Solution of the Transport Equations for two-Medium Slab Lattice with Anisotropic Scattering

The transport equations for two-medium slab lattice are studied. The neutron density distributions and the disadvantage factor for thermal neutrons in water moderated, highly enriched uranium and uranium oxide fuel lattices were obtained with its external surface with reflection coefficient R ? 1. Three coupled integral equations for the fluxes in the fuel and moderator are obtained. The Galerkin-method is used to solve the coupled integral equations for two-medium. The calculations are carried out for isotropic and anisotropic scattering in the moderator region and isotropic scattering in the fuel region. Numerical results are compared with the published calculations