bluuebunny/crossref_metadata_2025_split huggingface dataset:
Zirconoiate corrosion in dilute acidic and basic fluids at 180-700° C and 50 MPa
Authors: Malmström J. , Reusser E. , Giere R. , Lumpkin G. R. , Düggelin M. , Mathys D. , Guggenheim R. 
Published: April 2011

AbstractICorrosion of zirconolite in a closed system at elevated temperature and pressure for various fluid compositions has been investigated by several analytical techniques. Together with previous studies, the results indicate a weak corrosion below 250° C at 5 MPa, Above 400° C, secondary TiO2 phases crystallize in acidic and neutral fluids on the corroded surfaces and in near-surface pores, At 550° C and above, zirconolite starts to be replaced by perovskite and “calzirtite” in basic fluids. Our results indicate that zirconolite is reasonably durable in acidic. neutral and basic fluids up to 500° C and 50 MPa which corresponds to deep borehole conditions.
Corrosion Behavior of Pyrochlore-Rich Titanate Ceramics For Plutonium Disposition; Impurity Effects
Authors: Bakel A. J. , Zyryanov V. N. , Mertz C. J. , Buck E. C. , Chamberlain D. B. 
Published: April 2011

AbstractFour titanate ceramics were characterized and tested. These ceramics were similar to those proposed for Pu disposition, with Ce as a surrogate for Pu. The baseline ceramic contained Ti, U, Ca, Hf, Gd, and Ce, and was made up of only four crystalline phases: pyrochlore, zirconolite, rutile, and brannerite. The three other ceramics contained different amounts of impurities that are expected in the feed. Impurities are defined as any element other than Ti, U, Ca, Hf, Gd, and Ce. The ceramics that contained impurities contained different phases than the baseline. The addition of impurities led to the absence of brannerite and the presence of amorphous silicate, Ca-Al-Ti, and perovskite phases. The results from 3 day, 90°C MCC-I tests with impurity ceramics were significantly different than the results from tests with the baseline ceramic. Overall, the addition of impurities to these titanate ceramics altered the phase assemblages, which in turn, affected the corrosion behavior.
High-temperature melter tests for vitrificationof bnfl high-level nuclear wastes
Authors: Matlack Keith S. , Muller Isabelle S. , Hojaji Hamid , Pegg Ian L. , Ahearn Catherine , Scales Charles R. 
Published: April 2011

AbstractVitrification has become the international method of choice for the immobilization of highlevel nuclear wastes (HLW) from both commercial and defense fuel reprocessing activities. However, there are clear economic incentives for "next-generation" melters that might achieve increased waste loadings and higher throughput rates. In addition, BNFL is investigating potential immobilization processes for projected high-zirconium waste streams that would be generated from advanced nuclear fuel reprocessing processes. One approach towards addressing both of these challenges is waste vitrification at temperatures above the limits that are imposed on all currently operating HLW melters (around 1100 -1200°C). A unique high-temperature joule-heated melter that has been operating at the Vitreous State Laboratory for several years was used to conduct a matrix of tests designed to address these issues. The tests used simulants of two reprocessing waste streams: (i) A blend of "Magnox" and "Oxide" wastes, similar to those currently being vitrified at BNFL's Sellafield site; and (ii) The projected high-zirconium wastes. Each type of waste was investigated both as a wet slurry feed using chemical additives and as a dry calcined waste feed mixed with glass frit. Three target glass compositions were designed for these tests in order to permit processing over a wide range of temperatures. Data were collected on processing rates, off-gas composition, glass product composition and product quality, as well as on general processing behavior. In all cases, glass production rates increased rapidly with processing temperature, typically doubling for every 100°C, and decreased with increasing water content of the feed. The glass production rates per unit melt surface area were significantly higher than those typically observed for hot-wall induction melters operating under comparable conditions. Interestingly, off-gas emissions were much greater for the slurry feeds than for the comparable solid feeds. The resulting glass products were characterized for melt viscosity and electrical conductivity, leachability by both soxhlet and PCT procedures, glass transition temperature, and the amounts and types of any secondary phases that were present.
Study of crystallization mechanisms in the french nuclear waste glass
Authors: Orlhac X. , Fillet C. , Phalippou J. 
Published: April 2011

