Unit. Traditional methods of project management, which involve analyzing large numbers of single case economic evaluations, was found to be too cumbersome and inaccurate for an analysis of this scope. The decision analysis methodology utilized a statistical approach which resulted in a range of economic outcomes. Advantages of the decision analysis methodology included: a more organized approach to classification of decisions and uncertainties; a clear sensitivity method to identify the key uncertainties; an application of probabilistic analysis through the decision tree; and a comprehensive display of the range of possible outcomes for communication to decision makers. This range made it possible to consider the upside and downside potential of the options and to weight these against the Unit's strategies. Savings in time and manpower required to complete the study were also realized

Discovering the Effects-Endstate Linkage: Using Soft Systems Methodology to Perform EBO Mission Analysis

National Research Council Canada - National Science Library

Young, Jr, William E

2005-01-01

.... EBO mission analysis is shown to be more problem structuring than problem solving. A new mission analysis process is proposed using a modified version of Soft Systems Methodology to meet these challenges...

ASSESSMENT OF SEISMIC ANALYSIS METHODOLOGIES FOR DEEPLY EMBEDDED NPP STRUCTURES

International Nuclear Information System (INIS)

XU, J.; MILLER, C.; COSTANTINO, C.; HOFMAYER, C.; GRAVES, H. NRC.

2005-01-01

Several of the new generation nuclear power plant designs have structural configurations which are proposed to be deeply embedded. Since current seismic analysis methodologies have been applied to shallow embedded structures (e.g., ASCE 4 suggest that simple formulations may be used to model embedment effect when the depth of embedment is less than 30% of its foundation radius), the US Nuclear Regulatory Commission is sponsoring a program at the Brookhaven National Laboratory with the objective of investigating the extent to which procedures acceptable for shallow embedment depths are adequate for larger embedment depths. This paper presents the results of a study comparing the response spectra obtained from two of the more popular analysis methods for structural configurations varying from shallow embedment to complete embedment. A typical safety related structure embedded in a soil profile representative of a typical nuclear power plant site was utilized in the study and the depths of burial (DOB) considered range from 25-100% the height of the structure. Included in the paper are: (1) the description of a simplified analysis and a detailed approach for the SSI analyses of a structure with various DOB, (2) the comparison of the analysis results for the different DOBs between the two methods, and (3) the performance assessment of the analysis methodologies for SSI analyses of deeply embedded structures. The resulting assessment from this study has indicated that simplified methods may be capable of capturing the seismic response for much deeper embedded structures than would be normally allowed by the standard practice

Interpretive Phenomenological Analysis: An Appropriate Methodology for Educational Research?

Directory of Open Access Journals (Sweden)

Edward John Noon

2018-04-01

Full Text Available Interpretive phenomenological analysis (IPA is a contemporary qualitative methodology, first developed by psychologist Jonathan Smith (1996. Whilst its roots are in psychology, it is increasingly being drawn upon by scholars in the human, social and health sciences (Charlick, Pincombe, McKellar, & Fielder, 2016. Despite this, IPA has received limited attention across educationalist literature. Drawing upon my experiences of using IPA to explore the barriers to the use of humour in the teaching of Childhood Studies (Noon, 2017, this paper will discuss its theoretical orientation, sampling and methods of data collection and analysis, before examining the strengths and weaknesses to IPA’s employment in educational research.

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

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

International Nuclear Information System (INIS)

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

2010-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-06-01

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

Methodology for national risk analysis and prioritization of toxic industrial chemicals.

Science.gov (United States)

Taxell, Piia; Engström, Kerstin; Tuovila, Juha; Söderström, Martin; Kiljunen, Harri; Vanninen, Paula; Santonen, Tiina

2013-01-01

The identification of chemicals that pose the greatest threat to human health from incidental releases is a cornerstone in public health preparedness for chemical threats. The present study developed and applied a methodology for the risk analysis and prioritization of industrial chemicals to identify the most significant chemicals that pose a threat to public health in Finland. The prioritization criteria included acute and chronic health hazards, physicochemical and environmental hazards, national production and use quantities, the physicochemical properties of the substances, and the history of substance-related incidents. The presented methodology enabled a systematic review and prioritization of industrial chemicals for the purpose of national public health preparedness for chemical incidents.

