The Effect of the Flux Separability Approximation on Multigroup Neutron Transport

• Published in: Journal of Nuclear Engineering Vol. 2; no. 1; pp. 86 - 96
• Main Authors: Nelson, Adam G., Boyd, William, Romano, Paul K.
• Format: Journal Article
• Language: English
• Published: Basel MDPI AG 22.03.2021; MDPI
• Subjects: Angle; Application programming interface; Approximation; Availability; Computing time; Cross-sections; Eigenvalues; Energy; ENGINEERING; Flux Separability; Homogenization; Libraries; Mathematical analysis; Method of characteristics; MGXS; Monte Carlo; Multigroup; Neutron flux; OpenMC; Probability distribution; Resonance absorption; Solvers; Zirconium
• ISSN: 2673-4362; 2673-4362
• DOI: 10.3390/jne2010009

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.