• Journal of the Ergonomics Society of Korea
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• v.30 no.1
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• pp.99-108
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• 2011

Almost all companies have paid much attention to the safety management ranging from maintenance to operation even at the stage of designing in order to prevent accidents, but fatal accidents continue to increase throughout the world. In particular, it is essential to systematically prevent such fatal accidents as fire, explosion or leakage of toxic gas at factories in order to not only protect the workers and neighbors but also prevent economic losses and environmental pollution. Though it is well known that accident probability is very low in NPP(Nuclear Power Plants), the reason why many researches are still being performed about the accidents is the results may be so severe. HRA is the main process to make preparation for possibility of human error in designing of the NPP. But those techniques have some problems and limitation as follows; the evaluation sensitivity of those techniques are out of date. And the evaluation of human error is not coupled with the design process. Additionally, the scope of the human error which has to be included in reliability assessment should be expanded. This work focuses on the coincidence of human error and mechanical failure for some important performance shaping factors to propose a method for improving safety effectively of the process industries. In order to apply in these purposes into the thesis, I found 63 critical Performance Shaping Factors of the eight dimensions throughout studies that I executed earlier. In this study, various analysis of opinion of specialists(Personal Factors, Training, Knowledge or Experience, Procedures and Documentation, Information, Communications, HMI, Workplace Design, Quality of Environment, Team Factors) and the guideline for construction of PSF were accomplished. The selected method was AHP which simplifies objective conclusions by maintaining consistency. This research focused on the implementation process of PSF to evaluate the process of PSF at each phase. As a result, we propose an evaluation model of PSF as a tool to find critical problem at each phase and improve on how to resolve the problems found at each phase. This evaluation model makes it possible to extraction of PSF succesfully by presenting the basis of assessment which will be used by enterprises to minimize the trial and error of construction process of PSF.

• Nuclear Engineering and Technology
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• v.31 no.6
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• pp.8-22
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• 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
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• v.47 no.2
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• pp.187-195
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• 2015

Human-induced initiating events, also called Category B actions in human reliability analysis, are operator actions that may lead directly to initiating events. Most conventional probabilistic safety analyses typically assume that the frequency of initiating events also includes the probability of human-induced initiating events. However, some regulatory documents require Category B actions to be specifically analyzed and quantified in probabilistic safety analysis. An explicit modeling of Category B actions could also potentially lead to important insights into human performance in terms of safety. However, there is no standard procedure to identify Category B actions. This paper describes a systematic procedure to identify Category B actions for low power and shutdown conditions. The procedure includes several steps to determine operator actions that may lead to initiating events in the low power and shutdown stages. These steps are the selection of initiating events, the selection of systems or components, the screening of unlikely operating actions, and the quantification of initiating events. The procedure also provides the detailed instruction for each step, such as operator's action, information required, screening rules, and the outputs. Finally, the applicability of the suggested approach is also investigated by application to a plant example.

• Nuclear Engineering and Technology
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• v.41 no.10
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• pp.1361-1374
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• 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.

• Journal of the Korean Society of Safety
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• v.27 no.3
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• pp.111-116
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• 2012

This study is about the quantitative safety assessment of hydrogen station in Korea operating with on-site type. This was written by background information that before qualitative safety assessment to write. For the qualitative safety assessment method, the study used FMEA(failure mode & effect analysis) and HAZOP(hazard & operability), and adopted the FTA(fault tree analysis) as the quantitative safety assessment method. To write the FTA, we wrote FT by Top event that hydrogen leakage can be called most serious accident of hydrogen station. Each base event collect reliability data by reliability data handbook, THERP-HRA and estimation of the engineering. Assessment looked at the high frequency and the possible risk through Gate, Importance, m.cutsets analysis.

• Nuclear Engineering and Technology
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• v.40 no.5
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• pp.349-364
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• 2008

Over the years, probabilistic risk assessment (PRA) research activities conducted at the U.S. Nuclear Regulatory Commission (NRC) have played an essential role in support of the agency's move towards risk-informed regulation. These research activities have provided the technical basis for NRC's regulatory activities in key areas; provided PRA methods, tools, and data enabling the agency to meet future challenges; supported the implementation of NRC's 1995 PRA Policy Statement by assessing key sources of risk; and supported the development of necessary technical and human resources supporting NRC's risk-informed activities. PRA research aimed at improving the NRC's understanding of risk can positively affect the agency's regulatory activities, as evidenced by three case studies involving research on fire PRA, human reliability analysis (HRA), and pressurized thermal shock (PTS) PRA. These case studies also show that such research can take a considerable amount of time, and that the incorporation of research results into regulatory practice can take even longer. The need for sustained effort and appropriate lead time is an important consideration in the development of a PRA research program aimed at helping the agency address key sources of risk for current and potential future facilities.

• Journal of Korean Association for Spatial Structures
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• v.18 no.4
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• pp.61-68
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• 2018

Tensile load tests were conducted on High-Shear Ring Anchors (HRAs) after shear load had been applied to the HRAs, which had been developed to reduce the number of the anchors. Test variables include the embedment length of the rod and the width of the specimens and a total of 12 specimens were tested. Test results show that the HRAs pulled out due to bond failure or steel failure occurred in case that the HRAs were installed to the members with 300mm or greater width and the embedment length of 160mm (the actual embedment of rod is 140mm) or deeper. Except 4 HRAs showing steel failure of rod, the minimum and average of test-to-prediction by ACI 318-14 ratios are 1.18 and 1.79, respectively. The tensile strength of HRAs, after shear load was applied to the HRAs, can be safely evaluated by the minimum among the concrete breakout strength and bond strength with the actual embedment length of the rod.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.484-497
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• 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
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• v.55 no.8
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• pp.2897-2904
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• 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.

• Journal of the Korean Society of Safety
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• v.14 no.2
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• pp.62-69
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• 1999

The quantitative risk assessment and consequence analysis by accident scenario in the process of EPS(Expendable Poly Stylene) reaction process are conducted. And the decision making process is studied followed by selecting various alternatives to safety management and facility improvement. The result are as follows; 1) The object of decision making through comprehensive risk assessment are the scenario which can cause four major accident, which are made by process analysis, work analysis and hazard identification. 2) Frequency analysis of ETA, FTA, HRA and consequence analysis of accident to each have been conducted. The each frequency values are yielded $9.2{\times}10_{-5}/yr$ to scenarios $1, 8.2{\times}10^{-4}/yr$ to scenario 2, $4.5{\times}10^{-6}/yr$ scenario 3 and $1.8{\times}10^{-7}/yr$ to scenario 4. The each scenarios have been conducted consequence analysis. 3) The calculated values have been obtained 4.00 to scenario 1, 3.25 to scenario 2, 2.43 to scenario 3 and 1.34 to scenario 4, as the weight value had been applied to the quantitative and normalized criteria of all components. As a risk criteria, scenario 1 have been selected, which is the most dangerous scenario as a result of ranking the scenario. 4) According to the importance of FTA and contribute to scenario 1, the cost-benefit values are yielded $8.05\times10^5[￦/yr]$ to final alternative(Al), $1.55{\times}10^5[￦/yr]$ to final alternative(A2) and $2.32{\times}10^5[￦/yr]$ to final alternative(A3). As a result of final alternative(Al) has been selected, which is the most optimized alternative.