SUMMARY

The current generation of methods for whole-core neutronic reactor analysis is still based on diffusion theory and relies heavily on spatial homogenization. These methods work well for the operating generation of nuclear reactors but the trend towards more compositional heterogeneity in assembly and core of advanced and Generation IV reactors will push these methods beyond their accuracy limits. The main goal of this thesis was to develop an efficient three-dimensional whole core neutronic method/tool which is based solely on transport theory, does not de-couple the transport phenomena between nodes, does not rely on homogenization or discontinuity factors, contains an accurate self-contained flux reconstruction procedure and does not restrict the size of the coarse meshes. This method thus eliminates the errors associated with spatial homogenization and diffusion theory. These characteristics make the new method a flexible tool for a variety of reactor designs and spectra. A heterogeneous coarse mesh transport method was extended from two to three dimensions in Cartesian geometry and was adapted to reduce the strain on computational resources. The high efficiency of the method is achieved by decoupling the problem into smaller sub-volume elements (e.g. coarse meshes) and shifting the computation time to a priori calculations of response functions for the unique sub-volumes in the system. That is, the method takes advantage of the repeated structure found frequently in large problems such as nuclear reactor cores.