AbstractThe development of glass materials for long-term storage of high-level waste implies determining the glass thermal stability and notably assessing the risk of devitrification. Previous studies of the French nuclear waste glass have identified the crystalline phases and the domains in which they form, and have shown that devitrification is minimal. Modeling the long-term crystallization behavior requires an investigation of the thermodynamic and kinetic mechanisms liable to induce crystallization during cooling. The first step in this approach is to determine the nucleation and growth curves for each of the component phases. Crystals including CaMoO4, CeO2 and ZnCr2O4 were identified in glasses heat-treated at temperatures between 630°C and 1110°C, and the time and thermal dependence of the CaMoO4 morphology were evaluated. The nucleation and growth kinetics of these phases were determined by optical microscopy and SEM, and the impact of impurities was addressed by studying two glasses, with and without platinoid elements. The results indicated enhanced nucleation kinetics in glass containing platinoid elements. No induction time was observed before permanent nucleation in either of the glasses, and rapid saturation of nucleation kinetics—synonymous with the depletion of active centers of nucleation— was detected after a few hours. Furthermore, similar growth kinetics were observed in both glasses. The nucleation and growth curves coincided for all the phases. Peak values were much higher for nucleation than for growth kinetics, confirming the need for a thorough investigation of the mechanisms occurring in and below the glass transition range, i.e. in the non-equilibrium state.
A Mechanistic Model of Spent Fuel Dissolution, Secondary Mineral Precipitation, and Np Release
Authors: Chen Y. , Siegmann E. , Mattie P. , McNeish J. , Sevougian S. D. , Andrews R. 
Published: April 2011

AbstractA mechanistic spent fuel dissolution model has been developed, based on the generalreaction-transport code of AREST-CT. It considers the dissolution of spent fuel under flow conditions. The kinetic reactions of spent fuel dissolution and precipitation of schoepite, uranophane, soddyite, and Na-boltwoodite are included in the model. The results of model prediction are compared against the results of drip-tests that simulate the conditions that may occur in the Yucca Mountain Repository. Comparison shows that the modeling results match the laboratory observations very well and no contradiction has been found. It indicates that the model is a reasonably good representation of the real system.After validation, the model was used to investigate the release rate of Np from the dissolution of secondary uranyl minerals by examining various degrees of Np incorporation into secondary uranyl minerals. The predicted Np concentration in the aqueous phase is 3 orders of magnitude lower than the upper-bound of the Np solubility range currently used in DOE performance assessment analyses. It suggests that the Np solubility range currently used is too conservative and could be replaced with more realistic values.
Investigations of the Nizhnekanskiy Granitoid Massif (Middle Siberia, Russia) as a Promising Territory for Deep Geological Disposal of HLW: Results of Pre-Exploration Stages of the Work
Authors: Anderson E. B. , Shabalev S. I. , Savonenkov V. G , Lyubtseva E. F. , Rogozin Yu. M. 
Published: April 2011

AbstractTo decide the final fate of high-level radioactive wastes that are in Krasnoyarsk Land, Russia, and, most importantly the wastes from a weapons complex at the Mining-Chemical Combine (MCC, Krasnoyarsk-26), suitable sites for geological disposal of the wastes are being sought in Middle Siberia. The ultimate goal is to select one site for development of a repository for these wastes. As a result of complex geological-geophysical work completed to date, the two most promising sites are in the northern part of the Nizhnekanskiy granitoid massif (the Yenisei Ridge). These sites are approximately 25 km from the MCC. Based on geotectonic analysis and electrical surveying data, these sites are composed of low-permeability biotite granite that extends to relatively great depth. At present, both sites are undergoing more detailed studies to define the most suitable one for detailed geological exploration. Radiogeochemical studies showed that granitoid rock samples from the massif exhibited high retention for major long-lived radionuclides in the wastes. The general results of the investigations are presented.
Synthesis of Garnet/Perovskite-Based Ceramic for the Immobilization of Pu-Residue Wastes
Authors: Burakov B. E. , Anderson E. B. , Knecht D. A. , Zamoryanskaya M. A. , Strykanova E. E. , Yagovkina M. A. 
Published: April 2011

