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Predicting Chemical and Biochemical Properties Using the Abraham General Solvation Model

Description: Several studies were done to illustrate the versatillity of the Abraham model in mathematically describing the various solute-solvent interactions found in a wide range of different chemical and biological systems. The first study focused on using the solvation model to construct mathematical correlations describing the minimum inhibitory concentration of organic compounds for growth inhibition towards the three bacterial strains Porphyromonas gingivalis, Selenomonas artemidis, and Streptococcus sobrinus. The next several studies expand the practicallity of the Abraham model by predicting free energies of partition in chemical systems. The free energy studies expand the use of the Abraham model to other temperatures and properties by developing correlations for the enthalpies of solvation of gaseous solutes of various compounds dissolved in water, 1-octanol, hexane, heptane, hexadecane, cyclohexane, benzene, toluene, carbon tetrachloride, chloroform, methanol, ethanol, 1-butanol, propylene carbonate, dimethyl sulfoxide, 1,2-dichloroethane, N,N-dimethylformamide, tert-butanol, dibutyl ether, ethyl acetate, acetonitrile, and acetone. Also, a generic equation for linear alkanes is created for use when individual datasets are small. The prediction of enthalpies of solvation is furthered by modifying the Abraham model so that experimental data measured at different temperatures can be included into a single correlation expression. The temperature dependence is directly included in the model by separating each coefficient into an enthalpic and entropic component. Specifically, the final study describes the effects of temperature on the sorption coefficients of organic gases onto humic acid. The derived predicted values for each research study show a good correlation with experimental values.
Date: May 2009
Creator: Mintz, Christina
Partner: UNT Libraries

Comments on Solid-Liquid Phase Equilibrium and Phase Diagram for the Ternary o-Nitrobenzoic Acid + m-Nitrobenzoic Acid + Ethanol System

Description: Article commenting on a paper published in 2008 in the Journal of Chemical and Engineering Data discussing solid-liquid phase equilibrium and phase diagram for the ternary o-nitrobenzoic acid + m-nitrobenzoic acid + ethanol system.
Date: June 5, 2009
Creator: Acree, William E. (William Eugene)
Partner: UNT College of Arts and Sciences

Linear Free Energy Relationship Correlations for Room Temperature Ionic Liquids: Revised Cation-Specific and Anion-Specific Equation Coefficients for Predictive Applications Covering a Much Larger Area of Chemical Space

Description: Article discussing linear free energy relationship correlations for room temperature ionic liquids and revised cation-specific and anion-specific equation coefficients for predictive applications covering a much larger area of chemical space.
Date: March 23, 2009
Creator: Sprunger, Laura M.; Gibbs, Jennifer; Proctor, Amy; Acree, William E. (William Eugene); Abraham, M. H. (Michael H.); Meng, Yunjing et al.
Partner: UNT College of Arts and Sciences

Highly mismatched crystalline and amorphous GaN(1-x)As(x) alloys in the whole composition range

Description: Alloying is a commonly accepted method to tailor properties of semiconductor materials for specific applications. Only a limited number of semiconductor alloys can be easily synthesized in the full composition range. Such alloys are, in general, formed of component elements that are well matched in terms of ionicity, atom size, and electronegativity. In contrast there is a broad class of potential semiconductor alloys formed of component materials with distinctly different properties. In most instances these mismatched alloys are immiscible under standard growth conditions. Here we report on the properties of GaN1-xAsx, a highly mismatched, immiscible alloy system that was successfully synthesized in the whole composition range using a nonequilibrium low temperature molecular beam epitaxy technique. The alloys are amorphous in the composition range of 0.17<x<0.75 and crystalline outside this region. The amorphous films have smooth morphology, homogeneous composition, and sharp, well defined optical absorption edges. The band gap energy varies in a broad energy range from ~;;3.4 eV in GaN to ~;;0.8 eV at x~;;0.85. The reduction in the band gap can be attributed primarily to the downward movement of the conduction band for alloys with x>0.2, and to the upward movement of the valence band for alloys with x<0.2. The unique features of the band structure offer an opportunity of using GaN1-xAsx alloys for various types of solar power conversion devices.
Date: August 29, 2009
Creator: Yu, K. M.; Novikov, S. V.; Broesler, R.; Demchenko, I. N.; Denlinger, J. D.; Liliental-Weber, Z. et al.
Partner: UNT Libraries Government Documents Department

