• Proceedings of the Korean Operations and Management Science Society Conference
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• 2002.05a
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• pp.561-568
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• 2002

한국원자력연구소에서 개발중인 액체금속로인 KALIMER는 경수로나 증수로와 근본적으로 설계가 상이하므로 PSA 방법에 대한 새로운 접근방법을 개발해야 한다. 액체금속로 KALIMER에 대한 확률론적 안전성 평가 방법 (PSA, Probabilistic Safety Assessment) 관련 연구는 초기 사건의 도출 및 빈도계산 방법과 주요 계통의 신뢰성 예비 평가에 대한 것이다. 초기 사건이란 원전에 과도 현상을 유발하여 발전소 정지를 초래하는 모든 비정상 사건을 의미하는 것으로 PSA에서 사건 수목을 구성하는 데 기본이 되는 정보이다. 액체금속로는 기존의 경수로 및 중수로와는 전혀 다른 설계를 갖고 있으므로 액체금속로 특유의 초기 사건을 도출하는 방법 및 이들 초기 사건의 빈도를 계산하는 방법에 대한 연구를 수행하였다. KALIMER 주요 계통의 신뢰성 예비 평가를 수행하기 위하여 확률론적 안전성 평가에서 계통분석기법으로 널리 이용되는 고장수목분석의 절차와 방법에 대한 방법론을 선정하여 PSA 방법론을 개발하였다.

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.1065-1072
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• 2009

As the use of software increases at nuclear power plants (NPPs), the necessity for including software reliability and/or safety into the NPP Probabilistic Safety Assessment (PSA) rises. This work proposes an application procedure of software reliability growth models (RGMs), which are most widely used to quantify software reliability, to NPP PSA. Through the proposed procedure, it can be determined if a software reliability growth model can be applied to the NPP PSA before its real application. The procedure proposed in this work is expected to be very helpful for incorporating software into NPP PSA.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.103-110
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• 2021

In the development of a Risk Monitor probabilistic safety assessment (PSA) model from the basic PSA model of a nuclear power plant, the modeling of common-cause failure (CCF) is very important. At present, some approximate modeling methods are widely used, but there lacks criterion of modeling accuracy and error analysis. In this paper, aiming at ensuring the accuracy of risk assessment and minimizing the Risk Monitor PSA models size, we present three basic issues of CCF model resulted from the changes of a nuclear power plant configuration, put forward corresponding modeling methods, and derive accuracy criteria of CCF modeling based on minimum cut sets and risk indicators according to the requirements of risk monitoring. Finally, a nuclear power plant Risk Monitor PSA model is taken as an example to demonstrate the effectiveness of the proposed modeling method and accuracy criteria, and the application scope of the idea of this paper is also discussed.

• Communications for Statistical Applications and Methods
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• v.8 no.2
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• pp.453-464
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• 2001

A commercial nuclear power station contains at least tow emergency diesel generators(EDG) to control the risk of severe core damage during station blackout accidents. Therefore, the reliability of the EDG's to start and load-run on demand must be maintained at a sufficiently high level. Probabilistic safety assessments(PSA) are increasingly being used to quantify the public risk of operating potentially hazardous systems such as nuclear power reactors. In this paper, to perform PSA, we will introduce three different types of data and use Bayes procedure to estimate the error rate of nuclear power plant EDG, and using practical examples, illustrate which method is more reasonable in our situation.

• Journal of the Korean Society of Safety
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• v.35 no.1
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• pp.97-106
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• 2020

The objective of this paper is to suggest a new quantification method for multi-unit probabilistic safety assessment (PSA) that removes the overestimation error caused by the existing delete-term approximation (DTA) based quantification method. So far, for the actual plant PSA model quantification, a fault tree with negates have been solved by the DTA method. It is well known that the DTA method induces overestimated core damage frequency (CDF) of nuclear power plant (NPP). If a PSA fault tree has negates and non-rare events, the overestimation in CDF drastically increases. Since multi-unit seismic PSA model has plant level negates and many non-rare events in the fault tree, it should be very carefully quantified in order to avoid CDF overestimation. Multi-unit PSA fault tree has normal gates and negates that represent each NPP status. The NPP status means core damage or non-core damage state of individual NPPs. The non-core damage state of a NPP is modeled in the fault tree by using a negate (a NOT gate). Authors reviewed and compared (1) quantification methods that generate exact or approximate Boolean solutions from a fault tree, (2) DTA method generating approximate Boolean solution by solving negates in a fault tree, and (3) probability calculation methods from the Boolean solutions generated by exact quantification methods or DTA method. Based on the review and comparison, a new intersection removal by probability (IRBP) method is suggested in this study for the multi-unit PSA. If the IRBP method is adopted, multi-unit PSA fault tree can be quantified without the overestimation error that is caused by the direct application of DTA method. That is, the extremely overestimated CDF can be avoided and accurate CDF can be calculated by using the IRBP method. The accuracy of the IRBP method was validated by simple multi-unit PSA models. The necessity of the IRBP method was demonstrated by the actual plant multi-unit seismic PSA models.

