• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1672-1678
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• 2024

We introduce a creative approach combining machine learning with optimization techniques to enhance the optimization of the loading pattern (LP). Finding the optimal LP is a critical decision that impacts both the reload safety and the economic feasibility of the nuclear fuel cycle. While simulated annealing (SA) is a widely accepted technique to solve the LP optimization problem, it suffers from the drawback of high computational cost since LP optimization requires three-dimensional depletion calculations. In this note, we introduce a technique to tackle this issue by leveraging neural networks to filter out inappropriate patterns, thereby reducing the number of SA evaluations. We demonstrate the efficacy of our novel approach by constructing a machine learning-based optimization model for the LP data of the Korea Standard Nuclear Power Plant (OPR-1000).

• Journal of the Korean Society for Nondestructive Testing
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• v.28 no.2
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• pp.184-190
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• 2008

Various ECT techniques have been applied basically to assess the integrity of steam generator tithing in nuclear power plant. Among these techniques, the bobbin probe technique is applied generally to examine the volumetric flaws such as a crack-like defect and wear which is generally occurred on steam generator tubing, and additionally MRPC probe is used to examine closely tile top of tubesheet and bending regions due to the high possibility of cracking. Dent and bulge also may be formed on tube during installation process and operation of steam generator, but the dent and bulge indications greater than specific size criteria are recorded on examination report because these indications are not considered as flaw. These indications can be easily detected with bobbin probe and approximately sized with profile bobbin probe, but the size and shape can not be accurately verified. Accordingly, in this study, the $8{\times}1$ multi-array EC probe was designed to increase the measurement accuracy of the sectional profiling EC testing of tube. As a result, we would like to propose the application of $8{\times}1$ multi-array EC probe for the measurement of size and shape of profile change on steam generator tube in OPR-1000 nuclear power plant.

• Journal of Energy Engineering
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• v.21 no.4
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• pp.362-372
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• 2012

The steam generator tubes of nuclear power plants are pressure boundaries, and if tubes are leaked, the coolant with the radioactive materials was flowed out from the primary system to the secondary system and polluted the plant and the air. Recently most crack defects of tubes are stress corrosion cracks and these defects are located in expansion transition area, sludge pile-up region, and U-bend area. The most effective one of crack initiation factors in expansion transition area and U-bend area is the residual stress. According to the experiences of Korea standard nuclear plants(Optimized Power Reactor-1000), they had the stress corrosion cracks at the tube expansion transition area in early operating stage and especially lots of circumferential cracks were occurred. Therefore in this study, the distributions and conditions of residual stresses by tube expansion methods were compared and the dominant reason of a specific direction was examined.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3990-4002
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• 2021

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

• Nuclear industry
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• v.36 no.11
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• pp.29-44
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• 2016

1980~1990년대 OPR1000 기술 자립을 추진할 때도 그랬지만 한국은 원자력 기술 자립에 대한 도전이 선진국에 비해 늦었지만 과학기술자의 열정과 정부의 적극적인 지원으로 오늘날 원자력 선진국이 될 수 있었고, 원자력산업을 해외 수출 산업으로서 다양한 노력을 시도하고 있다. 특히 국내 가동 중인 원전은 외국과 차별되게 1기당 고장 정지율이 0.1건으로 외국 평균의 5.5건과 크게 대별된다. 또한 운전 신뢰성을 나타내는 발전소 가동률도 10% 이상 차로 월등히 높다. 한마디로 한국은 가장 원전의 기술 개발과 운영을 잘하고 있는 원전 선진국임을 자타가 인정하고 있다. 그러나 현재의 기술 수준에 머물면 미래 원전 기술에서는 다른 선진국 내지 중국, 인도 등 신흥국에 그 자리를 양보할 수밖에 없을 것이다. 미래 원자력이란 시대적 요건인 고유 안전성과 지속 가능성을 확보하고 경제성과 함께 핵확산 저항성이 전제되는 원자력 신기술로서 세계와의 경쟁 대상이다. 여기에 핵연료 자원의 유한성에 지속 가능성을 확보하기 위해서 우라늄 효율을 극대화하는 제4세대의 고속로 개발까지 우리나라는 선도적 위치로 가야 한다. 이 기술 개발 역시 출발은 늦었지만 적극적인 개발을 추진하고 있어 소듐고속로의 시현 원자로인 PGSFR을 2028년까지 완성하는 목표를 달성하면, 이를 근간으로 세계 선진국의 경쟁 대열에 나설 수 있다. 정부의 적극적인 지원이 선도적 위치에 갈 수 있는 지름길이다. 고속로 기술 개발과 관련하여 사용후핵연료(SF)의 국가 정책이 아직 확정되지 않아 재활용주기를 전제하고 있는 고속로 개발에 어려움을 주고 있다. 따라서 SF 부지를 2028년까지 확정하는 일정과 함께 국가 SF 정책이 조속히 확정되어야 한다.

