Browsing by Subject "nuclear fuel"
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Item Assessment of uranium-free nitride fuels for spent fuel transmutation in fast reactor systems(Texas A&M University, 2004-09-30) Szakaly, Frank JosephThe purpose of this work is to investigate the implementation of nitride fuels containing little or no uranium in a fast-spectrum nuclear reactor to reduce the amount of plutonium and minor actinides in spent nuclear fuel destined for the Yucca Mountain Repository. A two tier recycling strategy is proposed. Thermal spectrum transmutation systems converted from the existing LWR fleet were modeled for the first tier, and the Japanese fast reactor MONJU was used for the fast-spectrum transmutation. The modeling was performed with the Monteburns code. Transmutation performance was investigated as well as delayed neutron fraction, heat generation rates, and radioactivity of the spent material in the short and long term for the different transmutation fuel cycles. A two-tier recycling strategy incorporating fast and thermal transmutation with uranium-free nitride fuel was shown to reduce the long-term heat generation rates and radioactivity of the spent nuclear fuel inventory.Item Atomic Diffusion in the Uranium-50wt% Zirconium Nuclear Fuel System(2013-06-17) Eichel, DanielAtomic diffusion phenomena were examined in a metal-alloy nuclear fuel system composed of ?-phase U-50wt%Zr fuel in contact with either Zr-10wt%Gd or Zr-10wt%Er. Each alloy was fabricated from elemental feed material via melt-casting, and diffusion samples of nominal 1.5 mm thickness were prepared from the resulting alloy slugs. The samples were assembled into diffusion couples and annealed for periods of 14, 28, and 56 days at temperatures of 550?C, 600?C, and 650?C. Thus, the U-50Zr/Zr-10Er system and the U-50Zr/Zr-10Gd system were each annealed for three different time periods at each of three different temperatures, for an initial total of 18 diffusion interfaces that were to be studied. In practice, data was collected from only 12 of the 18 interfaces. At 650?C, the U-50wt%Zr alloy exists in the ?-phase region, which enabled the comparison of diffusion behavior between the ? phase and ? phase. Diffusion samples were examined by collecting composition profiles across the diffusion interface for each element via electron probe microanalysis. From the resulting experimental data diffusion coefficients were evaluated. Diffusion coefficients were found to be on the order of 10^-19 m2/s in the ?-phase systems, and 10^-17 m^2/s in the ?-phase systems. It was observed that atomic mobility of all diffusing species was generally greater in the U-50Zr/Zr-10Gd system than in the U-50Zr/Zr-10Er system; furthermore, it was found that diffusion rates were considerably higher above the phase transformation temperature into the ? phase, as indeed would be expected in the more open structure of the body-centered cubic ? phase, as compared to the hexagonal ?-phase U-Zr. However, values for diffusion coefficients measured in this study were considerably smaller than those found in past studies of ?-phase U-Zr, which are on the order of 10^-17 m^2/s. It is likely that diffusion was inhibited by the formation of stable metal oxides resulting from oxygen contamination; it is also possible that diffusion was suppressed by the presence of the erbium and gadolinium.Item Development of dual phase magnesia-zirconia ceramics for light water reactor inert matrix fuel(Texas A&M University, 2005-02-17) Medvedev, PavelDual phase magnesia-zirconia ceramics were developed, characterized, and evaluated as a potential matrix material for use in light water reactor inert matrix fuel intended for the disposition of plutonium and minor actinides. Ceramics were fabricated from the oxide mixture using conventional pressing and sintering techniques. Characterization of the final product was performed using optical microscopy, scanning electron microscopy, x-ray diffraction analysis, and energy-dispersive x-ray analysis. The final product was found to consist of two phases: cubic zirconia-based solid solution and cubic magnesia. Evaluation of key feasibility issues was limited to investigation of long-term stability in hydrothermal conditions and assessment of the thermal conductivity. With respect to hydrothermal stability, it was determined that limited degradation of these ceramics at 300^oC occurred due to the hydration of the magnesia phase. Normalized mass loss rate, used as a quantitative indicator of degradation, was found to decrease exponentially with the zirconia content in the ceramics. The normalized mass loss rates measured in static 300^oC de-ionized water for the magnesia-zirconia ceramics containing 40, 50, 60, and 70 weight percent of zirconia are 0.00688, 0.00256, 0.000595, 0.000131 g/cm2/hr respectively. Presence of boron in the water had a dramatic positive effect on the hydration resistance. At 300^oC the normalized mass loss rates for the composition containing 50 weight percent of zirconia was 0.00005667 g/cm2/hr in the 13000 ppm aqueous solution of the boric acid. With respect to thermal conductivity, the final product exhibits values of 5.5-9.5 W/(m deg) at 500^oC, and 4-6 W/(m deg) at 1200^oC depending on the composition. This claim is based on the assessment of thermal conductivity derived from thermal diffusivity measured by laser flash method in the temperature range from 200 to 1200^oC, measured density, and heat capacity calculated using rule of mixtures. Analytical estimates of the anticipated maximum temperature during normal reactor operation in a hypothetical inert matrix fuel rod based on the magnesia-zirconia ceramics yielded the values well below the melting temperature and well below current maximum temperatures authorized in light water reactors.