A5062917929- Nuclear Engineering Thermal-Hydraulics
- Nuclear reactor physics and engineering
- Nuclear Materials and Properties
- Fluid Dynamics and Mixing
- Heat Transfer and Boiling Studies
- Fluid Dynamics and Heat Transfer
- Heat transfer and supercritical fluids
- Combustion and Detonation Processes
- Risk and Safety Analysis
- Metallurgical Processes and Thermodynamics
- Engineering Applied Research
- Flow Measurement and Analysis
- Cyclone Separators and Fluid Dynamics
- Nuclear and radioactivity studies
- Graphite, nuclear technology, radiation studies
- Heat Transfer and Optimization
- Spacecraft and Cryogenic Technologies
- Fluid Dynamics and Turbulent Flows
- Minerals Flotation and Separation Techniques
- Drilling and Well Engineering
- Reservoir Engineering and Simulation Methods
- Fluid Dynamics and Thin Films
- Magnetic confinement fusion research
- Lattice Boltzmann Simulation Studies
- Marine and Coastal Research
Korea Atomic Energy Research Institute
2014-2023
Korea University of Science and Technology
2010-2022
Georgia Institute of Technology
2012
The phenomena of direct contact condensation (DCC) a steam jet submerged in water pool occur because the actuation discharging devices many industrial processes. There are practically two kinds technical concerns to consider. first is thermal mixing pool, and other thermo-hydraulically induced mechanical loads acting on structures relevant systems. inter-related can be well described only if local behavior condensing jets resultant turbulent understood. In this paper, DCC-related thermofluid...
A series of experiments were carried out to investigate phenomena from bubble nucleation lift-off for a subcooled boiling flow in vertical annulus channel. high-speed digital video camera was used capture the dynamics. The diameter and frequency evaluated terms heat flux, mass degree subcooling. fundamental features (i.e., variations across sites dependence on flux conditions) addressed based present observation. database built by gathering summarizing data Prodanovic et al., Situ...
The CUPID code has been developed at KAERI for a transient, three-dimensional analysis of two-phase flow in light water nuclear reactor components. It can provide both component-scale and CFD-scale simulation by using porous media or an open model flow. In this paper, recent advances the are presented three sections. First, domain decomposition parallel method implemented is described with efficiency test multiple processors. Then, coupling CUPID-MARS via heat structure introduced, where...
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,...
In order to measure the bubble departure frequency, a flow visualization system was set up on vertical annulus test section with heater rod by using high-speed camera. this study, we developed an efficient methodology of image processing for obtaining frequency data. Bubble nucleation investigated under various thermal hydraulic conditions water, which correspond pressures from 167 346 kPa, mass fluxes 214 1869 kg/m2s, heat 61 238 kW/m2, and subcooling degrees 7.5 23.4 K. The characteristics...
The program on thermal-hydraulic evaluation by testing and analysis (THETA) for the development licensing of new design features in APR1400 (Advanced Power Reactor-1400) is briefly introduced with a presentation research motivation typical results separate effect tests analyses major features. first part deals multi-dimensional phenomena related to safety APR1400. One area multidimensional behavior injection (SI) water reactor pressure vessel downcomer that uses direct type SI system. other...