• 산업경영시스템학회지
• /
• 제18권33호
• /
• pp.173-183
• /
• 1995

The Objective of the study was to determine the usefulness of human resource accounting(HRA) information in assisting financial analysis in their investment decisions. The objective achieved by an investigation through which the reporting of HRA, combined with demographic factors that are independent or interactive, affects the decisions of financial analysts regarding the estimation of the market price of a hypothetical company's common stocks. Two kinds of research were conducted to increase the reliability of the study at the same time. Two or three sets of financial statement were prepared. Each consists of balance sheet and income statement. The actual financial statement was modified to exclude personal bias and opinion.

• 한국안전학회지
• /
• 제21권3호
• /
• pp.137-142
• /
• 2006

Quantification of error possibility, in an HRA process, should be performed so that the result of the qualitative analysis can be utilized in other areas in conjunction with overall safety estimation results. And also, the quantification is an essential process to analyze the error possibility in detail and to obtain countermeasures for the errors through screening procedures. In previous studies for the quantification of error possibility, nominal values were assigned by the experts' judgements and utilized as corresponding probabilities. The values assigned by experts' experiences and judgements, however, require verifications on their reliability. In this study, the validity of new error possibility values in new MCR design was verified by using the Onisawa's model which utilizes fuzzy linguistic values to estimate human error probabilities. With the model of error probabilities are represented as analyst's estimations and natural language expression instead of numerical values. As results, the experts' estimation values about error probabilities are well agreed to the existing error probability estimation model. Thus, it was concluded that the occurrence probabilities of errors derived from the human error analysis process can be assessed by nominal values suggested in the previous studies. It is also expected that our analysis method can supplement the conventional HRA method because the nominal values are based on the consideration of various influencing factors such as PSFs.

• Nuclear Engineering and Technology
• /
• 제54권1호
• /
• pp.110-116
• /
• 2022

Human failure event (HFE) dependency analysis is a part of human reliability analysis (HRA). For efficient HFE dependency analysis, a maximum number of minimal cut sets (MCSs) that have HFE combinations are generated from the fault trees for the probabilistic safety assessment (PSA) of nuclear power plants (NPPs). After collecting potential HFE combinations, dependency levels of subsequent HFEs on the preceding HFEs in each MCS are analyzed and assigned as conditional probabilities. Then, HFE recovery is performed to reflect these conditional probabilities in MCSs by modifying MCSs. Inappropriate HFE dependency analysis and HFE recovery might lead to an inaccurate core damage frequency (CDF). Using the above process, HFE recovery is performed on MCSs that are generated with a non-zero truncation limit, where many MCSs that have HFE combinations are truncated. As a result, the resultant CDF might be underestimated. In this paper, a new method is suggested to incorporate HFE recovery into the MCS generation stage. Compared to the current approach with a separate HFE recovery after MCS generation, this new method can (1) reduce the total time and burden for MCS generation and HFE recovery, (2) prevent the truncation of MCSs that have dependent HFEs, and (3) avoid CDF underestimation. This new method is a simple but very effective means of performing MCS generation and HFE recovery simultaneously and improving CDF accuracy. The effectiveness and strength of the new method are clearly demonstrated and discussed with fault trees and HFE combinations that have joint probabilities.

• Nuclear Engineering and Technology
• /
• 제41권10호
• /
• pp.1361-1374
• /
• 2009

Due to the need of data for a Human Reliability Analysis (HRA), a number of data collection efforts have been undertaken in several different organizations. As a part of this effort, a human error analysis that focused on a set of simulator records on a Steam Generator Tube Rupture (SGTR) scenario was performed by using the Human Event Repository and Analysis (HERA) system. This paper summarizes the process and results of the HERA analysis, including discussions about the usability of the HERA system for a human error analysis of simulator data. Five simulated records of an SGTR scenario were analyzed with the HERA analysis process in order to scrutinize the causes and mechanisms of the human related events. From this study, the authors confirmed that the HERA was a serviceable system that can analyze human performance qualitatively from simulator data. It was possible to identify the human related events in the simulator data that affected the system safety not only negatively but also positively. It was also possible to scrutinize the Performance Shaping Factors (PSFs) and the relevant contributory factors with regard to each identified human event.

• 대한인간공학회:학술대회논문집
• /
• 대한인간공학회 1996년도 추계학술대회논문집
• /
• pp.111-118
• /
• 1996

THERP(Technique for Human Error Rate Prediction) 방법론은 원전의 확률론적 위험성 평가(PSA)시 운전원과 작업자의 정량적인 인간오류평가에 가장 널리 사용되고 있는 방법이다. HRA Handbook이라고도 불리는 이 모델은 운전원 행위를 시스템 부품의 한 요소로 가정하고 인간오류를 평가한다. 본 논문은 이 방법론을 이용하여 원전 등과 같이 위험시설물 중의 하나인 개스밸브 기지에서의 작업자 보수시 인적오류를 평가하고 기계적 오류와 함께 인적오류 의 기여도를 계산하였다. 본 눈문에서는 이 방법론이 원전, 개스밸브 기지 뿐만아니라 석유화 학 플랜트와 같은 위험시설물의 인적오류 평가에도 유연하게 사용될 수 있음을 보여주었다.

