MC²-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis Nuclear

Technical Report · Fri Aug 31 00:00:00 EDT 2018

DOI:https://doi.org/10.2172/1483949· OSTI ID:1483949

Lee, Changho ^[1]; Jung, Yeon Sang ^[1]; Yang, Won Sik ^[1]

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1. Argonne National Lab. (ANL), Argonne, IL (United States)

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.

• Research Organization: Argonne National Laboratory (ANL)
• Sponsoring Organization: USDOE Office of Nuclear Energy (NE), Nuclear Energy Advanced Modeling and Simulation (NEAMS)
• DOE Contract Number: AC02-06CH11357
• OSTI ID: 1483949
• Report Number(s): ANL/NE-11-41 Rev.3; 147840
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
97 MATHEMATICS AND COMPUTING