Stability of uranium incorporated into Fe(hydr)oxides under fluctuating redox conditions

Description: Reaction pathways resulting in uranium bearing solids that are stable (i.e., having limited solubility) under both aerobic and anaerobic conditions will limit dissolved concentrations and migration of this toxin. Here we examine the sorption mechanism and propensity for release of uranium reacted with Fe (hydr)oxides under cyclic oxidizing and reducing conditions. Upon reaction of ferrihydrite with Fe(II) under conditions where aqueous Ca-UO{sub 2}-CO{sub 3} species predominate (3 mM Ca and 3.8 mM CO{sub 3}-total), dissolved uranium concentrations decrease from 0.16 mM to below detection limit (BDL) after 5 to 15 d, depending on the Fe(II) concentration. In systems undergoing 3 successive redox cycles (15 d of reduction followed by 5 d of oxidation) and a pulsed decrease to 0.15 mM CO{sub 3}-total, dissolved uranium concentrations varied depending on the Fe(II) concentration during the initial and subsequent reduction phases - U concentrations resulting during the oxic 'rebound' varied inversely with the Fe(II) concentration during the reduction cycle. Uranium removed from solution remains in the oxidized form and is found both adsorbed on and incorporated into the structure of newly formed goethite and magnetite. Our 15 results reveal that the fate of uranium is dependent on anaerobic/aerobic conditions, aqueous uranium speciation, and the fate of iron.
Date: April 1, 2009
Creator: Stewart, B.D.; Nico, P.S. & Fendorf, S.
Partner: UNT Libraries Government Documents Department

A Modern Interpretation of the Barney Diagram for Aluminum Solubility in Tank Waste

Description: Experimental and modeling studies of aluminum solubility in Hanford tank waste have been developed and refined for many years in efforts to resolve new issues or develop waste treatment flowsheets. The earliest of these studies was conducted by G. Scott Barney, who performed solubility studies in highly concentrated electrolyte solutions to support evaporator campaign flowsheets in the 1970's. The 'Barney Diagram', a term still widely used at Hanford today, suggested gibbsite ({gamma}-Al(OH){sub 3}) was much more soluble in tank waste than in simple sodium hydroxide solutions. These results, which were highly surprising at the time, continue to be applied to new situations where aluminum solubility in tank waste is of interest. Here, we review the history and provide a modern explanation for the large gibbsite solubility observed by Barney, an explanation based on basic research that has been performed and published in the last 30 years. This explanation has both thermodynamic and kinetic aspects. Thermodynamically, saturated salt solutions stabilize soluble aluminate species that are minor components in simple sodium hydroxide solutions. These species are the aluminate dimer and the sodium-aluminate ion-pair. Ion-pairs must be present in the Barney simulants because calculations showed that there was insufficient space between the highly concentrated ions for a water molecule. Thus, most of the ions in the simulants have to be ion-paired. Kinetics likely played a role as well. The simulants were incubated for four to seven days, and more recent data indicate that this was unlikely sufficient time to achieve equilibrium from supersaturation. These results allow us to evaluate applications of the Barney results to current and future tank waste issues or flowsheets.
Date: December 16, 2009
Creator: Reynolds, J. G. & Reynolds, D. A.
Partner: UNT Libraries Government Documents Department

Solubility of 7-Chloro-2-methylamino-5-phenyl-3H-1,4-benzodiazepine-4-oxide, 7-Chloro-1,3-dihydro-1-methyl-5-phenyl-2H-1,4-benzodiazepin-2-one, and 7-Chloro-5-(2-chlorophenyl)-3-hydroxy-1,3-dihydro-1,4-benzodiazepin-2-one in (Propane-1,2-diol + Water) at a Temperature of 303.2 K

