MC2-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis.
RSICC is authorized to distribute MC2-3 only to DOE Labs and US Universities with DOE-related projects for the term of that project. All commercial rights are reserved and licensing requests should be directed to the Technology Development and Commercialization Division of technology transfer at Argonne National Laboratory. This software is export controlled under at least 0D999 as well as to sensitive countries.
AUXILIARY DATA LIBRARIES
In addition to the eight standard MC2 library files, MC2-3 requires the chi matrix data files of fissionable isotopes and the inelastic scattering data files. The PENDF files of NJOY are also used as an alternative to the reconstruction of resolved resonance cross sections (in particular, for verifying the reconstructed pointwise cross sections) or to self-shield the resonance-like cross sections of intermediate mass isotopes above the resonance energy.
Argonne National Laboratory, Argonne, Illinois.
Fortran 95; Unix, Linux, PC Windows (P00577MNYCP00).
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. The code is executable on UNIX, Linux, and PC Windows systems, and its library includes all isotopes of the ENDF/BVII.0 data.
5. METHOD OF SOLUTION
MC2-3 is a computer program for solving the neutron transport equation in homogeneous mixtures or one-dimensional geometries to determine a detailed spectrum for use in deriving multigroup cross sections for fast reactor calculations. The code is composed of eight major functional modules: input processor, unresolved resonance self-shielding, resolved resonance self-shielding, scattering matrix calculation, ultrafine group spectrum calculation, group collapsing, cross section file generation in the ISOTXS format, and optionally hyperfine group integral transport spectrum calculation. Different computational paths are employed, depending upon the problem geometry and the number of energy groups. The main data library required for MC2-3 is structured in eight distinct data files which are created by ETOE-2 in the file formats provided in documentation. The code also needs libraries for fission spectrum matrix data and inelastic scattering data. The PENDF files are called optionally, and the pointwise cross section files are reused for computational efficiency if available.
The limitations of MC2-3 are summarized to remind the user. Up-scattering is not considered, and no thermal scattering law is implemented. Thus MC2-3 is not applicable to the thermal energy range. Interaction between resolved and unresolved resonances is not taken into account. The (n,3n) reaction is not included. Inelastic scattering moments higher than one are not provided; the P1 moments are currently supplied by the external files which are generated by NJOY. Only homogeneous and one-dimensional slab and cylinder geometries are currently supported for unit cell calculations. The MC2 libraries have been tested and verified only for ENDF/B-VII.0 data, and those for ENDF/B-V are being tested because of methodological changes in MC2-3, although they were used successfully in MC2-2 for a long time. The other libraries such as ENDF/B-VI, JEF, and JENDL have not been tested yet.
All included benchmarking problems ran in about 30 minutes.
Developed using the Compaq Visual Fortran on the Microsoft Windows operating system (OS), the MC2-3 code can be installed and executed on the Windows, Macintosh, Unix, and Linux OS environments. The memory requirements depend upon the problem. The current version requires more than 1G byte of memory. The memory management system in the current version does not use scratch files to save memory. Thus more than 4G byte of memory may be required for large problems with many isotopes, hyperfine groups, and/or one-dimensional geometry.
A Fortran compiler is required to compile the included source code. Minor changes may be required for code compilation.
a) included in documentation
C. H. Lee and W. S. Yang, “MC2-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis,” ANL/NE-11-41 Rev.1 (January 2012).
C. H. Lee and M. A. Smith, “Installation Guide for MC2-3,” ANL/NE-12-16 (January 2012).
b) background information
L. C. Just, H. Henryson, II, A. S. Kennedy, S. D. Sparck, B. J. Toppel, and P. M. Walker, “The System Aspects and Interface Data Sets of the Argonne Reactor Computation (ARC) System,” ANL‑7711 (April 1971).
C. G. Stenberg and A. Lindeman, “The ARC System Cross Section Generation Capabilities, ARC‑MC2-2,” ANL‑7722 excerpt 571‑576 (June 1973).
“Standard Interface Files and Procedures for Reactor Physics Codes, Version III,” LA‑5486‑MS (February 1974).
L. C. Leal, C. G. Stenberg, and B. R. Chandler, “Completion of the Conversion of the MC2-2/SDX Codes from IBM to SUN (#3),” Argonne National Laboratory Internal Memorandum (May 27, 1993).
C. G. Stenberg and H. Henryson, II, “MC2-2 Information on IBM and CDC Implementation, Use of ENDF/B‑V Library, and Added Capabilities,” ANL/AP Memorandum (January 13, 1981).
The package contains source code, sample problems, documentation and reference material in tar format.
KEYWORDS: MULTIGROUP CROSS SECTION PROCESSING