bluuebunny/crossref_metadata_2025_split huggingface dataset:
The Role of [M+O−] Centers (M+– Group IA Ion) in the Activation of Methane on Metal Oxides
Authors: Lunsford J. H. , Lin C.-H. , Wang J.-X. , Campbell K. D. 
Published: March 2011

AbstractIf the ionic radii are compatible, alkali metal ions will substitute for the divalent metal ions in magnesium oxide, calcium oxide and zinc oxide. At high temperatures in the presence of molecular oxygen, centers of the type [M+O−] are formed, where MM+– Group IA ion. The O− ions in equilibrium with these centers are effective in H” abstraction from CH4, which is the first step in the oxidative dimerization reaction. At reaction temperatures >720°C alkali metal oxides formed on the catalytic surface may themselves become active centers. Alkali metal carbonates also may inhibit the activity of the host oxide and sinter the oxide, thereby eliminating corner sites which result in complete oxidation.
Characterization of Supported Metal Oxides by Laser Raman Spectroscopy: Supported Vanadium Oxide on Al2O3 and TiO2
Authors: Wachs Israel E. , Hardcastle Franklin D. , Chan Shirley S. 
Published: March 2011

AbstractThe interaction of supported vanadium oxide with Al2O3 and TiO2 substrates is examined with Raman spectroscopy. The Raman spectra of the supported vanadium oxide reveal that the strong interaction of the vanadium oxide with the Al2O3 and TiO2 supports results in the formation of an atomically dispersed surface vanadium oxide phase as well as supported crystalline V2O5. The relative concentrations of the atomically dispersed surface vanadium oxide and crystalline V2O5 depend on the vanadium oxide loading and the surface area of the oxide support.
Mdiecuiar Die Catmalysis: Hexane Ackmzation Over Pt/KL
Authors: Thuster Samjel J. , Steger Jwhn J. 
Published: March 2011

AbstractIt is proposed that the high aromatization selectivity of Pt supported in zeolite L is caused by the ability of the channels to collimate the flux of hexane molecules, leading to end-on adsorption at the Pt surface.
The Effect of Sodium Catalyst Dispersion on the Carbon Dioxide Gasification Rate
Authors: Li Jian , Adriaan R. P. , Heiningen Van 
Published: March 2011

AbstractThe CO2 gasification rate of black liquor char (b1c) is studied in a thermogravimetric analysis set-up at temperatures between 600 to 800° C. B1c is prepared via fast pyrolysis of the dry solids in spent liquor of the kraft wood pulping process. B1c gasification by CO2 is well described by Langmuir-Hinshelwood type kinetics. The gasification rate of blc is one order of magnitude larger than a high surface area activated carbon impregnated with 12% Na2CO3. Also, the gasification rate of b1c remains high at sodium-carbon ratios where the rate of Na2CO3 impregnated chars would be strongly reduced. With SEM-EDS mapping and line scans it is shown that the unique gasification properties of b1c are caused by a very fine distribution of sodium in the carbon structure.
Energetics of Adsorbed Dimers VIA Self-Consistent Scattering Theory
Authors: Feibelman Peter J. 
Published: March 2011

AbstractA new approach to the surface electronic structure problem, based on a self-consistent scattering theory of point defects, permits 1st-principles calculations for an isolated cluster of adatoms and defects on an otherwise perfect infinite crystalline surface. A first numerical application of the method explains important observations concerning the interaction and diffusion of adatom dimers on Field Ion Microscope tips. Further studies will shed light on questions such as: What impurity species migrate to what kind of surface defects? What is the contribution of substrate atom positional relaxation to adsorption and diffusion barrier energies?
Wipp Site Studies: Secondary Selenite Veins in the Rustler Formation and Dewey Lake Red Beds
Authors: Brookins D. G. , Lambert S. J. 
Published: March 2011

