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At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Mediated Fuel Cells: Soluble Redox Mediators and their Applications to Electrochemical Reduction of O2 and Oxidation of H2, Alcohols, Biomass, and Complex Fuels

Mediated fuel cells are electrochemical devices that produce power in a manner similar to that of conventional proton exchange membrane fuel cells (PEMFCs). They differ from PEMFCs in their use of redox mediators dissolved in liquid electrolyte to conduct oxidation of the fuel or reduction of the oxidant, typically O2, in bulk solution. The mediators transport electrons (and often protons) between the electrode and the catalysts or chemical reagents in solution. This strategy can help overcome many of the challenges associated with conventional fuel cells, including managing complex multi-phase reactions (as in O2 reduction) or the use of challenging or heterogeneous fuels, such as hydrocarbons, polyols and biomass. Mediators are also commonly used in enzymatic fuel cells, where direct electron transfer from the electrode to the enzymatic active site can be slow. This review provides a comprehensive survey of historical and recent mediated fuel cell efforts, including applications using chemical and enzymatic catalysts. Our research on in this field has been exploring a number of different topics and has been supported different funding sources: the Center for Molecular Electrocatalysis, an Energy Frontier Research Center funded by the U.S. Department of Energy, Office of Science, Office of Basic Energy Sciences (molecular mediators for catalytic oxygen reduction); the Great Lakes Bioenergy Research Center, DOE Office of Science DE-SC0018409 (research on biomass-based fuels; U.S. Department of Energy, Office of Science, Office of Basic Energy Sciences, Catalysis Program, DE-FG02-05ER15690 (copper-catalyzed oxidation reactions); and the NIH NIGMS, R01 GM100143 and R35 GM134929 (use of molecular mediators for organic chemical synthesis).

Anson, Colin W.↗
DOE OSTI2012

Distant protonated pyridine groups in water-soluble iron porphyrin electrocatalysts promote selective oxygen reduction to water

Fe(III)-meso-tetra(pyridyl)porphines are selective electrocatalysts for the reduction of dioxygen to water in aqueous acidic solution. The 2-pyridyl derivatives, both the triflate and chloride salts, are more selective than the isomeric 4-pyridyl complexes. The improved selectivity of is ascribed to the inward-pointing pyridinium groups acting as intramolecular proton relays. This research was supported as part of the Center for Molecular Electrocatalysis, an Energy Frontier Research Center funded by the US Department of Energy, Office of Science. Pacific Northwest National Laboratory is operated by Battelle for the US Department of Energy.

Matson, Benjamin D↗
DOE OSTI2024

Quantitative Correction of Ohmic Effects on Square Wave Voltammetry for High-Concentration Soluble-Soluble Redox Reactions in Molten Salts

Molten salts are of particular interest for a variety of industrial applications. As such, accurate characterization of species’ concentrations within the molten salt media is critical to ensure a well-controlled unit operation. Although electroanalytical tools are properly suited for precise in situ monitoring in these systems, uncompensated ohmic resistance R Ω can lead to erroneous, technique-dependent results. Using numerical simulations applied to a model system, this work will first illustrate then quantify the extent to which R Ω attenuates square wave voltammograms. This approach allows for post-experiment correction that leads to converging results across voltammetric techniques and facilitates accurate predictions of species’ concentrations.

Shaheen, Nora A. [Argonne National Laboratory (ANL↗
NASA NTRS1987

Soluble minerals in chemical evolution. II - Characterization of the adsorption of 5-prime-AMP and 5-prime-CMP on a variety of soluble mineral salts

The adsorption of 5-prime-AMP and 5-prime-CMP is studied in the saturated solutions of several mineral salts as a function of pH, ionic strength, and surface area of the solid salt. It is suggested that the adsorption which results from the binding between the nucleotide molecule and the salt surface is due to electrostatic forces. The adsorption is reversible in nature and decreases with increasing ionic strength.

