Nuclear data uncertainty propagation on spent fuel nuclide compositions

Nuclear data uncertainty propagation plays a growing role in neutronic simulation by giving confidence bounds to physical macroscopic parameters used for safety and design of nuclear systems. This study deals with nuclear data uncertainty propagation (cross-sections, fission spectrum, neutron multiplicities and fission yields) on the nuclide composition of two UO2 spent fuel samples of different burnup. In order to perturb the fission yields in a neutron lattice physics code, the perturbation methodology combines an external correlation matrix to the standard deviations available in the international evaluations. The resulting matrix is normalized to fit to physical constraints. Then the impacts of nuclear data uncertainties on the k-inf are assessed and the main contributors are highlighted. Two different sources of uncertainty for the cross-sections and the fission yields are compared showing the large effect of the input uncertainty. The actinides and fission products composition of the two samples after depletion are sorted according to the main uncertainty origin of their C/E (experimental, cross-sections, fission yields). Special attention is paid to the Nd-148 C/E comparisons as these results show the need for improvement of the Nd-147 capture uncertainties.