Evaluation of the irradiation-averaged fission yield for burnup determination in spent fuel assays

In order to derive the burnup of spent nuclear fuel from the concentration of selected fission products (typically the Nd isotopes and 137Cs), their irradiation-averaged ssion yields need to be known with suffcient accuracy, as they evolve with the changes in the actinide vector over the irradiation history. To obtain irradiation-averaged values, radiochemists often resort to robust generic methods { i.e., based on simple mathematical relations { that weight the fission yields according to the actinides contributing to fission, without performing core physics calculations. In order to assess the performance of those generic methods, a database of about 30 000 spent nuclear fuel inventories has been constructed from neutron transport and depletion simulations, covering a representative range of fuel enrichment, burnup, assembly designs and reactor types. When testing several existing methods for effective fission yield calculation, some inaccuracies were identified, originating from improper one-group cross-section parameters that do not accurately re ect resonance and self-shielding effects, and too crude approximations in the estimation of the actinide concentration evolution. Revised effective fission and absorption cross-section parameters are then proposed here, as a first improvement to the earlier burnup determination methods. As a second step, a novel method is proposed that reduces the error on their radiation-averaged fission yield values, and hence on burnup, while retaining a straightforward calculation scheme.