Description: Article on 7-chloro-2-methylamino-5-phenyl-3H-1, 4-benzodiazepine-4-oxide, 7-chloro-1, 3-dihydro-1-methyl-5-phenyl-2H-1, 4-benzodiazepin-2-one, and 7-chloro-5-(2-chlorophenyl)-3-hydroxy-1, 3-dihydro-1, 4-benzodiazepin-2-one in (propane-1, 2-diol + water) at a temperature of 303.2 K.
Date: July 31, 2009
Creator: Jouyban, Abolghasem; Shokri, Javad; Barzegar-Jalali, Mohammad; Hassanzadeh, Davoud; Acree, William E. (William Eugene); Ghafourian, Taravat et al.
Partner: UNT College of Arts and Sciences

Solubility of 5-(2-Chlorophenyl)-7-nitro-1,3-dihydro-1,4-benzodiazepin-2-one, 7-Chloro-1-methyl-5-phenyl-3H-1,4-benzodiazepin-2-one, and 6-(2,3-Dichlorophenyl)-1,2,4-triazine-3,5-diamine in the Mixtures of Poly(ethylene glycol) 600, Ethanol, and Water at a Temperature of 298.2 K

Description: Article on the solubility of 5-(2-chlorophenyl)-7-nitro-1, 3-dihydro-1, 4-benzodiazepin-2-one, 7-chloro-1-methyl-5-phenyl-3H-1, 4-benzodiazepin-2-one, and 6-(2,3-dichlorophenyl)-1, 2, 4,-triazine-3, 5-diamine in the mixtures of poly(ethylene glycol) 600, ethanol, and water at temperature of 298.2 K.
Date: October 23, 2009
Creator: Soltanpour, Shahla; Acree, William E. (William Eugene) & Jouyban, Abolghasem
Partner: UNT College of Arts and Sciences

Actinide (III) solubility in WIPP Brine: data summary and recommendations

Description: The solubility of actinides in the +3 oxidation state is an important input into the Waste Isolation Pilot Plant (WIPP) performance assessment (PA) models that calculate potential actinide release from the WIPP repository. In this context, the solubility of neodymium(III) was determined as a function of pH, carbonate concentration, and WIPP brine composition. Additionally, we conducted a literature review on the solubility of +3 actinides under WIPP-related conditions. Neodymium(III) was used as a redox-invariant analog for the +3 oxidation state of americium and plutonium, which is the oxidation state that accounts for over 90% of the potential release from the WIPP through the dissolved brine release (DBR) mechanism, based on current WIPP performance assessment assumptions. These solubility data extend past studies to brine compositions that are more WIPP-relevant and cover a broader range of experimental conditions than past studies.
Date: September 1, 2009
Creator: Borkowski, Marian; Lucchini, Jean-Francois; Richmann, Michael K. & Reed, Donald T.
Partner: UNT Libraries Government Documents Department

300 Area Uranium Stabilization Through Polyphosphate Injection: Final Report

Description: The objective of the treatability test was to evaluate the efficacy of using polyphosphate injections to treat uranium-contaminated groundwater in situ. A test site consisting of an injection well and 15 monitoring wells was installed in the 300 Area near the process trenches that had previously received uranium-bearing effluents. This report summarizes the work on the polyphosphate injection project, including bench-scale laboratory studies, a field injection test, and the subsequent analysis and interpretation of the results. Previous laboratory tests have demonstrated that when a soluble form of polyphosphate is injected into uranium-bearing saturated porous media, immobilization of uranium occurs due to formation of an insoluble uranyl phosphate, autunite [Ca(UO2)2(PO4)2•nH2O]. These tests were conducted at conditions expected for the aquifer and used Hanford soils and groundwater containing very low concentrations of uranium (10-6 M). Because autunite sequesters uranium in the oxidized form U(VI) rather than forcing reduction to U(IV), the possibility of re-oxidation and subsequent re-mobilization is negated. Extensive testing demonstrated the very low solubility and slow dissolution kinetics of autunite. In addition to autunite, excess phosphorous may result in apatite mineral formation, which provides a long-term source of treatment capacity. Phosphate arrival response data indicate that, under site conditions, the polyphosphate amendment could be effectively distributed over a relatively large lateral extent, with wells located at a radial distance of 23 m (75 ft) reaching from between 40% and 60% of the injection concentration. Given these phosphate transport characteristics, direct treatment of uranium through the formation of uranyl-phosphate mineral phases (i.e., autunite) could likely be effectively implemented at full field scale. However, formation of calcium-phosphate mineral phases using the selected three-phase approach was problematic. Although amendment arrival response data indicate some degree of overlap between the reactive species and thus potential for the formation of calcium-phosphate mineral phases (i.e., apatite ...
Date: June 30, 2009
Creator: Vermeul, Vincent R.; Bjornstad, Bruce N.; Fritz, Brad G.; Fruchter, Jonathan S.; Mackley, Rob D.; Newcomer, Darrell R. et al.
Partner: UNT Libraries Government Documents Department

