Neutron cross section measurements for BUC approaches

Abstract

Criticality safety analysis is required at various stages of the back-end of the fuel cycle, i.e. reprocessing, transport, storage and disposal of spent nuclear fuel (SNF). To account for the reduction in reactivity due to fuel burnup, the Burn-Up Credit (BUC) concept was introduced. Evidently, this concept depends on the quality of nuclear data, in particular the absorption cross sections of some key nuclides. A dedicated programme has been established at the GELINA facility of the JRC-Geel to produce accurate cross section data and validate the evaluated nuclear data libraries for neutron interactions with fission fragments that are relevant for a BUC approach. In this work, cross section data for 103Rh and 155Gd are presented and the results are compared with the main evaluation libraries, showing good agreement in the thermal energy region with ENDF/B-VIII.0 and JEFF-3.3, but not with JENDL-4.0 for 103Rh.