• Journal of the Ergonomics Society of Korea
• /
• v.20 no.2
• /
• pp.1-28
• /
• 2001

This paper deals with the selection of the important Influencing Factors (IFs) under accident management situations in nuclear power plants for use in the assessment of human errors. In order to achieve this goal, we collected two types of IF taxonomies, one is the full set IF list mainly developed for human error analysis. and the other is the IFs for human reliability analysis (HRA) in probabilistic safety assessment (PSA). Five sets of IF taxonomy among the full set IF list and ten sets of IF taxonomy among HRA methodologies were collected in the study. From the review and analysis of BRA IFs, we could obtain some insights for the selection of HRA IFs. By considering the situational characteristics of the accident management domain, candidate IFs are chosen. Finally, those IFs are structured hierarchically to be appropriate for the use in the assessment of human error under accident management situation. Three nuclear accidents such as TMI. Chernobyl and JCO were analysed to validate the proposed taxonomy.

• Journal of the Korean Society of Safety
• /
• v.35 no.1
• /
• pp.87-96
• /
• 2020

The purpose of this paper is to develop a fire HRA (Human Reliability Analysis) procedure for full power operation of domestic NPPs (Nuclear Power Plants). For the development of fire HRA procedure, the recent research results of NUREG-1921 in an effort to meet the requirements of the ASME/ANS PRA Standard were reviewed. The K-HRA method, a standard method for HRA of a domestic level 1 PSA (Probabilistic Safety Assessment) and fire related procedures in domestic NPPs were reviewed. Based on the review, a procedure for the fire HRA required for a domestic fire PSA based on the K-HRA method was developed. To this end, HRA issues such as new operator actions required in the event of a fire and complexity of fire situations were considered. Based on the four kinds of HFE (Human Failure Event) developed for a fire HRA in this research, a qualitative analysis such as feasibility evaluation was suggested. And also a quantitative analysis process which consists of screening analysis and detailed analysis was proposed. For the qualitative analysis, a screening analysis by NUREG-1921 was used. In this research, the screening criteria for the screening analysis was modified to reduce vague description and to reflect recent experimental results. For a detailed analysis, the K-HRA method and scoping analysis by NUREG-1921 were adopted. To apply K-HRA to fire HRA for quantification, efforts to modify PSFs (Performance Shaping Factors) of K-HRA to reflect fire situation and effects were made. For example, an absence of STA (Shift Technical Advisor) to command a fire brigade at a fire area is considered and the absence time should be reflected for a HEP (Human Error Probability) quantification. Based on the fire HRA procedure developed in this paper, a case study for HEP quantification such as a screening analysis and detailed analysis with the modified K-HRA was performed. It is expected that the HRA procedure suggested in this paper will be utilized for fire PSA for domestic NPPs as it is the first attempt to establish an HRA process considering fire effects.

• Nuclear Engineering and Technology
• /
• v.32 no.1
• /
• pp.88-98
• /
• 2000

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.

• Journal of the Korean Society of Safety
• /
• v.38 no.6
• /
• pp.72-78
• /
• 2023

The purpose of this paper is to derive a quantified approach for Operator Manual Actions (OMAs) based on the existing fire Human Reliability Analysis (HRA) methodology developed by the Korea Atomic Energy Research Institute (KAERI). The existing fire HRA method was reviewed, and supplementary considerations for OMA quantification were established through a comparative analysis with NUREG-1852 criteria and the review of the existing literature. The OMA quantification approach involves a timeline that considers the occurrence of Multiple Spurious Operations (MSOs) during a Main Control Room Abandonment (MCRA) determination and movement towards the Remote Shutdown Panel (RSP) in the event of a Main Control Room (MCR) fire. The derived failure probability of an OMA from the approach proposed in this paper is expected to enhance the understanding of its reliability. Therefore, it allows moving beyond the deterministic classification of "reliable" or "unreliable" in NUREG-1852. Also, in the event of a nuclear power plant fire where multiple OMAs are required within a critical time range, it is anticipated that the OMA failure probability could serve as a criterion for prioritizing OMAs and determining their order of importance.

• Proceedings of the KSR Conference
• /
• 2006.11b
• /
• pp.1140-1145
• /
• 2006

The railway human reliability analysis (R-HRA) plays a role of identifying and assessing human failure events in the framework of the probabilistic risk assessment (PRA) of the railway systems. This paper reviews three existing HRA methods including the K-HRA (THERP/ASEP-based) method, the HEART method, the RSSB-HRA method, and introduces a case study that was performed to select an appropriate method for a railway risk assessment. The case is the signal passed at danger (SPAD) events, which are caused from a variety of factors. From the case study, the strengths and limitations of each method were derived and compared with each other from the viewpoint of the applicability to the railway industry.

