In order to improve the efficiency of dry used nuclear fuel (UNF) storage, reduced reactivity methods are being developed for various reactor types and operating conditions. Sub-criticality must be maintained in the storage configuration and conservative computer simulations are used as the primary basis for loading the storage casks. Methodologies are now being developed to reduce the amount of modeling and computation in order to make conservative assessments of how densely fuel can be packed. The SCALE/TRITON (Standardized Computer Analyses for Licensing Evaluation / Transport Rigor Implemented with Time-dependent Operation for Neutronic Depletion) code system has been used to simulate boiling water reactor (BWR) operating conditions in order to investigate nuclide densities in UNF and how the use of control rod blades affect nuclide densities found in UNF. Rodded and unrodded operating cases for a fuel assembly have been used as bounding cases and are used as reference solutions in a piecewise data approximation methodology (PDA method). A variety of control blade insertion patterns have been used with the PDA method and simulated in TRITON in order to observe trends in nuclide densities with varying control blade use. The PDA method is compared with TRITON simulated data in order to evaluate the validity and accuracy of the PDA method. The PDA method gives very accurate results for fissile nuclides but is insufficient in treating densities as a function of burnup for fission products and fertile nuclides. Predicting nuclide densities from temporally balanced control blade insertion and withdrawal patterns is also a strength of the PDA method. The PDA method, however, is not capable of properly accounting for neutron spectral shifts and the behavior in nuclide densities brought about by the spectral shift or nuclide density saturation. Observing the causes for the shortcomings in the PDA method, a more robust methodology can be developed.

In the event that a radiological dispersal device (RDD) is detonated in the U.S. or near U.S. interests overseas, it will be crucial that the actors involved in the event can be identified quickly. If irradiated nuclear fuel is used as the dispersion material for the RDD, it will be beneficial for law enforcement officials to quickly identify where the irradiated nuclear fuel originated. One signature which may lead to the identification of the spent fuel origin is the isotopic composition of the RDD debris. The objective of this research was to benchmark a forward model methodology for predicting isotopic composition of spent nuclear fuel used in an RDD while at the same time optimizing the fidelity of the model to reduce computational time. The code used in this study was Monteburns-2.0. Monteburns is a Monte Carlo based neutronic code utilizing both MCNP and ORIGEN. The size of the burnup step used in Monteburns was tested and found to converge at a value of 3,000 MWd/MTU per step. To ensure a conservative answer, 2,500 MWd/MTU per step was used for the benchmarking process. The model fidelity ranged from the following: 2-dimensional pin cell, multiple radial-region pin cell, modified pin cell, 2D assembly, and 3D assembly. The results showed that while the multi-region pin cell gave the highest level of accuracy, the difference in uncertainty between it and the 2D pin cell (0.07% for 235U) did not warrant the additional computational time required. The computational time for the multiple radial-region pin cell was 7 times that of the 2D pin cell. For this reason, the 2D pin cell was used to benchmark the isotopics with data from other reactors. The reactors from which the methodology was benchmarked were Calvert Cliffs Unit #1, Takahama Unit #3, and Trino Vercelles. Calvert Cliffs is a pressurized water reactor (PWR) using Combustion Engineering 14??14 assemblies. Takahama is a PWR using Mitsubishi Heavy Industries 17??17 assemblies. Trino Vercelles is a PWR using non-standard lattice assemblies. The measured isotopic concentrations from all three of the reactors showed good agreement with the calculated values.