Generalized equivalence methods for 3D multi-group neutron transport
Name
Forget_Generalized Equivalence Methods for 3D Multi-group Neutron Transport.pdf
Description
Accepted version
Size
625.81 KB
Format
Adobe PDF
Checksum (MD5)
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Author(s)
Giudicelli, Guillaume Louis
Smith, Kord
Forget, Benoit
Date Issued
September 2017
Journal
Annals of Nuclear Energy
Publisher
Elsevier BV
Citation
Giudicelli, Guillaume Louis, Smith, Kord, and Forget, Benoit. "Generalized equivalence methods for 3D multi-group neutron transport." Annals of Nuclear Energy 112 (February 2018): 9-16 © 2017 Elsevier Ltd
Version
Author's final manuscript
Abstract
The process of generating multi-group cross section data to be used in full core 3D transport models requires not only accurate resonance self-shielding methods, but also some form of equivalence method in order to precisely preserve reaction rates of spectral geometry calculations. This paper presents extensions of the traditional concepts of local reaction rate preservation (common in discontinuity factor, SPH, and BBH methods), to derive a new state-of-the-art transport equivalence method that incorporates angular flux jump conditions that provide polar angle neutron current preservation. This method is tested on numerous fixed-source pin-cell problems by condensing fine energy resonance fluxes and cross sections. The method is demonstrated to precisely preserve all spectral geometry multi-group reaction rates as well as polar angle neutron currents for a wide range of cross section resonance heights, fuel pin diameters, coolant densities, and group energy widths. Keywords: Neutron transport; Self-shielding; Equivalence; Condensation; Discontinuity factors; Jump conditions
MIT Department
Massachusetts Institute of Technology. Department of Nuclear Science and Engineering
Terms of Use
Creative Commons Attribution-NonCommercial-NoDerivs License
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DOI of Published Version
https://doi.org/10.1016/j.anucene.2017.09.024
