Nuclear Technology / Volume 167 / Number 2 / August 2009 / Pages 268-287
Technical Paper / Fuel Cycle and Management / dx.doi.org/10.13182/NT09-A8963
This paper provides insights into the neutronic similarities between a representative high-capacity rail-transport cask containing typical pressurized water reactor (PWR) spent nuclear fuel assemblies and critical reactor state-points, referred to as commercial reactor critical (CRC) state-points. Forty CRC state-points from five PWRs were analyzed, and the characteristics of CRC state-points that may be applicable for validation of burnup-credit criticality safety calculations for spent fuel transport/storage/disposal systems were identified. The study employed cross-section sensitivity and uncertainty analysis methods developed at Oak Ridge National Laboratory and the TSUNAMI set of tools in the SCALE code system as a means to investigate neutronic similarity on an integral and nuclide-reaction-specific level. The results indicate that except for the fresh-fuel-core configuration, all analyzed CRC state-points are either highly similar, similar, or marginally similar to the representative high-capacity cask containing spent nuclear fuel assemblies with burnups ranging from 10 to 60 GWd/tU in terms of their shared uncertainty in keff due to cross-section uncertainties. On a nuclide-reaction-specific level, the CRC state-points provide significant coverage, in terms of neutronic similarity, for most of the actinides and fission products relevant to burnup credit. Hence, in principle, the evaluated CRC state-points could serve as part of a set of benchmark experiments for determining a bias and bias uncertainty to be applied to the calculated keff of a spent fuel transport/storage/disposal system to correct for approximations in computational methods and errors and uncertainties in nuclear data. Note, however, that an evaluation to quantify the uncertainties associated with various CRC modeling parameters (e.g., fuel isotopic compositions, physical characteristics of reactor core components, and reactor operating history information), which has relevance to the use of these critical configurations for bias determination, was not performed as part of this study.