• 대한안전경영과학회지
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• 제24권1호
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• pp.61-72
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• 2022

This study aims to propose a conceptual design of information displays for supporting responsive actions under severe accidents in Nuclear Power Plants (NPPs). Severe accidents in NPPs can be defined as accident conditions that are more severe than a design basis accident and involving significant core degradation. Since the Fukushima accident in 2011, the management of severe accidents is increasing important in nuclear industry. Dealing with severe accidents involves several cognitively complex activities, such as situation assessment; accordingly, it is significant to provide human operators with appropriate knowledge support in their cognitive activities. Currently, severe accident management guidelines (SAMG) have been developed for this purpose. However, it is also inevitable to develop information displays for supporting the management of severe accidents, with which human operators can monitor, control, and diagnose the states of NPPs under severe accident situations. It has been reported that Ecological Interface Design (EID) framework can be a viable approach for developing information displays used in complex socio-technical systems such as NPPs. Considering the design principles underlying the EID, we can say that EID-based information displays can be useful for dealing with severe accidents effectively. This study developed a conceptual design of information displays to be used in severe accidents, following the stipulated design process and principles of the EID framework. We particularly attempted to develop a conceptual design to make visible the principle knowledge to be used for coping with dynamically changing situations of NPPs under severe accidents.

• Nuclear Engineering and Technology
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• 제46권1호
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• pp.27-38
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• 2014

The Fukushima Dai-Ichi accident in 2011 has affected various aspects of the nuclear society worldwide. The accident revealed some problems in the conventional approaches used to ensure the safety of nuclear installations. To prevent such disastrous accidents in the future, we have to learn from them and improve the conventional approaches in a more systematic manner. In this paper, we will cover three issues. The first is to identify the key issues that affected the progress of the Fukushima Dai-Ichi accident greatly. We examine the accident from a defense-in-depth point of view to identify such issues. The second is to develop a more systematic approach to enhance the safety of nuclear installations. We reexamine nuclear safety from a risk point of view. We use the concepts of residual and unknown risks in classifying the risk space. All possible accident scenarios types are reviewed to clarify the characteristics of the identified issues. An approach is proposed to improve our conventional approaches used to ensure nuclear safety including the design of safety features and the safety assessments from a risk point of view. Finally, we address some issues to be improved in the conventional risk assessment and management framework and/or practices to enhance nuclear safety.

• Nuclear Engineering and Technology
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• 제47권1호
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• pp.74-84
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• 2015

If a transient occurs in a nuclear power plant (NPP), operators will try to protect the NPP by estimating the kind of abnormality and mitigating it based on recommended procedures. Similarly, operators take actions based on severe accident management guidelines when there is the possibility of a severe accident occurrence in an NPP. In any such situation, information about the occurrence time of severe accident-related events can be very important to operators to set up severe accident management strategies. Therefore, support systems that can quickly provide this kind of information will be very useful when operators try to manage severe accidents. In this research, the occurrence times of several events that could happen during a severe accident were predicted using support vector machines with short time variations of plant status variables inputs. For the preliminary step, the break location and size of a loss of coolant accident (LOCA) were identified. Training and testing data sets were obtained using the MAAP5 code. The results show that the proposed algorithm can correctly classify the break location of the LOCA and can estimate the break size of the LOCA very accurately. In addition, the occurrence times of severe accident major events were predicted under various severe accident paths, with reasonable error. With these results, it is expected that it will be possible to apply the proposed algorithm to real NPPs because the algorithm uses only the early phase data after the reactor SCRAM, which can be obtained accurately for accident simulations.

