bluuebunny/crossref_metadata_2025 huggingface dataset:
The Dependence of ETCH Pit Density on the Interfacial Oxygen Levels in Thin Silicon Layers Grown by Ultra High Vacuum Chemical Vapor Deposition
Authors: Tejwani Manu J. , Ronsheim Paul A.
Published: March 2011

For low temperature silicon epitaxy it is not only important to have an oxygen free environment during growth but also an initial silicon surface free of trace concentrations of oxygen, carbon and other impurities. Variations in the pre-clean process (using the standard ex-situ aqueous hydrofluoric acid dip) used for ultra high vacuum chemical vapor depostion (UHVCVD) of silicon, result in interfacial oxygen levels ranging from 2 × 1012atoms/cm2 to 1014atoms/cm2 as measured by secondary ion ion mass spectroscopy (SIMS). Using a dilute Schimmel etch we have delineated the dislocations in the thin silicon epitaxial layers grown by UHVCVD. Correlation of the etch pit density to the interfacial oxygen levels suggests a power law dependence. Plausibility arguments are presented to explain this power law dependence.
A Complementary Wafer Cleaning and Growth Process for Low Temperature, Defect Free, Selective Silicon Epitaxy
Authors: Fitch Jon T. , Denning Dean J.
Published: March 2011

Low temperature (<850°C) defect free selective silicon epitaxy has been achieved with a conventional barrel type reactor (base pressure -10−4 Torr) using complementary cleaning and growth processes: a wet multi-step oxidizing clean, and a novel non-steady state CVD growth process. With this combination of cleaning and growth processes, it is shown that the need for a high temperature (950-1000°C) insitu native oxide removal step, which may be incompatible with advanced VLSI process integration, is eliminated.
Novel Doping Process for Ultra-Shallow Junction: Rapid Vapor-Phase Direct Doping (RVD)
Authors: Kiyota Yukihiro , Onai Takahiro , Nakamura Tohru , Inada Taroh , Kuranouchi Atsushi , Hirano Yasuaki
Published: March 2011

Characteristics are shown for shallow boron-doped layers formed by a new doping method called Rapid Vapor-phase direct Doping (RVD) which is suitable for making shallow junctions of less than 50 nm. In this method, boron atoms are doped into Si from the vaporphase after the native oxide is removed in hydrogen. From the results obtained for time dependence of doping characteristics, it was found that the surface boron concentration increases almost proportionally to the doping time. This result means that the surface boron concentration is determined by the amount of supplied boron atoms. It is a unique characteristic of this method and a reason why shallow junctions can be fonrned. This method was applied to the base region of a bipolar transistor and produced a normal DC characteristics.
Antimony as A Passivant of Si(111) in the Si(111) (√3×√3)-Sb System
Authors: Hughes A. , Shen T-H. , Matthai C.C.
Published: March 2011

The electronic density of states (DOS) for the Si(111) (√3×√3)-Sb system has been calculated using the tight binding method in the Extended Hiickel Approximation. We find that there is a gap of about 0.8eV between the valence band maximum (VBM) and a surface state. This is in contrast with the case of the unreconstructed (lxl) surface where the Fermi level lies at the surface state.
Surface Cleaning Prior to Formation of Si/SiO2 Interfaces by Remote Plasma-Enhanced Chemical Vapor Deposition (RPECVD)
Authors: Lamb H. H. , Kalem S. , Bedge S. , Yasuda T. , Ma Y. , Lucovsky G.
Published: March 2011

Ex situ UV/O2 cleaning prior to SiO2 deposition by RPECVD results in an SiO2/Si interface with mid-gap Dit values 2-5 times higher than interfaces formed by in situ exposure of HF-etched wafers to plasma-generated atomic O. In situ exposures to plasma-generated atomic H and atomic O are each effective at removing carbon contamination acquired by the UV/O2 cleaned wafers during transfer and introduction to the RPECVD chamber. However, in situ exposure of the photochemical oxide layer to atomic O results in higher mid-gap Dit values, and in situ exposure to atomic H results in creation of dangling bond defects (Pb centers).
Influence of Surface Pre-Cleaning on Electrical Properties of Rapid Thermal Oxide and Rapid Thermal Chemical Vapor Deposition Oxide
Authors: Xu Xiaoli , Kuehn R. T. , Melzak J. M. , Hames G. A. , Wortman J. J. , Öztürk M. C. , Nemanich R. J. , Harris G. , Maher D.
Published: March 2011

