• Nuclear Engineering and Technology
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• 제41권4호
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• pp.545-548
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• 2009

Korean utility has utilized a Periodic Safety Review (PSR) that assesses the cumulative effects of plant aging, modifications, operating experience, technical developments, and site characteristics since 2000. In particular, the assessment and management of plant aging is one of the major areas in PSR. It includes identification of critical Systems, Structures, and Components (SSCs) for aging, assessment of aging effects, and implementation of aging management programs. Since the PSR system was introduced based on the atomic energy acts and related laws, PSRs of eight sets for 12 Nuclear Power Plants (NPPs) that have been operating more than 10 years have been completed. PSRs of two sets for 4 NPPs are currently being carried out. The utility has confirmed that domestic NPPs have been operated safely through these PSRs and have implemented the follow-up corrective activities to increase the nuclear safety. In this paper, the status of PSR implementation is discussed and improvement programs to conduct PSR follow-up corrective activities efficiently for NPPs are suggested based on experiences with aging assessments.

• 디지털융복합연구
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• 제17권8호
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• pp.47-60
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• 2019

본 연구의 목적은 네트워크 분석을 통하여 원자력 법제의 전체적인 법령조항 간의 구조관계를 분석하여 법적 체계의 정합성을 파악하고자 한다. 특히 원자력기술의 안전규제의 중심 법으로서의 역할을 하고 있는 "원자력안전법"의 법령 구조를 파악하여 안전관리에 있어서의 주요 규정에 대해 검토하고 원자력 안전관리 및 규제에 대한 입법적 개선방안을 제시한다. 동법의 법적체계의 구조적 문제점을 파악하여 입법 개선 방안을 제시하고 이를 통해 원자력기술 및 산업관련 정책수립 활동 과정에서 과도한 입법 활동을 줄이고, 제 개정의 필요성 시급성 여부를 결정하는 데 있어서 중요한 역할을 할 수 있을 것으로 기대한다. 본 연구는 또한 향후 타 과학기술의 정책분야에 적용하여 법률적 개선 방안을 마련하기 위한 가이드라인으로 활용될 수 있을 것으로 사료된다.

• Nuclear Engineering and Technology
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• 제45권2호
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• pp.165-178
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• 2013

A comprehensive review on the technical standards about human factors (HF) design and software reliability maintenance for digital instrumentation and control (I&C) and human-machine interface technology (HMIT) in Japanese light water reactor nuclear power plants (NPPs) was given in this paper mainly by introducing the relevant activities at the Japan Electric Association to set up many industrial standards within the traditional framework of nuclear safety regulation in Japan. In Japan, the Fukushima Daiichi accident that occurred on March 11, 2011 has great impact on nuclear regulation and nuclear industries where concerns by the general public about safety have heightened significantly. However for the part of HF design and software reliability maintenance of digital I&C and HMIT for NPP, the author believes that the past practice of Japanese activities with the related technical standards can be successfully inherited in the future, by reinforcing the technical preparedness for the prevention and mitigation against any types of severe accident occurrence.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.781-782
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• 2005

• 에너지공학
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• 제23권1호
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• pp.90-94
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• 2014

본 논문은 원자력발전소의 내환경기기검증(EQ)을 위한 HVAC 과도분석 방법에 대한 내용을 기술하고 있다. 분석 대상 격실은 비안전관련 HVAC 계통에 의해 공급되는 격실 중에 원자로 안전정지를 담당하는 중요기기가 위치한 구역/격실을 선정하였다. 그리고 해당 HVAC 계통이 공급되는 건물별로 HVAC 과도시 온도조건을 분석하였다. 본 분석을 위해서 GOTHIC 전산코드를 사용하였다. 온도분석 결과는 원자로 보조건물 환기계통(DVN)의 W315/W415 격실에서 $82.2^{\circ}C$로 가장 높은 온도값을 나타내며, 제어봉구동장치 전원공급건물 및 보조급수펌프실(DVG) 계통의 W229 (Auxiliary feedwater pump room) 격실에서 $52.7^{\circ}C$, 기기냉각건물 환기계통(DVI)의 전 격실에서 $42.9^{\circ}C$, 전기건물 주환기 계통(DVL)의 L207 (Hot workshop) 격실에서 $57.3^{\circ}C$를 각각 나타났다. 이러한 온도값은 일반적인 원전의 기기검증 제한값인 $171^{\circ}C$이하이므로 내환경검증 요건을 만족하는 온도이다.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.581-588
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• 2013

This paper deals with the Safety Analysis for $CANDU^{(R)}$ 6 nuclear reactors as affected by main Heat Transport System (HTS) aging. Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermal-hydraulic analytic models. Safety analyses ranging from the assessment of safety limits associated with the prevention of intermittent fuel sheath dryout for a slow Loss of Regulation (LOR) analysis and fission gas release after a fuel failure are summarized. Specifically for fission gas release, the thermal-hydraulic analysis for a fresh core and an 11 Effective Full Power Years (EFPY) aged core was summarized, leading to the most severe stagnation break sizes for the inlet feeder break and the channel failure time. Associated coolant conditions provide the input data for fuel analyses. Based on the thermal-hydraulic data, the fission product inventory under normal operating conditions may be calculated for both fresh and aged cores, and the fission gas release may be evaluated during the transient. This analysis plays a major role in determining possible radiation doses to the public after postulated accidents have occurred.

