Analysis of Fresh Fuel Critical Experiments Appropriate for Burnup Credit Validation

The ANS/ANS-8.1 standard requires that calculational methods used in determining criticality
safety limits for applications outside reactors be validated by comparison with appropriate critical
experiments. This report provides a detailed description of 34 fresh fuel critical experiments and
their analyses using the SCALE-4.2 code system and the 27-group ENDF/B-IV cross-section library.
The 34 critical experiments were selected based on geometry, material, and neutron interaction
characteristics that are applicable to a transportation cask loaded with pressurized-water-reactor spent
fuel. These 34 experiments are a representative subset of a much larger data base of low-enriched
uranium and mixed-oxide critical experiments. A statistical approach is described and used to obtain
an estimate of the bias and uncertainty in the calculational methods and to predict a confidence limit
for a calculated neutron multiplication factor. The SCALE-4.2 results for a superset of approximately
100 criticals are included in uncertainty analyses, but descriptions of the individual criticals are not
included.