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)

Understanding information exchange during disaster response: Methodological insights from infocentric analysis

Science.gov (United States)

Toddi A. Steelman; Branda Nowell; Deena. Bayoumi; Sarah. McCaffrey

2014-01-01

We leverage economic theory, network theory, and social network analytical techniques to bring greater conceptual and methodological rigor to understand how information is exchanged during disasters. We ask, "How can information relationships be evaluated more systematically during a disaster response?" "Infocentric analysis"—a term and...

Snapshot analysis for rhodium fixed incore detector using BEACON methodology

International Nuclear Information System (INIS)

Cha, Kyoon Ho; Choi, Yu Sun; Lee, Eun Ki; Park, Moon Ghu; Morita, Toshio; Heibel, Michael D.

2004-01-01

The purpose of this report is to process the rhodium detector data of the Yonggwang nuclear unit 4 cycle 5 core for the measured power distribution by using the BEACON methodology. Rhodium snapshots of the YGN 4 cycle 5 have been analyzed by both BEACON/SPINOVA and CECOR to compare the results of both codes. By analyzing a large number of snapshots obtained during normal plant operation. Reviewing the results of this analysis, the BEACON/SPNOVA can be used for the snapshot analysis of Korean Standard Nuclear Power (KSNP) plants

A methodology for uncertainty analysis of reference equations of state

DEFF Research Database (Denmark)

Cheung, Howard; Frutiger, Jerome; Bell, Ian H.

We present a detailed methodology for the uncertainty analysis of reference equations of state (EOS) based on Helmholtz energy. In recent years there has been an increased interest in uncertainties of property data and process models of thermal systems. In the literature there are various...... for uncertainty analysis is suggested as a tool for EOS. The uncertainties of the EOS properties are calculated from the experimental values and the EOS model structure through the parameter covariance matrix and subsequent linear error propagation. This allows reporting the uncertainty range (95% confidence...

Evaluation of end-user satisfaction among employees participating in a web-based health risk assessment with tailored feedback.

Science.gov (United States)

Vosbergen, Sandra; Laan, Eva K; Colkesen, Ersen B; Niessen, Maurice A J; Kraaijenhagen, Roderik A; Essink-Bot, Marie-Louise; Peek, Niels

2012-10-30

Web technology is increasingly being used to provide individuals with health risk assessments (HRAs) with tailored feedback. End-user satisfaction is an important determinant of the potential impact of HRAs, as this influences program attrition and adherence to behavioral advice. The aim of this study was to evaluate end-user satisfaction with a web-based HRA with tailored feedback applied in worksite settings, using mixed (quantitative and qualitative) methods. Employees of seven companies in the Netherlands participated in a commercial, web-based, HRA with tailored feedback. The HRA consisted of four components: 1) a health and lifestyle assessment questionnaire, 2) a biometric evaluation, 3) a laboratory evaluation, and 4) tailored feedback consisting of a personal health risk profile and lifestyle behavior advice communicated through a web portal. HRA respondents received an evaluation questionnaire after six weeks. Satisfaction with different parts of the HRA was measured on 5-point Likert scales. A free-text field provided the opportunity to make additional comments. In total, 2289 employees participated in the HRA program, of which 637 (27.8%) completed the evaluation questionnaire. Quantitative analysis showed that 85.6% of the respondents evaluated the overall HRA positively. The free-text field was filled in by 29.7 % of the respondents (189 out of 637), who made 315 separate remarks. Qualitative evaluation of these data showed that these respondents made critical remarks. Respondents felt restricted by the answer categories of the health and lifestyle assessment questionnaire, which resulted in the feeling that the corresponding feedback could be inadequate. Some respondents perceived the personal risk profile as unnecessarily alarming or suggested providing more explanations, reference values, and a justification of the behavioral advice given. Respondents also requested the opportunity to discuss the feedback with a health professional. Most people

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

Discourse analysis: A useful methodology for health-care system researches.

