Boiling heat transfer and critical heat flux (CHF)
- Study on Pool Boiling Heat Transfer Enhancement
- CHF and Boiling Heat Transfer Experiment and System Code Development using Refrigerant
- Nanofluid Coating for Accident Tolerant Fuel (ATF)
Design of analysis of new concepts nuclear safety systems
- Development of Hybrid Heat Pipe with Control Rod: Passive IN-core Cooling system (PINCs) for the nuclear power plant, UCAN (UNIST CANister) for spent nuclear fuel dry storage
- Safety Enhancement of SFR Fuel Assembly using New-Pattern Wire Wrap Spacer
- In-Vessel Retention through External Reactor Vessel Cooling (IVR-ERVC) using Liquid Metal Fin Concept
- Nuclear Power Plant Severe Accident Analysis and Simulation Code Study using MELCOR and SAS4A
- Rotational Mixing vane in the nuclear subchannel
Design and analysis of Generation IV reactors
- Sodium-cooled fast reactor (SFR) Safety Experiments to examine metal fuel fragmentation and its debris coolability
- Similarity technology of molten salt in single-phase natural/forced convection heat transfer
- Multiphysics simulation for Molten Salt Reactor (MSR): Analysis of Neutronics & Thermal-hydraulics coupled systems using OpenFOAM