Parametric Analysis of PWR Spent Fuel Depletion Parameters for Long-Term Disposal Criticality Safety

Utilization of burnup credit in criticality safety analysis for long-term disposal of spent
nuclear fuel allows improved design efficiency and reduced cost due to the large mass of fissile
material that will be present in the repository. Burnup-credit calculations are based on depletion
calculations that provide a conservative estimate of spent fuel contents (in terms of criticality
potential), followed by criticality calculations to assess the value of the effective neutron
multiplication factor (k eff ) for a spent fuel cask or a fuel configuration under a variety of
probabilistically derived events. In order to ensure that the depletion calculation is conservative,
it is necessary to both qualify and quantify assumptions that can be made in depletion models.
This report describes calculations performed at the Oak Ridge National Laboratory
(ORNL) to assess the relative effects of depletion modeling assumptions on the predicted value
of the neutron multiplication factor for an infinite lattice (k i n f) of pressurized-water-reactor
(PWR) fuel. Calculations are performed assuming an infinite lattice of a Westinghouse 17 H 17
fuel pin design. It is anticipated that trends identified for such fuel will be generally applicable
to specific cask designs; however, design-specific calculations are likely to be necessary in
licensing that design. Furthermore, because of significant differences between pressurized- and
boiling-water reactors (BWRs), results in this report are not applicable to the depletion analysis
of BWR fuels.