Science.gov (United States)

Yazdannik, Ahmadreza; Yousefy, Alireza; Mohammadi, Sepideh

2017-01-01

Discourse analysis (DA) is an interdisciplinary field of inquiry and becoming an increasingly popular research strategy for researchers in various disciplines which has been little employed by health-care researchers. The methodology involves a focus on the sociocultural and political context in which text and talk occur. DA adds a linguistic approach to an understanding of the relationship between language and ideology, exploring the way in which theories of reality and relations of power are encoded in such aspects as the syntax, style, and rhetorical devices used in texts. DA is a useful and productive qualitative methodology but has been underutilized within health-care system research. Without a clear understanding of discourse theory and DA it is difficult to comprehend important research findings and impossible to use DA as a research strategy. To redress this deficiency, in this article, represents an introduction to concepts of discourse and DA, DA history, Philosophical background, DA types and analysis strategy. Finally, we discuss how affect to the ideological dimension of such phenomena discourse in health-care system, health beliefs and intra-disciplinary relationship in health-care system.

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Summary of the Supplemental Model Reports Supporting the Disposal Criticality Analysis Methodology Topical Report

International Nuclear Information System (INIS)

Brownson, D. A.

2002-01-01

The Department of Energy (DOE) Office of Civilian Radioactive Waste Management (OCRWM) has committed to a series of model reports documenting the methodology to be utilized in the Disposal Criticality Analysis Methodology Topical Report (YMP 2000). These model reports detail and provide validation of the methodology to be utilized for criticality analyses related to: (1) Waste form/waste package degradation; (2) Waste package isotopic inventory; (3) Criticality potential of degraded waste form/waste package configurations (effective neutron multiplication factor); (4) Probability of criticality (for each potential critical configuration as well as total event); and (5) Criticality consequences. This purpose of this summary report is to provide a status of the model reports and a schedule for their completion. This report also provides information relative to the model report content and validation. The model reports and their revisions are being generated as a result of: (1) Commitments made in the Disposal Criticality Analysis Methodology Topical Report (YMP 2000); (2) Open Items from the Safety Evaluation Report (Reamer 2000); (3) Key Technical Issue agreements made during DOE/U.S. Nuclear Regulatory Commission (NRC) Technical Exchange Meeting (Reamer and Williams 2000); and (4) NRC requests for additional information (Schlueter 2002)

A SAS2H/KENO-V methodology for 3D fuel burnup analysis

International Nuclear Information System (INIS)

Milosevic, M.; Greenspan, E.; Vujic, J.

2002-01-01

An efficient methodology for 3D fuel burnup analysis of LWR reactors is described in this paper. This methodology is founded on coupling Monte Carlo method for 3D calculation of node power distribution, and transport method for depletion calculation in ID Wigner-Seitz equivalent cell for each node independently. The proposed fuel burnup modeling, based on application of SCALE-4.4a control modules SAS2H and KENO-V.a is verified for the case of 2D x-y model of IRIS 15 x 15 fuel assembly (with reflective boundary condition) by using two well benchmarked code systems. The one is MOCUP, a coupled MCNP-4C and ORIGEN2.1 utility code, and the second is KENO-V.a/ORIGEN2.1 code system recently developed by authors of this paper. The proposed SAS2H/KENO-V.a methodology was applied for 3D burnup analysis of IRIS-1000 benchmark.44 core. Detailed k sub e sub f sub f and power density evolution with burnup are reported. (author)

Development of methodology for the analysis of fuel behavior in light water reactor in design basis accidents

International Nuclear Information System (INIS)

Salatov, A. A.; Goncharov, A. A.; Eremenko, A. S.; Kuznetsov, V. I.; Bolnov, V. A.; Gusev, A. S.; Dolgov, A. B.; Ugryumov, A. V.

