djlee
Deokjung Lee
Associate professor
Tel: +82-52-217-2940
DJLee_CV_20171226
Major
Nuclear Reactor Physics; Reactor Core Design Analysis, and Methodology Development
Curriculum Vitae
2016~Present: Associate Professor, UNIST
2011~2015: Assistant Professor, UNIST
2009~2011: R&D Staff, Oak Ridge National Laboratory
2005~2009: Senior Nuclear Engineer, Studsvik Scandpower Inc.,
1995~2000: Technical Staff, Korea Electric Power Research Institute
2003: Ph. D. Nuclear Engineering, Purdue University
1995: M. S. Nuclear Engineering, Seoul National University
1993: B. S. Nuclear Engineering, Seoul National University
Awards/Honors/Memberships
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
  1. Nuclear Reactor Physics
    • Initial / Relaod 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
  2. Numerical Methods for Neutron Transport and Diffusion Analysis
    • Monte Carlo Methods
    • Method of Characteristics
    • Unified Nodal Method
    • Acceleration Techniques
  3. Advanced Reactor Designs
    • Light Water Reactors
    • Sodium-cooled Fast Reactors
    • Molten Salt Reactors
    • Small Modular Reactors
    • High Temperature Gas-cooled Reactors
  4. High Performance Computing
    • Parallel Computing by MPI and OpenMP
    • Supercomputing
  5. Lattice Physics Code Development
    • Monte Carlo Code
    • Deterministic Transport Analysis Code
    • Nodal Diffuion Transient Code
    • Operation Support Packages
  6. Reactor Core Simulator
    • Wester Services Company NPP simulator
    • NEL macro-/micro- NPP simulator
    • IAEA/KAERI NPP simulator
    • Load Follow Opertion 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