Abstract
A mathematical procedure is presented for the formulation of two typical approximation methods for solving the neutron transport (or diffusion) equation of a heterogeneous reactor, the homogenization method based on the unit cell model and the heterogeneous method of Feinberg and Galanin. Also the limitation of their validity is discussed. For these purposes the transport (diffusion) equation is transformed into a set of integral equations whose unknowns are boundary values of neutron flux at interfaces between the moderator and fuel or absorber lumps and whose kernels are Green's functions for homogeneous media composed of the moderator or lump's materials. This transformation also provides a new approximation technique.
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