Abstract
A simple method is derived to improve the accuracy of the predicted neutron flux in and around control rods by coarse mesh difference diffusion scheme in fast reactor criticality calculations. In a two-region slab lattice, a simple polynomial function is used to approximate the flux in the integral transport model, and the coefficients in the polynomial are determined by the variational method, From this approximate transport solution, the terms of net current into the control rods are corrected in the difference diffusion scheme before treatment by iteration in the criticality calculation. The results represent an appreciable improvement over those obtained with the conventional diffusion scheme, without any increase of computer time in the iteration. This improvement is confirmed in comparison with the results obtained with the usual fine mesh transport scheme as the standard reference.
