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Reactors designed by Argonne National Laboratory

Software

MC2-3 (Fast Reactor Cross Section Processing Codes)

 

Standard Code Description

  1. Name and Title of Program:
    MC23: A code to calculate fast neutron spectra and multigroup cross sections.
  2. Computer for Which Program is Designed and Other Machine Version Packages Available:
    Linux, Unix, Windows, Macintosh OS
  3. Description of Problem or Function:
    The MC2-3 code is a Multigroup Cross section generation Code for fast reactor analysis, developed by improving the resonance self-shielding and spectrum calculation methods of MC2-2 and integrating the one-dimensional cell calculation capabilities of SDX. The code solves the consistent P1 multigroup transport equation using basic neutron data from ENDF/B data files to determine the fundamental mode spectra for use in generating multigroup neutron cross sections. A homogeneous medium or a heterogeneous slab or cylindrical unit cell problem is solved in ultrafine (~2000) or hyperfine (~400,000) group levels. In the resolved resonance range, pointwise cross sections are reconstructed with Doppler broadening at specified isotopic temperatures. The pointwise cross sections are directly used in the hyperfine group calculation whereas for the ultrafine group calculation, self-shielded cross sections are prepared by numerical integration of the pointwise cross sections based upon the narrow resonance approximation. For both the hyperfine and ultrafine group calculations, unresolved resonances are self-shielded using the analytic resonance integral method. The ultrafine group calculation can also be performed for two-dimensional whole-core problems to generate region-dependent broad-group cross sections. Multigroup cross sections are written in the ISOTXS format for a user-specified group structure.
  4. Restrictions on the Complexity of the Problems:
    1D cylindrical, 2D, and 3D Cartesian and hexagonal geometries
  5. Typical Running Time:
    The running time depends on a problem size to solve. Use a single core for 1D cylindrical geometry and multi cores using OpenMP for 2D Cartesian and hexagonal geometries.
  6. Unusual Features of the Program:
    Neutron and gamma cross section generation. Optionally use the PENDF generated from NJOY.
  7. Related and Auxiliary Programs:
    Cross section library generation: ETOE-2, NJOY
  8. References:
    1. C. H. Lee and W. S. Yang, “MC2-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis,” ANL/NE-11/41 Rev.4 (2019)
    2. C. H. Lee and Y. S. Yang, “MC2-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis,” Nucl. Sci. Eng., 187, 268-290 (2017)
  9. Machine Requirements:
  10. Programming Languages Used:
    Fortran 90+.
  11. Operating System:
    Linux, Unix, Windows, Mac OS
  12. Other Programming or Operating Information or Restrictions:
  13. Name and Establishment of Author or Contributor:
    • C.H. Lee, Argonne National Laboratory
    • W.S. Yang, Purdue University
  14. Materials Available:
    Export controlled code based on 0D999. The code licensing should be requested to Argonne: Elizabeth K Jordan and/or Changho Lee.
  15. Sponsor:
    U.S. Department of Energy, Office of Nuclear Energy

Last Modified: Wed, June 3, 2020 5:06 PM

 

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