AbstractThe use of garnet/perovskite-based ceramic, with formula type (Y, Gd,..)3(AI, Ga,..)5O1212/(Y, Gd,..)(A1, Ga,..)O3, was tested for immobilizing plutonium residue wastes. Pu residue wastes originate from nuclear weapons production and can contain more than 50% of impurities including such elements as Am, Al, Mg, Ga, Fe, K, La, Na, Mo, Nd, Si, Ta, Ce, Ba, B, W, Zn, Zr, C and Cl. While for some of these residues, direct conversion to typical glass or ceramic forms may be difficult, ceramic forms based on durable actinide host-phases are preferred for Pu, Am and other actinides immobilization. Garnet/perovskite crystalline host-phases are chemically and mechanically durable and desirable for the incorporation of Pu and most of the impurity elements in the Pu residue wastes in the lattices of host-phases in the form of solid solutions. Experiments on the synthesis of garnet/perovskite ceramic samples were carried out using melting in air at temperatures from 1300°C (for samples doped with 10 wt.% Pu residue waste simulant) to 2000°C (for samples doped with 10 wt.% Ce or U). Samples were studied by XRD, SEM and cathodoluminescence techniques. It was found that the garnet phase can incorporate upto 6 wt.% Ce and up to 4.0-5.5 wt.% U, which is correlated with the increase of Ga content and decrease of Al content in the melt. In one of the features of the melt, the perovskite phase formation substitutes for the formation of garnet. The capacity of the perovskite lattice to accommodate Ce and U is higher than the capacity of garnet, reaching about 8 and 7 wt.%, respectively. It was shown that cathodoluminescence can be effectively used to determine the valence state of Ce and U, an important step to optimize the starting precursor preparation. Incase of U4+in the melt, the charge-compensating elements (Sn2+, Ca2+...) are needed to successfully incorporate U in the garnet lattice.
Numerical Simulation of the Interaction Between Granitic Groundwater, Engineered Clay Barrier and Iron Canister
Authors: Trotignon L. , Faure M.-H. , Cranga M. , Peycelon H. 
Published: April 2011

AbstractThe geochemical evolution of a clay engineered barrier subjected simultaneously to granitic groundwater on its external boundary and to a steady flux of iron released by canister corrosion on the other boundary is simulated for time periods up to 1000 y. A multisite ion-exchange model combined with dissolution/precipitation of accessory minerals (calcite, gypsum, quartz, goethite, pyrite, pyrrhotite, iron silicate, magnetite) is used to describe the geochemical evolution of the clay barrier. Diffusive transport of aqueous species and dissolved gases is simulated in a one dimension space. Retroaction of chemical reactions on the corrosion process of the canister is not included in the model. The temperature is assumed to be constant (25°C). Calculations are performed with the IMPACT geochemical coupled code. Two main scenarios are considered: (i) the reduction of sulfates (gypsum) is kinetically hindered, (ii) the reduction of sulfates is possible and leads to the precipitation of sulfides. Main results obtained are dealing with H2(gas) generation and control, pH transient and buffering effect of the clay barrier system. These simulations show that the hydrogen partial pressure in the clay barrier could possibly stay for long periods of time at a very low level due to interactions with mineral buffers or sinks. In the sulfate reduction scenario, a transient pH, xcursion is predicted at the vicinity of the canister. Future developpements of this work include the taking into account of a temperature gradient. Improvement of the geochemical model and reaction pathway is also necessary and will be done with the guidance of on going experimental work.
Modelling of the Interaction Between an Engineered Clay Barrier and Concrete Structures in a Deep Storage Vault
Authors: Trotignon L. , Peycelon H. , Cranga M. , Adenot F. 
Published: April 2011