Various surface pre-cleaning processes for rapid thermal in-situ polysilicon/oxide/silicon stacked gate formation have been evaluated. MOS capacitors have been fabricated to assess the effects of surface pre-cleaning on the quality of both Rapid Thermal Oxide (RTO) and Rapid Thermal Chemical Vapor Deposition (RTCVD) oxide. Measurement results have shown that, 1) High temperature (≥ 900 °C) rapid thermal cleaning in Ar, H2 or high vacuum (10−8 Torr) ambients can lead to MOS gates with high leakage current if RTO is used to form the gate oxide, 2) The standard Huang clean and ultra-violet ozone (UV/O3) treatments can improve the film quality for both deposited and thermally grown oxide, and 3) Compared with RTO, the breakdown field of the RTCVD oxide is less dependent on the surface pre-cleaning treatment. These results indicate that silicon wafer surface cleaning techniques typically used for silicon epitaxial processes are not necessarily applicable to oxide film formation in RTP reactors.
Numerical and Experimental Investigations on the Time Dependent Behaviour of a Salt Dome With a High-Level Waste Repository
Authors: Prij Jan , Vons Leo H.
Published: March 2011

Results are presented of in-situ measurements, performed in a 300 m deep dry-drilled borehole, in the ASSEmine. Convergence measurements at ambient as well as elevated temperatures and pressure measurements at elevated temperatures are discussed. Creep equations derived from these experiments are used for the numerical analysis of the time dependent behaviour of a salt dome with a HLW repository. The analyses show that the total stresses in the salt remain compressive with deviatoric components smaller than 3 MPa.
Tough – A Numerical Model for Nonisothermal Unsaturated Flow to Study Waste Canister Heating Effects
Authors: Pruess Karsten , Wang Joseph S.Y.
Published: March 2011

The physical processes modeled and the mathematical and numerical methods employed in a simulator for non-isothermal flow of water, vapor, and air in permeable media are briefly summarized. The simulator has been applied to study thermohydrological conditions in the near vicinity of high-level nuclear waste packages emplaced in unsaturated rocks. The studies reported here specifically address the question whether or not the waste canister environment will dry up in the thermal phase.
Numerical Simulation of Flow and Transport in Fractured Tuff
Authors: Travis B. J. , Hodson S. W. , Nuttall H. E. , Cook T. L. , Rundberg R. S.
Published: March 2011

The unsaturated, fractured tuff of Yucca Mountain in the Nevada Test Site is one of the potential sites for geologic storage of high-level radioactive waste. A modeling study of flow and transport in this geologically complex site is presented. Numerical models of mass and heat flow in conjunction with analytical solutions are being used for sensitivity and pathway analysis studies and to aid in design and interpretation of laboratory and field flow and transport tests in tuff.
Rock Joint Description and Modelling for Prediction of Near-Field Repository Performance
Authors: Barton Nick , Bakhtar Khosrow , Bandis Stavros
Published: March 2011

The shear behavior and normal closure behavior of rock joints have a major influence on the changes in joint conductivity to be expected during the life of a repository. Methods are described for obtaining the joint parameters required to model these conductivity perturbations. They include the joint roughness and the wall strength, which can both be obtained from simple tests on jointed drill core. Examples are shown that illustrate the shear stress-displacement, dilation and conductivity coupling that occurs when a joint of given size is subjected to shear under various levels of effective normal stress. Major increases in conductivity may occur. Examples are also shown that illustrate the stress-closure-conductivity coupling. This comprehensive joint behavior model indicates that important size-effects exist in shear behavior which are related to the joint spacing or effective in situ block size. The discrete modelling of near-field joint and fracture performance is essential for predicting the long-term integrity of a nuclear waste repository.
Waste Package Reliability / Nalysis
Authors: Pescatore C. , Sastre C.
Published: March 2011

