Laves-type intermetallic phases have been observed to be the dominant phases in a series of alloy compositions being designed for the immobilization of technetium in a metallic waste form. The dominant metals in the alloy compositions were Fe-Mo and Fe-Mo-Zr. The alloy composition, Fe-Mo-Zr, also contained Pd, Zr, Cr, and Ni. Both non-radioactive rhenium-containing and radioactive technetium-bearing alloy compositions were investigated. In the Fe-Mo series, the phases observed were Fe2Mo (C14 Laves phase) and ferrite in agreement with predictions. Both Tc and Re resided predominantly in the Laves phases. In the Fe-Mo-Zr system, the phases included hexagonal C14 with the composition (Fe,Cr)2Mo, cubic C15 phase with a (Fe,Ni)2Zr composition, and the hcp phase Pd2Zr. The observation of these phases was in agreement with predictions. Re was found in the C14 intermetallic, (Fe,Cr)2Mo. Technetium was also observed to be partitioned preferentially into the (Fe,Cr)2Mo phase; however, this phase exhibited a cubic structure consistent with the C15 structural type. The composition of Laves phases is influenced by both the atomic size and electro-negativity of the constituent elements. The long-term release behavior of technetium under nuclear waste disposal conditions may be more dependent on the corrosion characteristics of these individual Laves phases containing Tc than the other metallic phases.

Results for a radiolysis model sensitivity study of radiolytically produced H2O2 are presented as they relate to Spent (or Used) Light Water Reactor uranium oxide (UO2) nuclear fuel (UNF) oxidation in a low oxygen environment. The model builds on previous reaction kinetic studies to represent the radiolytic processes occurring at the nuclear fuel surface. Hydrogen peroxide (H2O2) is the dominant oxidant for spent nuclear fuel in an O2-depleted water environment. The most sensitive parameters have been identified with respect to predictions under typical conditions. As compared with the full model with about 100 reactions, it was found that only 30 to 40 of the reactions are required to determine [H2O2] to one part in 10–5 and to preserve most of the predictions for major species. This allows a systematic approach for model simplification and offers guidance in designing experiments for validation.

This paper examines the ability of electron energy-loss spectroscopy (EELS) combined with transmission electron microscopy (TEM) to detect both low concentrations of Np in a U matrix and to provide evidence for incorporation of Np in U(VI) phases. The case for U(VI) secondary minerals acting as solubility-controlling phases for Np in repository performance assessment models has not been fully established. Direct evidence for incorporation, rather than sorption, continues to be difficult to obtain. Detection of Np with TEM-EELS is hampered by the occurrence of a plural (multiple) scattering event from the more abundant U atoms (U-M5 + U- O4,5), that results in severe overlap on the Np-M5 edge at ∼3665 eV. By examining the energy âgap' between the Np-M5 and Np-M4 edges (184 eV), a method for observing Np independently of the plural scattering event has been established. Clear evidence of Np incorporation into synthetic studtite {(UO2)(O2)(H2O)2](H2O)2} and uranophane {Ca(UO2)2(SiO3OH)2(H2O)5} has been found with TEM-EELS. The EELS technique continues to remain an important tool for examining the potential for transuranic behavior in nuclear waste materials.

In this study, the oxidation state of uranium was investigated with Electron Energy-Loss Spectroscopy on the Transmission Electron Microscope. The oxygen K edge in a series of uranium-bearing phases was investigated and this technique was compared to an alternative method of using the relative intensities of the M4,5- edges of the actinides. Under controlled conditions, it is possible to obtain reliable oxidation state data on uranium solids; however, mixtures of oxidation states in a single compound cannot be distinguished with certainty.

We report results of experimental studies on the behavior of Np during aqueous corrosion of unirradiated Np-bearing U oxides. Np-doped U oxides were reacted in humid air at 90°C and 150°C for several weeks within sealed stainless-steel vessels. Reacted solids were examined by scanning and transmission electron microscopies (SEM and TEM), electron energy-loss spectroscopy (EELS), and X-ray powder diffraction (XRD). Dehydrated schoepite, (UO2)O0.25-z(OH)1.5+2z (0 ≤z ≤0.15), is the predominant U(VI) compound formed in these experiments. Preliminary EELS analysis on crushed grains verify that dehydrated schoepite formed at 150°C contains up to approximately 2 wt.% Np, corresponding to a maximum Np:U molar ratio of approximately 1:40. These are maximum values because the degree to which surface-sorbed Np is present on the grains analyzed is not yet known. Crystalline NpO2 also precipitated during these experiments, and the concentration of Np in dehydrated schoepite may represent the maximum amount of Np that can be incorporated into dehydrated schoepite under the experimental conditions.

