A Monte Carlo method of evaluating heterogeneous effects in plate-fueled reactors

Few-group nuclear cross sections for small plate-fueled, light and heavy water test reactors are frequently generated with unit cell models that contain a homogeneous mixture of fuel, cladding, and water. The heterogeneous unit cells do not need to be represented explicitly for neutronics calculations when the plate and coolant channel thicknesses are small compared with the mean-free-path of neutrons. However, neutron and photon heating calculations were performed with heterogeneous fuel models to predict accurately the heat deposited in the fuel meat, cladding, and coolant. Heat deposited in the coolant channels and outside the fuel elements does not have a direct impact on the peak fuel meat temperature but must be included in the total coolant system heat balance. The results of a heterogeneous Monte Carlo calculation that estimates the heat loads in different fuel regions are presented and the fact that similar homogeneous fuel models can be used for many calculations. The calculations presented here were performed on models of the Advanced Neutron Source (ANS) and the Massachusetts Institute of Technology Reactor 2 (MITR-2). The ANS is a small, 362-MW (fission), plate-fueled, heavy water reactor designed to produce an intense steady-state source of neutrons.