Proof of future performance of a complex system such as a high-level nuclear waste package over a period of hundreds to thousands of years cannot be had in the ordinary sense of the word. The general method of probabilistic reliability analysis could provide an acceptable framework to identify, organize, and convey the information necessary to satisfy the criterion of reasonable assurance of waste package performance according to the regulatory requirements set forth in 10 CFR 60. General principles which may be used to evaluate the qualitative and quantitative reliability of a waste package design are indicated and illustrated with a sample calculation of a repository concept in basalt.
Analytic Models for Assessing the Performance of Engineered Barriers in a Basalt Repository
Authors: Scott Joanna I. , Koplik Charles M.
Published: March 2011

Analytic models have been developed and used to study the performance of engineered barriers in a basalt repository with waste packages vertically emplaced below the storage room floors. Models have been developed for the groundwater flow field and for waste transport through the waste package backfill and up to the top of the storage room. The lowest release rates at the top of the storage room occur for a room backfill with high hydraulic conductivity (greater than the surrounding rock), high porosity and good sorption. Such a room backfill is much more effective than waste package backfill at delaying releases from the engineered system, especially for short-lived nuclides. A repository design with horizontal emplacement of waste packages has also been studied and found to have higher release rates, owing to a larger cross-section to groundwater flow and the absence of an overlying storage room.
Corrosion Behavior of Low-Carbon Steels in Grande Ronde Basalt Groundwater in the Presence of Basalt-Bentonite Packing
Authors: Anantatmula R.P. , Delegard C.H. , Fish R.L.
Published: March 2011

Three low-carbon steels (AISI 1006, AISI 1020, and AISI 1025) were tested at 150° and 250°C in Hanford Grande Ronde Basalt groundwater (9.75 pH) under anoxic conditions (≤0.1 mg/L oxygen in water) in packing (75 wt% basalt + 25 wt% bentonite) for 1 and 2 weeks. Testing was initially performed in air-equilibrated groundwater (oxic conditions) without packing at 150°C. The average corrosion under anoxic conditions with packing was at least a factor of seven lower than that for oxic conditions at 150°C. The data indicated that carbon composition apparently affects the corrosion of low-carbon steel under oxic conditions. Under anoxic conditions with packing, corrosion was independent of the carbon composition of the steel as evidenced by almost equal weight losses for all three steels at 150° and 250°C. The weight loss at 250°C was about the same as that at 150°C due to the formation of a very adherent layer of iron-rich clay on the surface of the steel at 250°C, reducing the rate at which corrosion proceeds. Pitting was not detected in any of the specimens. Assuming linear corrosion kinetics, extrapolation of the 2-week data to 1,000 yr resulted in approximately 12 mm penetration of the low-carbon steel in Grande Ronde Basalt groundwater under anoxic conditions with packing.
Reactions in the System Basalt/Simulated Spent Fuel/Water
Authors: Grandstaff D.E. , Mckeon G.L. , Moore E.L. , Ulmer G.C.
Published: March 2011

The Grande Ronde Basalts underlying the Hanford Site are being evaluated as a possible site for a high-level nuclear waste repository. Experiments, in which basalt from the Umtanun flow of the Grande Ronde Basalt and basalt with simulated spent fuel were reacted with synthetic Hanford groundwater, were conducted to determine steady state concentrations which can be used in radionuclide release-rate models. Tests were performed at temperatures of 100°, 200°, and 300°C; 30 MPa pressure, and a solution:solid mass ratio of 10:1 for durations up to 7,000 hr. Solution aliquots were extracted periodically during the experiments for analysis. The pH was measured at 250°C and recalculated to higher temperatures. In the basalt-water system the stable high-temperature pH values achieved were 7.2 (100°C), 7.5 (200°C), and 7.6 (300°C). Solution composition variations are due to mesostasis (glass) dissolution and precipitation of secondary phases. Solution measurements indicate a redox potential (Eh) of about -0.7 volts at 300°C. Secondary phases produced include silica, potassium feldspar, iron oxides, clays, scapolite, and zeolites. Tests in the basalt + simulated spent fuel + water systen show that calculated pH values stabilized near 7.6 (100°C), 7.2 (200°C), and 7.7 (300°C). At higher temperatures, solution concentrations were controlled by secondary phases similar to those found in basalt-water tests. Less than 1% of uranium, thorium, samarium, rhenium, cerium, and palladium were released to solution while somewhat higher amounts of iodine, molybdenum, and cesium were released. The UO2 component was unreactive; however, other components (e.g., cesium-bearing phases) were almost completely dissolved. Secondary phases incorporating radionuclide-analog elements include clays, palladium sulfide, powellite, coffinite, and a potassium-uranium silicate.
Investigation of Groundwater Composition in Relation to Spent Nuclear Fuel Disposal
Authors: Karlsson Fred
Published: March 2011

