Year
2003
Abstract
Many commercial nuclear power plants in Europe operate hybrid cores of low-enriched uranium (LEU) and mixed-oxide (MOX) fuel assemblies, complicating material verification and safeguards. Computational methods for nuclear fuel burnup and decay analysis provide a potentially powerful software tool for enhancing the capability and sensitivity of neutron and gamma assay techniques to verify spent nuclear fuel safeguards and discriminate between LEU and MOX fuel types. Predictive codes may be used to verify the fissile material compositions from declared records, and the predicted radiation sources may be compared with assembly measurements made using ION Fork detectors. Such analysis software can enhance verification capability compared with past practice that relied on empirical relationships to correlate the gross neutron and gamma signal with the burnup. However, in order to be of practical use to inspectors, such codes must be fast, accurate, and easy to use. Under a cooperative agreement between the U.S. Department of Energy and the European Atomic Energy Community (EURATOM), Oak Ridge National Laboratory is expanding the computational methods and data libraries in the existing ORIGEN-ARP code for the analysis of MOX and extended LEU assembly types. ORIGEN-ARP uses a unique cross-section interpolation scheme to prepare a problem-dependent library in a fast and automated manner, allowing a complete spent fuel burnup simulation to be performed in a small fraction of the time typically required by reactor codes that perform comparable tasks. The ORIGEN-ARP package has been integrated into a Windows graphically enhanced interface that requires minimal user input. Automated post-analysis data processing and plotting capabilities are provided. Calculated results obtained via these methods and data for the analysis of pressurized-waterreactor (PWR) MOX fuel have been validated against experimental data from the ARIANE International Program.