• 한국안전학회지
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• 제33권1호
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• pp.66-72
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• 2018

This study analyses the types, related operations, facilities, and causes of chemical accidents in Korea based on the RISCAD classification taxonomy. In addition, human error analysis was carried out employing different human error classification criteria. Explosion and fire were major accident types, and nearly half of the accidents occurred during maintenance operation. In terms of related facility, storage devices and separators were the two most frequently involved ones. Results of the human error-based analysis showed that latent human errors in management level are involved in many accidents as well as active errors in the field level. Action errors related to unsafe behavior leads to accidents more often compared with the checking behavior. In particular, actions missed and inappropriate actions were major problems among the unsafe behaviors, which implicates that the compliance with the work procedure should be emphasized through education/training for the workers and the establishment of safety culture. According to the analysis of the causes of the human error, the frequency of skill-based mistakes leading to accidents were significantly lower than that of rule-based and knowledge based mistakes. However, there was limitation in the analysis of the root causes due to limited information in the accident investigation report. To solve this, it is suggested to adopt advanced accident investigation system including the establishment of independent organization and improvement in regulation.

• Nuclear Engineering and Technology
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• 제26권2호
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• pp.168-178
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• 1994

사고관리란 사고발생시에 이용가능한 모든 자원, 즉 인원과 설비를 효율적으로 활용함으로써 발전소를 안전상태로 회복시키거나 사고의 피해를 완화시키기 위한 제반 활동을 말한다. 사고관리의 접근방식은 첫째, 후보사고관리방안의 사전 평가, 둘째, 효과적으로 적절한 조치를 수행하게 하는 세부 절차서의 개발, 그리고 셋째, 그러한 조치수행에 필요한 도구와 자원의 준비, 실현 가능한 원전 시스템의 변경등을 포함한다. 사고관리 전략을 평가할 때에는 그 전략의 효율성분만 아니라 부작용, 타당성, 필요한 정보, 기존 절차서와의 양립성 등을 종합적으로 고려하여야 한다. 이 논문의 목적은 여러가지 사고관리 전략을 모델링하고 평가하기위한 체제를 개발하기 위한 의사결정 수목과 영향도의 해석도구를 소개하는 것이다. 이 해석도구와 관련한 여러가지 특징들이 제시되었으며 이 해석도구에 근거하여 세워진 사고관리전략의 평가체제가 간단한 예제문제에 적용되었다.

• International Journal of Reliability and Applications
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• 제2권1호
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• pp.37-48
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• 2001

This paper presents a new dynamic human reliability analysis method and its application for quantifying the human error probabilities in implementing management action. For comparisons of current HRA methods with the new method, the characteristics of THERP, HCR, and SLIM-MAUD, which are most frequency used method in PSAs, are discussed. The action associated with implementation of the cavity flooding during a station blackout sequence is considered for its application. This method is based on the concepts of the quantified correlation between the performance requirement and performance achievement. The MAAP 3.0B code and Latin Hypercube sampling technique are used to determine the uncertainty of the performance achievement parameter. Meanwhile, the value of the performance requirement parameter is obtained from interviews. Based on these stochastic obtained, human error probabilities are calculated with respect to the various means and variances of the things. It is shown that this method is very flexible in that it can be applied to any kind of the operator actions, including the actions associated with the implementation of accident management strategies.

• Spatial Information Research
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• 제12권1호
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• pp.89-100
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• 2004

최근 급속한 도시의 팽창과 산업의 발전으로 인하여 가스시설이 급속히 확대됨에 따라 많은 도시가스업체에서는 가스관망 시설정보를 전산화하여 최신의 현황을 유지할 수 있는 가스시설물관리시스템을 개발하여 사용하고 있는 실정이다. 이러한 시스템에서는 가스시설물의 현황파악 및 유지관리를 위한 기본적인 정보만을 제공하고 가스 누출사고시 정확한 사태의 파악과 신속한 대책를 위한 제반 의사결정의 지원은 어려운 실정이다. 본 연구에서는 가스사고의 발생시 신속하고 체계적인 사고 대처를 위한 적용알고리즘의 정립을 통한 가스사고관리시스템을 구현하였다. 본 시스템에서는 1ㆍ2차 차단밸브의 산정을 통하여 가스 누출사고의 발생시 사고 관거의 파악과 함께 가스공급 중단 대상 관로의 파악과 밸브의 현황, 나아가 피해 수용가에 대한 현황 파악이 신속히 이루어 질 수 있도록 하였다. 이러한 가스사고관리시스템의 개발을 통하여 사고 발생시 신속한 대처방안을 고려하여 피해를 최소화 할 수 있다. 나아가 이러한 시스템의 활용을 통하여 가스업체의 업무 효율화에 기여하고 국민의 생명과 재산 보호에 기여도가 크리라 판단된다.

