In the past, criticality analysis of pressurized water reactor (PWR) fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. If credit is allowed for fuel burnup in the design of casks that are used in the transport of spent light water reactor fuel to a repository, the increase in payload can lead to a significant reduction in the cost of transport and a potential reduction in the risk to the public. A portion of the work has been performed at Oak Ridge National Laboratory (ORNL) in support of the U.S. Department of Energy (DOE) efforts to demonstrate a validation approach for criticality safety methods to be used in burnup credit cask design. To date, the SCALE code system developed at ORNL has been the primary computational tool used by DOE to investigate technical issues related to burnup credit. The SCALE code package is a well-established code system that has been widely used in away from reactor applications. Criticality safety analyses are performed via the criticality safety analysis sequences (CSAS) and spent-fuel characterization via the shielding analysis sequence (SAS2H). The SCALE 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B- V (fission products) data has been used for all calculations. The American National Standards Institute/American Nuclear Society (ANSI/ANS)-8.1 criticality safety standard requires validation and benchmarking of the calculational methods used in evaluating criticality safety limits for applications outside reactors by correlation against critical experiments that are applicable. Numerous critical experiments for fresh PWR-type fuel in storage and transport configurations exist and can be used as part of a validation database. However, there are no critical experiments with burned PWR-type fuel in storage and transport configurations. As an alternative, commercial reactors offer an excellent source of measured critical configurations. Part of the work that has been performed to date to validate the SCALE-4 code system for burnup credit applications using measured critical configurations includes:

  1. fresh fuel critical experiments having geometric and nuclear characteristics similar to PWR spent fuel in storage and transport configurations
  2. commercial PWR hot-zero-power and hot-full-power reactor critical configurations.
The ability to closely predict reactor critical conditions is important in the validation of a methodology for spent-fuel applications because input data are determined based on relatively little detail of reactor core operation. Such limited information is expected to be representative of data available when burnup credit calculations are being performed in the determination of optimum cask loadings. The results reported demonstrate the ability of the ORNL SCALE-4 methodology to predict a value of keff very close to the known value of 1.0, both for fresh fuel criticals and for the more complex reactor criticals. Beyond these results, additional work in the determination of biases and uncertainties is necessary prior to use in burnup credit applications.