Chan, Stephen↗
DOE OSTI2012

Evolution of Chemical Conditions and Estimated Plutonium Solubility in the Residual Waste Layer During Post-Closure Aging of Tank 18

This document updates the Eh-pH transitions from grout aging simulations and the plutonium waste release model of Denham (2007, Rev. 1) based on new data. New thermodynamic data for cementitious minerals are used for the grout simulations. Newer thermodynamic data, recommended by plutonium experts (Plutonium Solubility Peer Review Report, LA-UR-12-00079), are used to estimate solubilities of plutonium at various pore water compositions expected during grout aging. In addition, a new grout formula is used in the grout aging simulations and apparent solubilities of coprecipitated plutonium are estimated using data from analysis of Tank 18 residual waste. The conceptual model of waste release and the grout aging simulations are done in a manner similar to that of Denham (2007, Rev. 1). It is assumed that the pore fluid composition passing from the tank grout into the residual waste layer controls the solubility, and hence the waste release concentration of plutonium. Pore volumes of infiltrating fluid of an assumed composition are reacted with a hypothetical grout block using The Geochemist's Workbench{reg_sign} and changes in pore fluid chemistry correspond to the number of pore fluid volumes reacted. As in the earlier document, this results in three states of grout pore fluid composition throughout the simulation period that are termed Reduced Region II, Oxidized Region II, and Oxidized Region III. The one major difference from the earlier document is that pyrite is used to account for reducing capacity of the tank grout rather than pyrrhotite. This poises Eh at -0.47 volts during Reduced Region II. The major transitions in pore fluid composition are shown. Plutonium solubilities are estimated for discrete PuO2(am,hyd) particles and for plutonium coprecipitated with iron phases in the residual waste. Thermodynamic data for plutonium from the Nuclear Energy Agency are used to estimate the solubilities of the discrete particles for the three stages of pore fluid evolution. In Denham (2007, Rev. 1), the solubilities in the oxidized regions were estimated at Eh values in equilibrium with dissolved oxygen. Here, these are considered to be maximum possible solubilities because Eh values are unlikely to be in equilibrium with dissolved oxygen. More realistic Eh values are estimated here and plutonium solubilities calculated at these are considered more realistic. Apparent solubilities of plutonium that coprecipitated with iron phases are estimated from Pu:Fe ratios in Tank 18 residual waste and the solubilities of the host iron phases. The estimated plutonium solubilities are shown. Uncertainties in the grout simulations and plutonium solubility estimates are discussed. The primary uncertainty in the grout simulations is that little is known about the physical state of the grout as it ages. The simulations done here are pertinent to a porous medium, which may or may not be applicable to fractured grout, depending on the degree and nature of the fractures. Other uncertainties that are considered are the assumptions about the reducing capacity imparted by blast furnace slag, the effects of varying dissolved carbon dioxide and oxygen concentrations, and the treatment of silica in the simulations. The primary uncertainty in the estimates of plutonium solubility is that little is known about the exact form of plutonium in the residual waste. Other uncertainties include those inherent in the thermodynamic data, pH variations from those estimated in the grout simulations, the effects of the treatment of silica in the grout simulations, and the effect of varying total dissolved carbonate concentrations. The objective of this document is to update the model for solubility controls on release of plutonium from residual waste in closed F-Area waste tanks. The update is based on new information including a new proposed grout formulation, chemical analysis of Tank 18 samples and more current thermodynamic data for plutonium and grout minerals. In addition, minor changes to the modeling of the grout chemical evolution have been made. It should be noted that the intent is to provide bounding solubilities for plutonium to be used in Performance Assessment modeling rather than trying to identify an exact concentration of plutonium in pore fluids released from a tank at any given time. This document also considers suggestions and opportunities for improvement regarding the plutonium modeling assumptions from the Plutonium Solubility Peer Review Report, LA-UR-12-00079.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Evolution of chemical conditions and estimated solubility controls on radionuclides in the residual waste layer during post-closure aging of high-level waste tanks