2013-01-01

The report attempts to analyze the current experience of the safety fuel for light-water reactors (LWRs) under design-basis accident conditions in terms of its compliance with international requirements for licensing nuclear power plants. The components of fuel behavior analysis methodology in design basis accidents in LWRs were considered, such as classification of design basis accidents, phenomenology of fuel behavior in design basis accidents, system of fuel safety criteria and their experimental support, applicability of used computer codes and input data for computational analysis of the fuel behavior in accidents, way of accounting for the uncertainty of calculation models and the input data. A brief history of the development of probabilistic safety analysis methodology for nuclear power plants abroad is considered. The examples of a conservative approach to safety analysis of VVER fuel and probabilistic approach to safety analysis of fuel TVS-K are performed. Actual problems in development of the methodology of analyzing the behavior of VVER fuel at the design basis accident conditions consist, according to the authors opinion, in following: 1) Development of a common methodology for analyzing the behavior of VVER fuel in the design basis accidents, implementing a realistic approach to the analysis of uncertainty - in the future it is necessary for the licensing of operating VVER fuel abroad; 2) Experimental and analytical support to the methodology: experimental studies to identify and study the characteristics of the key uncertainties of computational models of fuel and the cladding, development of computational models of key events in codes, validation code on the basis of integral experiments

Methodology for thermal hydraulic conceptual design and performance analysis of KALIMER core

International Nuclear Information System (INIS)

Young-Gyun Kim; Won-Seok Kim; Young-Jin Kim; Chang-Kue Park

2000-01-01

This paper summarizes the methodology for thermal hydraulic conceptual design and performance analysis which is used for KALIMER core, especially the preliminary methodology for flow grouping and peak pin temperature calculation in detail. And the major technical results of the conceptual design for the KALIMER 98.03 core was shown and compared with those of KALIMER 97.07 design core. The KALIMER 98.03 design core is proved to be more optimized compared to the 97.07 design core. The number of flow groups are reduced from 16 to 11, and the equalized peak cladding midwall temperature from 654 deg. C to 628 deg. C. It was achieved from the nuclear and thermal hydraulic design optimization study, i.e. core power flattening and increase of radial blanket power fraction. Coolant flow distribution to the assemblies and core coolant/component temperatures should be determined in core thermal hydraulic analysis. Sodium flow is distributed to core assemblies with the overall goal of equalizing the peak cladding midwall temperatures for the peak temperature pin of each bundle, thus pin cladding damage accumulation and pin reliability. The flow grouping and the peak pin temperature calculation for the preliminary conceptual design is performed with the modules ORFCE-F60 and ORFCE-T60 respectively. The basic subchannel analysis will be performed with the SLTHEN code, and the detailed subchannel analysis will be done with the MATRA-LMR code which is under development for the K-Core system. This methodology was proved practical to KALIMER core thermal hydraulic design from the related benchmark calculation studies, and it is used to KALIMER core thermal hydraulic conceptual design. (author)

The Nuclear Organization and Management Analysis Concept methodology: Four years later

International Nuclear Information System (INIS)

Haber, S.B.; Shurberg, D.A.; Barriere, M.T.; Hall, R.E.

1992-01-01

The Nuclear Organization and Management Analysis Concept was first presented at the IEEE Human Factors meeting in Monterey in 1988. In the four years since that paper, the concept and its associated methodology has been demonstrated at two commercial nuclear power plants (NPP) and one fossil power plant. In addition, applications of some of the methods have been utilized in other types of organizations, and products are being developed from the insights obtained using the concept for various organization and management activities. This paper will focus on the insights and results obtained from the two demonstration studies at the commercial NPPs. The results emphasize the utility of the methodology and the comparability of the results from the two organizations

Methodologies for risk analysis in slope instability

International Nuclear Information System (INIS)

Bernabeu Garcia, M.; Diaz Torres, J. A.

