A 1D monoenergetic neutron transport benchmark in an infinite medium

Research output: Chapter in Book/Report/Conference proceedingConference contribution

1 Scopus citations

Abstract

Analytical or semi-analytical benchmarks for the neutron transport equation are relatively infrequent. Some may argue they are no longer necessary because of the enormous computing power and computational technology that is now available. While to some extent true, they can still provide valuable code verification and also serve to teach theoretical and numerical transport methods not taught by executing MATLAB, MAPLE or MATHEMATICA programs or Monte Carlo simulations. The focus of this presentation is on a new analytical solution technique for the solution of the 1D, monoenergetic Green's function for neutron transport. In this formulation, we consider the analytical solution to a threeterm recurrence for flux moments resulting in a semianalytical benchmark. We then apply the benchmark to assess the accuracy of the PN approximation leading to a rather unexpected result.

Original languageEnglish (US)
Title of host publicationRadiation Protection and Nuclear Technology Applications; Fuel Cycle, Radioactive Waste Management and Decommissioning; Computational Fluid Dynamics (CFD) and Coupled Codes; Reactor Physics and Transport Theory
PublisherAmerican Society of Mechanical Engineers (ASME)
ISBN (Electronic)9780791845943
DOIs
StatePublished - Jan 1 2014
Event2014 22nd International Conference on Nuclear Engineering, ICONE 2014 - Prague, Czech Republic
Duration: Jul 7 2014Jul 11 2014

Publication series

NameInternational Conference on Nuclear Engineering, Proceedings, ICONE
Volume4

Other

Other2014 22nd International Conference on Nuclear Engineering, ICONE 2014
CountryCzech Republic
CityPrague
Period7/7/147/11/14

ASJC Scopus subject areas

  • Nuclear Energy and Engineering

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    Ganapol, B. D. (2014). A 1D monoenergetic neutron transport benchmark in an infinite medium. In Radiation Protection and Nuclear Technology Applications; Fuel Cycle, Radioactive Waste Management and Decommissioning; Computational Fluid Dynamics (CFD) and Coupled Codes; Reactor Physics and Transport Theory (International Conference on Nuclear Engineering, Proceedings, ICONE; Vol. 4). American Society of Mechanical Engineers (ASME). https://doi.org/10.1115/ICONE22-30156