Groundwater from boreholes in granitic rock at six different sites in Sweden has been sampled and analyzed. Sensitive parameters such as redox potential, pH sulphide– and oxygen content have been measured with a field equipment. This is an integrated part of a program of geological, geophysical, geochemical and hydrogeological investigations with the final aim to select a suitable site for a high-level radioactive waste repository. According to present results deep granitic groundwaters are reducing due to the presence of iron(II) ions. The pH is normally ranging from 7 to 9. The total sulphide content is generally less than 0.5 mg/l. The normal alkalinity range is 90-275 mg/l. Copper, which has been suggested as canister material in the present concept for spent nuclear fuel disposal, is stable in this groundwater environment, except for a very slow sulphide corrosion and a limited initial attack of oxygen from the emplacement operations. The dissolution of the spent nuclear fuel matrix, UO2, is ultimately controlled by the total carbonate concentration. Carbonate content, pH and redox conditions will also be decisive for the mobilities of actinides and technetium. The conditions are generally favouring a high degree of retention for these species in the undisturbed deep groundwater rock environment.
Migration in a Single Fracture in Granitic Rock
Authors: Abelin Harald , Gidlund Jard , Moreno Luis , Neretnieks Ivars
Published: March 2011

An in situ experiment to study migration in a single fracture in granitic rock has been performed. Both sorbing and nonsorbing tracers have been used. Results from a tracer run with a nonsorbing tracer (Uranine) are presented. The breakthrough curves have been explored with a model which takes into account hydrodynamic dispersion as well as diffusion into the rock matrix. Diffusivities and porosities of the rock matrix measured in the laboratory are used to interpret the field data. A short description of the experimental design is given. This investigation is a part of the STRIPA PROJECT phase I.
Post Emplacement Environment of Waste Packages
Authors: Knauss K. G. , Oversby V. M. , Wolery T. J.
Published: March 2011

Experiments have been conducted as part of the Nevada Nuclear Waste Storage Investigations Project to determine the changes in water chemistry due to reaction of the Topopah Spring tuff with natural groundwater at temperatures up to 150°C. The reaction extent has been investigated as a function of rock-to-water ratio, temperature, reaction time, physical state of the samples, and geographic location of the samples within the tuff unit. Results of these experiments will be used to provide information on the water chemistry to be expected if a high level waste repository were to be constructed in the Topopah Spring tuff.
The NNWSI Waste form Testing Program
Authors: Oversby V. M.
Published: March 2011

A waste form testing program has been developed to ensure that the release rate of radionuclides from the engineered barrier system will meet NRC and EPA regulatory requirements. Waste form performance testing will be done under unsaturated, low water availability conditions which represent the expected repository conditions. Testing will also be done under conditions of total immersion of the waste form in repository-type water to cover the possibility that localized portions of the repository might contain standing water. Testing of reprocessed waste forms for CHLW and DHLW will use reaction vessels fabricated from Topopah Spring tuff. Chemical elements which are expected to show the highest release rates in the mildly oxidizing environment of the Topopah Spring tuff horizon at Yucca Mountain are Np and Tc. To determine the effect of residual canister material and of corrosion products from the canister/overpack, waste form testing will be done in the presence of these materials. The release rate of all radionuclides which are subject to NRC and EPA regulations will be measured, and the interactive effects of the released radionuclides and the rock reaction vessels will be determined. The testing program for spent fuel will determine the release rate from bare spent fuel pellets and from Zircaloy clad spent fuel where the cladding contains minor defects. A metal testing program for Zircaloy will establish the expected lifetime of the cladding material. Estimation of the state of cladding for fuel presently in reactor pool storage will provide baseline data for Zircaloy containment credit.
Characterization of Earth Materials Properties for Conceptual Design of an Exploratory Shaft, Richton Dome, Mississippi
Authors: Haag R. D. , Swanson O. E.
Published: March 2011

