Professor

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2004              Ph. D. (Nuclear Thermal-Hydraulic), Department of Nuclear Engineering, SNU 

2005 ~ 2013  Senior Researcher, Thermal-Hydraulic Safety Research Division, KAERI 

2010 ~ 2011   Visiting Researcher, CEA-Grenoble, France 

2013 ~ 2017   Assistant Professor, Department of Nuclear Engineering, SNU

2019              Visiting Researcher, ANL Illinois, USA

2017 ~ 2022  Associate professor, Department of Nuclear Engineering, SNU

2021 ~           Executive Editor of Nuclear Engineering and Technology

2022 ~           Professor, Department of Nuclear Engineering, SNU

1983              Ph. D. (Nuclear Thermal-Hydraulic), Department of Nuclear Engineering, RPI

1983 ~ 2017    Professor, Department of Nuclear Engineering, SNU 

1997 ~ 2001    Advisor, Nuclear Safety and Security Commission  

2002 ~ 2004  Director, Research Support Center, College of Engineering, SNU

2009 ~ 2010   President, Korea Nuclear Society 

2010 ~ 2012    Vice President, Korea Institute of Science & Technology Evaluation and Planning 

2012 ~ 2014    President, KEPCO International Nuclear Graduate School (KINGS)

2017 ~             Emeritus professor, Department of Nuclear Engineering, SNU  

Mixed convection heat transfer inside a rectangular riser in an air-cooled passive safety system

Experimental investigation of critical heat flux on a single heater rod under inclined and rolling conditions

Modelling of liquid film off-take in reactor vessel upper downcomer based on high-fidelity experiment and simulation

Development of mechanistic boiling heat partitioning model for horizontal tube under natural convection 

Study on the buoyancy-aided turbulent mixed convection inside the vertical rectangular riser of VHTR RCCS 

Experiment and analysis for improvement of crossflow loss coefficient for PMR200 Core 

Experimental study for development of wall and interfacial friction models in two-dimensional film flow 

Development of heat transfer model for horizontal u-shaped heat exchanger submerged in pool 

Experiment on critical heat flux under rolling condition for TAPINS-M code development 

Thermal-hydraulic analysis of a floating nuclear reactor under ocean conditions using MARS-KS moving reactor model

Coupling between subchannel module of CUPID and fuel code FINIX for multi-physics transient analysis of PWR core

Thermal-structural analysis of micro reactor core using coupled OpenFOAM and heat pipe code

Implementation of multi-dimensional flow analysis capability in MARS-KS motion model 

for marine reactor safety analysis

Establishment and validation of reactor core simulator for subchannel thermal-hydraulic and neutron transport coupled analysis 

Improvement of grid-directed cross flow in CUPID subchannel scale analysis module based on CFD calculation 

Numerical simulation of structure phase change under high heat flux condition coupled with nuclear system analysis code 

Improvement of dynamic motion model in MARS-KS for marine reactor thermal-hydraulic analysis 

Improvement of subchannel scale analysis capability of CUPID with grid-directed cross flow and fuel rod models 

Experimental measurement and force balance analysis of sliding vapor bubble behavior on a horizontal tube 

Extension of the component thermal-hydraulics analysis code CUPID toward sub-channel scale analysis of PWR reactor core 

Development of three-ring conductance sensor based on FPCB for measuring liquid film thickness under non-isothermal condition  

Enhancement and optimization on cooling performance of gas-based fusion reactor blanket using binary and ternary gas mixtures