Solving Neutron Transport Equation in the Reactor using the Intergral Average Derivative Method

Phan Huy Thien

Abstract: Space and time dependent neutron diffusion equations with many energy groups and taking into account delayed neutrons are nonlinear partial differential equations. IAD method and finite difference method for simplified equations to partial differential equations often, then be rewritten in matrix form. General solution of differential equations contain the exponential matrix of matrix coefficients. Specific methods applied IAD (average method integral derivative) as an example to determine the flux reactor radius spherical R reflector layer. R-R'

Keywords: The Neutron field in Reactor, Boltzman transport equation,