AbstractDefense-generated transuranic (TRU) waste will be stored at WIPP in the bedded halite of the Salado Formation (Permian), which is overlain by the impure Permian evaporites of the Rustler Formation and the Dewey Lake Red Beds. Both the Rustler and Dewey Lake contain abundant to less common secondary selenite veins of uncertain origin, and dissolution zones occur in the Rustler. The Rustler Formation also contains two dolomite aquifers, the Magenta and Culebra members. The purpose of this study is to determine whether vein selenite is locally derived, or has been introduced in moving groundwater solutions. We have used Sr isotopic studies and REE, U and other trace element data to address the problem. The Sr isotopic data show that neither the Salado nor Rustler anhydrites have exchanged with secondary sources of Sr, and this is supported by the REE and U data. Further, selenite veins from the Rustler possess Sr isotopic compositions identical to the Rustler, indicating local origin, and this is also supported by the U and REE data. Selenite veins from the Dewey Lake Red Beds possess Sr isotopic compositions closer to surface caliche deposits, and may contain near-surface derived Sr. U and REE data show more scatter and evidence for extrinsic sources than the vein/host-rock pairs from the Rustler. Sr isotopic compositions of the Magenta and Culebra indicate that the Magenta has undergone less water/rock interaction than the Culebra. Collectively, the data argue for more possible surface or near-surface water/rock interactions in the Dewey Lake Red Beds than in the Rustler Formation; such interactions in the Salado Formation have been minimal.
The Role Of Chemical Reaction In Waste-Form Performance
Authors: Chambré P. L. , Kang C. H. , Lee W. W.-L. , Pigford T. H. 
Published: March 2011

AbstractThe dissolution rate of waste solids in a geologic repository is a complex function of waste form geometry, chemical reaction rate, exterior flow field, and chemical environment. We present here an analysis to determine the steady-state mass transfer rate, over the entire range of flow conditions relevant to geologic disposal of nuclear waste. The equations for steady-state mass transfer with a chemical-reaction-rate boundary condition are solved by three different mathematical techniques which supplement each other. This theory is illustrated with laboratory leach data for borosilicate-glass and a spherical spent-fuel waste form under typical repository conditions. For borosilicate glass waste in the temperature range of 57°C to 250°C, dissolution rate in a repository is determined for a wide range of chemical reaction rates and for Peclet numbers from zero to well over 100, far beyond any Peclet values expected in a repository. Spent-fuel dissolution in a repository is also investigated, based on the limited leach data now available.
Simulation of Effects of Redox and Precipitation on Diffusion of Uranium Solution Species in Backfill
Authors: Carnahan C. L. 
Published: March 2011

AbstractThis investigation addresses the problem of prediction of the rate of migration of redox-sensitive solution species within packing and backfill materials under conditions of variable oxidation potential. Effects of changes of oxidation potential and precipitation of stable uranium compounds during diffusion of uranium from a region of high oxidation potential into a region of low oxidation potential were simulated numerically. Questions of particular interest addressed in the investigation were the existence of a moving “redox front” and the influence of precipitation-dissolution processes on uranium migration. The simulations showed that no expanding redox fronts existed at any simulated time up to 3.2×105 years (1013 s). In simulations where precipitation of stable solids was not allowed, variations of oxidation potential did not affect total uranium concentrations in solution. Concentration profiles could be predicted simply by diffusion of the (constant) source concentrations. In simulations where precipitation of stable solids was allowed, uraninite and calcium uranate accumulated at the source-transport domain interface, while coffinite penetrated further into the transport domain. Total uranium concentrations in regions of precipitation were determined by solubilities of the precipitated solids, and were six to seven orders of magnitude lower than those in the simulations without precipitation, throughout the domain of transport.
The Solubility and Sorption of Lead-210 and Carbon-14 in a Near-Field Environment
Authors: Bayliss S. , Ewart F. T. , Howse R. M. , Smith-Briggs J. L. , Thomason H. P. , Willmott H. A. 
Published: March 2011

AbstractThis paper reports the results of some recent experimental studies of the solubility and sorption behaviour of lead-210 and carbon-14 under cementitious near-field conditions.These studies have shown that under these conditions carbon-14 will have a maximum solubility limit of 10−4 M and that the distribution ratio, RD, will increase with increasing carbon-14 concentrations from 10−9 to 10−7 M. Not all of the carbon in the cement is available for exchange with carbon in the pore water. Differences in values of RD are observed between the two cement grout types studied, SRPC and OPC/BFS. Lead has been shown to have a maximum solubility limit of about 10−3 M at high pH. Good agreement is obtained between these measurements and thermodynamic modelling using the PHREEQE code. No differences were observed between lead solubilities under reducing or oxidising conditions at high pH values using the same phase separation techniques. Lead is particularly sensitive to the phase separation techniques employed. A factor of up to 250 difference is observed between 25000 and 30000 molecular weight cut-off filters. The values of RD for lead increase with decreasing lead concentrations and the values of RD for 10−3 M solutions are observed to be 500 mlg−1 for SRPC and 1300 mlg−1 for OPC/BFS.
High-Level Waste Dissolution and Secondary Mineral(S) Formation and Dissolution
Authors: Zavoshy S. J. 
Published: March 2011

