• 한국안전학회지
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• 제16권1호
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• pp.84-87
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• 2001

중대사고관리는 원전의 노심손상사고를 예방하거나 완화시키기 위하여 기존의 가용자원이나 시스템, 운전의 행위를 사용하는 것을 말한다. 제어실이나 기술지원반을 위하여 중대사고관리를 위하여 개발된 TRAIN(Training pRogram for Accident Management Program In Nuclear Power Plant)의 초기문자로 명명된 시스템을 본 논문에 소개하였다. TRAIN은 중대사고현상 KB(Knowledge Base)와 사고시나리오 KB, 제어도와 함께 사고관리 절차도 그리고 필요정보로 구성되어있으며 제어실이나 기술지원반에게 중대사고의 현상지식을 취득하게 하고, 발전소의 취약특성을 파악하게 하며, 상당한 스트레스하에서 주어진 문제를 해결하게 하는데 본 연구의 결과는 기여할 것이다.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.1-10
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• 2016

The number of nuclear power plants operating in Korea was 24 as of year 2015. Nine units out of 24 units have been operated for a period over 20 years. Kori unit 1 has been in operation for 40 years, and an extended operation for Wolsong unit 1 was decided in 2015. There has been reported some crackings in reactor internals in PWR have been reported in Europe, USA, Japan and Korea, and some of them were replaced with new one. Repair and replacement technologies for the reactor internals have been developing in order to meet the regulatory requirements for long term operation in Korea. The technologies will also be used for the exported nuclear units. It is required to review degradation history of the reactor internals worldwide as a part of the degradation management program development. Schematics of reactor internals designed and supplied by Westinghouse, Framatome and Combustion Engineering are described herein. Materials degradation history of reactor internals of PWR plants in USA, Japan and Europe is surveyed and summarized. Some events from Korean plants are also described. Aging management strategy for the internals is suggested.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 추계 학술대회 논문집
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• pp.25-31
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• 2005

원자력연구시설에 대한 해체비용 산정은 해체계획 수립하는 데 중요한 작업이다. 해체비용 산정은 해체활동 단계와 해체 시설의 구설요소에 맞게 해체작업을 분류하여 계산을 해야 한다. 본 논문에서는 원자력연구시렁 해체비용 산정을 위하여 해체작업 활동을 분류하고 비용자료의 기준이 되는 비용항목을 계층적으로 세분화하여 구성하는 방법과 작업지연을 유발하는 비용영향 요인인 작업 난이도 인자에 대한 산출방법을 마련하였다. 이렇게 함으로써 해체활동 단계 및 작업에 대한 비용 항목별 분류 및 산정이 가능할 뿐만 아니라 원자력연구시설 해체비용 산정 방법론 및 프로그램을 개발하는데 활용할 예정이다.

• 한국안전학회지
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• 제30권5호
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• pp.131-137
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• 2015

For the evaluation of safety margin of a nuclear power plant using a conservative methodology, the influence of applied assumptions such as initial conditions and boundary conditions needs to be assessed deliberately. Usually, a combination of the most conservative initial conditions is determined, and the safety margin for the transient is evaluated through the analysis for this conservative conditions. In existing conservative methodologies, a most-conservative condition is searched through the analyses for the maximum, minimum, and nominal values of the major parameters. In the present study, we investigates a new approach which can be applied to choose a most-conservative initial condition effectively when a best-estimate computer code and a conservative evaluation methodology are utilized for the evaluation of safety margin of transients. By constituting the band of various initial conditions using the random sampling of input parameters, the sensitivity study for various parameters are performed systematically. A method of sampling the value of control or operation parameters for a certain range is adopted by use of MOSAIQUE program, which enables to minimize the efforts for achieving the steady-state for various different conditions. A representative control parameter is identified, which governs the reactor coolant flow rate, pressurizer pressure, pressurizer level, and steam generator level, respectively. It is shown that an appropriate distribution of input parameter is obtained by adjusting the range and distribution of the control parameter.

• Nuclear Engineering and Technology
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• 제42권2호
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• pp.145-150
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• 2010

The paper is describing work being developed in the frame of a 5-year IAEA Coordinated Research Programme (CRP) started in late 2005. Participants gained knowledge of modelling methodologies and experience in the development and use of rather sophisticated simulation tools in support of site characterization and performance assessment calculations. These goals were achieved by a coordinated effort, in which the advantages and limitations of numerical models are examined and demonstrated through a comparative analysis of simplified, illustrative test cases. This knowledge and experience should help them address these issues in their own country's nuclear waste program. Coordination efforts during the first three years of the project aimed at enabling this transfer of expertise and maximizing the learning experience of the participants as a group. This was accomplished by identifying common interests of the participants (i.e., Process Modelling and Total System Performance Assessment methodology), and by defining complementary tasks that are solved by the members. Synthesis of all available results by comparative assessments is planned in the coming months. The project will be completed end of 2010. This paper is summarizing activities up to November 2009.

