COMPUTING WITH MCNP THE THERMAL FLUX DENSITY DUE TO THE DELAYED FISSION NEUTRON GENERATED IN PROCESSING THE CLADS OF BURNED-UP FUEL ROADS

Describe a two-step procedure to calculate, by monte Carlo method, the thermal neutron flux den-sity produced by delayed neutron from fission in processing the clads of spent fuels: first, to calculatethe number of delayed fission nuetrons produced by external neutron source and its spatial distributionover the basket containing the clads of spent fuel; Second, to calculate the thermal neutron flux densitygenerated by this delayed neutron source. We have made some subroutines to complete the calculationwith existing MCNP, such as generating the spatial distribution of delayed neutrons, sampling from thisdistribution, and counting the volume flux of the detectors by the method of statistial estimate. Thecomputational results show that the method of statistical estimate of volume flux is better than themethod of track length in solving the deep penetration problem.