Dissolution characteristics of mixed UO{sub 2} powders in J-13 water under saturated conditions

The Yucca Mountain Project/Spent Fuel program at Argonne National Laboratory is designed to determine radionuclide release rates by exposing high-level waste to repository-relevant groundwater. To gain experience for the tests with spent fuel, a scoping experiment was conducted at room temperature to determine the uranium release rate from an unirradiated UO{sub 2} powder mixture (14.3 wt % enrichment in {sup 235}U) to J-13 water under saturated conditions. Another goal set for the experiment was to develop a method for utilizing isotope dilution techniques to determine whether the dissolution rate of UO{sub 2} matrix is in accordance with an existing kinetic model. Results of these analyses revealed unequal uranium dissolution rates from the enriched and depleted portions of the powder mixture because of undisclosed differences between them. Although the presence of this inhomogeneity has precluded the application of the kinetic model, it also provided an opportunity to elaborate on the utilization of isotope dilution data in recognizing and quantifying such conditions. Detailed listings of uranium release and solution chemistry data are presented. Other problems commonly associated with spent fuel, such as the effectiveness of filtering media, the existence of uranium concentration peaks during early stages of the leach tests, the need for concentration corrections due to water replenishments of sample volumes, and experience derived from isotope dilution data are discussed in the context of the present results. 10 refs., 5 figs., 7 tabs

A Probabilistic Description of the Configurational Entropy of Mixing

This work presents a formalism to calculate the configurational entropy of mixing based on the identification of non-interacting atomic complexes in the mixture and the calculation of their respective probabilities, instead of computing the number of atomic configurations in a lattice. The methodology is applied in order to develop a general analytical expression for the configurational entropy of mixing of interstitial solutions. The expression is valid for any interstitial concentration, is suitable for the treatment of interstitial short-range order (SRO) and can be applied to tetrahedral or octahedral interstitial solutions in any crystal lattice. The effect of the SRO of H on the structural properties of the Nb-H and bcc Zr-H solid solutions is studied using an accurate description of the configurational entropy. The methodology can also be applied to systems with no translational symmetry, such as liquids and amorphous materials. An expression for the configurational entropy of a granular system composed by equal sized hard spheres is deduced

Argonne National Laboratory Reports

The Yucca Mountain Project/Spent Fuel program at Argonne National Laboratory is designed to determine radionuclide release rates by exposing high-level waste to repository-relevant groundwater. To gain experience for the tests with spent fuel, a scoping experiment was conducted at room temperature to determine the uranium release rate from an unirradiated uranium dioxide powder mixture (14.3 wt % enrichment in uranium-235) to J-13 water under saturated conditions. Another goal set for the experiment was to develop a method for utilizing isotope dilution techniques to determine whether the dissolution rate of uranium dioxide matrix is in accordance with an existing kinetic model. Results of these analyses revealed unequal uranium dissolution rates from the enriched and depleted portions of the powder mixture because of undisclosed differences between them. Although the presence of this inhomogeneity has precluded the application of the kinetic model, it also provided an opportunity to elaborate on the utilization of isotope dilution data in recognizing and quantifying such conditions. Detailed listings of uranium release and solution chemistry data are presented. Other problems commonly associated with spent fuel, such as the effectiveness of filtering media, the existence of uranium concentration peaks during early stages of the leach tests, the need for concentration corrections due to water replenishments of sample volumes, and experience derived from isotope dilution data are discussed in the context of the present results