VALIDATION OF SCALE-4 FOR BURNUP CREDIT APPLICATIONS

In the past, criticality analysis of pressurized water reactor (PWR) f uel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. If credit is allowed for fue l burnup in the design of casks that are used in the transport of spen t light wafer reactor fuel to a repository, the increase in payload ca n lead to a significant reduction in the cost of transport and a poten tial reduction in the risk to the public. A portion of the work has be en performed at Oak Ridge National Laboratory (ORNL) In support of the U.S. Department of Energy (DOE) efforts to demonstrate a validation a pproach for criticality safety methods to be used in burnup credit cas k design. To date, the SCALE code system developed at ORNL has been th e primary computational tool used by DOE to investigate technical issu es related to burnup credit. The SCALE code package is a well-establis hed code system that has been widely used in away from reactor applica tions. Criticality safety analyses are performed via the criticality s afety analysts sequences (CSAS) and spent-fuel characterization via th e shielding analysis sequence (SAS2H). The SCALE 27-group burnup libra ry containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) da ta 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 met hods used in evaluating criticality safety limits for applications out side reactors by correlation against critical experiments that are app licable. Numerous critical experiments for fresh PWR-type fuel in stor age and transport configurations exist and can be used as part of a va lidation database. However, there are no critical experiments with bur ned PWR-type fuel in storage and transport configurations. As an alter native, commercial reactors offer an excellent source of measured crit ical configurations. Part of the work that has been performed to date to validate the SCALE-4 code system for burnup credit applications usi ng measured critical configurations includes: 1. fresh fuel critical e xperiments 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 i n 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 representa tive of data available when burnup credit calculations are being perfo rmed in the determination of optimum cask loadings. The results report ed demonstrate the ability of the ORNL SCALE-4 methodology to predict a value of k(eff) very close to the known value of 1.0, both for fresh fuel criticals and for the more complex reactor criticals. Beyond the se results, additional work in the determination of biases and uncerta inties is necessary prior to use in burnup credit applications.