• 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.

• 한국안전학회지
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• 제31권5호
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• pp.177-186
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• 2016

When an event is occurred in a nuclear power plant (NPP), the NPP operator reports it referred by the regulation on reporting and public announcement of accidents and incidents. Some of the events do not need to be reported because they are not included in the reporting criteria of the regulation. However, it is necessary that they should be managed effectively because the accident can be occurred by the recurrence of a lot of them as precursors. Among the events not included in the reporting criteria of the regulation, near miss is the event that is not occurred but can generate a significant consequence. This can provide the cause of the event which does not result an accident. So, it is able to offer insightful knowledges to prevent higher level events about the function and process of NPP. The objective of this study is to analyze the issues of near miss events, prepare the defence against the risk, and improve the management process of NPP. To achieve it, this study performed to analyze the management structure and status of near miss events as well as the accident reporting system of the domestic and foreign regulation bodies. In case of Korea, the status was analyzed by quantitative data, licensee event reports and procedures. Based on these, we could find the causes that near miss events were not managed effectively. Then, systematic alternatives that reflected the perspective of man, technology and organization were drawn.

• 대한기계학회논문집B
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• 제37권10호
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• pp.961-969
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• 2013

본 연구에서는 상용 전산유체역학 소프트웨어인 ANSYS CFX V.13을 사용하여 정상상태, 비압축성, 등온으로 가정된 엇갈림 관 다발 및 일렬 관 다발 주위의 난류유동을 계산하였다. 계산 결과는 전산유체역학 우수사례 지침에 근거하여 격자크기, 대류항 차분법의 정확도 및 난류모델에 대한 민감도 연구에 활용되었고 실험 결과와 정량적으로 비교함으로써 우수사례 지침의 적용성을 평가하였다. 결론적으로 전산유체역학 우수사례 지침이 관 다발 유동 분야에서 상용 전산유체역학 소프트웨어의 예측성능 향상을 반드시 보증하지 않음을 확인하였다.

• Nuclear Engineering and Technology
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• 제50권7호
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• pp.997-1005
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• 2018

The regulatory periodic safety inspection system is one of the most important methods for confirming the safety of nuclear power plants and the defense in depth in nuclear safety is the most important basic means for accident prevention and mitigation. Recently, a new regulatory technology based on risk-informed and safety performance has been developed and used in advanced countries. However, since the domestic periodic inspection system is being used in the same way over 30 years, it is necessary to know how the inspection contributes to the safety confirmation of the nuclear power plants. In this study, the domestic periodic inspection system currently in use was analyzed from the perspective of defense in depth in nuclear safety. In addition, the analysis results were compared to the U.S. NRC's safety inspection system to obtain consistency and lessons in this study. As a result of analysis, the NRC's safety inspections were distributed almost evenly at the all levels of defense in depth, while in the case of domestic inspection, they were heavily focused on the level 1 of defense in depth. Therefore, it appeared urgent to improve the inspection system to strengthen the other levels of defense in depth in nuclear safety.

• Nuclear Engineering and Technology
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• 제52권1호
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• pp.69-77
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• 2020

After the Fukushima Daiichi accidents, public concerns on nuclear safety and the corresponding burden of nuclear power plant licensees are increasing. In order to secure public trust and enhance the rationality of current safety regulation, we develop a risk-informed decision making (RIDM) framework for the Korean regulatory body. By analyzing all the regulatory activities for nuclear power plants in Korea, eight action items are selected for RIDM implementation, with appropriate procedures developed for each. For two items in particular - the accident sequence precursor analysis (ASPA) and the significance determination process (SDP) - two customized risk evaluation software has been developed for field inspectors and probabilistic safety assessment experts, respectively. The effectiveness of the proposed RIDM framework is demonstrated by applying the ASPA procedure to 35 unplanned scrams and the SDP to 24 findings from periodic inspections.