• 대한안전경영과학회지
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• 제9권3호
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• pp.29-40
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• 2007

The state and its proper countermeasures of unreported industrial accidents in Gyungbuk area was investigated throughout a survey research. The goal of the study was to identify major factors that affect the number and causes of unreported accidents. Results from the survey indicated that significant number of unreported accidents exists especially in the small and medium sized industries. Types of the accidents, amount of increase in the insurance cost and level of the governmental enforcement due to the accidents were the major factors for deciding whether to report or not. These results suggested more compromising actions have to be taken by the government for revealing the present but unreported industrial accidents. A more efficient way for preventing the industrial accidents can be considered on the basis of true understanding about real industrial accident statistics.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.866-872
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• 2024

Design Extension Conditions (DEC)-A assessments of the operating nuclear power plants are generally considered for the purpose of getting additional safety demonstrations of their capability to undergo conditions that are generally more severe than DBAs by features implemented in the design and accident management measures. The pursued methodology is generally based upon Best Estimate approaches aiming at verifying that the safety limits in terms of integrity of the barriers against eventual large or early releases of radioactive material are fulfilled. These aspects are nowadays being experimentally and analytically addressed within the OECD/NEA experimental projects like the ATLAS and PKL series where a set of DEC-A experiments are considered. In this paper, experiments related to SBLOCA at the vessel upper head of the pressurized vessel of ATLAS and PKL are analytically assessed using the CATHARE code. These experiments includes issues related to common cause failure of the safety injection system and operator actions for preventing core excessive overheating. It is shown that, on the one hand, the safety features embedded in the design together with the operator actions are capable to prevent the progression towards a severe accident state and on the other hand, the code prediction capabilities for such scenario are generally good but still to be enhanced.

• Nuclear Engineering and Technology
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• 제50권3호
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• pp.356-367
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• 2018

A series of tests dedicated to station blackout (SBO) accident scenarios have been recently performed at the $Prim{\ddot{a}}rkreislauf-Versuchsanlage$ (primary coolant loop test facility; PKL) facility in the framework of the OECD/NEA PKL-3 project. These investigations address current safety issues related to beyond design basis accident transients with significant core heat up. This work presents a detailed analysis using the best estimate thermal-hydraulic code TRACE (v5.0 Patch4) of different SBO scenarios conducted at the PKL facility; failures of high- and low-pressure safety injection systems together with steam generator (SG) feedwater supply are considered, thus calling for adequate accident management actions and timely implementation of alternative emergency cooling procedures to prevent core meltdown. The presented analysis evaluates the capability of the applied TRACE model of the PKL facility to correctly capture the sequences of events in the different SBO scenarios, namely the SBO tests H2.1, H2.2 run 1 and H2.2 run 2, including symmetric or asymmetric secondary side depressurization, primary side depressurization, accumulator (ACC) injection in the cold legs and secondary side feeding with mobile pump and/or primary side emergency core coolant injection from the fuel pool cooling pump. This study is focused specifically on the prediction of the core exit temperature, which drives the execution of the most relevant accident management actions. This work presents, in particular, the key improvements made to the TRACE model that helped to improve the code predictions, including the modeling of dynamical heat losses, the nodalization of SGs' heat exchanger tubes and the ACCs. Another relevant aspect of this work is to evaluate how well the model simulations of the three different scenarios qualitatively and quantitatively capture the trends and results exhibited by the actual experiments. For instance, how the number of SGs considered for secondary side depressurization affects the heat transfer from primary side; how the discharge capacity of the pressurizer relief valve affects the dynamics of the transient; how ACC initial pressure and nitrogen release affect the grace time between ACC injection and subsequent core heat up; and how well the alternative feeding modes of the secondary and/or primary side with mobile injection pumps affect core quenching and ensure stable long-term core cooling under controlled boiling conditions.