• Proceedings of the KIEE Conference
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• 2011.07a
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• pp.1999-2000
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• 2011

본 연구의 목적은 발전소의 정상적인 출력운전을 위해 필요한 주요 계통의 기능에 영향을 미쳐 발전소 불시정지를 유발할 수 있는 핵심 기기, 즉, 발전정지유발기기의 설치 개소를 체계적인 방법을 통하여 정밀 분석하고, 해당 기기의 고장모드와 그 영향을 검토하여 이를 방지하기 위한 대책을 수립하도록 하는 것이다. 발전정지유발기기의 평가는 발전소 종사자로 하여금 가동 중 발전소에서 발생 가능한 발전정지 영향기기와 그들의 상호관계를 이해하고, 정량적 평가를 통해 해당기기들의 발전소 발전정지 영향을 시각적으로 확인하여 불시 발전정지를 예방할 수 있는 대응 논리를 인지할 수 있도록 하는데 그 목적이 있다. 원자로냉각재계통에 대한 발전정지유발기기(SPV, Single Point Vulnerability)를 분석하기 위해 고장모드영향분석(FMEA, Failure Mode Effect Analysis)을 수행하고 상세 고장수목을 개발하여 통합단위의 계통 분석을 수행하였다. 분석결과 원자로냉각재계통의 발전정지유발기기는 원자로냉각재 펌프와 가압기 주살수 밸브의 제어회로에 집중되어 있는 것으로 나타났다.

• Journal of Sensor Science and Technology
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• v.25 no.3
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• pp.229-234
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• 2016

Self-powered neutron detector (SPND) is being widely used to monitor the reactor core of the nuclear power plants. The SPND contains a neutron-sensitive metallic emitter surrounded by a ceramic insulator. Currently, the vanadium (V) SPND has been being developed to be used in OPR1000 nuclear power plants. Some Monte Carlo simulations were accomplished to calculate the initial sensitivity of vanadium emitter material and alumina insulator with a cylindrical geometry. An MCNP code was used to simulate some factors (neutron self-shielding factor and beta escape probability from the emitter) and space charge effect of an insulator necessary to calculate the sensitivity of vanadium detector. The simulation results were compared with some theoretical and experimental values. The method presented here can be used to analyze the optimum design of the vanadium SPND and contribute to the development of TMI (Top-mount In-core Instrumentation) which might be used in the SMART and SMR.

• Journal of the Korean Society of Safety
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• v.31 no.6
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• pp.129-134
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• 2016

Standard design of APR+(advanced power reactor plus) was certified at 2014 by Korea regulatory body. Based on the experience gained from OPR1000 and APR1400, the APR1400 was being developed as a 1,500MWe class reactor using Korean technologies for design code, reactor coolant pump, and man-machine interface system. APR+ has been basically designed to have the seismic design basis of safe shutdown earthquake (SSE) 0.3g, a 4-train safety concept based on N+2 design philosophy, and a passive auxiliary feedwater system (PAFS). Also, safety issues on the Fukushima-type accidents have been extensively reviewed and applied to enhance APR+ safety. APR+ provides higher reliability and safety against tsunami and earthquake. The purpose of this paper is to implement probabilistic safety assessment considering these design features and to analyze sensitivity of core damage frequency for large loss of coolant accident of APR+.

• Journal of Radiation Protection and Research
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• v.42 no.2
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• pp.106-113
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• 2017

Background: Interfacing-system loss-of-coolant-accident (ISLOCA) has been identified as the most hazardous accident scenario in the typical PWR plants. The present study as an effort to improve the knowledge of the source term to the environment during ISLOCA focuses on an improvement of the estimation method. Materials and Methods: The improvement was performed to take into account an effect of broken pipeline and auxiliary building structures relevant to ISLOCA. An estimation of the source term to the environment was for the OPR-1000 plants by MELOCR code version 1.8.6. Results and Discussion: The key features of the source term showed that the massive amount of fission products departed from the beginning of core degradation to the vessel breach. Conclusion: The release amount of fission products may be affected by the broken pipeline and the auxiliary building structure associated with release pathway.

• Nuclear Engineering and Technology
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• v.46 no.6
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• pp.797-802
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• 2014

This study investigated the thermal hydraulic issues in the Containment Filtered Venting System (CFVS) for a long operating time using the MELCOR computer code. The modeling of the CFVS, including the models for pool scrubbing and the filter, was added to the input file for the OPR-1000, and a Station Blackout (SBO) was chosen as an accident scenario. Although depressurization in the containment building as a primary objective of the CFVS was successful, the decontamination feature by scrubbing and filtering in the CFVS for a long operating time could fail by the continuous evaporation of the scrubbing solution. After the operation of the CFVS, the atmosphere temperature in the CFVS became slightly above the water saturation temperature owing to the release of an amount of steam with high temperature from the containment building to the scrubbing solution. Reduced pipe diameters at the inlet and outlet of the CFVS vessel mitigated the evaporation of scrubbing water by controlling the amount of high-temperature steam and the water saturation temperature.