• 한국산업안전학회:학술대회논문집
• /
• 한국안전학회 2002년도 춘계 학술논문발표회 논문집
• /
• pp.346-355
• /
• 2002

THERP (Technique for Human .Error Rate Prediction) 방법론은 원전의 확률론적 위험성 평가(PSA)시 운전원과 작업자의 인간오류평가에 가장 널리 사용되고 있는 방법이다. HRA Handbook이라고도 불리는 이 모델은 운전원 행위를 시스템 부품의 한 요소로 가정하고 인간오류를 평가한다. 본 논문은 이 방법론을 이용하여 원전 등과 같이 위험시설물 중의 하나인 개스밸브기지에서의 작업자 보수시 인적오류를 평가하고 기계적 오류와 합께 인적오류의 기여도를 계산하였다 본 방법론은 원전, 개스밸브기지 뿐만아니라 석유화학 플랜트와 같은 위험시설물의 인적오류 평가에도 유연하게 사용될 수 있음을 보여주었다.(중략)

• Nuclear Engineering and Technology
• /
• 제55권8호
• /
• pp.2897-2904
• /
• 2023

The Standardized Plant Analysis Risk-Human Reliability Analysis (SPAR-H) method is a widely used method in human reliability analysis (HRA). Performance shaping factors (PSFs) refer to the factors that may influence human performance and are used to adjust nominal human error probabilities (HEPs) in SPAR-H. However, the PSFs are assumed to be independent, which is unrealistic and can lead to unreasonable estimation of HEPs. In this paper, a new method is proposed to handle the dependencies among PSFs in SPAR-H to obtain more reasonable results. Firstly, the dependencies among PSFs are analyzed by using decision-making trial and evaluation laboratory (DEMATEL) method. Then, PSFs are assigned different weights according to their dependent relationships. Finally, multipliers of PSFs are modified based on the relative weights of PSFs. A case study is illustrated that the proposed method is effective in handling the dependent PSFs in SPAR-H, where the duplicate calculations of the dependent part can be reduced. The proposed method can deal with a more general situation that PSFs are dependent, and can provide more reasonable results.

• Nuclear Engineering and Technology
• /
• 제53권2호
• /
• pp.484-497
• /
• 2021

This paper proposes an extended time uncertainty analysis approach in Level 2 human reliability analysis (HRA) considering severe accident management (SAM) strategies. The method is a time-based model that classifies two time distribution functions-time required and time available-to calculate human failure probabilities from delayed action when implementing SAM strategies. The time required function can be obtained by the combination of four time factors: 1) time for diagnosis and decision by the technical support center (TSC) for a given strategy, 2) time for strategy implementation mainly by the local emergency response organization (ERO), 3) time to verify the effectiveness of the strategy and 4) time for portable equipment transport and installation. This function can vary depending on the given scenario and includes a summation of lognormal distributions and a choice regarding shifting the distribution. The time available function can be obtained via thermal-hydraulic code simulation (MAAP 5.03). The proposed approach was applied to assess SAM strategies that use portable equipment and safety depressurization system valves in a total loss of component cooling water event that could cause reactor vessel failure. The results from the proposed method are more realistic (i.e., not conservative) than other existing methods in evaluating SAM strategies involving the use of portable equipment.

• Nuclear Engineering and Technology
• /
• 제31권6호
• /
• pp.8-22
• /
• 1999

This study was performed to comparatively evaluate selected Human Reliability Analysis (HRA) methods which mainly focus on cognitive error analysis, and to derive the requirement of a new human error analysis (HEA) framework for Accident Management (AM) in Nuclear Power Plants (NPPs). In order to achieve this goal, we carried out a case study of human error analysis on an AM task in NPPs. In the study we evaluated three cognitive HEA methods, HRMS, CREAM and PHECA, which were selected through the review of the currently available seven cognitive HEA methods. The task of reactor cavity flooding was chosen for the application study as one of typical tasks of AM in NPPs. From the study, we derived seven requirement items for a new HEA method of AM in NPPs. We could also evaluate the applicability of three cognitive HEA methods to AM tasks. CREAM is considered to be more appropriate than others for the analysis of AM tasks, HRMS is also applicable to the error analysis of AM tasks. But, PHECA is regarded less appropriate for the predictive HEA technique as well as for the analysis of AM tasks. In addition to these, the advantages and disadvantagesofeachmethodaredescribed.

• Nuclear Engineering and Technology
• /
• 제52권6호
• /
• pp.1180-1187
• /
• 2020

A team of operators is required for nuclear power plant operation, and communication between the operators is an important aspect of the team's ability to successfully carry out tasks. It has been difficult to evaluate the quality of this communication though, and as the relationship between communication quality and team performance has yet to be clarified, it has not been applied to most human reliability analysis (HRA) methodologies. This study investigates the relationship between the quality of communication and team performance using data from a full-scope training simulator of a digital main control room (MCR). Two important characteristics of communication were considered to determine quality: each operator's ability to self-confirm the status of a given task in a digital MCR, and the type of communication, as divided into 1-way, 2-way, and 3-way between operators. To measure team performance, the concept of an unsafe act was employed, which is defined as a human error that has the potential to negatively affect plant safety. Analysis results showed that the communication quality and team performance were related to each other. With this more clearly defined relationship, the results of this study can be applied to related performance shaping factors to improve HRA.