Criticality calculations for heterogeneous reactor media are usually based on a single‐cell calculation off, the thermal utilization, followed by a ``homogenized'' treatment of the macroscopic neutron migration. A changed approach is here advocated, in which diffusion theory is applied only to the multiply‐connected moderator portions of the reactor, the fuel is treated by the use of multiplying boundary conditions, and no arbitrary averaging need be done. As an example, the one‐dimensional case of plane fuel slabs in moderator is treated, and the material buckling characteristic of the fundamental mode is obtained using two‐group theory. The equation for this case readily allows a comparison between neutron migration lengths parallel to and perpendicular to the fuel slabs: considerable anistropy is found, in magnitude greatly dependent on fuel slab separation.