EFRT M-12 Issue Resolution: Solids Washing

Description: Pacific Northwest National Laboratory (PNNL) has been tasked by Bechtel National Inc. (BNI) on the River Protection Project-Hanford Tank Waste Treatment and Immobilization Plant (RPP-WTP) project to perform research and development activities to resolve technical issues identified for the Pretreatment Facility (PTF). The Pretreatment Engineering Platform (PEP) was designed, constructed, and operated as part of a plan to respond to issue M12, “Undemonstrated Leaching Processes.” The PEP is a 1/4.5-scale test platform designed to simulate the WTP pretreatment caustic leaching, oxidative leaching, ultrafiltration solids concentration, and slurry washing processes. The PEP replicates the WTP leaching processes using prototypic equipment and control strategies. Two operating scenarios were evaluated for the ultrafiltration process (UFP) and leaching operations. The first scenario has caustic leaching performed in the UFP-VSL-T01A/B ultrafiltration feed vessels, identified as Integrated Test A. The second scenario has caustic leaching conducted in the UFP-VSL-T02A ultrafiltration feed preparation vessel, identified as Integrated Test B. Washing operations in PEP Integrated Tests A and B were conducted successfully as per the approved run sheets. However, various minor instrumental problems occurred, and some of the process conditions specified in the run sheet were not met during the wash operations, such as filter-loop flow-rate targets not being met. Five analytes were selected based on full solubility and monitored in the post-caustic-leach wash as successful indicators of washing efficiency. These were aluminum, sulfate, nitrate, nitrite, and free hydroxide. Other analytes, including sodium, oxalate, phosphate, and total dissolved solids, showed indications of changing solubility; therefore, they were unsuitable for monitoring washing efficiency. In the post-oxidative-leach wash, two analytes with full solubility were selected as suitable indicators of washing efficiency. These were chromium and oxalate. Other analytes, including sodium, manganese, nitrate, and total dissolved solids, showed indications of changing solubility; therefore, they were unsuitable for monitoring washing efficiency. An ...
Date: August 14, 2009
Creator: Baldwin, David L.; Schonewill, Philip P.; Toth, James J.; Huckaby, James L.; Eslinger, Paul W.; Hanson, Brady D. et al.
Partner: UNT Libraries Government Documents Department

Filtration and Leach Testing for PUREX Cladding Sludge and REDOX Cladding Sludge Actual Waste Sample Composites

