Prediction of the neutrons subcritical multiplication using the diffusion hybrid equation with external neutron sources

Silva, Adilson Costa da; Silva, Fernando Carvalho da; Martinez, Aquilino Senra

doi:10.1016/j.anucene.2010.10.021

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Multigroup Neutron Diffusion Equation1

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Cited by 9 publications

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“…These methods calculate with great precision the eigenvalue and neutron flux in each node as seen in the comparison with some reference methods (usually fine-mesh calculations). Amongst them, we point the nodal expansion method (NEM) [6][7][8] that uses the continuity equation and Fick's Law [9]. The neutron continuity equation and Fick's Law in one dimension, two groups of energy and stationary state are expressed by the following equations, respectively:…”

Section: Multigroup Neutron Diffusion Equationmentioning

confidence: 99%

Reconstruction of the Neutron Flux in a Slab Reactor

Silva

Martinez²,

Gonçalves³

2012

WJNST

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In this electronic article we use the one-dimensional multigroup neutron diffusion equation to reconstruct the neutron flux in a slab reactor from the nuclear parameters of the reactor, boundary and symmetry condition, initial flux and .The diffusion equation was solved analytically for one single homogeneous fuel region and for two regions considering fuel and reflector. To validate the method proposed, the results obtained in this article were compared using reference methods found in the literature.

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Section: Multigroup Neutron Diffusion Equationmentioning

confidence: 99%