• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.128.1-128.1
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• 2003

• Journal of Korean Society of Industrial and Systems Engineering
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• v.25 no.6
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• pp.23-35
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• 2002

The problem of system reliability is very important issue in the digitalized nuclear power plant, because the failure of its system brings about extravagant economic loss, environment destruction, and fatal damage of human. Therefore the purpose of this study has developed the reliability evaluation model through the scoring model by the quantitative and qualitative factors in order to justify the evaluation considering the advanced safety factors in the Advanced Pressurized water Reactor 1400MWe(APR 1400MWe) under uncertainty. Especially, the qualitative factors considering the human, information control, and quality factors for the systematic and rational justification have been closely analyzed. The proposed model can be simply applied in real fields in order to minimize the industrial accidents in the digitalized nuclear power plant.

• Journal of Energy Engineering
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• v.25 no.1
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• pp.43-47
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• 2016

KHNP CRI has developed APR+ nuclear power plant since 2007, which is GEN III+ model with 1500 MWe capacity. To develop safer nuclear power plant than APR1400, we investigated advanced design features of ALWR being constructed in Korea and being developed/constructed in foreign countries. We applied the advanced design features and lessons learned from Fukushima accident to develop APR+ standard design suitable for both domestic construction and overseas construction business. One economic assessments have performed during safety design improvement phase(2013.1 ~ 2015.12) of APR+. The result of the economic analysis for APR+ safety inhancement design showed that APR+ N-th plant is about 39.2% more economical than coal-fired 1,000MW power plant. Also APR+ plant is more cost advantage over foreign advanced nation ALWRs.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.691-708
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• 2009

In order to analyze thermal hydraulic phenomena during a DVI (Direct Vessel Injection) line break LOCA (Loss-of-Coolant Accident) in the APR1400 (Advanced Power Reactor 1400 MWe), we performed experimental studies with the SNUF (Seoul National University Facility), a reduced-height and reduce-pressure integral test loop with a scaled down APR1400. We performed experiments dealing with eight test cases under varied tests. As a result of the experiment, the primary system pressure, the coolant temperature, and the occurrence time of the downcomer seal clearing were affected significantly by the thermal power in the core and the SI flow rate. The break area played a dominant role in the vent of the steam. For our analytical investigation, we used the MARS code for simulation of the experiments to validate the calculation capability of the code. The results of the analysis showed good and sufficient agreement with the results of the experiment. However, the analysis revealed a weak capability in predicting the bypass flow of the SI water toward the broken DVI line, and it was insufficient to simulate the streamline contraction in the broken side. We, hence, need to improve the MARS code.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1651-1664
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• 2023

It is very difficult to capture the multi-dimensional phenomena such as asymmetric flow and temperature distributions with the one-dimensional (1D) model, obviously, due to its inherent limitation. In order to overcome such a limitation of the 1D representation, many state-of-the-art system codes have equipped a three-dimensional (3D) component for multi-dimensional analysis capability. In this study, a standard multi-dimensional analysis model of APR1400 (Advanced Power Reactor 1400) has been developed using TRACE (TRAC/RELAP Advanced Computational Engine). The entire reactor pressure vessel (RPV) of APR1400 has been modeled using a single 3D component. The fuels in the reactor core have been described with detailed and coarse representations, respectively, to figure out the impact of the fuel description. Using both 3D RPV models, a comparative analysis has been performed postulating a double-ended guillotine break at a cold leg. Based on the results of comparative analysis, it is revealed that both models show no significant difference in general plant behavior and the model with coarse fuel model could be used for faster transient analysis without reactor kinetics coupling. The analysis indicates that the asymmetric temperature and flow distributions are captured during the transient, and such nonuniform distributions contribute to asymmetric quenching behaviors during blowdown and reflood phases. Such asymmetries are directly connected to the figure of merits in the LBLOCA analysis. Therefore, it is recommended to employ a multi-dimensional RPV model with a detailed fuel description for a realistic safety analysis with the consideration of the spatial configuration of the reactor core.