• 자원ㆍ환경경제연구
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• 제26권2호
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• pp.139-172
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• 2017

본 연구는 국제원자력기구(International Atomic Energy Agency: IAEA)의 원자로정보시스템(Power Reactor Information System: PRIS)에서 제공하는 원자로 운전실적지표를 이용하여 원전의 안전성에 미치는 기술적 비기술적 요인의 역할을 분석한다. 이를 위해 안전성의 척도로 원전의 고장정지에 따른 발전손실률 (FLR: forced loss rate)을 사용했다. 1970년부터 2015년까지 전 세계에서 운영된 모든 원전으로 구성된 패널자료를 통해 분석한 결과, 기존 연구와 마찬가지로 원전의 전반적인 기술수준과 정비기술수준이 향상될수록 FLR이 하락하는 사실을 확인했다. 하지만 1986년 체르노빌 원전 사고 이후 기술적 요인이 통제된 FLR은 유의적으로 상승했다. 이는 원자력발전사업자가 체르노빌 사고 이후 원자력안전을 위해 보다 많은 기회비용을 지불하고 있다고 해석된다.

• Nuclear Engineering and Technology
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• 제44권6호
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• pp.663-672
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• 2012

As software based digital I&C (Instrumentation and Control) systems are used more prevalently in nuclear plants, enhancement of software dependability has become an important issue in the area of nuclear I&C systems. Critical attributes of software dependability are safety and reliability. These attributes are tightly related to software failures caused by faults. Software testing and V&V (Verification and Validation) activities are hence important for enhancing software dependability. If the risky modules of safety-critical software can be predicted, it will be possible to focus on testing and V&V activities more efficiently and effectively. It should also make it possible to better allocate resources for regulation activities. We propose a prediction technique to estimate risky software modules by adopting machine learning models based on software complexity metrics. An empirical study with various machine learning algorithms was executed for comparing the prediction performance. Experimental results show SVMs (Support Vector Machines) perform as well or better than the other methods.

• Nuclear Engineering and Technology
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• 제47권3호
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• pp.240-252
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• 2015

In the original licensing application for the prototype fast-breeder reactor, MONJU, the event progression during an unprotected loss of flow (ULOF), which is one of the technically inconceivable events postulated beyond design basis, was evaluated. Through this evaluation, it was confirmed that radiological consequences could be suitably limited even if mechanical energy was released. Following the Fukushima-Daiichi accident, a new nuclear safety regulation has become effective in Japan. The conformity of MONJU to this new regulation should hence be investigated. The objectives of the present study are to conduct a preliminary evaluation of ULOF for MONJU, reflecting the knowledge obtained after the original licensing application through CABRI experiments and EAGLE projects, and to gain the prospect of in-vessel retention for the conformity of MONJU to the new regulation. The preliminary evaluation in the present study showed that no significant mechanical energy release would take place, and that thermal failure of the reactor vessel could be avoided by the stable cooling of disrupted-core materials. This result suggests that the prospect of in-vessel retention against ULOF, which lies within the bounds of the original licensing evaluation and conforms to the new nuclear safety regulation, will be gained.

• Nuclear Engineering and Technology
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• 제37권3호
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• pp.207-220
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• 2005

Nuclear power plants (NPP) are complex socio-technological systems that rely on the success of both hardware and human components. Empirical studies of plant operating experience show that human errors are important contributors to accidents and incidents, and that organizational factors play an important role in creating contexts for human errors. Current probabilistic safety assessments (PSA) do not explicitly model the systematic contribution of organizational factors to safety. As some countries, like the United States, are moving towards increased use of risk information in the regulation and operation of nuclear facilities, PSA quality has been identified as an area for improvement. The modeling of human errors, and underlying organizational weaknesses at the root of these errors, are important sources of uncertainty in existing PSAs and areas of on-going research. This paper presents a review of research into the following questions: Is there evidence that organizational factors are important to NPP safety? How do organizations contribute to safety in NPP operations? And how can these organizational contributions be captured more explicitly in PSA? We present a few past incidents that illustrate the potential safety implications of organizational deficiencies, some mechanisms by which organizational factors contribute to NPP risk, and some of the methods proposed in the literature for performing root-cause analyses and including organizational factors in PSA.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2011년도 제42회 하계학술대회
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• pp.290-291
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• 2011

전력산업이 발전함에 따라 전력품질 요구도 증가하고 있으나 고조파왜형률에 대한 국제기준이 제조사와 운영자의 입장에 따라 계산방법이 상충되는 부분이 현실적으로 존재하고 있는 것으로 검토되었다. 본 논문에서는 원자력발전소에 사용되는 축전지 및 충전기의 시험 및 관리규격인 ANSI/IEEE 519와 NEMA PE-5에 대한 THD 계산방식에 대한 차이점을 설명하고 현장 규제경험 사례를 통해 바람직한 관리 방향을 제시하고자 한다.