• Nuclear Engineering and Technology
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• 제46권4호
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• pp.481-488
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• 2014

An analysis of the uncertainty quantification related to LBLOCA using the Monte-Carlo calculation has been performed and compared with the tolerance level determined by the Wilks' formula. The uncertainty range and distribution of each input parameter associated with the LOCA phenomena were determined based on previous PIRT results and documentation during the BEMUSE project. Calulations were conducted on 3,500 cases within a 2-week CPU time on a 14-PC cluster system. The Monte-Carlo exercise shows that the 95% upper limit PCT value can be obtained well, with a 95% confidence level using the Wilks' formula, although we have to endure a 5% risk of PCT under-prediction. The results also show that the statistical fluctuation of the limit value using Wilks' first-order is as large as the uncertainty value itself. It is therefore desirable to increase the order of the Wilks' formula to be higher than the second-order to estimate the reliable safety margin of the design features. It is also shown that, with its ever increasing computational capability, the Monte-Carlo method is accessible for a nuclear power plant safety analysis within a realistic time frame.

• Nuclear Engineering and Technology
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• 제41권6호
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• pp.849-858
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• 2009

Risk caused by safety-critical instrumentation and control (I&C) systems considerably affects overall plant risk. As digitalization of safety-critical systems in nuclear power plants progresses, a risk model of a digitalized safety system is required and must be included in a plant safety model in order to assess this risk effect on the plant. Unique features of a digital system cause some challenges in risk modeling. This article aims at providing an overview of the issues related to the development of a static fault-tree-based risk model. We categorize the complicated issues of digital system probabilistic risk assessment (PRA) into four groups based on their characteristics: hardware module issues, software issues, system issues, and safety function issues. Quantification of the effect of these issues dominates the quality of a developed risk model. Recent research activities for addressing various issues, such as the modeling framework of a software-based system, the software failure probability and the fault coverage of a self monitoring mechanism, are discussed. Although these issues are interrelated and affect each other, the categorized and systematic approach suggested here will provide a proper insight for analyzing risk from a digital system.

• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2660-2666
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• 2020

Korea successfully achieved energy independence in the shortest period of time from being the poorest country in terms of energy 50 years ago through steady development of nuclear technology. In the past, the nuclear industry has been driven through government-centered policy development, public institution-based research, and industrial facility and infrastructure construction. Consequently, South Korea became a nuclear energy powerhouse exporting nuclear power plants to the UAE, surpassing the level of domestic technological independence. However, in recent years, the nuclear industry in Korea has experienced a decline in new plant construction since the Fukushima accident in Japan, which caused changes in public perspectives regarding nuclear power plant operation, more stringent safety standards on the operation of nuclear power plants, and a shift in governmental energy policy. These changes are expected to change the domestic nuclear industry ecosystem. Therefore, in this study, we investigate the priority of technology development investment from the perspective of experts in private nuclear power companies, shifting the focus from government-led nuclear R&D policies. To establish a direction in nuclear technology development, a survey was conducted by applying an analytic hierarchy analysis to experts who have worked in nuclear power plants for more than 15 years. The analysis items of focus were the 3 attributes of strategic importance, urgency, and business feasibility of four major fields related to nuclear energy: nuclear safety, decommissioning, radioactive waste management, and strengthening industrial competitiveness.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.2096-2106
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• 2023

Rotating machinery is widely applied in important equipment of nuclear power plants (NPPs), such as pumps and valves. The research on intelligent fault diagnosis of rotating machinery is crucial to ensure the safe operation of related equipment in NPPs. However, in practical applications, data-driven fault diagnosis faces the problem of small and imbalanced samples, resulting in low model training efficiency and poor generalization performance. Therefore, a deep convolutional conditional generative adversarial network (DCCGAN) is constructed to mitigate the impact of imbalanced samples on fault diagnosis. First, a conditional generative adversarial model is designed based on convolutional neural networks to effectively augment imbalanced samples. The original sample features can be effectively extracted by the model based on conditional generative adversarial strategy and appropriate number of filters. In addition, high-quality generated samples are ensured through the visualization of model training process and samples features. Then, a deep convolutional neural network (DCNN) is designed to extract features of mixed samples and implement intelligent fault diagnosis. Finally, based on multi-fault experimental data of motor and bearing, the performance of DCCGAN model for data augmentation and intelligent fault diagnosis is verified. The proposed method effectively alleviates the problem of imbalanced samples, and shows its application value in intelligent fault diagnosis of actual NPPs.