M.Phil. (Energy Studies) / Nodal diffusion methods are often used to calculate the distribution of neutrons in a nuclear reactor core. They require few-group homogenized neutron cross sections for every heterogeneous sub-region of the core. The homogenized cross sections are pre-calculated at various reactor states and represented in a way that facilitates the reconstruction of cross sections at other possible states. In this study a number of such representations were built for the homogenized cross sections of a MOX (mixed oxide) fuel assembly via hierarchical Lagrange interpolation on Clenshaw-Curtis sparse grids. These cross sections were represented as a function of various thermal hydraulic and material composition parameters of a pressurized water reactor core (i.e. burnup, soluble boron concentration, fuel temperature, moderator temperature and moderator density), which are generally referred to as state parameters. Representations were produced for the homogenized cross sections of a number of individual isotopes, as well as the e ective (lumped) cross section of all the materials in the assembly. This was done for both two and six energy groups. Additionally, two sets of state parameter intervals were considered for each of the group structures. The first set of intervals was chosen to correspond to conditions that may be encountered during day-to-day reactor operations. The second set of intervals was chosen to be applicable to the simulation of accident scenarios and therefore have wider ranges for fuel temperature, moderator temperature and moderator density.