AbstractA mathematical model of waste dissolution that includes the formation of secondary solid precipitate(s) explicitly is formulated. For a static type of experiment, the model indicates that the concentration of the highly soluble species compared to the concentration of the main component of the matrix is equal to their corresponding density in the matrix (i.e., congruent release). For low solubility nuclides, the solubility limit of precipitate(s) may be exceeded (depending on the waste composition) and a supersaturated solution is temporarily achieved. From the model, the concentration of radionuclides and mass flux at the waste form surface in a geologic repository is obtained. Numerical results (Sr-90) indicate that the liquid-surface concentration rises very rapidly after dissolution starts. In the first week, the solubility limit of Sr precipitate(s) is exceeded and Sr solid precipitate(s) will be formed. Subsequently the concentration decreases and becomes almost equal to the solubility limit and remains so for up to 1 year. It decreases later due to radioactive decay. After the first week, the surface concentration and surface mass flux are independent of the retardation coefficient throughout the leaching process. The mass transfer rate is determined by the matrix dissolution rate and not by the solubility of the precipitate.
Effects of Radiation on the Oxidation Potential of Salt Brine
Authors: Gray W. J. 
Published: March 2011

AbstractTwo different types of experiments have shown that alpha and gamma radiation each increase the redox potential of salt brines. Other experiments have shown that the decrease in potential indicated by platinum electrodes in irradiated brines is due to the response of the electrodes to the hydrogen generated from the resulting water decomposition. Platinum electrodes indicated no change in redox potential of solutions containing equimolar concentrations of K3Fe(CN)5 and K4Fe(CN)5 when they were gamma irradiated in quartz containers; when stainless steel containers were used, the solutions reacted with the stainless steel but gamma radiation had no significant effect on that reaction or the solution composition. A gamma radiation dose rate of 2.4 × 104 rad/h measurably increased the uniform corrosion rate of mild steel in salt brines at 75°C, but at 150°C the increased thermal corrosion rate obscured any effect of even higher gamma radiation dose rates.
The Effect of Waste Package Components on Radionuclides Released from Spent Fuel Under Hydrothermal Conditions
Authors: Rawson Shirley A. , Neal William L. , Burnell James R. 
Published: March 2011

AbstractThe Basalt Waste Isolation Project has conducted a series of hydrothermal experiments to characterize waste/barrier/rock interactions as a part of its study of the Columbia River basalts as a potential medium for a nuclear waste repository. Hydrothermal tests of 3–15 months duration were performed with light water reactor spent fuel and simulated groundwater, in combination with candidate container materials (low-carbon steel or copper) and/or basalt, in order to evaluate the effect of waste package materials on spent fuel radionuclide release behavior. Solutions were filtered through 400 and 1.8 nm filters to distinguish colloidal from dissolved species. In all experiments, 14C, 129I, and 137Cs occurred only as dissolved species, whereas the actinides occurred in 400 nm filtrates primarily as spent fuel particles. Actinide concentrations in 1.8 nm filtrates were below detection in steel-bearing experiments. In the system spent fuel + copper, apparent time-invariant concentrations of 14C and 137Cs were obtained, but in the spent fuel + steel system, the concentrations of 14C and 137Cs increased gradually throughout the experiments. In experiments containing basalt or steel + basalt, 137Cs concentrations decreased with time. In tests with copper + basalt, 14C and 129I concentrations attained time-invariant values and 137Cs concentrations decreased. Concentrations for the actinides and fission products measured in these experiments were below those calculated from Federal regulations governing radionuclide release.
Summary of Results From the Series 2 and Series 3 NNWSI Bare Fuel Dissolution Tests
Authors: Wilson C. N. 
Published: March 2011

