This report describes a novel approach developed at the Oak Ridge National Laboratory
(ORNL) for the estimation of the uncertainty in the prediction of the neutron multiplication factor
for spent nuclear fuel. This technique focuses on burnup credit, where credit is taken in criticality
safety analysis for the reduced reactivity of fuel irradiated in and discharged from a reactor.
Validation methods for burnup credit have attempted to separate the uncertainty associated with

Slides - 2014 WM Symposia, March 2-6, 2014, Phoenix, AZ

A comprehensive, integrated data and analysis tool—the Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS) —is being developed for the US DOE Office of Nuclear Energy (DOE-NE) Nuclear Fuels Storage and Transportation Planning Project (NFST). The overarching goal of UNF-ST&DARDS is to provide a comprehensive controlled source of technical data integrated with key analysis capabilities to characterize inputs to the overall US waste management system from reactor power production through ultimate disposition.

The objective of this calculation is to provide the uncertainty term for fission product and minor actinides which contributes to the determination of the critical limit for burnup credit calculations. The scope of this calculation covers PWR and BWR spent nuclear fuel. This activity supports the Criticality Department's validation of burnup credit. The intended use of these results is in future Criticality Department calculations and analyses.