A. Professor
Deokjung Lee
Professor
Tel: +82-52-217-2940
Research interests:
Nuclear Reactor Physics
Reactor Core Design Analysis
Methodology Development
Education
2003: Ph.D., Nuclear Engineering, Purdue University, USA
1994: M.S., Nuclear Engineering, Seoul National University, South Korea
1992: B.S., Nuclear Engineering, Seoul National University, South Korea
Career
2022~Present: Professor, UNIST
2022~2024: Advisory Committee, Nuclear Safety and Security Commission, South Korea
2016~Present: CEO (founder) of Advanced Nuclear Technology and Services Co., Ltd.
2023~2024: Director of International Collaboration, Korea Nuclear Society
2021~2023: Director of Public Relation, Korea Nuclear Society
2016~2021: Associate Professor, UNIST
2011~2015: Assistant Professor, UNIST
2009~2011: R&D Staff, Oak Ridge National Laboratory, USA
2005~2009: Senior Nuclear Engineer, Studsvik Scandpower Inc., USA
2003~2004: Post-Doctoral Research Associate, Purdue University, USA
1995~2000: R&D Staff, Korea Electric Power Research Institute
Awards/Honors/Memberships
The Chairman of Balanced Development Committee appointed by the President of South Korea H. E. Yoon Suk Yeol, 2022
Award from the Minister of the Ministry of Science and ICT "Meritorious Deed to National Development in the Field of Nuclear Energy Promotion", 2020.
M&C 2017 Organization Committee
PHYSOR 2016 Organization Committee
Member, Joint Benchmark Committee of Mathematics and Computation, Reactor Physics, and Radiation & Shielding, American Nuclear Society, 2010~2012
Member, Reactor Physics Division Program Committee, American Nuclear Society, 2008~2011
Organization Committee for PHYSOR 2012 (Technical Program Co-chair)
Organization Committee for PHYSOR 2004 (Publication Committee)
The best paper award at ANS winter meeting, Washington, D.C., Nov. 2004
Research Interests
Nuclear Reactor Physics
Initial / Reload Reactor Core Design
Steady State and Transient Core Analysis
Resonance Treatment
On-the-fly Doppler Broadening
Safety Parameter Evaluations
Neutron Cross Section Measurements
Neutron Scattering Experiments
Numerical Methods for Neutron Transport and Diffusion Analysis
Monte Carlo Methods
Method of Characteristics
Unified Nodal Method
Acceleration Techniques
Advanced Reactor Designs
Light Water Reactors
Sodium-cooled Fast Reactors
Molten Salt Reactors
Small Modular Reactors
High Temperature Gas-cooled Reactors
High Performance Computing
Parallel Computing by MPI and OpenMP
Supercomputing
Lattice Physics Code Development
Monte Carlo Code
Deterministic Transport Analysis Code
Nodal Diffusion Transient Code
Operation Support Packages
Reactor Core Simulator
Wester Services Company NPP simulator
NEL macro-/micro- NPP simulator
IAEA/KAERI NPP simulator
Load Follow Operation Simulator
Ongoing Research Topics
1. Development of MOC Transport code
Multi-group Cross-Section Generation
- Pin-based pointwise energy slowing-down method
- Equivalence theory for structure material
- Resonance upscattering correction
- Enhanced neutron current method
- Resonance interference factor library method (option)
- Inflow transport correction
Transport Solver
- Method of Characteristics
- T-Y optimum quadrature sets
- Assembly modular ray tracing method
- Direct neutron path linking method
- P0~P5 scattering source treatment
- Coarse mesh finite difference acceleration
Depletion
- Matrix exponential method
- Chebyshev rational approximation method
- Chain with ~1400 isotopes
- Predictor/corrector
Few-group Constants Generation
- Discontinuity factor
- Two-group cross-sections
- Critical spectrum with fundamental mode calculation
2. Development of 3D Nodal Diffusion code
- Development of two-step approach reactor core design code
- Nodal diffusion theory and pin-by-pin SP3 solver steady/transient state calculation
- Engineering feature and multi-cycle calculation capability for practical reactor core design
- High fidelity multi-scale multi-physics coupling with TH subchannel, fuel performance, water chemistry, system codes for AOA/RIA analysis
- Advanced reactor core (SFR/VHTR/LWR/MSR) design and analysis capability
- Development of real time reactor simulator
- Parallel computing capability for pin-by-pin calculation
- Verification and Validation with IAEA ADS Benchmark, Analysis of KUCA Experiments: Phase I
3. Reactor Design
- Low-boron commercial PWR design
- Non-boron Small Modular PWR design
- Ultra-long Cycle Sodium Cooled Fast Reactor Design
- Long-life Lead Cooled Fast Reactor for Icebreacker
- Loading pattern search for long life mixed-cycle operation in PWR
- Load Follow Operation Method development based on Smart Grid
4. Fast Reactor Analysis
- Nuclear cross section library generation for fast reactors simulation in deterministic code system
- IAEA-CRP CEFR Benchmark Problem Validation
- ABR-1000 core benchmark problem verification
5. Development of Monte-Carlo code
- High fidelity multi-physics simulation (CTF, FRAPCON etc.)
- Cross section processing with On The Fly (OTF) doppler broadening
- Multi-group constant generation
- Generalized Perturbation Theory (GPT)
- Variance reduction technique
- Neutron/Gamma transport simulation
- Source term calculation
- Cask analysis
6. Multi-physics coupling
- Development of multi-physics coupled reactor core analysis code system
- Coupling with subchannel T/H codes
- Coupling with T/H system codes
- Coupling with fuel performance codes
- Coupling with water chemistry codes
- Nuclear reactor core safety analysis (RIA, LOCA etc.)
7. Sensitivity and Uncertainty Analysis
- Sensitivity coefficient calculation for various response functions such as the effective multiplication factor, reaction rate ratios and bilinear ratios
- Uncertainty quantification as a function of covariance data library
- Verification for UAM-LWR and UAM-SFR benchmark
8. AI & Simulator
- AI training data generation using nuclear power plant simulator
- Integrated code system for the safety analysis of spent fuel storage cask