A5081917429- Heat Transfer and Boiling Studies
- Nuclear Engineering Thermal-Hydraulics
- Fluid Dynamics and Mixing
- Nuclear Materials and Properties
- Nuclear reactor physics and engineering
- Heat transfer and supercritical fluids
- Metallurgical Processes and Thermodynamics
- Fluid Dynamics and Heat Transfer
- Heat Transfer and Optimization
- Flow Measurement and Analysis
- Combustion and Detonation Processes
- Cyclone Separators and Fluid Dynamics
- Engineering Applied Research
- Risk and Safety Analysis
- Advanced Sensor Technologies Research
- Heat Transfer Mechanisms
- Spacecraft and Cryogenic Technologies
- Minerals Flotation and Separation Techniques
- Marine and Coastal Research
- Aerosol Filtration and Electrostatic Precipitation
- Lattice Boltzmann Simulation Studies
- Drilling and Well Engineering
- Fluid Dynamics and Thin Films
- Nuclear and radioactivity studies
- Environmental Science and Water Management
Pusan National University
2015-2024
Korea Atomic Energy Research Institute
2005-2018
Busan National University of Education
2013
Jeju National University
2012
Abstract In this paper, the interfacial flow structure of subcooled water boiling in a subchannel 3 × rod bundles is presented. The 9 rods are positioned quadrangular assembly with diameter 8.2mm and pitch distance 16.6 mm. Local void fraction, area concentration, velocity, Sauter mean diameter, liquid velocity have been measured using conductivity probe Pitot tube 20 locations inside one subchannels. A total 53 conditions considered experimental dataset at atmospheric pressure mass rate,...
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...
The thermal-hydraulic integral effect test (IET) program is being progressed by the Korea Atomic Energy Research Institute. This paper presents an overview of IET program; scientific design characteristics facility; ATLAS, which under construction; and experimental analytical validation works. ATLAS facility has following characteristics: (a) a 1/2-height, 1/288-volume, full-pressure simulation APR1400, (b) geometrical similarity with including 2 (hot legs) × 4 (cold reactor coolant loops,...
We have developed a heat transfer correlation for saturated flow boiling of water in helically coiled tube. Initially, we collected experimental data encompassing broad spectrum thermal-hydraulic conditions and geometric configurations, examined the influences key dimensionless parameters, such as convection number number. The analysis showed that observed trend aligns with previous studies on within straight Also, investigated influence centrifugal force acting fluid tube confirmed its...
As passive safety features for nuclear power plants receive increasing attention, various studies have been conducted to develop systems 3rd-generation (GEN-III) that are driven by systems. The Passive Auxiliary Feedwater System (PAFS) is one of several being designed the Advanced Power Reactor Plus (APR+), and extensive complete its design verify feasibility. Because PAFS removes decay heat from reactor core under transient accident conditions, it necessary evaluate removal capability...
This paper briefly introduces recent progress in thermal-hydraulic R&Ds, which is mainly being performed at KAERI, for the APR+ (Advanced Power Reactor plus) development. The main R&D items reactor are associated directly with efforts to introduce new safety concepts standard design developments, currently Republic of Korea. activities reported here cover and severe accident areas experimental and/or analytical ways. They include: (1) advancement optimization injection system, (2)...