• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2002.05a
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• pp.135-141
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• 2002

An optimization of KTF thorium fuel assembly design was performed on the basis of the design parameter studies. Optimization goals ware to make the core have both proliferation resistance and fuel cycle economics. Four kinds of proliferation resistance indexes were used; SNS, TG, BCM, Toxicity. A new index, FEI was regarded as a limiting index for the maximization of fuel cycle economics. Optimized thorium fuel design was applied for APR-1400 reactor core. Nuclear core design procedures were examined to solve the thorium fuel reactor problems. It was shown that heterogeneous thorium fuel core option is acceptable in safety and economics aspects.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1508-1515
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• 2022

In the present study, the cavity module of the MELCOR code is used for the simulation of molten corium concrete interaction (MCCI) during the late phase of postulated large break loss of coolant (LB-LOCA) accident in the APR1400 reactor design. Using the molten corium composition data from previous MELCOR Simulation of APR1400 under LB-LOCA accident, the ex-vessel phases of the accident sequences with long-term MCCI are recalculated with stand-alone cavity package of the MELCOR code to investigate the impact of water ingression and melt eruption models which were hitherto absent in MELCOR code. Significant changes in the MCCI behaviors in terms of the heat transfer rates, amount of gases released, and maximum cavity ablation depths are observed and reported in this study. Most especially, the incorporation of these models in the new release of MELCOR code has led to the reduction of the maximum ablation depth in radial and axial directions by ~38% and ~32%, respectively. These impacts are substantial enough to change the conclusions earlier reached by researchers who had used the older versions of the MELCOR code for their studies. and it could also impact the estimated cost of the severe accident mitigation system in the APR1400 reactor.

• Journal of the Korean Society of Systems Engineering
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• v.19 no.2
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• pp.32-45
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• 2023

Implementing Severe Accident Management (SAM) strategies is crucial for enhancing a nuclear power plant's resilience and safety against severe accidents conditions represented in the analysis of Station Blackout (SBO) event. Among these critical approaches, the In-Vessel Retention (IVR) through External Reactor Vessel Cooling (IVR-ERVC) strategy plays a key role in preventing vessel failure. This work is designed to evaluate the efficacy of the IVR strategy for a high-power density reactor APR1400. The APR1400's plant is represented and simulated under steady-state and transient conditions for a station blackout (SBO) accident scenario using the computer code, ASYST. The APR1400's thermal-hydraulic response is analyzed to assess its performance as it progresses toward a severe accident scenario during an extended SBO. The effectiveness of emergency operating procedures (EOPs) and severe accident management guidelines (SAMGs) are systematically examined to assess their ability to mitigate the accident. A group of associated key phenomena selected based on Phenomenon Identification and Ranking Tables (PIRT) and uncertain parameters are identified accordingly and then propagated within DAKOTA Uncertainty Quantification (UQ) framework until a statistically representative sample is obtained and hence determine the uncertainty bands of key system parameters. The Systems Engineering methodology is applied to direct the progression of work, ensuring systematic and efficient execution.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2010.10a
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• pp.433-438
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• 2010

Reactor vessel internals comprehensive vibration assessment program(RVI CVAP) is one of the necessary tests to ensure the safety of nuclear power plants. RVI CVAP of U.S. Nuclear Regulatory Commission Regulatory Guide 1.20(U.S. NRC R.G. 1.20) consists of the analysis, measurement, and inspection. One of the core technologies of the measurement program for RVI CVAP is to select suitable sensors. We analyzed RVI design data of Palo Verde nuclear generating station(U.S.) and Yonggwang nuclear generating station(Korea) and investigated measuring sensors used in both of them; moreover, we investigated sensors used for measurement of RVI CVAP for the last 20 years throughout the world. Based on these results, we selected the most suitable sensors for RVI CVAP in Advanced Power Reactor 1400(APR1400).

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.25 no.9
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• pp.599-605
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• 2015

The revised U.S. Nuclear Regulatory Commission(NRC), Regulatory Guide(RG) 1.20, rev.3 requires the evaluation of the potential adverse effects from pressure fluctuations and vibrations on piping and components for the reactor coolant, steam, feedwater, and condensate systems. Detailed vibration analyses for the systems attached to the steam generator are very difficult, because these piping systems are very complicated. This paper suggests a screening method for the flow-induced vibration of acoustic resonances and pump-induced vibration of the piping systems attached to the steam generator in order to conduct the APR1400 comprehensive vibration assessment program. This paper seeks to address the areas such as potential vibration sources, and methods to prevent the occurrence of acoustic resonances and pump-induced vibration of piping systems attached to the steam generator, for conducting the APR1400 comprehensive vibration assessment program. The screening method in this paper will be used to estimate the flow-induced vibration of the piping systems attached to the steam generator for the APR1400.