2014-01-01

This paper is an approach to the different methodologies used in conducting landslide risk maps so that the reader can get a basic knowledge about how to proceed in its development. The landslide hazard maps are increasingly demanded by governments. This is because due to climate change, deforestation and the pressure exerted by the growth of urban centers, damage caused by natural phenomena is increasing each year, making this area of work a field of study with increasing importance. To explain the process of mapping a journey through each of the phases of which it is composed is made: from the study of the types of slope movements and the necessary management of geographic information systems (GIS) inventories and landslide susceptibility analysis, threat, vulnerability and risk. (Author)

Reporting and methodological quality of survival analysis in articles published in Chinese oncology journals.

Science.gov (United States)

Zhu, Xiaoyan; Zhou, Xiaobin; Zhang, Yuan; Sun, Xiao; Liu, Haihua; Zhang, Yingying

2017-12-01

Survival analysis methods have gained widespread use in the filed of oncology. For achievement of reliable results, the methodological process and report quality is crucial. This review provides the first examination of methodological characteristics and reporting quality of survival analysis in articles published in leading Chinese oncology journals.To examine methodological and reporting quality of survival analysis, to identify some common deficiencies, to desirable precautions in the analysis, and relate advice for authors, readers, and editors.A total of 242 survival analysis articles were included to be evaluated from 1492 articles published in 4 leading Chinese oncology journals in 2013. Articles were evaluated according to 16 established items for proper use and reporting of survival analysis.The application rates of Kaplan-Meier, life table, log-rank test, Breslow test, and Cox proportional hazards model (Cox model) were 91.74%, 3.72%, 78.51%, 0.41%, and 46.28%, respectively, no article used the parametric method for survival analysis. Multivariate Cox model was conducted in 112 articles (46.28%). Follow-up rates were mentioned in 155 articles (64.05%), of which 4 articles were under 80% and the lowest was 75.25%, 55 articles were100%. The report rates of all types of survival endpoint were lower than 10%. Eleven of 100 articles which reported a loss to follow-up had stated how to treat it in the analysis. One hundred thirty articles (53.72%) did not perform multivariate analysis. One hundred thirty-nine articles (57.44%) did not define the survival time. Violations and omissions of methodological guidelines included no mention of pertinent checks for proportional hazard assumption; no report of testing for interactions and collinearity between independent variables; no report of calculation method of sample size. Thirty-six articles (32.74%) reported the methods of independent variable selection. The above defects could make potentially inaccurate

Source apportionment and sensitivity analysis: two methodologies with two different purposes

Science.gov (United States)

Clappier, Alain; Belis, Claudio A.; Pernigotti, Denise; Thunis, Philippe

2017-11-01

This work reviews the existing methodologies for source apportionment and sensitivity analysis to identify key differences and stress their implicit limitations. The emphasis is laid on the differences between source impacts (sensitivity analysis) and contributions (source apportionment) obtained by using four different methodologies: brute-force top-down, brute-force bottom-up, tagged species and decoupled direct method (DDM). A simple theoretical example to compare these approaches is used highlighting differences and potential implications for policy. When the relationships between concentration and emissions are linear, impacts and contributions are equivalent concepts. In this case, source apportionment and sensitivity analysis may be used indifferently for both air quality planning purposes and quantifying source contributions. However, this study demonstrates that when the relationship between emissions and concentrations is nonlinear, sensitivity approaches are not suitable to retrieve source contributions and source apportionment methods are not appropriate to evaluate the impact of abatement strategies. A quantification of the potential nonlinearities should therefore be the first step prior to source apportionment or planning applications, to prevent any limitations in their use. When nonlinearity is mild, these limitations may, however, be acceptable in the context of the other uncertainties inherent to complex models. Moreover, when using sensitivity analysis for planning, it is important to note that, under nonlinear circumstances, the calculated impacts will only provide information for the exact conditions (e.g. emission reduction share) that are simulated.

