Estimates of axial leakage and void reactivity in a CANDU reactor

Estimating the axial leakage in a Canada deuterium uranium (CANDU) fuel channel has been of ongoing interest because of the perceived significance of this parameter to an accurate calculation of the void effect. To verify the accuracy of such estimates, a benchmark-quality calculation of void reactivity in a standard CANDU fuel channel has been performed using the Monte Carlo code MCNP4. The result of a similar calculation for an infinite lattice of 37-element CANDU fuel bundles has been reported elsewhere. In this paper, estimates of reactivity and reactivity difference for two different arrangements of 37-element fresh CANDU fuel at both operating and voided conditions in the presence of axial leakage are discussed. It is emphasized that these results are valid for a highly idealized reactor model and cannot be directly applied to an operating reactor