• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.850-855
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• 1998

소형 냉각재상실사고 발생 후 고압안전주입이 작동하지 않는 경우, 국내 원자력발전소의 확률론적 안전성평가 (Probabilistic Safety Assessment: PSA) 에서 고려한 일차측 급속냉각 (Aggressive Cool Down of Reactor Coolant System)의 수행 가능성에 대한 논란이 있다. PSA분석 결과에 의하면, 일차측 급속냉각을 위해서는 운전원 조치가 전체 노심손상빈도에 큰 영향을 주고 있음을 보여주지만, 현재 작성되어 있는 국내 원자력발전소의 비상 운전절차서에 따르면 PSA 모델시 가정된 성공기준으로 일차측 급속냉각의 수행에 실패할 가능성이 매우 높은 것으로 판단된다. 이에 따라 본 논문에서는 소형 냉각재상실사고로 인한 노심 손상빈도 측면에서 PSA에서 사용한 일차측 급속냉각 성공기준과 인간오류에 대하여 민감도분석을 수행하였다. 또한 열수력학적 분석을 통해 일차측 급속냉각의 타당성과 성공기준을 재검토했다. 이 결과 일차측 급속냉각의 수행 가능성 여부와 노심 손상빈도에 미치는 영향을 도출하였고 일차측 급속냉각의 성공적 수행을 위한 새로운 성공기준을 제시한다.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.1983-1989
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• 2020

The human reliability analysis is a method by which, in general terms, the human impact to the safety and risk of a nuclear power plant operation can be modelled, quantified and analysed. It is an indispensable element of the PSA process within the nuclear industry nowadays. The paper herein presents a sensitivity study of the human reliability analysis performed on a real nuclear power plant-specific probabilistic safety assessment model. The analysis is performed on a pre-selected set of post-initiator operator actions. The purpose of the study is to investigate the impact of these operator actions on the plant risk by altering their corresponding human error probabilities in a wide spectrum. The results direct the fact that the future effort should be focused on maintaining the current human reliability level, i.e. not letting it worsen, rather than improving it.

• 한국안전학회지
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• 제28권3호
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• pp.123-128
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• 2013

The Advanced Power Reactor Plus(APR+), which is a GEN III+ reactor based on the APR1400, is being developed in Korea. In order to enhance the safety of the APR+, a passive auxiliary feedwater system(PAFS) has been adopted in the APR+. The PAFS replaces the conventional active auxiliary feedwater system(AFWS) by introducing a natural driving force mechanism while maintaining the system function of cooling the primary side and removing the decay heat. As the PAFS completely replaces the conventional AFWS, it is required to verify the cooling capacity of PAFS for the core damage frequency(CDF) evaluation. For this reason, this paper discusses the cooling performance of the PAFS during transient accidents. The test case and scenarios were picked from the result of the sensitivity analysis in APR+ Probabilistic Safety Assessment(PSA). The analysis was performed by the best estimate thermal-hydraulic code, RELAP5/.MOD3.3. This study shows that the plant maintains the stable state without the core damages under the given test scenarios. The results of PSA considering this analysis' results shows that the CDF values are decreased. The analysis results can be used for more realistic and accurate performance of a PSA.

• 한국지진공학회논문집
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• 제14권4호
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• pp.17-22
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• 2010

본 연구에서는 지진해일에 의한 원자력발전소의 확률론적 안전성 평가를 위하여 필수적으로 도출해야 하는 지진해일 재해도 곡선을 도출하기 위한 연구를 수행하였다. 1900년도 이후에 기록된 동해안에서의 지진해일 기록과 1900년도 이전에 역사지진기록에서 찾을 수 있는 지진해일 기록을 이용하여 지진해일에 의한 최대파고에 대한 재현주기를 산정하고자 하였다. Power law, upper-truncated power law 그리고 지수함수에 의해서 추세선을 작성하였으며 그 결과를 비교하였다. 동해안에서 발생한 지진해일의 기록이 10건 내외에 불과하므로 기록에 의한 지진해일 재해도 곡선추정 연구에 제한이 있으나 국내에는 지진해일의 재해도곡선 추정에 관한 연구가 전무한 현실이므로 지진해일 확률론적 안전성 평가를 위한 초석을 놓은 것으로 판단된다.

