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DOE OSTI · 1483949

MC 2 -3: Multigroup Cross Section Generation Code for Fast Reactor Analysis Nuclear

Lee, Changho·Jung, Yeon Sang·Yang, Won Sik
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Abstract

The MC 2 -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 MC 2 -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.

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Lee, Changho, Jung, Yeon Sang, Yang, Won Sik. 2018-08-31. MC 2 -3: Multigroup Cross Section Generation Code for Fast Reactor Analysis Nuclear. https://doi.org/10.2172/1483949

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