AbstractThe degradation of concrete structures in a storage vault will release an alkaline plume influencing the geochemical evolution of clay engineered barriers or plugs. The studied configuration has a representative scale and is composed of a rectangular clay barrier interfaced on one side with a cement block and on the other side with an external granitic field. A clay model including major and accessory minerals is used; cement material is supposed to be made of portlandite and CSH minerals. The clay barrier and the cement block are assumed to be initially saturated and in equilibrium with their respective interstitial water. Diffusive transport of aqueous species coupled to chemistry is simulated in a one dimension space up to 10000 y. The feed-back of the porosity evolution on the transport properties is not taken into account. The temperature is supposed to be constant (25°C). Calculations are performed with the TRIO-EF coupled transport-chemistry code.The main results deal with the evolution of the pH profile across the clay barrier, the dissolution of clay minerals and the precipitation of new solid phases in the clay such as albite, gibbsite and CSH. These simulations show that the pH excursion includes only a minor part of the clay barrier at 10 000 y and that the CSH precipitation retards the alkaline plume progression. The appearance of local accumulations of newly precipitated minerals might reduce strongly the porosity. Future developpements of this work include: (i) the completion of the geochemical model by introducing other minerals in concrete (other CSH phases and sulfo-aluminates) and ion exchange sites within the clay, (ii) the implementation of a precipitation/dissolution model in order to describe on a realistic way the porosity profile evolution and possible clogging phenomena.
Reaction of Montmorillonite in Alkaline Solution at 60°C, 90°C, 120°C and 180°C
Authors: Amaya Takayuki , Shimojo Mikio , Fujihara Hiroshi , Yokoyama Katsuhiko 
Published: April 2011

AbstractThe reaction of montmorillonite was investigated. Three kinds of bentonites with different montmorillonite composition were mixed with 0.3M NaOH solution and 0.3M Ca(OH)2 slurry. They were immersed at 60°C, 90°C, 120°C and 180°C for one month, three months and six months. The concentrations of the soluble ions were measured and the bentonites were analyzed quantitatively after the immersion. 50% of the montmorillonite was reacted within two weeks at greater than 90°C. Montmorillonite reacts less when mixed with Si-minerals. It extensively reacted in 0.3M Ca(OH)2 slurry. These results suggest that the reaction mechanism of the montmorillonite in alkaline solution was dominantly Si dissolution, and would decrease by controlling the concentration of Si ion. The cement / bentonite system under Si saturated conditions is discussed.
Time Scales for Dissolution of Calcite Fracture Fillings and Implications for Saturated Zone Radionuclide Transport at Yucca Mountain, Nevada
Authors: Winterle James R. , Murphy William M. 
Published: April 2011

AbstractAn analysis was performed to estimate time scales for dissolution of calcite fracture fillings in the fractured tuff aquifer that underlies Yucca Mountain (YM), Nevada, where groundwater is chemically undersaturated with respect to calcite. The impetus for this analysis originates from speculation that undissolved calcite in the saturated zone is evidence for limited diffusive exchange between fracture and matrix waters. Assuming that matrix diffusion is the rate-limiting process, the time scale for dissolution of calcite fracture fillings depends on the amount of calcite initially deposited, the distance between flowing fractures, the degree of chemical disequilibrium, and the rate of diffusion. Assuming geochemistry of J-13 well water in free-flowing fractures, estimated time scales for complete dissolution of matrix-entrapped calcite range from about 104yr for a 2 mm-thick deposit located I m from a flowing fracture, to over 107 yr for a 2 cm-thick deposit located 100 m from a flowing fracture. We conclude that, given the geochemical and hydrologic characteristics observed at YM, the persistence of calcite minerals over geologic time scales in aquifers where flowing water is under-saturated with calcite does not necessarily preclude matrix diffusion as a dilution mechanism. However, our model suggests that the effective spacing between flowing fractures may be large enough to diminish the overall benefit of matrix diffusion to proposed high-level waste repository performance.
Passivation Behavior of Mild Steel Used for Nuclear:Waste Disposal Package
Authors: Fukaya Yuichi , Akashi Masatsune 
Published: April 2011

AbstractMild steel has been investigated as one of the overpack materials for geological disposal of high-level nuclear waste. The present paper discussed the passivation behavior of mild steel using an electrochemical approach to evaluate the critical condition for passivation. The critical current density for passivation, icris, of mild steel in 0.1 mol /L [HCO3 + CO32−] solution at room temperature was found to decrease with increasing pH of the solution. The depassivation pH, pHd, at room temperature was determined to be about 7.5 by comparisons between icris and the diffusion-limiting current density, iL02, for dissolved oxygen of 8.1 ppm in equilibrium with atmospheric oxygen. The anodic current density of mild steel in the solution of pH = 8 at 90°C was also measured and was found to decrease rapidly with time, leading to the absence of a critical current density for passivation. Thus, mild steel can be self-passivated in the system. An empirically determined corrosion phase diagram is provided for mild steel in 0.1 mol/L [HCO3− + CO32−] solution at 90°C, based on these findings and some published data.
A Modeling Study of General Corrosion of Copper Overpack for Geological Isolation of High-Level Radioactive Waste
Authors: Honda A. , Taniguchi N , Ishikawa H. , Kawasaki M. 
Published: April 2011