A new methodology of spatial cross-correlation analysis.

Science.gov (United States)

Chen, Yanguang

2015-01-01

Spatial correlation modeling comprises both spatial autocorrelation and spatial cross-correlation processes. The spatial autocorrelation theory has been well-developed. It is necessary to advance the method of spatial cross-correlation analysis to supplement the autocorrelation analysis. This paper presents a set of models and analytical procedures for spatial cross-correlation analysis. By analogy with Moran's index newly expressed in a spatial quadratic form, a theoretical framework is derived for geographical cross-correlation modeling. First, two sets of spatial cross-correlation coefficients are defined, including a global spatial cross-correlation coefficient and local spatial cross-correlation coefficients. Second, a pair of scatterplots of spatial cross-correlation is proposed, and the plots can be used to visually reveal the causality behind spatial systems. Based on the global cross-correlation coefficient, Pearson's correlation coefficient can be decomposed into two parts: direct correlation (partial correlation) and indirect correlation (spatial cross-correlation). As an example, the methodology is applied to the relationships between China's urbanization and economic development to illustrate how to model spatial cross-correlation phenomena. This study is an introduction to developing the theory of spatial cross-correlation, and future geographical spatial analysis might benefit from these models and indexes.

Comparison Of Irms Delhi Methodology With Who Methodology On Immunization Coverage

Directory of Open Access Journals (Sweden)

Singh Padam

1996-01-01

Full Text Available Research question: What are the merits of IRMS Model over WHO Model for Coverage Evaluation Survey? Which method is superior and appropriate for coverage evolution survey of immunization in our setting? Objective: To compare IRMS Delhi methodology with WHO methodology on Immunization Coverage. Study Design: Cross-Sectional Setting: Urban and Rural both. Participants: Mothers& Children Sample Size: 300 children between 1-2 years and 300 mothers in rural areas and 75 children and 75 mothers in urban areas. Study Variables: Rural, Urban, Cast-Group, Size of the stratum, Literacy, Sex and Cost effectiveness. Outcome Variables: Coverage level of immunization. Analysis: Routine Statistical Analysis. Results: IRMS developed methodology scores better rating over WHO methodology, especially when coverage evolution is attempted in medium size villages with existence of socio-economic seggregation-which remains the main characteristic of the Indian villages.

Methodology for thermal-hydraulics analysis of pool type MTR fuel research reactors

International Nuclear Information System (INIS)

Umbehaun, Pedro Ernesto

2000-01-01

This work presents a methodology developed for thermal-hydraulic analysis of pool type MTR fuel research reactors. For this methodology a computational program, FLOW, and a model, MTRCR-IEAR1 were developed. FLOW calculates the cooling flow distribution in the fuel elements, control elements, irradiators, and through the channels formed among the fuel elements and among the irradiators and reflectors. This computer program was validated against experimental data for the IEA-R1 research reactor core at IPEN-CNEN/SP. MTRCR-IEAR1 is a model based on the commercial program Engineering Equation Solver (EES). Besides the thermal-hydraulic analyses of the core in steady state accomplished by traditional computational programs like COBRA-3C/RERTR and PARET, this model allows to analyze parallel channels with different cooling flow and/or geometry. Uncertainty factors of the variables from neutronic and thermalhydraulic calculation and also from the fabrication of the fuel element are introduced in the model. For steady state analyses MTRCR-IEAR1 showed good agreement with the results of COBRA-3C/RERTR and PARET. The developed methodology was used for the calculation of the cooling flow distribution and the thermal-hydraulic analysis of a typical configuration of the IEA-R1 research reactor core. (author)

Analysis of core damage frequency from internal events: Methodology guidelines: Volume 1

International Nuclear Information System (INIS)

Drouin, M.T.; Harper, F.T.; Camp, A.L.

