Multigroup Cross-Section Generation with the OpenMC Monte Carlo Particle Transport Code

Boyd, William; Nelson, Adam G.; Romano, Paul K.; Shaner, Samuel Christopher; Forget, Benoit; Smith, Kord

doi:10.1080/00295450.2019.1571828

Cited by 27 publications

(5 citation statements)

References 8 publications

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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%

Validation of NEAMS Tools Using MSRE Data

Fei

Shahbazi

Fang

et al. 2022

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Section: Griffin and Sam Modelmentioning

confidence: 99%

Validation of NEAMS Tools Using MSRE Data

Fei

Shahbazi

Fang

et al. 2022

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“…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 Effect of the Flux Separability Approximation on Multigroup Neutron Transport

Nelson

Boyd

Romano

2021

JNE

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The angular dependence of flux-weighted multigroup cross sections is commonly neglected when generating multigroup libraries. The error of this flux separability approximation is typically not isolated from other error sources due to a lack of availability of library generation and corresponding solvers that cannot relax this approximation. These errors can now be isolated and quantified with the availability of a multigroup Monte Carlo transport and multigroup library-generation capability in the OpenMC Monte Carlo transport code. This work will discuss relevant details of the OpenMC implementation, provide an example case useful for detailing the type of errors one can expect from making the flux separability approximation, and end with more realistic problems which show the impact of the approximation and highlight how it can strongly arise from an energy-dependent resonance absorption effect. Since the angle-dependence is intrinsically linked to the energy group structure, these examples also show that relaxing the flux separability approximation with angle-dependent cross sections could be used to reduce either the fine-tuning required to set a multigroup energy structure for a specific reactor type or the number of energy groups required to obtain a desired level of accuracy for a given problem. This trade-off could increase the costs of generating multigroup cross sections, and has the potential to require more memory for storing the multigroup library during the transport calculations, but it can significantly reduce the computational time required since the runtime of a discrete ordinates or method of characteristics neutron transport solver scales roughly linearly with the number of groups.

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“…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%