COMBINE7.1 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program
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 Lab. (INL), Idaho Falls, ID (United States)
- Sponsoring Organization:
- DOE - NE
- DOE Contract Number:
- DE-AC07-05ID14517
- OSTI ID:
- 1031657
- Report Number(s):
- INL/EXT-08-14729; TRN: US1200186
- Country of Publication:
- United States
- Language:
- English
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COMBINE7.0 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program
NJOY; neutron and photon crosssections from ENDF/B. [CDC7600; American National Standard X3. 9-1966 FORTRAN language is used with a few exceptions, such as CDC overlay commands, mixed-mode arithmetic, and expressions as array indices. Machine-dependent functions such as timing, input-output, and character manipulation are isolated in special subroutines. Changes required for IBM implementation are denoted by 'CIBM' comment cards. Some loops have been rearranged to facilitate vectorization]
Related Subjects
ADJOINT FLUX
APPROXIMATIONS
COMPUTER CODES
CROSS SECTIONS
DIFFUSION
ENERGY RANGE
FORTRAN
HEATING
INTERPOLATION
NEUTRON TRANSPORT
NEUTRONS
NUCLEAR DATA COLLECTIONS
PHOTON TRANSPORT
PHOTONS
RESONANCE
SELF-SHIELDING
SPECTRA
TRANSPORT
TRANSPORT THEORY
WEIGHTING FUNCTIONS
COMBINE
NGNP Methods