Abstract
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.
Highlights
The SCALE-XSProc [1] multigroup (MG) cross section processing procedure includes two methods to process problem-dependent self-shielded MG cross sections [2] for transport calculations: (1) a Bondarenko method [3], and (2) a pointwise (PW) slowing down calculation
The standard SCALE-XSProc procedure processes effective MG neutron cross sections and scattering matrices for use in MG neutron transport calculations, using the Bondarenko approach with the AMPX MG library and performing the PW slowing down calculations
Recent improvements discussed in this study have resolved multiple issues regarding thermal and fast reactor analyses
Summary
The SCALE-XSProc [1] multigroup (MG) cross section processing procedure includes two methods to process problem-dependent self-shielded MG cross sections [2] for transport calculations: (1) a Bondarenko method [3], and (2) a pointwise (PW) slowing down calculation. In this method, MG data at problem-specific conditions are interpolated from the tabulated self-shielding factors in the MG library. Several areas for improvement were identified in the SCALE-XSProc MG cross section processing: (1) resonance self-shielding calculations within the unresolved resonance range (URR) [5], (2) the 5 eV thermal cut-off energy for the free gas model, (3) on-the fly adjustments to the thermal scattering matrix, (4) normalization of the PW neutron flux, and (5) multigroup structure [5,6]. This work has reduced a significant amount of the difference compared with CE Monte Carlo by developing a new analytic probability table method for the URR, increasing the cut-off energy to 10 eV, incorporating a new procedure to explicitly reconstruct thermal scattering matrices, increasing robustness of the flux normalization, and improving the energy group structure for any advanced reactor analysis
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