[en] The NRU and NRX research reactors at Chalk River Nuclear Laboratories are very heterogeneous. Analyses of these reactors are performed using ''equivalent'' axially averaged homogeneous cell parameters for every lattice site. This allows a 2-D neutron diffusion code to be used to calculate flux distributions, fuel powers and eigenvalues. The axially averaged parameters are obtained from the code CELDATA which calculates ''equivalent'' homogeneous parameters for a cylindrical cell surrounded by two concentric environment annuli using neutron diffusion and perturbation theories in R, Z geometry. Reaction rates, neutron flux distributions and eigenvalues calculated by CELDATA, for both the axially heterogeneous central cell and its homogeneous ''equivalent'' are compared for three examples. Some of the NRU loops contain bundles with azimuthal variations in the outer fuel ring. An empirical method is presented whereby the individual element powers for a bundle of this type can be obtained by making corrections to the outer element powers calculated for a bundle with elements of an average composition. The method is tested against an experiment performed in the ZED-2 reactor. (author)