Abstract
SCALE 6 includes several problem-independent multigroup (MG) libraries that were processed from the evaluated nuclear data file ENDF/B using a generic flux spectrum. The library data must be self-shielded and corrected for problem-specific spectral effects for use in MG neutron transport calculations. SCALE 6 computes problem-dependent MG cross sections through a combination of the conventional Bondarenko shielding-factor method and a deterministic continuous-energy (CE) calculation of the fine-structure spectra in the resolved resonance and thermal energy ranges. The CE calculation can be performed using an infinite medium approximation, a simplified two-region method for lattices, or a one-dimensional discrete ordinates transport calculation with pointwise cross-section data. This paper describes the SCALE-resonance self-shielding methodologies, including the deterministic calculation of the CE flux spectra using pointwise nuclear data, the method for using CE spectra to produce problem-specific MG cross sections for various configurations (including doubly heterogeneous lattices). It also presents results of verification and validation studies.