Description: A testing program evaluating actual tank waste was developed in response to Task 4 from the M-12 External Flowsheet Review Team (EFRT) issue response plan (Barnes and Voke 2006). The test program was subdivided into logical increments. The bulk water-insoluble solid wastes that are anticipated to be delivered to the Hanford Waste Treatment and Immobilization Plant (WTP) were identified according to type such that the actual waste testing could be targeted to the relevant categories. Under test plan TP RPP WTP 467 (Fiskum et al. 2007), eight broad waste groupings were defined. Samples available from the 222S archive were identified and obtained for testing. Under this test plan, a waste testing program was implemented that included: • Homogenizing the archive samples by group as defined in the test plan. • Characterizing the homogenized sample groups. • Performing parametric leaching testing on each group for compounds of interest. • Performing bench-top filtration/leaching tests in the hot cell for each group to simulate filtration and leaching activities if they occurred in the UFP2 vessel of the WTP Pretreatment Facility. This report focuses on a filtration/leaching test performed using two of the eight waste composite samples. The sample groups examined in this report were the plutonium-uranium extraction (PUREX) cladding waste sludge (Group 3, or CWP) and reduction-oxidation (REDOX) cladding waste sludge (Group 4, or CWR). Both the Group 3 and 4 waste composites were anticipated to be high in gibbsite, thus requiring caustic leaching. WTP RPT 167 (Snow et al. 2008) describes the homogenization, characterization, and parametric leaching activities before benchtop filtration/leaching testing of these two waste groups. Characterization and initial parametric data in that report were used to plan a single filtration/leaching test using a blend of both wastes. The test focused on filtration testing of the waste and caustic leaching for ...
Date: March 2, 2009
Creator: Shimskey, Rick W.; Billing, Justin M.; Buck, Edgar C.; Casella, Amanda J.; Crum, Jarrod V.; Daniel, Richard C. et al.
Partner: UNT Libraries Government Documents Department

Fate of Contaminants in Contact with West Valley Grouts

Description: The objective of the work described here is to determine to what extent a variety of contaminants, including fission products, actinides, and RCRA elements are sequestered by the two grout formulations. The conceptual model for this study is as follows: a large mass of grout having been poured into a high-level waste tank is in the process of aging and weathering for thousands of years. The waste remaining in the tank will contain radionuclides and other contaminants, much of which will adhere to tank walls and internal structures. The grout will encapsulate the contaminants. Initially the grout will be well sequestered, but over time rainwater and groundwater will gain access to it. Ultimately, the grout/waste environment will be an open system. In this condition water will move through the grout, exposing it to O{sub 2} and CO{sub 2} from the air and HCO{sub 3}{sup -} from the groundwater. Thus we are considering an oxic environment containing HCO{sub 3}{sup -}. Initially the solubility of many contaminants, but not all, will be constrained by chemistry dominated by the grout, primarily by the high pH, around 11.8. This is controlled and buffered by the portland cement and blast furnace slag components of the grout, which by themselves maintain a solution pH of about 12.5. Slowly the pH will diminish as Ca(OH){sub 2} and KOH dissolve, are carried away by water, and CaCO{sub 3} forms. As these conditions develop, the behavior of these elements comes into question. In our conceptual model, although the grout is formulated to provide some reducing capacity, in order to be conservative this mechanism is not considered. In addition to solubility constraints imposed by pH, the various contaminants may be incorporated into a variety of solid phases. Some may be incorporated into newly forming compounds as the grout sets and ...
Date: July 1, 2009
Creator: Fuhrmann,M. & Gillow, J.
Partner: UNT Libraries Government Documents Department

Implementation of equilibrium aqueous speciation and solubility (EQ3 type) calculations into Cantera for electrolyte solutions.

Description: In this report, we summarize our work on developing a production level capability for modeling brine thermodynamic properties using the open-source code Cantera. This implementation into Cantera allows for the application of chemical thermodynamics to describe the interactions between a solid and an electrolyte solution at chemical equilibrium. The formulations to evaluate the thermodynamic properties of electrolytes are based on Pitzer's model to calculate molality-based activity coefficients using a real equation-of-state (EoS) for water. In addition, the thermodynamic properties of solutes at elevated temperature and pressures are computed using the revised Helgeson-Kirkham-Flowers (HKF) EoS for ionic and neutral aqueous species. The thermodynamic data parameters for the Pitzer formulation and HKF EoS are from the thermodynamic database compilation developed for the Yucca Mountain Project (YMP) used with the computer code EQ3/6. We describe the adopted equations and their implementation within Cantera and also provide several validated examples relevant to the calculations of extensive properties of electrolyte solutions.
Date: June 1, 2009
Creator: Moffat, Harry K. & Jove-Colon, Carlos F.
Partner: UNT Libraries Government Documents Department

