This report proposes and documents a computational benchmark for the estimation of the
additional reactivity margin available in spent nuclear fuel (SNF) from fission products and minor
actinides in a burnup-credit storage/transport environment, relative to SNF compositions
containing only the major actinides. The benchmark problem/configuration is a generic burnupcredit
cask designed to hold 68 boiling water reactor (BWR) spent nuclear fuel assemblies. The
purpose of this computational benchmark is to provide a reference configuration for the

This report has been prepared to support technical discussion of and planning for future
research supporting implementation of burnup credit for boiling-water reactor (BWR) spent fuel
storage in spent fuel pools and storage and transport cask applications. The review and
discussion in this report are based on knowledge and experience gained from work performed
in the United States and other countries, including experience with burnup credit for
pressurized-water reactor (PWR) spent fuel. Relevant physics and analysis phenomena are

Pressurized water reactor (PWR) burnup credit validation is
demonstrated using the benchmarks for quantifying fuel reactivity
decrements, published as Benchmarks for Quantifying Fuel Reactivity
Depletion Uncertainty, Electric Power Research Institute (EPRI)
report 1022909. This demonstration uses the depletion module
TRITON (Transport Rigor Implemented with Time-Dependent
Operation for Neutronic Depletion) available in the SCALE 6.1
(Standardized Computer Analyses for Licensing Evaluations) code

This report presents a best-estimate probabilistic risk assessment (PRA) to quantify the frequency of criticality accidents during railroad transportation of spent nuclear fuel casks. The assessment is of sufficient detail to enable full scrutiny of the model logic and the basis for each quantitative parameter contributing to criticality accident scenario frequencies. The report takes into account the results of a 2007 peer review of the initial version of this probabilistic risk assessment, which was published as EPRI Technical Report 1013449 in December 2006.

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

The purpose of this calculation report, Range of Applicability and Bias Determination for Postclosure
Criticality of Commercial Spent Nuclear Fuel, is to validate the computational method used to perform
postclosure criticality calculations. The validation process applies the criticality analysis methodology
approach documented in Section 3.5 of the Disposal Criticality Analysis Methodology Topical Report.1
The application systems for this validation consist of waste packages containing transport, aging, and

This report investigates trends in the radiological decay properties and changes in relative nuclide importance associated with increasing enrichments and burnup for spent LWR fuel as they affect the areas of criticality safety, thermal analysis (decay heat), and shielding analysis of spent fuel transport and storage casks. To facilitate identifying the changes in the spent fuel compositions that most directly impact these application areas, the dominant nuclides in each area have been identified and ranked by importance.

In the 1980s, a series of critical experiments referred to as the Haut Taux de Combustion (HTC)
experiments was conducted by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) at the
experimental criticality facility in Valduc, France. The plutonium-to- uranium ratio and the isotopic
compositions of both the uranium and plutonium used in the simulated fuel rods were designed to be
similar to what would be found in a typical pressurized-water reactor fuel assembly that initially had an

The purpose of this study is to provide insights into the neutronic similarities that may exist between a
generic cask containing typical spent nuclear fuel assemblies and commercial reactor critical (CRC) state-
points. Forty CRC state-points from five pressurized-water reactors were selected for the study and the
type of CRC state-points that may be applicable for validation of burnup credit criticality safety
calculations for spent fuel transport/storage/disposal systems are identified. The study employed cross-

Taking credit for the reduced reactivity of spent nuclear fuel (SNF) in criticality analyses is referred to as burnup credit (BUC). Criticality safety evaluations require validation of the computational methods with critical experiments that are as similar as possible to the safety analysis models, and for which the keff values are known. This poses a challenge for validation of BUC criticality analyses, as critical experiments with actinide and fission product (FP)

This paper presents an evaluation of the amount of burnup credit needed for high-density casks to
transport the current U.S. inventory of commercial spent nuclear fuel (SNF) assemblies. A prototypic
32-assembly cask and the current regulatory guidance were used as bases for this evaluation.
By comparing actual pressurized-water-reactor (PWR) discharge data (i.e., fuel burnup and initial
enrichment specifications for fuel assemblies discharged from U.S. PWRs) with actinide-only-based

As part of the plutonium waste form development and down-select process, repository analyses have been conducted to evaluate the long-term performance of these forms for repository acceptance. Intact and degraded mode criticality analysis of mixed oxide (MOX) spent fuel is presented in Volume I, while Volume II presents the evaluations of the waste form containing plutonium immobilized in a ceramic matrix.

Analytical methods, described in this report, are used to
systematically determine experimental fuel sub-batch
reactivities as a function of burnup. Fuel sub-batch reactivities
are inferred using more than 600 in-core pressurized water
reactor (PWR) flux maps taken during 44 cycles of operation
at the Catawba and McGuire nuclear power plants. The
analytical methods systematically search for fuel sub-batch
reactivities that minimize differences between measured and
computed reaction rates, using Studsvik Scandpower’s