• 수산해양교육연구
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• 제29권3호
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• pp.920-931
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• 2017

Since the export of Korean-type APR 1400 in 2009 to the UAE, Korea has been achieved management performance, quality inspections, training, nuclear fuel exports for the nuclear power plant. Despite this apparent growth, there are lacking of the research on the marine transportation of radioactive waste. And the terrible accident at the Japan nuclear power plant in 2011 has caused another reconsideration such as emergency response training and plan, reinforcement of safety regulation. According to the Korean government aims to rebuild the appropriate regulation, training, education that is necessary in order to ensure the safety of marine transportation of radioactive waste. Therefore, this study analyzed the various problems identified by the team of experts for the radioactive waste and marine field, the investigation of relevant legal basis, the need for emergency response training for the person in charge of radioactive waste and suggested the simulation-based interactive curriculum during the process of safety verification related to the marine transport of mid- and low-level radioactive waste generated at the Yeon-ggwang nuclear power(Hanbit) plant in 2015.

• Nuclear Engineering and Technology
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• 제23권3호
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• pp.285-298
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• 1991

전형적인 정성적 퍼지형태의 입력데이타를 가진, 주어진 사고진행사건수목의 일부분에 대하여 퍼지집합이론(fuzzy set theory)의 응용 예를 먼저 보여주고, 이 예를 통해서 퍼지집합이론을 사고 진행사건수목에 적용하기 위해 적절한 계산알고리즘을 찾아내고 또 예를 들어 설명하였다. 그리고, 간단한 예제에 사용한 계산절차를 많은 현상학적 불확실성 인자를 포함한 아주 복잡한 사고진행사건수목 즉, 최근 Zion 발전소 위험도평가(PRA)에 사용된 전형적인 발전소 손상군의 하나인‘SEC’에 응용해서 적용하였다. 퍼지집합이론으로 평가한 계산값들의 퍼지평균치들은 최근 통계적 PRA 평가 방법론으로 얻는 값들의 평균치와 거의 같은 결과를 보여주고 있다. 본 논문의 주요목적은 부정확하고 또 정성적인 분기점확률이나 또는 많은 현상학적 불확실성 인자들을 가진 사고진행사건수목들에 이 퍼지집합이론을 적용하기 위한 공식적 계산절차를 제공하는데 있다.

• 시스템엔지니어링학술지
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• 제11권1호
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• pp.1-7
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• 2015

The purpose of this paper is to present the results of reengineering of the YGN 3&4 (Yonggwang Nuclear Power Plant, Units 3&4) SDS (Safety Depressurization System) retrofit design and to make recommendations for the improvement in design and design procedure implementing the Systems Engineering (SE) process. YGN 3&4 is a basic model for OPR1000 (the Korean standard 1000 MWe plant). The basic model, herein, represents the reference plant for the OPR1000 development. In the middle of the YGN 3&4 construction, the Korean Nuclear Regulatory Body requested a retrofit of this plant with a means to rapidly depressurize the plant in conformance with a severe accident mitigation requirement. For the reengineering of the SDS in YGN 3&4, V-model and functional and physical architectures have been developed. A SE decision making method has been used for the selection of SDS valves. Finally, recommendations have been made to improve OPR1000 design for the improved operation and enhanced safety.

• 대한산업공학회지
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• 제16권2호
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• pp.135-147
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• 1990

A nuclear power plant can be viewed as a large complex man-machine system where high system reliability is obtained by ensuring that sub-systems are designed to operate at a very high level of performance. The chance of severe accident involving at least partial core-melt is very low but once it happens the consequence is very catastrophic. The prediction of risk in low probability, high-risk incidents must be examined in the contest of general engineering knowledge and operational experience. Engineering knowledge forms part of the prior information that must be quantified and then updated by statistical evidence gathered from operational experience. Recently, Bayesian procedures have been used to estimate rate of accident and to predict future risks. The Bayesian procedure has advantages in that it efficiently incorporates experts opinions and, if properly applied, it adaptively updates the model parameters such as the rate or probability of accidents. But at the same time it has the disadvantages of computational complexity. The predictive distribution for the time to next incident can not always be expected to end up with a nice closed form even with conjugate priors. Thus we often encounter a numerical integration problem with high dimensions to obtain a predictive distribution, which is practically unsolvable for a model that involves many parameters. In order to circumvent this difficulty, we propose a method of approximation that essentially breaks down a problem involving many integrations into several repetitive steps so that each step involves only a small number of integrations.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2534-2546
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• 2021