Radionuclide Partitioning in an Underground Nuclear Test Cavity

Description: In 2004, a borehole was drilled into the 1983 Chancellor underground nuclear test cavity to investigate the distribution of radionuclides within the cavity. Sidewall core samples were collected from a range of depths within the re-entry hole and two sidetrack holes. Upon completion of drilling, casing was installed and a submersible pump was used to collect groundwater samples. Test debris and groundwater samples were analyzed for a variety of radionuclides including the fission products {sup 99}Tc, {sup 125}Sb, {sup 129}I, {sup 137}Cs, and {sup 155}Eu, the activation products {sup 60}Co, {sup 152}Eu, and {sup 154}Eu, and the actinides U, Pu, and Am. In addition, the physical and bulk chemical properties of the test debris were characterized using Scanning Electron Microscopy (SEM) and Electron Microprobe measurements. Analytical results were used to evaluate the partitioning of radionuclides between the melt glass, rubble, and groundwater phases in the Chancellor test cavity. Three comparative approaches were used to calculate partitioning values, though each method could not be applied to every nuclide. These approaches are based on: (1) the average Area 19 inventory from Bowen et al. (2001); (2) melt glass, rubble, and groundwater mass estimates from Zhao et al. (2008); and (3) fission product mass yield data from England and Rider (1994). The U and Pu analyses of the test debris are classified and partitioning estimates for these elements were calculated directly from the classified Miller et al. (2002) inventory for the Chancellor test. The partitioning results from this study were compared to partitioning data that were previously published by the IAEA (1998). Predictions of radionuclide distributions from the two studies are in agreement for a majority of the nuclides under consideration. Substantial differences were noted in the partitioning values for {sup 99}Tc, {sup 125}Sb, {sup 129}I, and uranium. These differences are attributable to ...
Date: January 9, 2009
Creator: Rose, T P; Hu, Q; Zhao, P; Conrado, C L; Dickerson, R; Eaton, G F et al.
Partner: UNT Libraries Government Documents Department

DEVELOPMENT OF GLASS COMPOSITIONS TO IMMOBILIZE ALKALI, ALKALINE EARTH, LANTHANIDE AND TRANSITION METAL FISSION PRODUCTS FROM NUCLEAR FUEL REPROCESSING

Description: The Advanced Fuel Cycle Initiative (AFCI) waste management strategy revolves around specific treatment of individual or groups of separated waste streams. A goal for the separations processes is to efficiently manage the waste to be dispositioned as high level radioactive waste. The Advanced Fuel Cycle Initiative (AFCI) baseline technology for immobilization of the lanthanide (Ln) and transition metal fission product (TM) wastes is vitrification into a borosilicate glass. A current interest is to evaluate the feasibility of vitrifying combined waste streams to most cost effectively immobilize the wastes resulting from aqueous fuel reprocessing. Studies showed that high waste loadings are achievable for the Ln only (Option 1) stream. Waste loadings in excess of 60 wt % (on a calcined oxide basis) were demonstrated via a lanthanide borosilicate (LaBS) glass. The resulting glasses had excellent relative durability as determined by the Product Consistency Test (PCT). For a combined Ln and TM waste stream glass (Option 2), noble metal solubility was found to limit waste loading. However, the measured PCT normalized elemental releases for this glass were at least an order of magnitude below that of Environmental Assessment (EA) glass. Current efforts to evaluate the feasibility of vitrifying combined Ln, TM, alkali (Cs is the primary radionuclide of concern) and alkaline earth (Sr is the primary radionuclide of concern) wastes (Option 3) have shown that these approaches are feasible. However, waste loading limitations with respect to heat load (Cs/Sr loading), molybdenum solubility and/or noble metal solubility will likely be realized and must be considered in determining the cost effectiveness of these approaches.
Date: June 24, 2009
Creator: Marra, J. & Billings, A.
Partner: UNT Libraries Government Documents Department

Novel imaging techniques, integrated with mineralogical, geochemical and microbiological characterizations to determine the biogeochemical controls on technetium mobility in FRC sediments

