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Cassell, J.S. and Williams, M.M.R. (2004) A Solution of Neutron Diffusion Equation for a Hemispherical with Mixed Boundary Conditions. Annals of Nuclear Energy, 31, 1987-2004.
https://doi.org/10.1016/j.anucene.2004.04.008

has been cited by the following article:

• TITLE: Estimates of the Fast and Termal Flux in Blanket of Critical Reactors by Using Multi-Group Methods

  AUTHORS: Aybaba Hançerlioğullari, Aslı Kurnaz, Yosef G. Ali Madee, Ltfei A. Abdalsmd, Salem A. A. Shufat, Khaled M. Elhadad, Hand Hadia Almezogi, Mansur Mohamed Ali Mansur

  KEYWORDS: Critical Reactor, Neutron Diffusion Equation, Mcnp, Multi-Group Method, Simulation

  JOURNAL NAME: Open Journal of Applied Sciences, Vol.7 No.2, February 28, 2017

  ABSTRACT: In this study, based differential equations methods are used to solve equations because these methods are dependent on boundary value data more than other mathematical equations. We have calculated neutron flux, criticality and geometrical eigenvalue by using multi-group method and solving the neutron diffusion equation for finite and infinite cylindrical and spherical reactors in this study. For the calculation of the total neutron flux cross sections, we need the neutron diffusion equation. Thus, we have established the relationship between neuron flow and cross-section of neuron depending on neutron energy. Critical calculations have been made by comparing the results with MNCP (montecarlo n-partical) simulation methods. For necessary computer calculations, the programme, Wolfram-Matematica-7 has been used.