• 한국안전학회지
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• 제35권1호
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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.

• 한국안전학회지
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• 제34권5호
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• pp.159-166
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• 2019

The seismic PSA is to probabilistically estimate the potential damage that a large earthquake will cause to a nuclear power plant. It integrates the probabilistic seismic hazard analysis, seismic fragility analysis, and system analysis and is utilized to identify seismic vulnerability and improve seismic capacity of nuclear power plants. Recently, the seismic risk of domestic multi-unit nuclear power plant sites has been evaluated after the Great East Japan Earthquake and Gyeongju Earthquake in Korea. However, while the currently available methods for system analysis can derive basic required results of seismic PSA, they do not provide the detailed results required for the efficient improvement of seismic capacity. Therefore, for in-depth seismic risk evaluation, improved system analysis method for seismic PSA has become necessary. This study develops a system analysis method that is not only suitable for multi-unit seismic PSA but also provides risk information for the seismic capacity improvements. It will also contribute to the enhancement of the safety of nuclear power plants by identifying the seismic vulnerability using the detailed results of seismic PSA. In addition, this system analysis method can be applied to other external event PSAs, such as fire PSA and tsunami PSA, which require similar analysis.

• Nuclear Engineering and Technology
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• 제34권4호
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• pp.331-343
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• 2002

We performed an importance analysis of In-Service Testing (157) components for Ulchin Unit 3 using the integrated evaluation method for categorizing component safety significance developed in this study. The developed method is basically aimed at having a PSA expert perform an importance analysis using PSA and its related information. The importance analysis using the developed method is initiated by ranking the component importance using quantitative PSA information. The importance analysis of the IST components not modeled in the PSA is performed through the engineering judgment, based on the expertise of PSA, and the quantitative and qualitative information for the 157 components. The PSA scope for importance analysis includes not only Level 1 and 2 internal PSA but also Level 1 external and shutdown/low power operation PSA. The importance analysis results of valves show that 167 (26.55%) of the 629 IST valves are HSSCs and 462 (73.45%) are LSSCs. Those of pumps also show that 28 (70%)of the 40157 pumps are HSSCs and 12 (30%) are LSSCs.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1234-1245
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• 2018

The risk of multi-unit nuclear power plants (NPPs) at a site has received considerable critical attention recently. However, current probabilistic safety assessment (PSA) procedures and computer code do not support multi-unit PSA because the traditional PSA structure is mostly used for the quantification of single-unit NPP risk. In this study, the main purpose is to develop a multi-unit Level 2 PSA method and apply it to full-power operating six-unit OPR1000. Multi-unit Level 2 PSA method consists of three steps: (1) development of single-unit Level 2 PSA; (2) extracting the mapping data from plant damage state to source term category; and (3) combining multi-unit Level 1 PSA results and mapping fractions. By applying developed multi-unit Level 2 PSA method into six-unit OPR1000, site containment failure probabilities in case of loss of ultimate heat sink, loss of off-site power, tsunami, and seismic event were quantified.

• Nuclear Engineering and Technology
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• 제43권3호
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• pp.271-278
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• 2011

When performing fire PSA in a nuclear power plant, an event mapping method, using an internal event PSA model, is widely used to reduce the resources used by fire PSA model development. Feasible initiating events and component failure events due to fire are identified to transform the fault tree (FT) for an internal event PSA into one for a fire PSA using the event mapping method. A surrogate event or damage term method is used to condition the FT of the internal PSA. The surrogate event or the damage term plays the role of flagging whether the system/component in a fire compartment is damaged or not, depending on the fire being initiated from a specified compartment. These methods usually require explicit states of all compartments to be modeled in a fire area. Fire event scenarios, when using explicit identification, such as surrogate or damage terms, have two problems: (1) there is no consideration of multiple fire propagation beyond a single propagation to an adjacent compartment, and (2) there is no consideration of simultaneous fire propagations in which an initiating fire event is propagated to multiple paths simultaneously. The present paper suggests a fire propagation equation to identify all possible fire event scenarios for an explicitly treated fire event scenario in the fire PSA. Also, a method for separating fire events was developed to make all fire events a set of mutually exclusive events, which can facilitate arithmetic summation in fire risk quantification. A simple example is given to confirm the applicability of the present method for a $2{\times}3$ rectangular fire area. Also, a feasible asymptotic approach is discussed to reduce the computational burden for fire risk quantification.

• Nuclear Engineering and Technology
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• 제39권3호
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• pp.215-220
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• 2007

The existing flooding Probabilistic Safety Analysis(PSA) was updated to reflect the Korean plant specific operating experience data into the flooding frequency to improve the PSA quality. Both the Nuclear Power Experience(NPE) database and the Korea Nuclear Pipe Failure Database(NuPIPE) databases were used in this study, and from these databases, only the Pressurized Water Reactor(PWR) data were used for the flooding frequencies of the flooding areas in the primary auxiliary building. With these databases and a Bayesian method, the flooding frequencies for the flooding areas were estimated. Subsequently, the Core Damage Frequency(CDF) for the flooding PSA of the Ulchin(UCN) unit 3 and 4 plants based on the Korean Standard Nuclear Power Plant(KSNP) internal full-power PSA model was recalculated. The evaluation results showed that sixteen flooding events are potentially significant according to the screening criterion, while there were two flooding events exceeding the screening criterion of the existing UCN 3 and 4 flooding PSA. The result was compared with two kinds of cases: (1) the flooding frequency and CDF from the method of the existing flooding PSA with the PWR and Boiled Water Reactor(BWR) data of the NPE database and the Maximum Likelihood Estimate(MLE) method and (2) the flooding frequency and CDF with the NPE database(PWR and BWR data), NuPIPE database, and a Bayesian method. From the comparison, a difference in CDF results was revealed more clearly between the CDF from this study and case (2) than between case (1) and case (2). That is, the number of flooding events exceeding the screen criterion further increased when only the PWR data were used for the primary auxiliary building than when the Korean specific data were used.

