TY - GEN
T1 - Improvement of scale-xsproc multigroup cross section processing based on the centrm pointwise slowing down calculation
AU - Kim, Kang Seog
AU - Holcomb, Andrew M.
AU - Bostelmann, Friederike
AU - Wiarda, Dorothea
AU - Wieselquist, William
N1 - Publisher Copyright:
© The Authors, published by EDP Sciences. This is an open access article distributed under the terms of the Creative Commons Attribution License 4.0 (http://creativecommons.org/licenses/by/4.0/).
PY - 2020
Y1 - 2020
N2 - The SCALE-XSProc multigroup (MG) cross section processing procedure based on the CENTRM pointwise slowing down calculation is the primary procedure to process problem-dependent self-shielded MG cross sections and scattering matrices for neutron transport calculations. This procedure supports various cell-based geometries including slab, 1-D cylindrical, 1-D spherical and 2-D rectangular configurations and doubly heterogeneous particulate fuels. Recently, this procedure has been significantly improved to be applied to any advanced reactor analysis covering thermal and fast reactor systems, and to be comparable to continuous energy (CE) Monte Carlo calculations. Some reactivity bias and reaction rate differences have been observed compared with CE Monte Carlo calculations, and several areas for improvement have been identified in the SCALE-XSProc MG cross section processing: (1) resonance self-shielding calculations within the unresolved resonance range, (2) 10 eV thermal cut-off energy for the free gas model, (3) on-the-fly adjustments to the thermal scattering matrix, (4) normalization of the pointwise neutron flux, and (5) fine MG energy structure. This procedure ensures very accurate MG cross section processing for high-fidelity deterministic reactor physics analysis for various advanced reactor systems.
AB - The SCALE-XSProc multigroup (MG) cross section processing procedure based on the CENTRM pointwise slowing down calculation is the primary procedure to process problem-dependent self-shielded MG cross sections and scattering matrices for neutron transport calculations. This procedure supports various cell-based geometries including slab, 1-D cylindrical, 1-D spherical and 2-D rectangular configurations and doubly heterogeneous particulate fuels. Recently, this procedure has been significantly improved to be applied to any advanced reactor analysis covering thermal and fast reactor systems, and to be comparable to continuous energy (CE) Monte Carlo calculations. Some reactivity bias and reaction rate differences have been observed compared with CE Monte Carlo calculations, and several areas for improvement have been identified in the SCALE-XSProc MG cross section processing: (1) resonance self-shielding calculations within the unresolved resonance range, (2) 10 eV thermal cut-off energy for the free gas model, (3) on-the-fly adjustments to the thermal scattering matrix, (4) normalization of the pointwise neutron flux, and (5) fine MG energy structure. This procedure ensures very accurate MG cross section processing for high-fidelity deterministic reactor physics analysis for various advanced reactor systems.
KW - Multigroup cross section processing
KW - SCALE-XSProc
KW - Slowing down calculation
UR - http://www.scopus.com/inward/record.url?scp=85108451294&partnerID=8YFLogxK
U2 - 10.1051/epjconf/202124702011
DO - 10.1051/epjconf/202124702011
M3 - Conference contribution
AN - SCOPUS:85108451294
T3 - International Conference on Physics of Reactors: Transition to a Scalable Nuclear Future, PHYSOR 2020
SP - 194
EP - 201
BT - International Conference on Physics of Reactors
A2 - Margulis, Marat
A2 - Blaise, Partrick
PB - EDP Sciences - Web of Conferences
T2 - 2020 International Conference on Physics of Reactors: Transition to a Scalable Nuclear Future, PHYSOR 2020
Y2 - 28 March 2020 through 2 April 2020
ER -