Abstract
Taking credit for the reduced reactivity of spent nuclear fuel in criticality analyses is referred to as burnup credit. Criticality safety evaluations employing burnup credit require validation of the depletion and criticality calculation methods and computer codes with available measurement data. To address the issues of burnup credit criticality validation, the U.S. Nuclear Regulatory Commission initiated a project with Oak Ridge National Laboratory to (1) develop and establish a validation approach for commercial spent nuclear fuel criticality safety evaluations based on best-available data and methods and (2) apply the approach on representative spent nuclear fuel storage and transport systems and conditions to demonstrate its usage and applicability, as well as to provide reference bias and bias uncertainty results. This report describes an approach for establishing depletion code bias and bias uncertainty in terms of a reactivity difference based on comparison of measured and calculated nuclide concentrations.