1987-09-01

NUREG-1150 examines the risk to the public from a selected group of nuclear power plants. This report describes the methodology used to estimate the internal event core damage frequencies of four plants in support of NUREG-1150. In principle, this methodology is similar to methods used in past probabilistic risk assessments; however, based on past studies and using analysts that are experienced in these techniques, the analyses can be focused in certain areas. In this approach, only the most important systems and failure modes are modeled in detail. Further, the data and human reliability analyses are simplified, with emphasis on the most important components and human actions. Using these methods, an analysis can be completed in six to nine months using two to three full-time systems analysts and part-time personnel in other areas, such as data analysis and human reliability analysis. This is significantly faster and less costly than previous analyses and provides most of the insights that are obtained by the more costly studies. 82 refs., 35 figs., 27 tabs

Biomass Thermogravimetric Analysis: Uncertainty Determination Methodology and Sampling Maps Generation

Science.gov (United States)

Pazó, Jose A.; Granada, Enrique; Saavedra, Ángeles; Eguía, Pablo; Collazo, Joaquín

2010-01-01

The objective of this study was to develop a methodology for the determination of the maximum sampling error and confidence intervals of thermal properties obtained from thermogravimetric analysis (TG), including moisture, volatile matter, fixed carbon and ash content. The sampling procedure of the TG analysis was of particular interest and was conducted with care. The results of the present study were compared to those of a prompt analysis, and a correlation between the mean values and maximum sampling errors of the methods were not observed. In general, low and acceptable levels of uncertainty and error were obtained, demonstrating that the properties evaluated by TG analysis were representative of the overall fuel composition. The accurate determination of the thermal properties of biomass with precise confidence intervals is of particular interest in energetic biomass applications. PMID:20717532

Heart rate asymmetry follows the inspiration/expiration ratio in healthy volunteers

International Nuclear Information System (INIS)

Klintworth, Anne; Ajtay, Zénó; Paljunite, Alina; Szabados, Sándor; Hejjel, László

2012-01-01

Heart rate asymmetry (HRA) quantifies the uneven distribution of points above and below the identity-line in a Poincaré plot of RR-intervals. The authors investigated if HRA could be influenced by the inspiration/expiration ratio. Healthy volunteers (n = 18) were studied in the supine position at 4.5 s metronome breathing. ECG and breathing signals were recorded for 360 s at each breathing pattern: inspiration controlled, inspiration/expiration controlled (1:2, 1:1, 2:1 ratio), inspiration controlled again. Time domain, frequency domain and Poincaré plot heart rate variability (HRV) analysis with Porta's and Guzik's indices were performed on 300 s tachograms. There were no statistically significant differences in time domain, frequency domain and standard Poincaré plot parameters during the various breathing patterns, whereas Porta's and Guzik's indices significantly rose at 1:1 and 2:1 compared to physiological 1:2 breathing. There were no significant differences in the HRA parameters between the first and the last runs. In our population the inspiration/expiration ratio significantly influenced HRA, but not standard HRV parameters. Positive correlation of Guzik's and Porta's index reflects reciprocal changes of the number of points and their dispersion in the accelerating and decelerating sets of RR-intervals. HRA-analysis can be a promising method for investigating cardiovascular regulation/health particularly with further spreading of wearable monitors. (paper)

The decade 1989-1998 in Spanish psychology: an analysis of research in statistics, methodology, and psychometric theory.