This document provides information specific to H-Area waste tanks that enables a flow and transport model with limited chemical capabilities to account for varying waste release from the tanks through time. The basis for varying waste release is solubilities of radionuclides that change as pore fluids passing through the waste change in composition. Pore fluid compositions in various stages were generated by simulations of tank grout degradation. The first part of the document describes simulations of the degradation of the reducing grout in post-closure tanks. These simulations assume flow is predominantly through a water saturated porous medium. The infiltrating fluid that reacts with the grout is assumed to be fluid that has passed through the closure cap and into the tank. The results are three stages of degradation referred to as Reduced Region II, Oxidized Region II, and Oxidized Region III. A reaction path model was used so that the transitions between each stage are noted by numbers of pore volumes of infiltrating fluid reacted. The number of pore volumes to each transition can then be converted to time within a flow and transport model. The bottoms of some tanks in H-Area are below the water table requiring a different conceptual model for grout degradation. For these simulations the reacting fluid was assumed to be 10% infiltrate through the closure cap and 90% groundwater. These simulations produce an additional four pore fluid compositions referred to as Conditions A through D and were intended to simulate varying degrees of groundwater influence. The most probable degradation path for the submerged tanks is Condition C to Condition D to Oxidized Region III and eventually to Condition A. Solubilities for Condition A are estimated in the text for use in sensitivity analyses if needed. However, the grout degradation simulations did not include sufficient pore volumes of infiltrating fluid for the grout to evolve to Condition A. Solubility controls for use in a flow and transport model were estimated for 27 elements in each of the chemical stages generated in the grout simulations plus local groundwater. The grout simulations were run with the initial infiltrating fluid in equilibrium with atmospheric oxygen to account for degradation of the reduction capacity of the grout. However, a lower Eh was used in pore fluids in the oxidizing conditions used to estimate solubilities to be more consistent with measured Eh values and natural systems. Solubilities of plutonium are affected by this decision, but those of other elements are not. In addition, the baseline for H-Area tanks is that they will be washed with oxalic acid prior to being filled with grout. Hence, oxalate was included in the pore fluids by assuming equilibrium with calcium oxalate. Solubility estimates were done by equilibrating a solubility controlling phase for each element with the pore fluid compositions using The Geochemist’s Workbench®. Condition B pore fluids are similar to Condition D. Therefore, solubilities for Condition B were not estimated, but assumed to be the same as in Condition D. In general solubility controlling phases were selected to bias solubilities to higher values. Several elements had no solubility controls and solubility estimates for other elements were omitted because the elements had short half-lives or were present in residual waste in very low amounts. For these it is recommended that release from the tank be instantaneous when the tank liner is breached. There is considerable uncertainty in this approach to enabling a flow and transport model to account for variable waste release. Yet, it is also flexible and requires much less computing time than a fully coupled reactive transport model. This allows some of the uncertainty to be addressed by multiple flow and transport sensitivity cases. Some of the uncertainties are addressed within this document. These include uncertainty in infiltrate composition, grout mineralogy, and disposition of certain components during the simulations. Uncertainty in the solubility estimates is addressed in part by examining sensitivity of solubilities for four key elements to uncertainty in thermodynamic data, pH, Eh, total inorganic carbon concentration, and oxalate concentration. In addition, the solubilities for the elements estimated here are compared to two other compilations of solubilities in cementitious materials.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

NdPO 4 solubility and aqueous Neodymium speciation in supercritical fluids: An experimental study at 500–700 °C and 1.7 kbar

