A5063843367- Heat Transfer and Boiling Studies
- Heat transfer and supercritical fluids
- Nuclear Engineering Thermal-Hydraulics
- Nuclear reactor physics and engineering
- Nuclear Materials and Properties
- Fluid Dynamics and Mixing
- Heat Transfer and Optimization
- Combustion and Detonation Processes
- Heat Transfer Mechanisms
- Fluid Dynamics and Heat Transfer
- Metallurgical Processes and Thermodynamics
- Risk and Safety Analysis
- Cyclone Separators and Fluid Dynamics
- Nuclear and radioactivity studies
Korea Atomic Energy Research Institute
2016-2024
Pusan National University
2015-2021
The subchannel of a research reactor used to generate high power density is designed be narrow and rectangular comprises plate-type fuels operating under downward flow conditions. Critical heat flux (CHF) crucial parameter for estimating the safety nuclear fuel; hence, this should accurately predicted. Here, machine learning applied prediction CHF in channel. Although can effectively analyze large amounts complex data, its application CHF, particularly channels, remains challenging because...
A condensation heat transfer model is essential to accurately predict the performance of passive containment cooling system (PCCS) during an accident in advanced light water reactor. However, most existing models tend very well for a specific range thermal-hydraulic conditions. In this study, new correlation coefficient (HTC) presented using machine learning technique. To secure sufficient training data, large number pseudo data were produced by ten models. Then, neural network was...
In a multidimensional two-phase flow analysis, bubble size significantly affects interfacial transfer terms such as mass, momentum, and energy. With regard to bubbly flow, the application of simple correlation-type model presents certain advantages, including short calculation times ease usage. this study, we propose semi-theoretical correlation developed from steady state number density transport equation for predicting distribution local using computational fluid dynamics (CFD) code. The...
Owing to spatial effects and vortex flow, flow in research reactors that use plate-type fuels can be maldistributed the parallel channels of core, which significantly impacts reactor safety. In this study, core an advanced was measured a real-scale facility under various hydraulic conditions. For measurement, integrated pressure lines were embedded mockups 22 fuel assemblies six fission molybdenum assemblies. Each assembly mockup individually calibrated obtain relationship between drop rate....
Conceptual design was carried out on test section for studying heat transfer characteristics of narrow rectangular channel research reactor application. Thermal hydraulic requirements were specified through literature study mid- to high-powered reactors currently under operation or design. Engineering then in streamlined fashion with help in-house code and computational fluid dynamics analysis. Considering various constraints uncertainties terms engineering hot factors, sensitivity studies...
본 연구에서는 전산유체역학(Computational Fluid Dynamics, CFD) 해석코드를 사용하여 마스트집합체의 열수력적 안전성에 대한 연구를 수행하였다. 이를 위해 자연대류 벤치마크 문제를 선정하여 CFD 코드의 물리모델을 선정 및 해석 능력을 검증하고 이용하여 마스트집합체에 열전달 해석을 Betts et al.의 사각 수직공동에서 난류 실험결과를 대상으로 수행하여 조건에 적용하기 위한 모델로 표준 <TEX>$k-{\omega}$</TEX> 모델을 선정하였다. 이렇게 도출된 난류모델을 CFD코드에 적용하여 Bates al.에 의해 수행된 PNL(Pacific Northwest Laboratory)의 <TEX>$2{\times}6$</TEX> 번들 실험과 이에 Kwon MATRA, Fluent 해석과 비교 계산을 CFD코드의 부수로조건 검증하였다. 최종적으로 마스트집합체 핵연료 집합체에 결과 수조 내부 부수로 내에서 안정적인 유동이 발생함을 확인하였으며, 유동...
요르단 수출용 연구용 원자로(JRTR, Jordan Research and Training Reactor)에 사용되는 판형핵연료(Plate-type-fuel)의 부수로를 모의하는 좁은 사각 유로에서 하향류 유동 조건의 임계열유속(CHF, Critical Heat Flux) 실험 연구가 수행되었다. 실험은 원자로 인허가 요건을 만족하는 조건에서 수행되었으며 크기가 다른 두 가지 시험부를 이용하였다. 시험부는 각각 피션몰리(Fission moly) 우라늄 타겟의 부수로와 판형핵연료의 모의하며 모두 원형과 같은 크기로 제작되었다. 각 시험부에 대해 임계열유속 실험이 이를 통해 임계열유속에 영향을 주는 변수를 분석하였다. 그리고 실험결과를 기존의 모델과 비교하였으며 모델의 예측 정확도를 제시하였다. 유로 내 조건에서의 예측에 대한 기존 적용성을 Experimental investigation was carried out on the CHF(Critical under downward...
Conceptual design was carried out on test section for studying heat transfer characteristics of narrow rectangular channel research reactor application. Thermal hydraulic requirements were specified through literature study mid- to high-powered reactors currently under operation or design. Engineering then in streamlined fashion with help in-house code and computational fluid dynamics analysis. Considering various constraints uncertainties terms engineering hot factors, sensitivity studies...