• 한국가스학회지
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• 제26권1호
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• pp.7-19
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• 2022

일반적으로 수소의 정제는 화학적, 물리적인 방법을 통해 수행한다. 여러 종류의 정제방법 중 현재는 정제 용량 및 경제성이 가장 우수한 PSA(Pressure Swing Adsorption)를 이용한 정제방법이 가장 널리 사용되고 있다. 국내도 대부분 PSA를 이용하여 자동차 및 발전용 수소 연료전지 등에 사용하는 수소를 정제하고 있다. 기존 석유화학 단지중심의 부생수소는 운송 등의 어려움이 있다. 정부는 도시가스 공급망과 연계하여 소비지에서 직접 수소를 생산하는 수소추출기 설치 계획하고 있으며, 기업들도 이와 관련된 연구 및 실증 설비를 속속 설치하고 있는 실정이다. 유럽 등은 최근 PSA와 관련된 안전기준을 마련하여 시공 및 운영단계에서 체계적인 안전관리를 위한 노력을 기울이고 있으나, 국내는 PSA와 관련된 안전기준 마련이 아직까지는 미흡하다. 본 연구에서는 기존 PSA를 운영하고 있는 회사의 설문 및 위험성평가를 통해 기존설비의 문제점을 파악하고, 국외 기술기준에 이를 포함한 국내 기술기준을 작성하여 신규설치 및 기존 운영되고 있는 PSA시스템의 안전을 도모하고자 한다.

• 조명전기설비학회논문지
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• 제23권10호
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• pp.36-44
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• 2009

본 논문에서는 PSA(position solar algorithm)을 이용하여 태양광 추적시스템의 발전 효율을 분석하였다. 태양의 위치 추적시스템은 자연환경 조건에 무관한 태양광 발전시스템에 매우 효과적으로 필요하다. 프로그램 방식의 태양광 추적시스템은 구름이나 대기 조건에 의해 일사량이 급하게 변할 경우에도 오동작 없이 정확하게 태양을 추적을 할 수 있다. 따라서, 본 논문에서는 더욱 정확하게 태양의 위치를 추적하기 위한 PSA 알고리즘을 제시하고, 제시한 알고리즘을 이용하여 태양광 발전시스템의 효율을 분석한다. 또한 적용된 알고리즘에 의해 제어된 고도각 및 방위각을 한국 천문연구원에서 제공된 데이터와 비교한다. 본 논문에서는 고도각 및 방위각 제어의 오차와 적용된 알고리즘의 발전효율을 분석하고 결과를 통하여 적용된 알고리즘의 타당성을 입증한다.

• 한국안전학회지
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• 제24권5호
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• pp.63-68
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• 2009

The urgent VAI method development is required since "The Act of Physical Protection and Radiological Emergency that is established in 2003" requires an evaluation of physical threats in nuclear facilities and an establishment of physical protection in Korea. The VAI methodology is developed to (1) make a sabotage model by reusing existing fire/flooding/pipe break PSA models, (2) calculate MCSs and TEPSs, (3) select the most cost-effective TEPS among many TEPSs, (4) determine the compartments in a selected TEPS as vital areas, and (5) provide protection measures to the vital areas. The developed VAI methodology contains four steps, (1) collecting the internal level 1 PSA model and information, (2) developing the fire/flood/pipe rupture model based on level 1 PSA model, (3) integrating the fire/flood/pipe rupture model into the sabotage model by JSTAR, and (4) calculating MCSs and TEPS. The VAT process is performed through the VIPEX that was developed in KAERI. This methodology serves as a guide to develop a sabotage model by using existing internal and external PSA models. When this methodology is used to identify the vital areas, it provides the most cost-effective method to save the VAI and physical protection costs.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1297-1306
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• 2019

Probabilistic safety assessments (PSA) have been used for several decades to visualize the risk level of commercial nuclear power plants (NPPs). Since the role of a human reliability analysis (HRA) is to provide human error probabilities for safety critical tasks to support PSA, PSA quality is strongly affected by HRA quality. Therefore, it is important to understand the underlying limitations or problems of HRA techniques. For this reason, this study conducted a survey among 14 subject matter experts who represent the HRA community of domestic Korean NPPs. As a result, five significant HRA issues were identified: (1) providing a technical basis for the K-HRA (Korean HRA) method, and developing dedicated HRA methods applicable to (2) diverse external events to support Level 1 PSA, (3) digital environments, (4) mobile equipment, and (5) severe accident management guideline tasks to support Level 2 PSA. In addition, an HRA method to support multi-unit PSA was emphasized because it plays an important role in the evaluation of site risk, which is one of the hottest current issues. It is believed that creating such a catalog of prioritized issues will be a good indication of research direction to improve HRA and therefore PSA quality.