Oxidative dissolution of spent UO2 fuel in vapor and dripping groundwater at 90°C occurs via general corrosion at fragment surfaces. Dissolution along fuel-grain boundaries is also evident in samples contacted by the largest volumes of groundwater, and corroded grain boundaries extend at least 20 or 30 grains deep (> 200 μm), possibly throughout mm-sized fragments. Apparent dissolution of fuel along defects that intersect grain boundaries has produced 50 to 200 nm-diameter dissolution pits that penetrate 1–2 μm into each grain, giving rise to a “worm-like” texture along fuel-grain boundaries. Sub-micrometer-sized fuel shards are common between fuel grains and may contribute to the reactive surface area of fuel exposed to groundwater. Outer surfaces of reacted fuel fragments develop a fmne-grained layer of corrosion products adjacent to the fuel (5–15 μm thick). A more coarsely crystalline layer of corrosion products commonly covers the fine-grained layer, the thickness of which varies considerably among samples (from less than 5 μm to greater than 40 μm). The thickest and most porous corrosion layers develop on fuel fragments exposed to the largest volumes of groundwater. Corrosion-layer compositions depend strongly on water flux, with uranyl oxy-hydroxides predominating in vapor experiments, and alkali and alkaline earth uranyl silicates predominating in high drip-rate experiments. Low drip-rate experiments exhibit a complex assemblage of corrosion products, including phases identified in vapor and high drip-rate experiments.

Secondary phases that form during the corrosion of nuclear waste forms may influence both the rate of waste form dissolution and the release of radionuclides [1]. The identification of these phases is critical in developing models for the corrosion behavior of nuclear waste forms. In particular, the secondary uranyl (VI) minerals that form during waste form alteration may control uranium solubility and release of radionuclides incorporated into these phases [2].

The U6+ cation in uranyl minerals is almost always present as a linear (UO2)2+ ion [3]. This uranyl (Ur) ion is coordinated by four, five, or six anions (ϕ) in the equatorial plane resulting in the formation of square (Urϕ4), pentagonal (Urϕ5), and hexagonal (Urϕ6) bipyramids, respectively [3]. These bipyramid polyhedra may polymerize to form complex infinite sheet structures. The linking of Urϕ5 is observed in a number of uranyl minerals formed during waste glass and spent fuel corrosion [2,4], such as weeksite [Na,K(UO2)2(Si205)3*4H2O] and β-uranophane [Ca[(UO2)(SiO3OH)]2*5H2O].

The alteration behavior of UO2 pellets following their reaction under unsaturated drip-test conditions, at 90°C, for time periods of up to 10 years has been examined by solid phase and leachate analyses. Sample reactions were characterized by preferential dissolution of grain boundaries between the original press-sintered UO2 granules comprising the samples, development of a polygonal network of open channels along the intergrain boundaries, and spallation of surface granules that had undergone severe grain boundary corrosion. The development of a dense mat of alteration phases after two years of reaction trapped loose granules, resulting in reduced rates of particulate uranium release. The paragenetic sequence of alteration phases that formed on the present samples was similar to that observed in surficial weathering zones of natural uraninite (UO2) deposits, with alkali and alkaline earth uranyl silicates representing the long-term solubility-limiting phases for uranium in both systems.

Drip tests designed to replicate the synergistic interactions between waste glass, repository groundwater, water vapor, and sensitized 304L stainless steel in the potential Yucca Mountain Repository have been ongoing in our laboratory for over ten years. Results will be presented from three sets of these drip tests: two with actinide-doped glasses, and one with a fully-radioactive glass. Periodic sampling of these tests have revealed trends in actinide release behavior that are consistent with their entrainment in colloidal material when as-cast glass is reacted. Results from vapor hydrated glass show that initially the actinides are completely dissolved in solution, but as the reaction proceeds, the actinides become suspended in solution. Sequential filtering and alpha spectroscopy of colloid-bearing leachate solutions indicate that more than 80% of the plutonium and americium are bound to particles that are captured by a 0.1 μm filter, while less than 10% of the neptunium is stopped by a 0.1 μm filter. Analytical transmission electron microscopy has been used to examine particles from leachate solutions and to identify several actinide-bearing phases which are responsible for the majority of actinide release during glass corrosion.

The weathering of glass is reviewed by examining processes that affect the reaction of commercial, historical, natural, and nuclear waste glass under conditions of contact with humid air and slowly dripping water, which may lead to immersion in nearly static solution. Radionuclide release data from weathered glass under conditions that may exist in an unsaturated environment are presented and compared to release under standard leaching conditions. While the comparison between the release under weathering and leaching conditions is not exact, due to variability of reaction in humid air, evidence is presented of radionuclide release under a variety of conditions. These results suggest that both the amount and form of radionuclide release can be affected by the weathering of glass.