Regular Article

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

¹ Belgian Nuclear Research Centre (SCK CEN), Boeretang, 200, 2400 Mol, Belgium
² Federal Agency for Nuclear Control (FANC), Rue du Marquis 1 Box 6A, 1000 Brussels, Belgium

^* e-mail: goversk@hotmail.com

Received: 17 January 2022
Received in final form: 20 April 2022
Accepted: 25 July 2022
Published online: 13 September 2022

Abstract

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 fission yields need to be known with sufficient 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 reflect 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.