• Nuclear Engineering and Technology
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• 제33권6호
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• pp.596-604
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• 2001

This paper quantitatively presents the effects of important factors of the probabilistic safety assessment (PSA) of safety-critical digital systems. The result which is quantified using fault tree analysis methodology shows that these factors remarkably affect the system safety. In this paper we list the factors which should be represented by the model for PSA. Based on the PSA experience, we select three important factors which are expected to dominate the system unavailability. They are the avoidance of common cause failure, the coverage of fault tolerant mechanisms and software failure probability. We Quantitatively demonstrate the effect of these three factors. The broader usage of digital equipment in nuclear power plants gives rise to the safety problems. Even though conventional PSA methods are immature for applying to microprocessor-based digital systems, practical needs force us to apply it because the result of PSA plays an important role in proving the safety of a designed system. We expect the analysis result to provide valuable feedback to the designers of digital safety- critical systems.

• 대한안전경영과학회지
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• 제13권3호
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• pp.45-53
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• 2011

The research proposes seven elimination rules of redundant gates and blocks in Fault Tree Analysis (FTA) and Reliability Block Diagram (RBD). The computational complexity of cut sets and path sets is NP-hard. In order to reduce the complexity of Minimal Cut Set (MCS) and Minimal Path Set (MPS), the paper classifies generation algorithms. Moreover, the study develops six implementation steps which reflect structural importance (SI) and reliability importance (RI) from Reliability Centered Maintenance (RCM) that a priority of using the functional logic among components is to reduce (improve) the system unavailability (or availability). The proposed steps include efficient generation of state structure function by Rare Event Enumeration (REA). Effective use of importance measures, such as SI and ill measures, is presented based on the number and the size of MCS and MPS which is generated from the reference[5] of this paper. In addition, numerical examples are presented for practitioners to obtain the comprehensive understanding of six steps that is proposed in this research.

• Nuclear Engineering and Technology
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• 제26권1호
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• pp.9-18
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• 1994

월성 원자력발전소의 안전계통은 비상사태시에만 작동하는 3분의 2논리로 구성되어 있다. 그들의 작동성을 보증하기 위해 이 안전계통은 주기적으로 점검되어진다. 본연구에서 사람의 실수가 고려되어진 3분의 2논리 구성 시스템에서의 불이용도가 계산되어졌다. 그리고 우리는 시험기간중에 사람의 실수또는 기계의 고장으로 인해 발전정지를 일으킬 확률을 구했다. 우리는 이 불이용도와 발전정지를 일으킬 확률을 둘다 고려하여 적정한 최적점검주기를 계산하였다. 이렇게 얻어진 점검주기와 현재 사용되는 점검주기를 비교하면 사람의 실수를 최소(8.24 $\times$ $10^{-6}$ )로 보았을때 최적점검주기는 현재 사용되는 점검주기 보다 조금 짧았고 사람의 실수를 최대 (4.44 $\times$ $10^{-4}$ )로 보았을 때 최적점검주기는 현재 사용하는 점검 주기보다 다소 긴 것으로 계산되어졌다.

• Nuclear Engineering and Technology
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• 제20권1호
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• pp.47-53
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• 1988

본 연구의 목적은 충전/안전주입 펌프 최소순환관으로부터 안전주입신호(s-신호)를제거 함에 따른 원자력 5,6,7,8호기의 고압안전주입계통(HHSIS)의 신뢰도를 분석, 평가하는 것이다. 계산은 s-신호를 제거한 경우와 제거하지 않은 경우에 대하여 각각 수행되었다. 각 경우에 대하여 s-신호 발생시 고압안전주입계통의 이용불능도와 충전/안전주입 펌프의 파손확률이 계산되었다. 계산결과에 따르면, s-신호를 제거함에 따라 고압안전주입계통의 이용불능도는 미세하게 증가하였으며 반면에 충전/안전주입 펌프의 파손확률은 크게 감소하였다. 따라서 여러가지 측면에서 충전/안전주입 펌프의 최소순환관으로부터 s-신호를 제거하고 운전하는 것이 합당하다는 것이 밝혀졌으며, 고압안전주입계통의 이용불능도를 줄이기 위하여 운전절차를 개선하고 운전원의 훈련 및 교육을 강화할 것을 추천한다.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.764-775
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• 2020

Periodic safety reviews (PSRs) are conducted on operating nuclear power plants (NPPs) and have been mandated also for research reactors in Korea, in response to the Fukushima accident. One safety review tool, the probabilistic safety assessment (PSA), aims to identify weaknesses in the design and operation of the research reactor, and to evaluate and compare possible safety improvements. However, the PSA for research reactors is difficult due to scarce data availability. An important element in the analysis of research reactors is the reactor protection system (RPS), with its functionality and importance. In this view, we consider that of the AGN-201K, a zero-power reactor without forced decay heat removal systems, to demonstrate a risk-informed safety improvement study. By incorporating risk- and safety-significance importance measures, and sensitivity and uncertainty analyses, the proposed method identifies critical components in the RPS reliability model, systematically proposes potential safety improvements and ranks them to assist in the decision-making process.