AbstractThis paper describes a modeling study for general corrosion of copper which is a candidate material for high-level radioactive waste overpacks. The model is a mixed-potential model combined with diffusive transport of reactants and reaction products. The rest potential and corrosion rate of copper in aerated solution were measured while controlling the thickness of a diffusive solution layer on the copper surface using a rotating-disk electrode. Experimental data were used for validation of the model.
Corrosion Testing of Stainless Steel-Zirconium:Metal Waste Forms
Authors: Abraham D. P. , Simpson L. J. , Devries M. J. , Mcdeavitt S. M. 
Published: April 2011

AbstractStainless steel-zirconium (SS-Zr) alloys have been developed as waste forms for the disposal of metallic waste generated during the electrometallurgical treatment of spent nuclear fuel. The waste forms incorporate irradiated cladding hulls, components of the alloy fuel, noble metal fission products, and actinide elements. The baseline waste form is a stainless steel- 15 wt% zirconium (SS-15Zr) alloy. This article presents microstructures and some of the corrosion studies being conducted on the waste form alloys. Electrochemical corrosio n, immersion corrosion, and vapor hydration tests have been performed on various alloy compositions to evaluate corrosion behavior and resistance to selective leaching of simulated fission products. The SS-Zr waste forms immobilize and retain fission products very effectively and show potential for acceptance as high-level nuclear waste forms.
High-Rate Growth of Stable a-Si:H
Authors: Takagi T. , Hayashi R. , Payne A. , Futako W. , Nishimoto T. , Takai M. , Kondo M. , Matsuda A. 
Published: April 2011

AbstractCorrelation between the gas phase species in silane plasma measured by mass spectrometry and the properties of hydrogenated amorphous silicon (a-Si:H) films deposited by plasma enhanced chemical vapour deposition (PECVD) has been investigated. We have especially been interested in the higher-order silane related species in the plasma, whose contribution to the film growth is considered to be the cause of light-induced degradation in the film quality, especially at high growth rate. In this study, we varied excitation frequency, gas pressure and power density to vary the growth rates of a-Si:H films ranging from 2 Å/s to 20 Å/s.Molecular density ratio of trisilane, representative of higher silane related radicals, to monosilane has shown a clear correspondence to the fill factor after light soaking of Schottky cells fabricated on the resulting films.
Analysis of Plasma Properties and Deposition of Amorphous Silicon Alloy Solar Cells Using Very High Frequency Glow Discharge
Authors: Yan B. , Yang J. , Guha S. , Gallagher A. 
Published: April 2011

AbstractPositive ionic energy distributions in modified very-high-frequency (MVHF) and radio frequency (RF) glow discharges were measured using a retarding field analyzer. The ionic energy distribution for H2 plasma with 75 MHz excitation at a pressure of 0.1 torr has a peak at 22 eV with a half-width of about 6 eV. However, with 13.56 MHz excitation, the peak appears at 37 eV with a much broader half-width of 18 eV. The introduction of SiH4 to the plasma shifts the distribution to lower energy. Increasing the pressure not only shifts the distribution to lower energy but also broadens the distribution. In addition, the ionic current intensity to the substrate is about five times higher for MVHF plasma than for RF plasma. In order to study the effect of ion bombardment, the deposition of a-Si alloy solar cells using MVHF was investigated in detail at different pressures and external biases. Lowering the pressure and negatively biasing the substrate increases ion bombardment energy and results in a deterioration of cell performance. It indicates that ion bombardment is not beneficial for making solar cells using MVHF. By optimizing the deposition conditions, a 10.8% initial efficiency of a-Si/a-SiGe/SiGe triple-junction solar cell was achieved at a deposition rate of 0.6 nm/sec.