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A code to calculate fast neutron spectra and multigroup cross sections

Standard Code Description

  1. Coding Language and Computing Platforms
    Fortran source code for Linux, Microsoft Windows, Macintosh 
  2. Description of Purpose
    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 the MC2-2. MC2-3 includes the one-dimensional cell calculation capabilities of SDX such that SDX is no longer needed. MC2-3 solves the consistent P1 multigroup transport equation using basic neutron data from ENDF/B processed by ETOE to determine the fundamental mode spectra for use in generating multigroup neutron cross sections. A homogeneous mixture, slab, or cylindrical unit cell problem can be 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.
  3. Typical Running Time
    The running time is generally less than 20 minutes on a modern computing workstation noting that MC2 is a serial application
  4. 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.
  5. Primary Authors
    • C.H. LeeArgonne National Laboratory
    • W.S. Yang, Purdue University
  6. Materials Available
    The source code and compilation instructions are provided. Precompiled executables for Linux and Macintosh are available. Documentation on methodology and installation is provided along with all of the verification test cases. Contact nera-​software@​anl.​gov for licensing and distribution information.
  7. Sponsor
    U.S. Department of Energy, Office of Nuclear Energy