AbstractThe Nevada Nuclear Waste Storage Investigations (NNWSI) Project is studying dissolution and radionuclide release behavior of spent nuclear fuel in Nevada Test Site groundwater. Specimens prepared from pressurized water reactor (PWR) fuel rod segments were tested for multiple cycles in J-13 well water. The Series 2 tests were run in unsealed silica vessels under ambient hot cell air (25°C) for five cycles for a total of 34 months. The Series 3 tests were run in sealed stainless steel vessels at 25°C and 85°C for three cycles for a total of 15 months. Selected summary results from Series 2 and Series 3 tests with bare fuel specimens are reported.Actinide concentrations tended to saturate and then often decreased during test cycles. Uranium concentrations in later test cycles ranged from 1 to 2 μg/ml in the Series 2 Tests versus about 0.1 to 0.4 μg/ml in Series 3 with the lowest concentrations occurring in the 85°C tests. Formation of a calciumuranium-silicate phase identified as uranophane in the 85°C Series 3 Tests is thought to have limited uranium concentration in these tests. Americium-241, Pu-239 and Pu-240 activities measured in filtered solution decreased to less than 1 pCi/ml in the 85°C tests. Preferential release of fission products Cs, I, Sr and Tc, and activation product C-14, was indicated relative to the actinides. Tc-99 and Cs-137 activities measured in solution after Cycle 1 increased linearly with time, with the rate of increase greater at 85°C than at 25°C. Continuous preferential release of soluble fission products is thought to result primarily from the dissolution of fine particles of fission product phases concentrated on grain boundaries.
Pourbaix Diagram for The Prediction of Waste Glass Durability in Geologic Environments
Authors: Jantzen Carol M. 
Published: March 2011

AbstractDissolution of nuclear waste glass occurs by corrosion mechanisms similar to those of metallurgical and mineralogic systems albeit on different time scales. The effects of imposed pH and oxidation potential (Eh) conditions existing in natural environments on metals and minerals have been quantatively and phenomenologically described in compendiums of Pourbaix (pH-potential) diagrams. Construction of Pourbaix diagrams to quantify the response of nuclear waste glasses to repository specific pH and Eh conditions is demonstrated. The expected long-term effects of groundwater contact on the durability of nuclear waste glasses can then be unified.
Chemical Corrosion of Cogema Glass R7T7 in High Saline Brines
Authors: Lutze Werner , Müller R. , Montserrat W. 
Published: March 2011

AbstractIn the Federal Republic of Germany, the salt dome formation at Gorleben has been selected as a possible site for final disposal of HLW. The waste form will be borosilicate glass, e. g. the French Cogema glass R7T7. The corrosion of this glass has been investigated in three reference salt brines (brine 1: high Mg, brine 2: Ca and Mg, and brine 3: high Na) at three different temperatures, 110°, 150°, and 190°C and three S/V ratios, 10, 100, and 1000m−1. Results are presented for brine 1 and are compared qualitatively with those obtained for the other brines. As in deionized water and pure NaCl, the corrosion rate decreases when silica saturation is approached, but the reaction does not cease. The data after silica saturation were used to calculate upper limiting, i. e. time-independent corrosion rates. The long-term data indicate that a transport process through the growing surface layer may become rate controlling, and that the rate may further decrease after silica saturation.
Chemical Corrosion of Lead-Iron Phosphate Glass
Authors: Lutze Werner , Schubert P. 
Published: March 2011

AbstractThis study provides experimental data on the corrosion of lead-iron phosphate (LIP) glass (6.4 wt% LWR waste loading) and the first comparison with the dissolution kinetics of borosilicate (BS) nuclear waste glass (all experiments at 90°C). Based on previous experiments the hypothesis was made that the alteration phenomena and the corrosion mechanism are analogous to what is known for BS glasses. The corrosion rate was found to be constant, 0.06gm−2d−1, at high flow rates (MCC 5 type test), based on total mass loss and on ion conductivity measurements in the leachate. In an MCC 1 type experiment, doped glasses (32p, 210Pb and 137Cs, respectively) were corroded and the release of activity into solution monitored. The release rates were 0.05gm−2 d−1 in the beginning, but decreased drastically when the solution became saturated with respect to Pb(OH)2 and Pb3 (PO4)2. Glass corrosion continued at a low rate as indicated by Cs activity measurements. The higher chemical durability of the LIP glass vs. BS glasses in DI water is a result of a smaller initial corrosion rate (10 to 10O×) and the fact that the saturation concentration is lower for LIP glass, i. e. it takes less dissolved glass (10 to 50×) to reach saturation than in the case of BS glass.
Corrosion of Lead and Lead Alloys in Simulated Repository Environments
Authors: Goodwin Frank E. , Guenther Richard J. 
Published: March 2011

