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Creep strain has been identified as the dominant failure mode for commercial spent nuclear fuel cladding during dry storage, including the vacuum drying phase. It could also be important during the early period of repository closure. A statistical analysis of creep failure during these three phases was performed. Statistical analysis is an important tool for predicting fuel behavior and the distributions can be modified for specific applications. A burnup distribution (rod average = 44 MWd/kgU, range = 2 to 75 MWd/kgU) was assumed and a distribution of rod properties, including stress was developed. The Murty creep correlation was selected after comparing six different correlations with results from five different experiments. It was then modified to better predict irradiated cladding creep data. Creep failure criteria is a Complementary Cumulative Distribution Function (CCDF) based on 52 failure tests. The fuel rods are exposed to three consecutive temperature histories that are typical of what is to be expected: 24 hours of vacuum drying, 20 years dry storage, and 1000 years of repository thermal history. Each phase has a peak temperature, treated as an independent variable, and temperature history taken from the literature. Uncertainties in the temperatures and strain rate are included. The radial temperature distribution across the waste package is also modeled. For the first phase, vacuum drying, rod failures start to occur at about 550°C and exceed 1% failure at 600°C. With a peak vacuum drying temperature of 430°C, rods begin to fail during dry storage when the peak temperature reaches 400°C and approached a 1% failure level at 450°C. With representative peak temperatures of 430°C for drying and 350°C for dry storage, rod failures start to occur during repository closure at a peak cladding temperature of 390°C. They reached 1% at about 430°C. In the current repository design, the cladding temperatures are below 210°C and rod failures from creep are not expected

A mechanistic spent fuel dissolution model has been developed, based on the generalreaction-transport code of AREST-CT. It considers the dissolution of spent fuel under flow conditions. The kinetic reactions of spent fuel dissolution and precipitation of schoepite, uranophane, soddyite, and Na-boltwoodite are included in the model. The results of model prediction are compared against the results of drip-tests that simulate the conditions that may occur in the Yucca Mountain Repository. Comparison shows that the modeling results match the laboratory observations very well and no contradiction has been found. It indicates that the model is a reasonably good representation of the real system.After validation, the model was used to investigate the release rate of Np from the dissolution of secondary uranyl minerals by examining various degrees of Np incorporation into secondary uranyl minerals. The predicted Np concentration in the aqueous phase is 3 orders of magnitude lower than the upper-bound of the Np solubility range currently used in DOE performance assessment analyses. It suggests that the Np solubility range currently used is too conservative and could be replaced with more realistic values.