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A. Professor

    
 Deokjung Lee
 Associate professor
 Tel: +82-52-217-2940



 

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

 

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