• Journal of the Korean Society of Safety
• /
• v.18 no.2
• /
• pp.104-118
• /
• 2003

It has been criticized that conventional human reliability analysis (HRA) methodologies for probabilistic safety assessment (PSA) have been focused on the quantification of human error probability (HEP) without detailed analysis of human cognitive processes such as situation assessment or decision-making which are crticial to successful response to emergency situations. This paper introduces a new human reliability analysis (HRA) methodology, AGAPE-ET (A guidance And Procedure for Human Error Analysis for Emergency Tasks), focused on the qualitative error analysis of emergency tasks from the viewpoint of the performance of human cognitive function. The AGAPE-ET method is based on the simplified cognitive model and a taxonomy of influencing factors. 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, overall human error analysis process is designed considering the cognitive demand of the required task. The application to an emergency task shows that the proposed method is useful to identify task vulnerabilities associated with the performance of emergency tasks.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1997.05a
• /
• pp.611-616
• /
• 1997

The Level 1 probabilistic safety assessment(PSA) for Wolsong(WS) 2/3/4 nuclear power plant(NPPs) in design stage is performed using the methodologies being equivalent to PWR PSA. Accident sequence evaluation program(ASEP) human reliability analysis(HRA) procedure and technique for human error rate prediction(THERP) are used in HRA of WS 2/3/4 NPPs PSA. The purpose of this paper is to introduce the procedure and methodology of HRA in WS 2/3/4 NPPs PSA. Also, this paper describes the interim results of importance analysis for human actions modeled in WS 2/3/4 PSA and the findings and recommendations of administrative control of secondary control area from the view of human factors.

• Nuclear Engineering and Technology
• /
• v.45 no.2
• /
• pp.151-164
• /
• 2013

A key input for the assessment of Human Error Probabilities (HEPs) with Human Reliability Analysis (HRA) methods is the evaluation of the factors influencing the human performance (often referred to as Performance Shaping Factors, PSFs). In general, the definition of these factors and the supporting guidance are such that their evaluation involves significant subjectivity. This affects the repeatability of HRA results as well as the collection of HRA data for model construction and verification. In this context, the present paper considers the TAsk COMplexity (TACOM) measure, developed by one of the authors to quantify the complexity of procedure-guided tasks (by the operating crew of nuclear power plants in emergency situations), and evaluates its use to represent (objectively and quantitatively) task complexity issues relevant to HRA methods. In particular, TACOM scores are calculated for five Human Failure Events (HFEs) for which empirical evidence on the HEPs (albeit with large uncertainty) and influencing factors are available - from the International HRA Empirical Study. The empirical evaluation has shown promising results. The TACOM score increases as the empirical HEP of the selected HFEs increases. Except for one case, TACOM scores are well distinguished if related to different difficulty categories (e.g., "easy" vs. "somewhat difficult"), while values corresponding to tasks within the same category are very close. Despite some important limitations related to the small number of HFEs investigated and the large uncertainty in their HEPs, this paper presents one of few attempts to empirically study the effect of a performance shaping factor on the human error probability. This type of study is important to enhance the empirical basis of HRA methods, to make sure that 1) the definitions of the PSFs cover the influences important for HRA (i.e., influencing the error probability), and 2) the quantitative relationships among PSFs and error probability are adequately represented.

• Proceedings of the ESK Conference
• /
• 1996.04a
• /
• pp.113-127
• /
• 1996

In Human Reliability Analysis(HRA), the uncertainties involved in many factors that affect human reliability have to be represented as the quantitative forms. Conventional probability- based human reliability theory is used to evaluate the effect of those uncertainties but it is pointed out that the actual human reliability should be different from that of conventional one. Conventional HRA makes use of error rates, however, it is difficult to collect data enough to estimate these error rates, and the estimates of error rates are dependent only on engineering judgement. In this paper, the error possibility that is proposed by Onisawa is used to represent human reliability, and the error possibility is obtained by use of fuzzy reasoning that plays an important role to clarify the relation between human reliability and human error. Also, assuming these factors are connected to the top event through Fault Tree structure, the influence and correlation of these factors are measured by fuzzy operation. When a fuzzy operation is applied to Fault Tree Analysis, it is possible to simplify the operation applying the logic disjuction and logic conjuction to structure function, and the structure of human reliability can be represented as membership function of the top event. Also, on the basis of the the membership function, the characteristics of human reliability can be evaluated by use of the concept of pattern recognition.

• Journal of the Korea Society of Computer and Information
• /
• v.11 no.6 s.44
• /
• pp.261-267
• /
• 2006

In order to perform a probabilistic safety assessment (PSA), it requires a large number of data for various fields. And the quality of a PSA results have become more important thing of the risk assessment. As part of enhancing the PSA qualify, Korea Atomic Energy Research Institute is developing a full power Human Reliability Analysis (HRA) calculator to manage human failure events (HFEs) and to calculate the diagnosis human error probabilities and execution human error probabilities. This paper introduces the development process and an overview of a standard HRA method for nuclear power plants. The study was carried out in three stages; 1) development of the procedures and rules for a standard HRA method. 2) design of a system structure, 3) development of the HRA calculator.