MC2-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis
Author:
Affiliation:
1. Nuclear Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Lemont, Illinois 60439
2. Purdue University, School of Nuclear Engineering, 400 Central Drive, West Lafayette, Indiana 47907
Funder
U.S. Department of Energy
Publisher
Informa UK Limited
Subject
Nuclear Energy and Engineering
Link
https://www.tandfonline.com/doi/pdf/10.1080/00295639.2017.1320893
Reference14 articles.
1. Efficient Methods for the Treatment of Resonance Cross Sections
2. Nuclear Reactor Physics
3. Methods for Processing ENDF/B-VII with NJOY
4. ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology
5. A Rigorous Pole Representation of Multilevel Cross Sections and Its Practical Applications
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