Woodruff School of Mechanical Engineering

Faculty Candidate Seminar


Advancement of multi-group cross section processing and resonance self-shielding methodology for multi-physics whole core simulations


Dr. Kang Seog Kim


Oak Ridge National Laboratory


Thursday, September 17, 2015 at 11:00:00 AM


Boggs Building, Room 3-47


Dr. Bojan Petrovic


High-fidelity multi-physics whole core simulation code packages are under development for which both multigroup (MG) deterministic and continuous energy (CE) Monte Carlo transport codes have been considered for neutronics simulation. However, although there is a significant success in on-the-fly CE cross section interpolation for specific temperatures and depletion capability in Monte Carlo codes, Monte Carlo simulation for actual reactor core with Thermal-Hydraulic (T-H) feedback and burnup is still very far from practical even with massively parallel supercomputing. Currently the only successful approach is a deterministic 2D/1D synthetic method by utilizing about 50 group cross section library in which 2-D radial MOC transport and 1-D axial nodal diffusion or SPn calculations are performed to nonlinearly update neutron leakage correction factors at a frame of 3-D coarse mesh finite difference method. The accuracy of this deterministic whole core neutronics simulation is determined by both spatial/angular discretizations and MG cross sections. The latter is more challenging due to coarse energy group and dominant in determining accuracy, and requires special cross section processing procedure and resonance self-shielding methodology to achieve high accuracy. Intensive research and development has been performed to develop a coarse group cross section library including resonance data and the associated resonance self-shielding methodology for high-fidelity whole core transport simulations with T-H feedback and burnup. This presentation focuses on an overall cross section library processing procedure and resonance self-shielding methods based on traditional approach and author’s new developments to enhance computational accuracy and performance.


Dr. Kang Seog Kim is a senior R&D staff at Oak Ridge National Laboratory joined reactor physics group on 2011 where he is working on the AMPX/SCALE methodology development, leading accuracy and performance improvement of the CASL core simulator MPACT and responsible for the MPACT cross section library. He started his career at Korea Atomic Energy Research Institute (KAERI) working in reload core design for PWRs, development of the SMART small and modular reactor (SMR) and development of various reactor physics analysis code packages for more than 20 years. His background is a strong combination of academic and industrial. His expertise is in computational reactor physics including nuclear data processing, resonance self-shielding methodology, diffusion/transport theory, transport lattice methodology, 2-step transport/nodal diffusion reactor physics analysis procedure, multi-physics whole core transport simulator, uncertainty evaluation, in-core fuel management and radiation shielding. He developed various state-of-the art reactor physics analysis codes for cross section library processing, slowing down, transport lattice, burnup and whole core simulation. His experience on reactor specific methodology and application includes Light Water Reactor, SMR, High Temperature Gas-Cooled Reactor, Super Critical Water Reactor and Research Reactor. He earned Bachelors and Master degrees in physics from Yonsei University in S. Korea and completed his Ph.D. degree in nuclear engineering at Oregon State University during his employment at KAERI. He published more than 120 journal articles and conference papers covering most of reactor physics areas from nuclear data to nuclear power plant applications. His research interests include computational reactor physics methodologies, code development and applications for the current and future needs of nuclear industry and government.


Refreshments will be served.