Improved 3D modeling of CANDU reactors using transport-fitted diffusion coefficients

To model the neutronic physics behavior of the core in CANDU pressure tube type heavy water reactors with natural-uranium fuel, two levels of calculations are required. Initially, lattice-level transport calculations are carried out to obtain, with high detail and accuracy, the flux distribution inside the lattice cell and composition of the nuclear fuel. Lattice calculations use many (30- 180) energy groups and detailed geometric information to model the fuel channel and the fuel contained within. Once the lattice calculations are complete, the fuel compositions obtained can be used to generate cell-homogenized macroscopic cross-sections condensed to two energy groups, for use in full-core diffusion calculations. Two-group cell-homogenized cross-sections work to acceptable levels of accuracy in most full-core configurations. However, challenges appear when modelling the neutron flux at the fuel-reflector interface (at the boundary of the reactor). This work aims to improve the neutron flux estimates obtained in three-dimensional diffusion calculations by using diffusion coefficients fitted to transport results. It will be shown that significant improvements (>10%) can be made for modeling the neutron physics at the core-reflector interface. (author)