• Journal of the Korean Society of Safety
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• v.18 no.2
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• pp.98-103
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• 2003

Based on ASME probabilistic risk assessment (PRA) and NEI PRA peer review guidance, we evaluate a human reliability analysis (HRA) in probabilistic safety assessment (PSA) for Korea standard nuclear power plants, Ulchin Unit 3&4, to improve it performed at under design. The HRA for Ulchin Unit 3&4 is assessed as higher than Grade I based on ASME PRA standard and as higher than Grade 2 based on NEI PRA peer review guidance. The major items to be improved identified through the evaluation process are the documentation, the systematic human reliability analysis, the participitation of operators in the works and review of HRA. We suggest the guidance on the identification and qualitative screening analysis for pre-accident human errors and solve some items to be improved using the suggested guidance.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.138-145
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• 2019

With the application of digital technology to safety-critical infrastructures, cyber-attacks have emerged as one of the new dangerous threats. In safety-critical infrastructures such as a nuclear power plant (NPP), a cyber-attack could have serious consequences by initiating dangerous events or rendering important safety systems unavailable. Since a cyber-attack is conducted intentionally, numerous possible cases should be considered for developing a cyber security system, such as the attack paths, methods, and potential target systems. Therefore, prior to developing a risk-informed cyber security strategy, the importance of cyber-attacks and significant critical digital assets (CDAs) should be analyzed. In this work, an importance analysis method for cyber-attacks on an NPP was proposed using the probabilistic safety assessment (PSA) method. To develop an importance analysis framework for cyber-attacks, possible cyber-attacks were identified with failure modes, and a PSA model for cyber-attacks was developed. For case studies, the quantitative evaluations of cyber-attack scenarios were performed using the proposed method. By using quantitative importance of cyber-attacks and identifying significant CDAs that must be defended against cyber-attacks, it is possible to develop an efficient and reliable defense strategy against cyber-attacks on NPPs.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3324-3335
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• 2022

Several countries show interest in the Generation-IV power reactor innovative small module (PRISM), including: Canada, Japan, Korea, Saudi Arabia and the United Kingdom. Generation IV International Forum (GIF) has recommended the utilizing of probabilistic safety assessment (PSA) in evaluating the safety of Generation-IV reactors. This paper reviews the PSA performed for PRISM using SAPHIRE 7.27 code. This work shows that the core damage frequency (CDF) of PRISM for a single module is estimated by 8.5E-8/year which is lower than the Generation-IV target that is 1E-6 core damage per year. The social risk of PRISM (likelihood of latent cancer fatality) with evacuation is estimated by 9.0E-12/year which is much lower than the basic safety objective (BSO) that is 1E-7/year. The social risk without evacuation is estimated by 1.2E- 11/year which is also much lower than the BSO. For the individual risk (likelihood of prompt fatality), it is concluded that it can be considered negligible with evacuation (1.0E-13/year). Assuming no evacuation, the individual risk is 2.7E-10/year which is again much lower than the BSO. In comparison with other PSAs performed for similar sodium fast reactors (SFRs), it shows that PRISM concept has the lowest CDF.

• Journal of the Korean Society of Safety
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• v.33 no.6
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• pp.136-143
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• 2018

In this study, we conducted a pilot study on the multiple spurious operations (MSO) analysis in the fire probabilistic safety assessment (PSA) of domestic nuclear power plant (NPP) to identify the degree of influence of the operator actions used in the MSO mitigation strategies. The MSO scenario of the domestic reference NPP selected for this study is refueling water tank (RWT) drain down event. It could be caused by spurious operations of the containment spray system (CSS) of the reference NPP. The RWT drain down event can be stopped by the main control room (MCR) operator actions for stopping the operation of CSS pump or closing the CSS motor operated valve if the containment spray actuation signal (CSAS) is spuriously actuated. Outside the MCR, it can be stopped by operator actions for closing the CSS manual valves or motor operated valve or stopping the operation of CSS pump. The quantification result of a fire PSA model that takes into account all recovery actions for the RWT drain down event lead to risk reduction by about 95%, compared with quantification result of fire PSA model without considering them. Among the various operator actions, the recovery action for the spurious CSAS operations and the operator action for the manual valve are identified as the most important operator actions. This study quantitatively showed the extent to which the operator actions used as MSO countermeasures have affected the fire PSA quantification results. In addition, we can see the rank of importance among the operator recovery actions in quantitative terms.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1255-1265
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• 2018

The need for a PSA (Probabilistic Safety Assessment) for a multi-unit at a site is growing after the Fukushima accident. Many countries have been studying issues regarding a multi-unit PSA. One of these issues is the problem of many combinations of accident sequences in a multi-unit PSA. This paper deals with the methodology and software to quantify a PSA scenarios for a multi-unit site. Two approaches are developed to quantify a multi-unit PSA. One is to use a minimal cut set approach, and the other is to use a Monte Carlo approach.