COMBINE7.1 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

Technical Report · Thu Sep 01 04:00:00 EDT 2011

DOI:https://doi.org/10.2172/1031657· OSTI ID:1031657

Yoon, Woo Y; Nigg, David W

COMBINE7.1 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.1 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 fine-group cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B3 or B1 zero-dimensional approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko self-shielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. COMBINE7.1 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constants may be output in any of several standard formats including INL format, ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a one-dimensional (1-D) discrete-ordinate transport code, is incorporated into COMBINE7.1. As an option, the 167 fine-group constants generated by zero-dimensional COMBINE portion in the program can be used to calculate regionwise spectra in the 1-D ANISN portion, all internally to reflect the 1-D transport correction. The regionwise spectra are then used to generate mutigroup regionwise neutron constants. The 1-D neutron transport can be performed up to three stages, e.g., from a TRISO fuel to PEBBLE to 1-D full core wedge. In addition, COMBINE7.1 has now the capability of adjoint flux calculation through the 1-D ANISN transport. Photon transport capability is also added. For this, a photon production and photo-atomic cross section library, MATNG.LIB, was generated in MATXS format through NJOY code. The photon production cross section matrix is of 167 neutron - 18 photon groups. Photo-atomic cross sections, including heating, are in 18 energy groups.

• Research Organization: Idaho National Laboratory (INL)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
• DOE Contract Number: AC07-05ID14517
• OSTI ID: 1031657
• Report Number(s): INL/EXT-08-14729
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
ADJOINT FLUX
APPROXIMATIONS
COMBINE
COMPUTER CODES
CROSS SECTIONS
DIFFUSION
ENERGY RANGE
FORTRAN
HEATING
INTERPOLATION
NEUTRON TRANSPORT
NEUTRONS
NGNP Methods
NUCLEAR DATA COLLECTIONS
Nuclear Criticality Safety Program (NCSP)
PHOTON TRANSPORT
PHOTONS
RESONANCE
SELF-SHIELDING
SPECTRA
TRANSPORT
TRANSPORT THEORY
WEIGHTING FUNCTIONS