Characterization of Material from Wells 299-W10-35 (C7573) and 299-W14-74 (C7024)

The objective of this work was to characterize material accumulating on wells 299-W10-35 (C7573) and 299-W14-74 (C7024) to determine the type of material (i.e., chemical or biological) and, if the material is biological, to identify the microorganisms present

Epsilon Metal Summary Report FY 2011

The Epsilon-metal ({var_epsilon}-metal) phase was selected in FY 2009 as a potential waste form to for immobilizing the noble metals found in the undissolved solids + aqueous stream, and the soluble Tc from ion-exchange process, each resulting from proposed aqueous reprocessing. {var_epsilon}-metal phase is observed in used nuclear fuel and the natural reactors of Oklobono in Gabon, where the long-term corrosion behavior was demonstrated. This makes {var_epsilon}-metal a very attractive waste form. Last fiscal year, {var_epsilon}-metal was successfully fabricated by combining the five-metals, Mo, Ru, Rh, Pd and Re (surrogate for Tc), into pellets followed by consolidation with an arc melter. The arc melter produced fully dense samples with the epsilon structure. However, some chemistry differences were observed in the microstructure that resulted in regions rich in Re and Mo, and others rich in Pd, while Ru and Rh remained fairly constant throughout. This year, thermal stability (air), and corrosion testing of the samples fabricated by arc melting were the main focus for experimental work. Thermal stability was measured with a differential scanning calorimeter - thermogravimetric analyzer, by both ramp heating as well as step heating. There is clear evidence during the ramp heating experiment of an exothermic event + a weight loss peak both beginning at {approx}700 C. Step heating showed an oxidation event at {approx}690 C with minimal weight gain that occurs just before the weight loss event at 700 C. The conclusion being that the e-metal begins to oxidize and then become volatile. These findings are useful for considering the effects of voloxidation process. Three different pellets were subjected to electrochemical testing to study the corrosion behavior of the epsilon-metal phase in various conditions, namely acidic, basic, saline, and inert. Test was done according to an interim procedure developed for the alloy metal waste form. First an open circuit potential was measured, followed by linear polarization sweeps. The linear polarization sweep range was the Tafel equation was fit to the linear polarization sweep data to determine the corrosion rate of each pellet in each test solution. The average calculated corrosion rates of the three pellets according to solution conditions were: -1.91 x 10{sup -4} mm/yr (0.001 M NaOH), -1.48 x 10{sup -3} mm/yr (0.01 M NaCl), -8.77 x 10{sup -4} mm/yr (0.001 M H{sub 2}SO{sub 4}), -2.09 x 10{sup -3} mm/yr (0.001 M NaOH + 0.01 M NaCl), and -1.54 x 10{sup -3} mm/yr (0.001 M H{sub 2}SO{sub 4} + 0.01 M NaCl). Three single-pass flow through (SPFT) test were conducted at a flow rate of 10 ml/day, at 90 C, and pH of 2.5, 7.0, and 9.0 for up to 322 days. Results of the tests indicate that dissolution rates were 5 x 10{sup -4} g m{sup 2} d{sup -1} at pH 9.0, 1.2 x 10{sup -4} g m{sup -2} d{sup -1} at pH 7.0, and 2 x 10{sup -4} g m{sup -2} d{sup -1} at pH 2.5. The sample used for the pH 7.0 SPFT test contains extra Re compared to samples used for the other two SPFT test, which came from a single pellet. The corrosion data measured this year indicate that the {var_epsilon}-metal phase is chemically durable. The two chemically different phases, but structurally the same, behave differently during dissolution according to the microstructure changes observed in both the electrochemical and in SPFT test. Characterization of the test specimens after testing suggests that the dissolution is complex and involves oxidative dissolution followed by precipitation of both oxide and metallic phases. These data suggest that the dissolution in the electrochemical and SPFT tests is different; a process that needs further investigation

Effect of Concrete Wasteform Properties on Radionuclide Migration

The objective of this investigation was to initiate numerous sets of concrete-soil half-cell tests to quantify 1) diffusion of I and Tc from concrete into uncontaminated soil after 1 and 2 years, 2) I and Re (set 1) and Tc (set 2) diffusion from fractured concrete into uncontaminated soil, and 3) evaluate the moisture distribution profile within the sediment half-cell. These half-cells will be section in FY2009 and FY2010. Additionally, 1) concrete-soil half-cells initiated during FY2007 using fractured prepared with and without metallic iron, half of which were carbonated using carbonated, were sectioned to evaluate the diffusion of I and Re in the concrete part of the half-cell under unsaturated conditions (4%, 7%, and 15% by wt moisture content), 2) concrete-soil half cells containing Tc were sectioned to measure the diffusion profile in the soil half-cell unsaturated conditions (4%, 7%, and 15% by wt moisture content), and 3) solubility measurements of uranium solid phases were completed under concrete porewater conditions. The results of these tests are presented

Effect of Concrete Waste Form Properties on Radionuclide Migration

Assessing long-term performance of Category 3 waste cement grouts for radionuclide encasement requires knowledge of the radionuclide-cement interactions and mechanisms of retention (i.e., sorption or precipitation) the mechanism of contaminant release, the significance of contaminant release pathways, how waste form performance is affected by the full range of environmental conditions within the disposal facility, the process of waste form aging under conditions that are representative of processes occurring in response to changing environmental conditions within the disposal facility, the effect of waste form aging on chemical, physical, and radiological properties and the associated impact on contaminant release. This knowledge will enable accurate prediction of radionuclide fate when the waste forms come in contact with groundwater. Numerous sets of tests were initiated in fiscal years (FY) 2006-2009 to evaluate (1) diffusion of iodine (I) and technetium (Tc) from concrete into uncontaminated soil after 1 and 2 years, (2) I and rhenium (Re) diffusion from contaminated soil into fractured concrete, (3) I and Re (set 1) and Tc (set 2) diffusion from fractured concrete into uncontaminated soil, (4) evaluate the moisture distribution profile within the sediment half-cell, (5) the reactivity and speciation of uranium (VI) (U(VI)) compounds in concrete porewaters, (6) the rate of dissolution of concrete monoliths, and (7) the diffusion of simulated tank waste into concrete