• Journal of Energy Engineering
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• v.23 no.4
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• pp.236-239
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• 2014

KHNP CRI has been developing APR+ nuclear power plant since 2007, which is GEN III+ model with 1500 MWe capacity. To develop safer and more economical nuclear power plant than APR1400, we investigated advanced design features of ALWR(advanced light water reactor) being constructed in Korea and being developed/constructed in foreign countries. We applied the advanced design features and lessons learned from Fukushima accident to develop APR+ standard design suitable for both domestic construction and overseas construction business. Three economic assessments have performed during standard design phase of APR+. The result of the 3th(final) economic analysis for APR+ standard design showed that APR+ N-th plant was about 23% more economical than coal-fired 1,000MW power plant.

• Proceedings of the KIEE Conference
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• 2007.10a
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• pp.305-306
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• 2007

본 논문은 차세대 원전 신고리 #3, 4호기 NSSS(Nuclear Steam Supply System) 제어계통의 검증시스템을 개발하기 위한 설계개념에 관한 것이다. 차세대 원전 신고리 #3, 4호기는 KHNP(Korea Hydro & Nuclear Power Co., Ltd.)가 개발한 APR1400(Advanced Power Reactor 1400 [MWe])을 적용하는 최초의 원자력 발전소이다. APR1400은 3세대 원자력발전소로 인정받고 있으며, APR1400 원자력발전소의 안전한 운영을 위하여 I&C(Instrumentation and Control)시스템이 디지털 표준 플랫폼으로 설계되었다[2]. 특히, 차세대 원전 신고리 #3, 4호기의 비안전계통(제어 감시 및 경보계통)은 WEC (Westinghouse Electric Company)의 DCS(Distributed Control System) 상용 단일 플랫폼으로 구성될 예정이다. 우리는 신고리 #3, 4호기의 제어계통 중에서 NSSS(Nuclear Steam Supply System) 제어계통의 검증시스템을 개발하기 위하여 Stimulated Simulator의 방법론을 적용하여 "Simulator"라는 설계 개념을 정립하였다. 현재 원자력발전소 NSSS 제어계통의 DCS Stimulator 개발을 위하여 차세대 원전 신고리 #3, 4호기에 시설될 WEC의 DCS와 Simulation 서버 그리고 I/O 설비를 구축 중에 있으며, 원자력발전소 현장 기기 모델링 소프트웨어와 I/O 설비간의 인터페이스를 위한 동신 소프트웨어도 개발하고 있다.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.1
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• pp.18-27
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• 2019

The Advanced Power Reactor 1400 (APR1400) Steam Generator (SG) uses alloy 690 as a tube material and SA-508 Grade 3 Class 1 as a tubesheet material to form tube-to-tubesheet joint through hydraulic expansion process. In this paper, the residual stresses in the SG tube-to-tubesheet contact area was investigated by applying Model-Based System Engineering (MBSE) methodology and the V-model. The use of MBSE transform system description into diagrams which clearly describe the logical interaction between functions hence minimizes the risk of ambiguity. A theoretical and Finite Element Methodology (FEM) was used to assess and compare the residual stresses in the tube-to-tubesheet contact area. Additionally, the axial strength of the tube to tubesheet joint based on the pull-out force against the contact joint force was evaluated and recommended optimum autofrettage pressure to minimize residual stresses in the transition zone given. A single U-tube hole and tubesheet with ligament thickness was taken as a single cylinder and plane strain condition was assumed. An iterative method was used in FEM simulation to find the limit autofrettage pressure at which pull-out force and contact force are of the same magnitude. The joint contact force was estimated to be 20 times more than the pull-out force and the limit autofrettage pressure was estimated to be 141.85MPa.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.124-124
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• 2001

• Proceedings of the KSME Conference
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• 2003.04a
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• pp.49-54
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• 2003

Most of the operating pressurized water reactors (PWRs)has chosen Inconel 600 as steam generator tubing. The long-term operation of steam generators showed that the use of this material induced localized corrosion damages. The current trend is using Inconel 690 as a tube material for the replacement steam generators. Based on the current trend, we have chosen Inconel 690 for the advanced Power Reactor 1400 (APR1400) steam generator tube material. In this paper, we examined the technical consideration in this modification: the effect of chemical composition, thermal conductivity, corrosion resistance and wear characteristics