2019
DOI: 10.1080/00295450.2019.1571828
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Multigroup Cross-Section Generation with the OpenMC Monte Carlo Particle Transport Code

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Cited by 34 publications
(5 citation statements)
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“…The axial length of the fuel lattice is discretized into 20 segments. The MGXS module [29] of the OpenMC code [30] is employed to generate the multigroup (20 groups) neutron cross section based on a 2-dimensional slice of the MSRE fuel lattice. The vacuum boundary condition is applied on the top and bottom boundary.…”
Section: Griffin and Sam Modelmentioning
confidence: 99%
“…The axial length of the fuel lattice is discretized into 20 segments. The MGXS module [29] of the OpenMC code [30] is employed to generate the multigroup (20 groups) neutron cross section based on a 2-dimensional slice of the MSRE fuel lattice. The vacuum boundary condition is applied on the top and bottom boundary.…”
Section: Griffin and Sam Modelmentioning
confidence: 99%
“…For each of these models, a CE OpenMC simulation with ENDF/B-VII.1 cross sections will be used to generate the MGXS for the given model, group structure, and with or without the FSA. All MGXS libraries developed herein will use the OpenMC MGXS Python API interface and its methods discussed in [2]. Tracklength estimators are used for all reaction rates and fluxes, except for generating the scattering matrices and fission energy spectra that use analog estimators for the relevant reaction rates and flux.…”
Section: Evaluation Of Flux Separability Approximation Errorsmentioning
confidence: 99%
“…OpenMC is a Monte Carlo particle transport code developed with an emphasis on reactor analysis calculations [1]. The software and its accompanying Python application programming interface (API) has the ability to generate multigroup cross sections (MGXS) [2] from the continuous-energy (CE) calculation. OpenMC also contains a multigroup (MG) transport solver [3].…”
Section: Introductionmentioning
confidence: 99%
“…The OpenMC continuous energy Monte Carlo code Romano and Forget (2013) was employed to generate multigroup cross sections, and reference multi-group reaction rates and fluxes, for the second test case. The openmc.mgxs Python module Boyd et al (2017) was used to tally multi-group cross 1 Equal volume radial rings were used in the fuel; equally spaced radial rings were used in the moderator. 2014) was employed to model the second test case benchmark using the MGXS generated by OpenMC.…”
Section: Simulation Toolsmentioning
confidence: 99%