• Nuclear Engineering and Technology
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• 제36권6호
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• pp.497-511
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• 2004

A PSV (pressurizer safety valve) popping test carried out in the early phases of a refueling outage may trigger a test-induced LOCA(loss of coolant accident) if a PSV fails to fully close and is stuck in a partially open position. According to a KSNP (Korea standard nuclear power plant) low power and shutdown PSA (probabilistic safety assessment), the failure of a high pressure safety injection (HPSI) accompanied by the failure of a PSV to fully close was identified as a dominant accident sequence with a significant impact on low power and shutdown risks (LPSR). In this study, we aim to investigate and verify a new means for mitigating this type of accident using a thermal-hydraulic analysis. In particular, we explore the applicability of an aggressive cool-down combined with operator actions. The results of the various sensitivity studies performed there will help reduce LPSR and improve Refueling outage safety.

• Nuclear Engineering and Technology
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• 제36권1호
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• pp.64-72
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• 2004

The purpose of this study is to develop a system dynamics model for the assessment of the organizational and human factors in a nuclear power plant which contribute to nuclear safety. Previous studies can be classified into two major approaches. One is the engineering approach using tools such as ergonomics and Probability Safety Assessment (PSA). The other is the socio-psychology approach. Both have contributed to find organizational and human factors and to present guidelines to lessen human error in plants. However, since these approaches assume that the relationship among factors is independent they do not explain the interactions among the factors or variables in Nuclear Power Plants. To overcome these restrictions, a system dynamics model, which can show cause and effect relationships among factors and quantify the organizational and human factors, has been developed. Handling variables such as the degree of leadership, the number of employees, and workload in each department, users can simulate various situations in nuclear power plant organization. Through simulation, users can get insights to improve safety in plants and to find managerial tools in both organizational and human factors.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.499-504
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• 1996

Current Probabilistic Safety Assessment (PSA) techniques are not usually utilized for day-to-day applications in nuclear power plants. The major reason for this anomaly is the complexity of plant models developed for PSA studies and the multitude of resulting fault trees. This impediment can be overcome by the use of simplified plant models. However, oversimplified models usually result in loss of valuable information and therefore. simplification approaches have to be used judiciously in order to achieve accurate and meaningful results. For this reason. development of an appropriate simplification approach must be performed using extreme caution followed with results verification in sequence as well as system levels. If there are no significant differences between the simplified and the original models, the simplified model can be efficiently used in the application of the PSA. This paper presents a methodology for how to develop a suitable simplification technique and the results of its verification for sample systems and sequences. The results show that the utilization of simplified plant models will significantly reduce the number of fault trees with no significant loss of accuracy.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.570-575
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• 1996

1단계 확률론적 안전성 평가 (Level 1 Probabilistic Safety Assessment, PSA)를 수행할 때 나타나는 보조계통 고장 수목간의 순환 논리는 사고 경위 정량화를 위하여 해결되어야만 한다. 기존의 PSA에서는 이를 위하여 별도의 고장 수목을 다시 작성하였으나, 이 방법을 사용하기 위하여서는 보조계통 간의 관계를 검토하여야 하며, 이에 따른 별도의 고장 수목을 작성하여야 하는 등 추가적인 작업이 요구된다. 또한 기존 방법은 일부 최소 단절군이 생성되지 않는 약점을 갖고 있다. 이에 따라 한국원자력연구소에서는 해석적으로 순환 논리를 푸는 방법을 개발하였으며, 이를 PSA용 코드인 KIRAP 코드에 구축하였다. 이에 따라 기존 방법의 약점을 극복하고 고장 수목간의 순환 논리를 자동으로 풀 수 있게 되었다. 본 논문에서는 개발된 해석적 방법을 설명하며, 또한 이 방법을 실제 PSA에 적용하며 나타난 여러 현상에 대하여 살펴본다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.759-764
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• 1997

본 논문에서는 영광 3,4호기의 초기 부분충수 운전중 정지냉각 상실 사건에 대하여 확률론적 안전성평가(Probabilistic Safety Assessment; PSA)를 수행하였다. 1단계 PSA 결과인 노심손상빈도에 크게 영향을 끼치는 인간행위는 THERP(technique for human error rate prediction)를 사용하여 평가하였고, 사고경위는 KIRAP(KAERI integrated reliability analysis code package)을 이용하여 정량화하였다. 영광 3,4호기의 부분충수 운전중 정지냉각 상실 사건에 대한 예비적인 PSA 결과, 노심손상 빈도는 1.43E-6로 평가되었고 노심손상 빈도에 주요하게 기여하는 것은 원자로 냉각재 보충에 대한 운전원의 진단 실패로 나타났다. 노심손상빈도를 감소하는 방안의 하나는 운전원의 진단오류 확률을 낮추기 위해 노심손상까지의 운전원 여유시간을 확장하는 것이다. 그러나 보다 구체적인 결과는 분석에 필요한 여러 가지 자료들을 검토하고 PSA를 다시 수행해야 얻을 수 있을 것으로 판단된다.

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