• 한국안전학회지
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• 제27권5호
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• pp.224-228
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• 2012

According to the results of Probabilistic Safety Assessment(PSA) for a Nuclear Power Plant(NPP), an Emergency Power Supply(EPS) system has been considered as one of the most important safety system. Especially, the interests in the reliability of the EPS system have been increased after the severe accidents of Fukushima Daiichi. Firstly, we performed the risk assessment and the importance analysis of the EPS system based on the PSA models of the reference plant, which is the Korean standard NPP type. Considering a portable Diesel Generator(DG) system as the reliability reinforcement of the EPS system, we modified the PSA models and performed the risk impact assessment and the importance analysis. Although the reliability of the potable DG could be about 20% of the reliability of the alternative AC DG, we identified that Core Damage Frequency(CDF) was decreased by at least 4.6%. In addition, the risk impacts due to the unavailability of the EPS system on CDF were decreased.

• 한국신뢰성학회지:신뢰성응용연구
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• 제8권2호
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• pp.101-111
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• 2008

Good diagnostics improves both the safety and system unavailability of digital safety systems. The measure of a diagnostic capability is called the Coverage Factor. Because the Failure Modes, Effects and Diagnostic Analysis (FMEDA) provides information on the failure rates and failure mode distributions necessary to calculate a diagnostic coverage factor for a component, the FMEDA can be used as a useful tool to calculate it. Through performing FMEDA on a digital signal processor (DSP) board used in a digital safety system, it is shown that some components of the DSP board can be replaced or improved to satisfy the required diagnostic coverage. That is, the FMEDA can serve as a useful verification tool to design a diagnostic capability for the DSP board.

• Nuclear Engineering and Technology
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• 제56권1호
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• pp.48-55
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• 2024

The safety-related raw water system's strong operational condition supports the radiation defense and biological shield of nuclear plant containment structures. Gaps and failures in maintaining proper working condition of main equipment like pump were among the most common causes of unavailability of safety related raw water systems. We integrated the advanced data analytics tools to evaluate the maintenance records of water systems and gave special consideration to deficiencies related to pump. We utilized maintenance data over a three-and-a-half-year period to produce metrics like MTBF, MTTF, MTTR, and failure rate. The visual analytic platform using tableau identified the efficacy of maintenance & deficiency in the safety raw water systems. When the number of water quality violation was compared to the other O&M deficiencies, it was discovered that water quality violations account for roughly 15% of the system's deficiencies. The pumps were substantial contributors to the deficit. Pump availability was predicted and optimized with real time data using response surface method. The prediction model was significant with r-squared value of 0.98. This prediction model can be used to predict forth coming pump failures in nuclear plant.

• 시스템엔지니어링학술지
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• 제12권2호
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• pp.91-99
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• 2016

The safety of nuclear power plants has received much attention; this safety largely depends on the continuous availability of electrical energy source during all modes of nuclear power plant operation. A station blackout (SBO) describes the loss of the off-site electric power, the failure of the emergency diesel generators, and the unavailability of the alternate AC (AAC) power. Consequently, all systems that are AC powered such as the safety injection, shutdown cooling, component cooling water, and essential service water systems are unavailable. The aim of this study is to investigate the deficiencies of the existing alternatives for coping with an extended SBO for APR1400 design. The method is analyzing the existing deficiencies and proposing an optimal solution for the NPP design during the extended SBO. This study, established a new passive system, called passive decay heat removal system (PDHRS), using systems engineering approach.

• 한국안전학회지
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• 제29권4호
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• pp.199-207
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• 2014

In this paper, to systematically assess the abandonment risk of main control room (MCR) fire, fire simulations with Fire Dynamics Simulator were performed and abandonment probabilities were estimated for the MCR bench-board fire of domestic reference nuclear power plant. The fire simulation scenarios performed in this study included propagating and non-propagating fires of the MCR bench-board, and the availability and unavailability of heating, ventilation, and air conditioning system (HVACS). The following results were obtained. First, temperature was the major abandonment impact factor for the MCR bench-board fire if the HVACS was available and optical density was that if the HVACS was unavailable. Second, the fire scenario contributing the MCR bench-board fire abandonment risk was identified to be only the propagating fire. Third, it was confirmed that the abandonment probability of the MCR bench-board fire for domestic reference nuclear power plant could be reduced by using the fire modeling.