AbstractA series of corrosion tests were sponsored by the International Lead Zinc Research Organization, Inc. and conducted by Battelle Northwest to obtain information regarding the suitability of lead for applications in the containment of high-level nuclear waste. Two grades of unalloyed lead and two lead alloys were exposed to basalt, salt, and tuff simulated repository environments for times up to 1 year and temperatures from 55 to 200°C. Information was obtained on the corrosion rates, corrosion products, and concentration of lead in the test liquids under these conditions, which were intended to simulate the anticipated environments in candidate U.S. nuclear waste repositories. Leachates from the 12-month tests were analyzed in order to prepare synthetic leachates that were used to conduct scoping evaluations of the transport of lead in basalt and tuff environments.
Rapid Particle Size Analysis of Fly Ash with a Commercial Laser Diffraction Instrument
Authors: Diamond Sidney 
Published: March 2011

ABSTRACTParticle size distributions of a suite of fly ashes previously analyzed by aclassical sedimentary proced ure have been redetermined using a commercial laser-based particle size measurement system. For most of the fly ashes the concordance between the two sets of data was unexpectedly close, considering the different physical principles involved in the two types of measurement. Laser diffraction instruments (“laser sizers”) appear to have the potential for extremely rapid and accurate size distribution measurement for fly ashes to be used in commercial applications. Such determinations, easily carried out on a routine basis, could serve as important elements in future specifications for acceptance and quality assurance programs.
Sulphate Durability of Plain and Fly Ash Mortars
Authors: Day R. L. , Ward M. A. 
Published: March 2011

ABSTRACTSulphate expansion and strength tests were performed on mortars containing equal proportions of Type 10 (ASTM Type I) Portland cement and fly ash as the cementitious material. Two intermediate-Ca subbituminous ashes, one high-Ca subbituminous ash and one bituminous ash were tested. Control mortars manufactured with Types 10 and Type 50 (Type V ASTM) cements and one mortar containing 50% replacement of cement with limestone powder were also examined. The period and temperature of curing prior to sulphate exposure and the concentration and pH of the attacking solutions were other test parameters. Results indicate that the two intermediate-Ca subbituminous ashes from Alberta Canada, combined with Type 10 cement produce mortars that show excellent long-term strength, acceptable short term strength, and are highly resistant to attack by aggressive sulphate solutions. In some cases durability of these mortars was better than the control mortar made with Type 50 cement. On the other hand, mortar made using the high-Ca ash showed very high susceptibility to sulphate attack; at the same time, mortars made with this ash showed good strength development. Qualitative x-ray diffraction analysis indicates the presence of more lime, magnesium and iron bearing crystalline compounds in the high-Ca ash than in the Alberta ashes. These appear to be responsible for the poor performance in sulphate environments.
Low Cement/High Fly Ash Concretes: Their Properties and Response to Chemical Admixtures
Authors: Dodson V. H. 
Published: March 2011

ABSTRACTIn practice, the amount of fly ash added to portland cement concrete varies depending upon the desired end properties of the concrete. Generally, when a given portland cement concrete is redesigned to include fly ash, between 10 and 50% of the cement is replaced by a volume of fly ash equal to that of the cement. Sometimes as much as twice the volume of the cement replaced, although 45.4 kg (100 lbs) of cement will only produce enough calcium hydroxide during its reaction with water to react with about 9 kg (20 lbs) of a typical fly ash. The combination of large amounts of certain fly ashes with small amounts of portland cement in concrete has been found to produce surprisingly high compressive strengths, which cannot be accounted for by the conventional “pozzolanic reaction”. Ratios of cement to fly ash as high as 1:15 by weight can produce compressive strengths of 20.7 MPa (3,000 psi) at I day and over 41.4 MPa (6,000 psi) at 28 days. Methods of identifying these “hyperactive” fly ashes along with some of the startling results, with and without chemical admixtures are described.