Description: The objective of this research program was to take a highly multidisciplinary approach to define the biogeochemical factors that control technetium (Tc) mobility in FRC sediments. The aim was to use batch and column studies to probe the biogeochemical conditions that control the mobility of Tc at the FRC. Background sediment samples from Area 2 (pH 6.5, low nitrate, low {sup 99}Tc) and Area 3 (pH 3.5, high nitrate, relatively high {sup 99}Tc) of the FRC were selected (http://www.esd.ornl.gov/nabirfrc). For the batch experiments, sediments were mixed with simulated groundwater, modeled on chemical constituents of FRC waters and supplemented with {sup 99}Tc(VII), both with and without added electron donor (acetate). The solubility of the Tc was monitored, alongside other biogeochemical markers (nitrate, nitrite, Fe(II), sulfate, acetate, pH, Eh) as the 'microcosms' aged. At key points, the microbial communities were also profiled using both cultivation-dependent and molecular techniques, and results correlated with the geochemical conditions in the sediments. The mineral phases present in the sediments were also characterized, and the solid phase associations of the Tc determined using sequential extraction and synchrotron techniques. In addition to the batch sediment experiments, where discrete microbial communities with the potential to reduce and precipitate {sup 99}Tc will be separated in time, we also developed column experiments where biogeochemical processes were spatially separated. Experiments were conducted both with and without amendments proposed to stimulate radionuclide immobilization (e.g. the addition of acetate as an electron donor for metal reduction), and were also planned with and without competing anions at high concentration (e.g. nitrate, with columns containing Area 3 sediments). When the columns had stabilized, as determined by chemical analysis of the effluents, we used a spike of the short-lived gamma emitter {sup 99m}Tc (50-200 MBq; half life 6 hours) and its mobility was monitored using a {gamma}-camera. ...
Date: February 3, 2009
Creator: Lloyd, Jonathan R.
Partner: UNT Libraries Government Documents Department

SLUDGE CHARACTERIZATION AND SRAT SIMULATIONS USING A NITRITE-FREE SLUDGE SIMULANT

Description: Understanding catalytic hydrogen generation is fundamental to the safe operation of the Defense Waste Processing Facility (DWPF) Chemical Process Cell (CPC). Two Sludge Receipt and Adjustment Tank (SRAT) simulations were completed at the Aiken County Technology Laboratory (ACTL) of the Savannah River National Laboratory (SRNL) using a nitrite-free starting simulant. One simulation was trimmed with Rh and Hg and the other with Ru and Hg. The two noble metals were trimmed at the upper end of the recent Rh-Ru-Hg study. Mercury was trimmed at 1.5 wt% in the total solids. Excess acid comparable in quantity to that in the recent Rh-Ru-Hg matrix study was used. In spite of the favorable conditions for hydrogen generation, virtually no hydrogen production was observed during either SRAT simulation. The Rh test result confirmed the postulated significance of nitrite ion to the catalytic reactions producing hydrogen in CPC testing with normal DWPF sludge simulants. As for Ru, however, previous testing has shown that Ru activated for hydrogen generation only after nitrite destruction. Therefore, Ru could have potentially been catalytically active from the start of the nitrite-free SRAT test, but no such activity was seen. The nitrite-free Ru test result suggests that the intermediate form detected in the bead-frit melter feed preparation Ru solubility profiles was some form of nitro-Ru complex. The nitro-Ru complex is apparently not catalytically active for hydrogen generation but is a precursor to the catalytically active form (presumably a different complex not involving nitrite ligands). Removing nitrite ion from the system prevented the Ru catalyst precursor from forming and consequently blocked formation of the catalytically active form. These results, along with the results of a simulation in which sodium nitrite was metered into the SRAT to prevent ligand substitution reactions that occur during nitrite destruction from occurring in order to reduce hydrogen ...
Date: December 17, 2009
Creator: Koopman, D.
Partner: UNT Libraries Government Documents Department

