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

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COMBINE7.0 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 COMBINE7.0 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 finegroup 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 ... continued below

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Yoon, Woo Y. & Nigg, David W. September 1, 2008.

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COMBINE7.0 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 COMBINE7.0 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 finegroup 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 B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko selfshielding 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. The fine group cross sections in the thermal energy range are replaced by those selfshielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.0 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 constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a onedimensional, discrete-ordinate transport code, is incoprated into COMBINE7.0. As an option, the 167 fine-group constants generated by COMBINE portion in the program can be used to cacluate regionwise spectra in the ANISN portion, all internally to reflect the one-dimensional transport correction. Results for the criticality validation calculations are included as a part of verification and validation.

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  • Report No.: INL/EXT-08-14729
  • Grant Number: DE-AC07-05ID14517
  • DOI: 10.2172/1031660 | External Link
  • Office of Scientific & Technical Information Report Number: 1031660
  • Archival Resource Key: ark:/67531/metadc845374

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  • September 1, 2008

Added to The UNT Digital Library

  • May 19, 2016, 3:16 p.m.

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  • Dec. 8, 2016, 2:59 p.m.

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Yoon, Woo Y. & Nigg, David W. COMBINE7.0 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program, report, September 1, 2008; Idaho Falls, Idaho. (digital.library.unt.edu/ark:/67531/metadc845374/: accessed May 21, 2018), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.