A key aspect in the formation of rare earth elements (REE) deposits is the role of REE transport as aqueous REE complexes in supercritical hydrothermal solutions, where the nature of the aqueous complex is controlled by solution composition, temperature and pressure. Despite chloride being considered as one of the most abundant transporting ligands in magmatic-hydrothermal fluids, experimental investigations on the stability of aqueous REE chloride complexes are scarce above 300 °C. In this study, synthetic NdPO 4 crystals were reacted with non-saline and saline (0, 0.05 and 0.5 mNaCl), acidic (0.01 mHCl) aqueous solutions in a series of solubility experiments conducted at 500–700 °C and 1.7 kbar, where the solubilities were determined using a stable Nd isotope ( 145 Nd isotope spike) dilution technique. NdPO 4 solubility ranges between 28 ppm and 10,858 ppm, where solubility increases with both temperature and salinity. At 500 °C, log mNdPO 4 increases from –3.93 to –1.60 and there is a strong correlation between NdPO 4 solubility and NaCl concentrations (slope of 1.2 ± 0.3), indicating stabilization of the Nd chloride aqueous complexes with a stoichiometry corresponding to NdCl 2+ . At 600 °C, this correlation is weaker (slope of 0.4, log mNdPO 4 increases from –2.63 to –1.88) indicating the stabilization of both Nd chloride and hydroxyl species controlling solubility. At 700 °C, NdPO 4 solubility is largely independent of NaCl concentration indicating that solubility is controlled by Nd hydroxyl complexes, where stoichiometry suggests the neutral Nd(OH) 3 0 species is dominant. The solubility product (Ksp) of NdPO4 is derived from experimental data with the relation: log K sp = -41.81 – 0.057T – 20987/T, with T temperature in Kelvin. Comparison of the measured Nd phosphate solubility to thermodynamic predictions using the available Helgeson-Kirkham-Flowers equation of state parameters for aqueous Nd complexes indicate that predictions are up to three orders of magnitude lower compared to experimental observations. This discrepancy is most pronounced in saline solutions, suggesting that thermodynamic properties of the REE chloride species in supercritical fluids require revision. Numerical simulations of fluid-rock interaction between acidic, saline fluids and a Strange Lake felsic mineral assemblage demonstrates that NdPO 4 solubility predictions from models are four to six orders of magnitude lower than those calculated based on empirical fits from experiments, which suggests that acidic, saline fluids may play an important role in mobilizing large amounts of light REE from 450 to 700 °C.

58 GEOSCIENCES↗
DOE OSTI2016

Water Solubility of Plutonium and Uranium Compounds and Residues at TA-55

Understanding the water solubility of plutonium and uranium compounds and residues at TA-55 is necessary to provide a technical basis for appropriate criticality safety, safety basis and accountability controls. Individual compound solubility was determined using published solubility data and solution thermodynamic modeling. Residue solubility was estimated using a combination of published technical reports and process knowledge of constituent compounds. The scope of materials considered includes all compounds and residues at TA-55 as of March 2016 that contain Pu-239 or U-235 where any single item in the facility has more than 500 g of nuclear material. This analysis indicates that the following materials are not appreciably soluble in water: plutonium dioxide (IDC=C21), plutonium phosphate (IDC=C66), plutonium tetrafluoride (IDC=C80), plutonium filter residue (IDC=R26), plutonium hydroxide precipitate (IDC=R41), plutonium DOR salt (IDC=R42), plutonium incinerator ash (IDC=R47), uranium carbide (IDC=C13), uranium dioxide (IDC=C21), U 3 O 8 (IDC=C88), and uranium filter residue (IDC=R26). This analysis also indicates that the following materials are soluble in water: plutonium chloride (IDC=C19) and uranium nitrate (IDC=C52). Equilibrium calculations suggest that PuOCl is water soluble under certain conditions, but some plutonium processing reports indicate that it is insoluble when present in electrorefining residues (R65). Plutonium molten salt extraction residues (IDC=R83) contain significant quantities of PuCl 3 , and are expected to be soluble in water. The solubility of the following plutonium residues is indeterminate due to conflicting reports, insufficient process knowledge or process-dependent composition: calcium salt (IDC=R09), electrorefining salt (IDC=R65), salt (IDC=R71), silica (IDC=R73) and sweepings/screenings (IDC=R78). Solution thermodynamic modeling also indicates that fire suppression water buffered with a commercially-available phosphate buffer would significantly reduce the solubility of PuCl 3 by the precipitation of PuPO 4 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