Nuclear emergency preparedness and response is an essential part to ensure the safety of nuclear power plant (NPP). Key support technologies of nuclear emergency decision-making usually consist of accident diagnosis, source term estimation, accident consequence assessment, and protective action recommendation. Source term estimation is almost the most difficult part among them. For example, bad communication, incomplete information, as well as complicated accident scenario make it hard to determine the reactor status and estimate the source term timely in the Fukushima accident. Subsequently, it leads to the hard decision on how to take appropriate emergency response actions. Hence, this paper aims to develop a method for rapid source term estimation to support nuclear emergency decision making in pressurized water reactor NPP. The method aims to make our knowledge on NPP provide better support nuclear emergency. Firstly, this paper studies how to build a Bayesian network model for the NPP based on professional knowledge and engineering knowledge. This paper presents a method transforming the PRA model (event trees and fault trees) into a corresponding Bayesian network model. To solve the problem that some physical phenomena which are modeled as pivotal events in level 2 PRA, cannot find sensors associated directly with their occurrence, a weighted assignment approach based on expert assessment is proposed in this paper. Secondly, the monitoring data of NPP are provided to the Bayesian network model, the real-time status of pivotal events and initiating events can be determined based on the junction tree algorithm. Thirdly, since PRA knowledge can link the accident sequences to the possible release categories, the proposed method is capable to find the most likely release category for the candidate accidents scenarios, namely the source term. The probabilities of possible accident sequences and the source term are calculated. Finally, the prototype software is checked against several sets of accident scenario data which are generated by the simulator of AP1000-NPP, including large loss of coolant accident, loss of main feedwater, main steam line break, and steam generator tube rupture. The results show that the proposed method for rapid source term estimation under nuclear emergency decision making is promising.

• 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.

• 대한전기학회:학술대회논문집
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• 대한전기학회 1998년도 하계학술대회 논문집 C
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• pp.1170-1172
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• 1998

The electrical power of nuclear power plant consists of Safety related power systems and Non-Safety related power systems. The safety related power systems are designed to have sufficient capacity to safely shut down the unit and to mitigate the effects of an accident assuming loss of off-site power. This paper presents the operation scheme of the safety related power system for several plant conditions in Korean Next Generation Reactor and reviews the diversity of power supply to the safety related bus.

• Nuclear Engineering and Technology
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• 제29권1호
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• pp.25-34
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• 1997

This study presents a new approach for expert opinion elicitation process to assess an uncertainty inherent in accident management. The need to work with rare event and limited data in accident management leads analysis to use expert opinions extensively. Unlike the conventional approach using point-valued probabilities, the study proposes the concept of fuzzy probability to represent expert opinion. The use of fuzzy probability has an advantage over the conventional approach when an expert's judgment is used under limited dat3 and imprecise knowledge. The study demonstrates a method of combining and propagating fuzzy probabilities. finally, the proposed methodology is applied to the evaluation of the probability of a bottom head failure for the flooded case in the Peach Bottom BWR nuclear power plant.

• Nuclear Engineering and Technology
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• 제36권5호
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• pp.430-449
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• 2004

A quantitative model for a control system that integrates human operators, systems, and their interactions is developed based on discrete functions. After identifying the major entities and the key factors that are important to each entity in the control system, a quantitative analysis to estimate the recovery failure probability from an abnormal state is performed. A numerical analysis based on assumed values of related variables shows that this model produces reasonable results. The concept of 'relative sensitivity' is introduced to identify the major factors affecting the reliability of the control system. The analysis shows that the hardware factor and the design factor of the instrumentation system have the highest relative sensitivities in this model. T도 probability of human operators performing incorrect actions, along with factors related to human operators, are also found to have high relative sensitivities. This model is applied to an analysis of the TMI-2 nuclear power plant accident and systematically explains how the accident took place.

• Nuclear Engineering and Technology
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• 제49권3호
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• pp.459-465
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• 2017

The goal of this study is to evaluate the effective thermal conductivity and diffusivity of containment walls as heat sinks or passive cooling systems during nuclear power plant (NPP) accidents. Containment walls consist of steel reinforced concrete, steel liners, and tendons, and provide the main thermal resistance of the heat sinks, which varies with the volume fraction and geometric alignment of the rebar and tendons, as well as the temperature and chemical composition. The target geometry for the containment walls of this work is the standard Korean NPP OPR1000. Sample tests and numerical simulations are conducted to verify the correlations for models with different densities of concrete, volume fractions, and alignments of steel. Estimation of the effective thermal conductivity and diffusivity of the containment wall models is proposed. The Maxwell model and modified Rayleigh volume fraction model employed in the present work predict the experiment and finite volume method (FVM) results well. The effective thermal conductivity and diffusivity of the containment walls are summarized as functions of density, temperature, and the volume fraction of steel for the analysis of the NPP accidents.