• Nuclear Engineering and Technology
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• 제52권3호
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• pp.647-653
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• 2020

Stainless steel is used commonly in nuclear applications for shielding radiation, so in this study, three different types of new stainless steel samples were designed and developed. New stainless steel compound ratios were determined by using Monte Carlo Simulation program Geant 4 code. In the sample production, iron (Fe), nickel (Ni), chromium (Cr), silicium (Si), sulphur (S), carbon (C), molybdenum (Mo), manganese (Mn), wolfram (W), rhenium (Re), titanium (Ti) and vanadium (V), powder materials were used with powder metallurgy method. Total macroscopic cross sections, mean free path and transmission number were calculated for the fast neutron radiation shielding by using (Geant 4) code. In addition to neutron shielding, the gamma absorption parameters such as mass attenuation coefficients (MACs) and half value layer (HVL) were calculated using Win-XCOM software. Sulfuric acid abrasion and compressive strength tests were carried out and all samples showed good resistance to acid wear and pressure force. The neutron equivalent dose was measured using an average 4.5 MeV energy fast neutron source. Results were compared to 316LN type stainless steel, which commonly used in shielding radiation. New stainless steel samples were found to absorb neutron better than 316LN stainless steel at both low and high temperatures.

• 대한인간공학회지
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• 제32권1호
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• pp.37-45
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• 2013

Objective: This study describes the change of operator performance in drastic change of illumination level, and proposes an alternative method to cope with it. Background: The control standard of illumination for nuclear power plants(NPPs) is based on the set of limit criteria for maintaining a specific illumination level. However, there is a possibility to cause human errors according to the psychological and physiological influences to operators in the situation of drastic change of illumination such as SBO(Station Black Out), so a basic study is necessary to review the current approach. Method: We assessed the visual fatigue, subjective work load and task performance according to the three illumination situations(Normal Illumination, Emergency Illumination, and Drastic Change of Illumination). Result: Research finding shows that there are not significant differences in task performance between normal illumination (1,000lx level) and emergency illumination (100lx level), only if beyond the dark adaptation limit. However, subjective work load on mental demand and visual fatigue show a potential challenge to visual performance in drastic change of illumination. Conclusion/Application: Several trials can complement this challenge in NPPs by applying 3-way communication, enhancing readability of procedures, and managing the visual factors affecting the operators' performance through a Visual Environment Management Program including visual health aspects, etc.

• 설비공학논문집
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• 제26권3호
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• pp.130-136
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• 2014

The present study is performed to set up the framework of cooling tower performance predictions. The performance of mechanical forced draft cooling tower is directly related to the state of a nuclear power plant system, such as the condenser and evaporator. The main parameters related to the state of systems are as follows : wet bulb temperature, dry bulb temperature and absolute humidity. The performance evaluation of cooling tower must be considered at the power plant design. In this study, the toolkit developed by the American Cooling Tower Industry association (CTI) has been used for the framework construction. In order to validate the framework, it is being applied to the cooling tower constructed for the U.S. Nuclear Power Plant. The test results have shown good agreements with the cold water temperature on the cooling tower performance curves provided by manufacturers.

• Advances in Computational Design
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• 제2권3호
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• pp.211-223
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• 2017

This research presented a numerical and experimental study on the seismic performance of first-generation base-isolated and fixed-base nuclear power plants (NPP). Three types of the base isolation system were applied to rehabilitate the first-generation nuclear power plants: frictional pendulum (FP), high-damping rubber (HDR) and lead-rubber (LR) base isolation. Also, an Excel program was proposed for the design of the abovementioned base isolators in accordance with UBC 97 and the Japan Society of Base Isolation Regulation. The seismic assessment was performed using the pushover and nonlinear time history analysis methods in accordance with the FEMA 356 regulation. To validate the adequacy of the proposed design procedure, two small-scale NPPs were constructed at Universiti Teknologi Malaysia's structural laboratory and subjected to a pushover test for two different base conditions, fixed and HDR-isolated base. The results showed that base-isolated structures achieved adequate seismic performance compared with the fixed-base one, and all three isolators led to a significant reduction in the containment's tension, overturning moment and base shear.

• 대한기계학회논문집A
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• 제34권4호
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• pp.495-499
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• 2010

이 연구에서는 우리나라 CANDU형 원전의 계속운전 규제지침을 개발하였다. CANDU 600 로형인 월성 1호기는 2012년에 30년 설계수명에 도달한다. 설계수명이후에 계속운전을 원하는 원전 사업자는 11개 인자를 포함하는 주기적안전성 평가 보고서를 제출해야 하며, 추가로 다음 사항에 대한 보고서를 제출해야 한다. (1) 경년열화 관리에 대한 범위 및 사전 검토 결과 (2) 경년열화 관리 프로그램 (3) 계속 운전기간을 포함하는 시간제한 경년열화 수명평가, (4) 운전경험 및 주요 연구결과의 반영. 이 연구에서는 상기 항목에 대한 54개의 규제지침이 개발되었다.