MC ^{2} 3: Multigroup Cross Section Generation Code for Fast Reactor Analysis
Abstract
This paper presents the methods and performance of the MC2 3 code, which is a multigroup crosssection generation code for fast reactor analysis, developed to improve the resonance selfshielding and spectrum calculation methods of MC2 2 and to simplify the current multistep schemes generating regiondependent broadgroup cross sections. Using the basic neutron data from ENDF/B data files, MC2 3 solves the consistent P1 multigroup transport equation 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 (2082) or hyperfine (~400 000) group levels. In the resolved resonance range, pointwise cross sections are reconstructed with Doppler broadening at specified temperatures. The pointwise cross sections are directly used in the hyperfine group calculation, whereas for the ultrafine group calculation, selfshielded 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 selfshielded using the analytic resonance integral method. The ultrafine group calculation can also be performed for a twodimensional wholecore problem to generate regiondependent broadgroup cross sections. Verification tests have been performed using themore »
 Authors:

 Nuclear Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Lemont, Illinois 60439
 Purdue University, School of Nuclear Engineering, 400 Central Drive, West Lafayette, Indiana 47907
 Publication Date:
 Research Org.:
 Argonne National Lab. (ANL), Argonne, IL (United States)
 Sponsoring Org.:
 USDOE Office of Nuclear Energy  Nuclear Energy Advanced Modeling and Simulation (NEAMS)
 OSTI Identifier:
 1412700
 DOE Contract Number:
 AC0206CH11357
 Resource Type:
 Journal Article
 Journal Name:
 Nuclear Science and Engineering
 Additional Journal Information:
 Journal Volume: 187; Journal Issue: 3; Journal ID: ISSN 00295639
 Publisher:
 American Nuclear Society  Taylor & Francis
 Country of Publication:
 United States
 Language:
 English
 Subject:
 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; Multigroup cross section; fast reactor analysis; resonance selfshielding
Citation Formats
Lee, Changho, and Yang, Won Sik. MC 2 3: Multigroup Cross Section Generation Code for Fast Reactor Analysis. United States: N. p., 2017.
Web. doi:10.1080/00295639.2017.1320893.
Lee, Changho, & Yang, Won Sik. MC 2 3: Multigroup Cross Section Generation Code for Fast Reactor Analysis. United States. doi:10.1080/00295639.2017.1320893.
Lee, Changho, and Yang, Won Sik. Fri .
"MC 2 3: Multigroup Cross Section Generation Code for Fast Reactor Analysis". United States. doi:10.1080/00295639.2017.1320893.
@article{osti_1412700,
title = {MC 2 3: Multigroup Cross Section Generation Code for Fast Reactor Analysis},
author = {Lee, Changho and Yang, Won Sik},
abstractNote = {This paper presents the methods and performance of the MC2 3 code, which is a multigroup crosssection generation code for fast reactor analysis, developed to improve the resonance selfshielding and spectrum calculation methods of MC2 2 and to simplify the current multistep schemes generating regiondependent broadgroup cross sections. Using the basic neutron data from ENDF/B data files, MC2 3 solves the consistent P1 multigroup transport equation 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 (2082) or hyperfine (~400 000) group levels. In the resolved resonance range, pointwise cross sections are reconstructed with Doppler broadening at specified temperatures. The pointwise cross sections are directly used in the hyperfine group calculation, whereas for the ultrafine group calculation, selfshielded 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 selfshielded using the analytic resonance integral method. The ultrafine group calculation can also be performed for a twodimensional wholecore problem to generate regiondependent broadgroup cross sections. Verification tests have been performed using the benchmark problems for various fast critical experiments including Los Alamos National Laboratory critical assemblies; ZeroPower Reactor, ZeroPower Physics Reactor, and Bundesamt für Strahlenschutz experiments; Monju startup core; and Advanced Burner Test Reactor. Verification and validation results with ENDF/BVII.0 data indicated that eigenvalues from MC2 3/DIF3D agreed well with Monte Carlo NParticle5 MCNP5 or VIM Monte Carlo solutions within 200 pcm and regionwise onegroup fluxes were in good agreement with Monte Carlo solutions.},
doi = {10.1080/00295639.2017.1320893},
journal = {Nuclear Science and Engineering},
issn = {00295639},
number = 3,
volume = 187,
place = {United States},
year = {2017},
month = {6}
}
Works referenced in this record:
Methods for Processing ENDF/BVII with NJOY
journal, December 2010
 MacFarlane, R. E.; Kahler, A. C.
 Nuclear Data Sheets, Vol. 111, Issue 12
ENDF/BVII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology
journal, December 2006
 Chadwick, M. B.; Obložinský, P.; Herman, M.
 Nuclear Data Sheets, Vol. 107, Issue 12