Validation of Minor Actinide Cross Sections by Studying Samples Irradiated for 492 Days at the Dounreay Prototype Fast Reactor - II: Burnup Calculations

To evaluate neutron cross-section data of minor actinides (MAs), separated actinide samples and dosimetry samples were irradiated at the Dounreay Prototype Fast Reactor for 492 effective full-power days. Irradiated samples were analyzed both at Oak Ridge National Laboratory and at Japan Atomic Energy Research Institute (JAERI). This independent duplication has resulted in the generation of reliable radiochemical analysis data. Based on the burnup calculations of major actinide (²³⁵U and ²³⁹Pu) and dosimetry samples, the neutron flux distribution and the flux level were adjusted at the locations where MA samples were irradiated. The burnup calculations were carried out for MAs using the determined flux distribution and flux level. The calculated results were compared with the experimental data. A brief description of sample preparation and irradiation and a detailed discussion of radiochemical analysis at JAERI are given in a companion paper. The current paper discusses the burnup calculations and the validation of MA cross-section data in evaluated nuclear data libraries.