Neutron Behavior in Cluster-Type Fuel Lattices, (II)

Theory and Analysis of Experiment

抄録

A multigroup method of calculation is presented for describing the neutron behavior in a clustertype fuel lattice. It solves the integral transport equation by a semi-analytical method proposed in a previous paper for calculating collision probabilities in the lattice of a clustered fuel element. Using only fundamental nuclear data, it gives space and energy dependent neutron flux. The method has been programmed for HITAC-5020F (computer code named CLUSTER-III).
The accuracy of the method has been tested by comparing the calculation with the experiment described in Part (I) of this paper. The lattices are 28-pin clusters of UO₂ or PuO₂+UO₂ fuel pins, with heavy- or light-water moderators and with light-water coolant containing varying void ratios. The quantities studied are micro-parameters, reaction distributions in energy and space, thermal disadvantage factors and the multiplication factors. It is found that the calculated results are generally in good agreement with experiment, typically within 10% for micro-parameters and thermal disadvantage factor, and within 1% for the multiplication factor.