• Proceedings of the KIEE Conference
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• 2008.10b
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• pp.429-430
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• 2008

본 논문은 전기출력이 1400 MWe급으로 개발된 첨단 원자력 발전소인 APR1400(신형겨수로 1400) 제어계통에 적용되는 디지털시스템의 설계 및 성능 검증을 위해 개발 중인 디지털제어계통 검증시스템에 관한 것이다. APR1400 디지털제어계통은 발전소 출력 제어 및 안전운전과 관련 된 중요 기능들을 수행하며, 기존 원자력발전소와 달리 단일 디지털 Platform을 적용하고, Multi-Loop 개념과 네트워크을 적용하여 Controller와 케이블 수량을 줄인 특징을 가지고 있다. 이와 같을 설계는 지금가지 원자력발전소에는 적용된 적이 없기 때문에 사용자 측면에서는 디지털 제어 계통 설계 및 성능 관점에서의 검증을 위한 시스템이 요구되었다. 현재는 APR1400 시뮬레이터(발전소 모델링을 통한 모의시스템)를 이용한 검증시스템을 1차적으로 구축한 상태에 있으며, 시스템 전체 시험을 진행 중에 있다. 특히, 이번에 개발 중인 검증시스템은 구성이 간단하고 사용이 편리한 장점을 지니고 있을 뿐만 아니라 다양한 고장상황을 재현해 봄으로써 디지털제어계통의 성능을 확인해 볼 수 있는 특징을 보유하고 있다. 본 검증시스템의 활용방안으로는 첫째, 계통설계의 구현 가능성 관점에서의 확인시험을 수행하는 방안, 둘째, 발전소 시운전 착수 전 시운전요원 교육에 활용하는 방안, 셋째, 발전소 설계 변경 필요 시 설계 변경에 따른 영향 파악, 넷째, 디지털제어계통 유지보수 기술 습득 등에 효과적으로 활용 할 수 있을 것으로 본다. AFR1400 디지털제어계통은 현재 건설 중인 신고리 3,4호기 원자력발전소에 적용될 예정이며, 향후에는 해외 원자력 수출을 위한 기반기술로 활용될 수 있을 것으로 확신한다.

• Journal of Energy Engineering
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• v.14 no.1
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• pp.54-61
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• 2005

The APR1400 is an Advanced Pressurized Water Reactor with 3983 MWt power, 2×4 loops, and direct vessel injection system. The Fluidic Device (FD) is adopted to regulate the safety injection flow rate in a Safety Injection Tank (SIT) of APR1400. The performance of a newly designed fluidic Device is evaluated by analyzing a Large Break Loss-of-Coolant Accident (LBLOCA) using TRAC-M/F90, version 3.782. The analysis results show that the TRAC-M code reasonably predicts the important phenomena of blowdown, refill and reflood phases of LBLOCA. The sensitivity studies about gas/water volume changes in a SIT and K factor changes in a SI system were also done to understand the important phenomena with a Fluidic Device in APR1400.

• Proceedings of the Earthquake Engineering Society of Korea Conference
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• 2001.09a
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• pp.67-74
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• 2001

This paper presents the Safe Shutdown Earthquake(SSE) input motion for the seismic design of the Advanced Power Reactor 1400(APR1400). The Design Ground Response Spectra(DGRS) far the SSE is based on the design spectrum specified in regulatory Guide(RG) 1.60 of U.S. Nuclear Regulatory Commission(US NRC), anchored to a Peak Ground Acceleration(PGA) of 0.3g and enriched in the high frequency range. This SSE seismic input motion is to be applied to the seismic analysis as the free-field seismic motion at the ground surface of both the rock and generic soil sites fur APRI1400. The enrichment for APR1400 seismic input motion is performed considering the current US NRC regulations, the seismic hazard studies performed by the Lawrence Livermore National Laboratory (LINL) and Electric Power Research Institute(EPRI) for the Central and Eastern United States nuclear power plant sites, and the seismic input motions used in the design certifications of the three existing U.S. advanced standard plants. It is represented by a set of DGRS and the accompanying Target Power Spectral Density(PSD) Function in both the horizontal and vertical directions.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2011.04a
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• pp.821-826
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• 2011

A basic concept for selection criteria of measuring places of RVI CVAP is to determine measuring places and sensors based on the results of the hydraulic and structural analysis for RVI CVAP in APR1400. In addition, there is the important selection criteria to determine measuring places for measurement of RVI CVAP ; the first is to choose measuring places according to U.S. NRC R.G. 1.20, the second is to select measuring places by RVI design review, the third is to option on the basis of measurement results of SYSTEM 80, the forth is to decide using review results on a design change of a reactor and the last is to determine using the review on the possibility of installation/removal of sensors and structures for the measurement. We developed selection criteria of measuring places for RVI CVAP in APR1400 and this will be directly applied to the measurement program for RVI CVAP.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.128.1-128
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• 2001