Interdiffusion in Diffusion Couples: U-Mo v. Al and Al-Si

Description: Interdiffusion and microstructural development in the U-Mo-Al system was examined using solid-tosolid diffusion couples consisting of U-7wt.%Mo, U-10wt.%Mo and U-12wt.%Mo vs. pure Al, annealed at 600°C for 24 hours. The influence of Si alloying addition (up to 5 wt.%) in Al on the interdiffusion microstructural development was also examined using solid-to-solid diffusion couples consisting of U-7wt.%Mo, U-10wt.%Mo and U-12wt.%Mo vs. pure Al, Al-2wt.%Si, and Al-5wt.%Si annealed at 550°C up to 20 hours. Scanning electron microscopy (SEM), transmission electron microscopy (TEM) and electron probe microanalysis (EPMA) were employed to examine the development of a very fine multiphase intermetallic layer. In ternary U-Mo-Al diffusion couples annealed at 600°C for 24 hours, interdiffusion microstructure varied of finely dispersed UAl3, UAl4, U6Mo4Al43, and UMo2Al20 phases while the average composition throughout the interdiffusion zone remained constant at approximately 80 at.% Al. Interdiffusion microstructure observed by SEM/TEM analyses and diffusion paths drawn from concentration profiles determined by EPMA appear to deviate from the assumption of “local thermodynamic equilibrium,” and suggest that interdiffusion occurs via supersaturated UAl4 followed by equilibrium transformation into UAl3, U6Mo4Al43, UAl4 and UMo2Al20 phases. Similar observation was made for U-Mo vs. Al diffusion couples annealed at 550°C. The addition of Si (up to 5 wt.%) in Al significantly reduced the thickness of the intermetallic layer by changing the constituent phases of the interdiffusion zone developed in U-Mo vs. Al-Si diffusion couples. Specifically, the formation of (U,Mo)(Al,Si)3 with relatively large solubility for Mo and Si, along with UMo2Al20 phases was observed along with disappearance of U6Mo4Al43 and UAl4 phases. Simplified understanding based on U-Al, U-Si, and Mo-Si binary phase diagrams is discussed in the light of the beneficial effect of Si alloying addition.
Date: November 1, 2009
Creator: D. D. Keiser, Jr.; Perez, E.; Yao, B. & Sohn, Y. H.
Partner: UNT Libraries Government Documents Department

Transuranic Contamination in Sediment and Groundwater at the U.S. DOE Hanford Site

Description: A review of transuranic radionuclide contamination in sediments and groundwater at the DOE’s Hanford Site was conducted. The review focused primarily on plutonium-239/240 and americium-241; however, other transuranic nuclides were discussed as well, including neptunium-237, plutonium-238, and plutonium-241. The scope of the review included liquid process wastes intentionally disposed to constructed waste disposal facilities such as trenches and cribs, burial grounds, and unplanned releases to the ground surface. The review did not include liquid wastes disposed to tanks or solid wastes disposed to burial grounds. It is estimated that over 11,800 Ci of plutonium-239, 28,700 Ci of americium-241, and 55 Ci of neptunium-237 have been disposed as liquid waste to the near surface environment at the Hanford Site. Despite the very large quantities of transuranic contaminants disposed to the vadose zone at Hanford, only minuscule amounts have entered the groundwater. Currently, no wells onsite exceed the DOE derived concentration guide for plutonium-239/240 (30 pCi/L) or any other transuranic contaminant in filtered samples. The DOE derived concentration guide was exceeded by a small fraction in unfiltered samples from one well (299-E28-23) in recent years (35.4 and 40.4 pCi/L in FY 2006). The primary reason that disposal of these large quantities of transuranic radionuclides directly to the vadose zone at the Hanford Site has not resulted in widespread groundwater contamination is that under the typical oxidizing and neutral to slightly alkaline pH conditions of the Hanford vadose zone, transuranic radionuclides (plutonium and americium in particular) have a very low solubility and high affinity for surface adsorption to mineral surfaces common within the Hanford vadose zone. Other important factors are the fact that the vadose zone is typically very thick (hundreds of feet) and the net infiltration rate is very low due to the desert climate. In some cases where transuranic radionuclides have ...
Date: August 20, 2009
Creator: Cantrell, Kirk J.
Partner: UNT Libraries Government Documents Department