Science.gov (United States)

García-Pérez, M A

2001-11-01

This paper presents an analysis of research published in the decade 1989-1998 by Spanish faculty members in the areas of statistical methods, research methodology, and psychometric theory. Database search and direct correspondence with faculty members in Departments of Methodology across Spain rendered a list of 193 papers published in these broad areas by 82 faculty members. These and other faculty members had actually published 931 papers over the decade of analysis, but 738 of them addressed topics not appropriate for description in this report. Classification and analysis of these 193 papers revealed topics that have attracted the most interest (psychophysics, item response theory, analysis of variance, sequential analysis, and meta-analysis) as well as other topics that have received less attention (scaling, factor analysis, time series, and structural models). A significant number of papers also dealt with various methodological issues (software, algorithms, instrumentation, and techniques). A substantial part of this report is devoted to describing the issues addressed across these 193 papers--most of which are written in the Spanish language and published in Spanish journals--and some representative references are given.

A Methodology for the Development of RESTful Semantic Web Services for Gene Expression Analysis.

Directory of Open Access Journals (Sweden)

Gabriela D A Guardia

Full Text Available Gene expression studies are generally performed through multi-step analysis processes, which require the integrated use of a number of analysis tools. In order to facilitate tool/data integration, an increasing number of analysis tools have been developed as or adapted to semantic web services. In recent years, some approaches have been defined for the development and semantic annotation of web services created from legacy software tools, but these approaches still present many limitations. In addition, to the best of our knowledge, no suitable approach has been defined for the functional genomics domain. Therefore, this paper aims at defining an integrated methodology for the implementation of RESTful semantic web services created from gene expression analysis tools and the semantic annotation of such services. We have applied our methodology to the development of a number of services to support the analysis of different types of gene expression data, including microarray and RNASeq. All developed services are publicly available in the Gene Expression Analysis Services (GEAS Repository at http://dcm.ffclrp.usp.br/lssb/geas. Additionally, we have used a number of the developed services to create different integrated analysis scenarios to reproduce parts of two gene expression studies documented in the literature. The first study involves the analysis of one-color microarray data obtained from multiple sclerosis patients and healthy donors. The second study comprises the analysis of RNA-Seq data obtained from melanoma cells to investigate the role of the remodeller BRG1 in the proliferation and morphology of these cells. Our methodology provides concrete guidelines and technical details in order to facilitate the systematic development of semantic web services. Moreover, it encourages the development and reuse of these services for the creation of semantically integrated solutions for gene expression analysis.

A Methodology for the Development of RESTful Semantic Web Services for Gene Expression Analysis.

Science.gov (United States)

Guardia, Gabriela D A; Pires, Luís Ferreira; Vêncio, Ricardo Z N; Malmegrim, Kelen C R; de Farias, Cléver R G

2015-01-01

Gene expression studies are generally performed through multi-step analysis processes, which require the integrated use of a number of analysis tools. In order to facilitate tool/data integration, an increasing number of analysis tools have been developed as or adapted to semantic web services. In recent years, some approaches have been defined for the development and semantic annotation of web services created from legacy software tools, but these approaches still present many limitations. In addition, to the best of our knowledge, no suitable approach has been defined for the functional genomics domain. Therefore, this paper aims at defining an integrated methodology for the implementation of RESTful semantic web services created from gene expression analysis tools and the semantic annotation of such services. We have applied our methodology to the development of a number of services to support the analysis of different types of gene expression data, including microarray and RNASeq. All developed services are publicly available in the Gene Expression Analysis Services (GEAS) Repository at http://dcm.ffclrp.usp.br/lssb/geas. Additionally, we have used a number of the developed services to create different integrated analysis scenarios to reproduce parts of two gene expression studies documented in the literature. The first study involves the analysis of one-color microarray data obtained from multiple sclerosis patients and healthy donors. The second study comprises the analysis of RNA-Seq data obtained from melanoma cells to investigate the role of the remodeller BRG1 in the proliferation and morphology of these cells. Our methodology provides concrete guidelines and technical details in order to facilitate the systematic development of semantic web services. Moreover, it encourages the development and reuse of these services for the creation of semantically integrated solutions for gene expression analysis.