300 Area Treatability Test: Laboratory Development of Polyphosphate Remediation Technology for In Situ Treatment of Uranium Contamination in the Vadose Zone and Capillary Fringe

This report presents results from bench-scale treatability studies conducted under site-specific conditions to optimize the polyphosphate amendment for implementation of a field-scale technology demonstration to stabilize uranium within the 300 Area vadose and smear zones of the Hanford Site. The general treatability testing approach consisted of conducting studies with site sediment and under site conditions, to develop an effective chemical formulation and infiltration approach for the polyphosphate amendment under site conditions. Laboratory-scale dynamic column tests were used to 1) quantify the retardation of polyphosphate and its degradation products as a function of water content, 2) determine the rate of polyphosphate degradation under unsaturated conditions, 3) develop an understanding of the mechanism of autunite formation via the reaction of solid phase calcite-bound uranium and aqueous polyphosphate remediation technology, 4) develop an understanding of the transformation mechanism, the identity of secondary phases, and the kinetics of the reaction between uranyl-carbonate and -silicate minerals with the polyphosphate remedy under solubility-limiting conditions, and 5) quantify the extent and rate of uranium released and immobilized based on the infiltration rate of the polyphosphate remedy and the effect of and periodic wet-dry cycling on the efficacy of polyphosphate remediation for uranium in the vadose zone and smear zone

Characterization of Material from Wells 299-W10-35 (C7573) and 299-W14-74 (C7024)

The objective of this work was to characterize material accumulating on wells 299-W10-35 (C7573) and 299-W14-74 (C7024) to determine the type of material (i.e., chemical or biological) and, if the material is biological, to identify the microorganisms present

Summary of Uranium Solubility Studies in Concrete Waste Forms and Vadose Zone Environments

One of the methods being considered for safely disposing of Category 3 low-level radioactive wastes is to encase the waste in concrete. Concrete encasement would contain and isolate the waste packages from the hydrologic environment and act as an intrusion barrier. The current plan for waste isolation consists of stacking low-level waste packages on a trench floor, surrounding the stacks with reinforced steel, and encasing these packages in concrete. These concrete-encased waste stacks are expected to vary in size with maximum dimensions of 6.4 m long, 2.7 m wide, and 4 m high. The waste stacks are expected to have a surrounding minimum thickness of 15 cm of concrete encasement. These concrete-encased waste packages are expected to withstand environmental exposure (solar radiation, temperature variations, and precipitation) until an interim soil cover or permanent closure cover is installed and to remain largely intact thereafter. Any failure of concrete encasement may result in water intrusion and consequent mobilization of radionuclides from the waste packages. This report presents the results of investigations elucidating the uranium mineral phases controlling the long-term fate of uranium within concrete waste forms and the solubility of these phases in concrete pore waters and alkaline, circum-neutral vadose zone environments

Summary of Uranium Solubility Studies in Concrete Waste Forms and Vadose Zone Environments

One of the methods being considered for safely disposing of Category 3 low-level radioactive wastes is to encase the waste in concrete. Concrete encasement would contain and isolate the waste packages from the hydrologic environment and act as an intrusion barrier. The current plan for waste isolation consists of stacking low-level waste packages on a trench floor, surrounding the stacks with reinforced steel, and encasing these packages in concrete. These concrete-encased waste stacks are expected to vary in size with maximum dimensions of 6.4 m long, 2.7 m wide, and 4 m high. The waste stacks are expected to have a surrounding minimum thickness of 15 cm of concrete encasement. These concrete-encased waste packages are expected to withstand environmental exposure (solar radiation, temperature variations, and precipitation) until an interim soil cover or permanent closure cover is installed and to remain largely intact thereafter. Any failure of concrete encasement may result in water intrusion and consequent mobilization of radionuclides from the waste packages. This report presents the results of investigations elucidating the uranium mineral phases controlling the long-term fate of uranium within concrete waste forms and the solubility of these phases in concrete pore waters and alkaline, circum-neutral vadose zone environments

Radionuclide Retention in Concrete Waste Forms

Assessing long-term performance of Category 3 waste cement grouts for radionuclide encasement requires knowledge of the radionuclide-cement interactions and mechanisms of retention (i.e., sorption or precipitation); the mechanism of contaminant release; the significance of contaminant release pathways; how waste form performance is affected by the full range of environmental conditions within the disposal facility; the process of waste form aging under conditions that are representative of processes occurring in response to changing environmental conditions within the disposal facility; the effect of waste form aging on chemical, physical, and radiological properties; and the associated impact on contaminant release. This knowledge will enable accurate prediction of radionuclide fate when the waste forms come in contact with groundwater. The information presented in the report provides data that 1) quantify radionuclide retention within concrete waste form materials similar to those used to encapsulate waste in the Low-Level Waste Burial Grounds (LLBG); 2) measure the effect of concrete waste form properties likely to influence radionuclide migration; and 3) quantify the stability of uranium-bearing solid phases of limited solubility in concrete