Preliminary Exploratory Shaft design studies have been performed for representative sites in all salt basins within the NWTS Program. These studies have been based on data which are not site specific. Earth materials overlying the Richton Dome were characterized by analysis of geotechnical and hydrological data that had been acquired for site selection purposes. Data sets were reorganized, reinterpreted and evaluated in light of published empirical correlations, known constituative relations, experience with other sites, and engineering judgment.

Geotechnical properties were assessed from geophysical logs, lithologic sample descriptions, and limited blowcount, grain size and pump test data. These properties included grain size, plasticity, unit weight, moisture content, bulk density, porosity, shear strength, elasticity, permeability, and saturation. Additionally, chemical and thermal properties were estimated and the local hydrologic flow properties were addressed.

The analyses allowed heretofor unrecognized lithologic material groupings (defineable layers and sublayers) to be identified based on similarities in physical properties. Subsurface conditions, as interpreted, pose no unique excavation problems. However, the analysis identified some potential issues which had not been previously recognized and gave confidence that other previously assumed potential problems may not exist. Future work for subsequent design should focus on shaft wall stability, ground-water inflow, expansive clays, and the impacts of long-term exposure of the materials.
Geohydrology Surrounding a Potential High-Level Nuclear Waste Repository in the Paradox Basin, Utah
Authors: Brandstetter Albin , Kroitoru Levy , Andrews Robert W. , Thackston John W.
Published: March 2011

The Gibson Dome area in the Paradox Basin in southeastern Utah (Figure 1) has been identified as a potential location for a high-level nuclear waste repository on the basis of an adequate thickness of bedded salt formations at desirable depths, suitable topography for surface facilities, few known archaeological sites, less resource potential than otherwise similar areas, and long-term geologic and tectonic stability [1]. The area appears also suitable from a geohydrologic viewpoint, on the basis of data collected and analyses performed to date. The regional geohydrologi information was obtained from three deep boreholes drilled as part of the Department of Energy's exploration program and from other existing sources. The upper, near-surface, geologic formations include both regionally continuous water-bearing formations and locally perched ground waters that discharge into nearby surface streams and into the Colorado River. These are underlain by the Paradox formation, which consists of a series of essentially impermeable salt layers and of nonsalt interbeds with negligible ground water. Below the Paradox salts, the formations of interest with respect to repository safety include regionally continuous water-bearing formations, with the Leadville limestone being the principal water-transmitting unit. Flows in all water-bearing formations are essentially horizontal. If a vertical connection were established through a potential repository, hydraulic gradients indicate that the flow would first be downward from the upper to the lower formations and then laterally, principally in the Leadville formation. Surface discharge of Leadville formation ground water from the Gibson Dome area has not been detected. There are some indications that minor leakage could occur into the Colorado River as close as Cataract Canyon, about 20 to 25 km (10 to 15 miles) from a potential repository location in Davis Canyon, or into the Colorado River in Marble Canyon (Arizona), about 240 km (150 miles) to the southwest. The computed ground-water flow time in the Leadville formation from below Davis Canyon is about 90,000 years for the first 10 km (6 miles) and about 200,000 years to below Cataract Canyon. The ground-water flow time is estimated to exceed 500,000 years to Marble Canyon. Groundwater flow from a repository to these locations is unlikely, however, since water would first have to penetrate the essentially impermeable salt layers before it would reach the Leadville formation.
The Geochemistry of the Castile Brines: Implications for their Origin and Impact on the Wipp Site(1)
Authors: Coons W. E. , Meyer D. , Olsen R. L. , Register J. K.
Published: March 2011

Pressurized brine reservoirs have been found in the Castile anhydrite which underlies the proposed Waste Isolation Pilot Plant. The major and minor component chemistry of the brines and the isotopic abundances of the brines and coexisting minerals have been determined and evaluated. Thermodynamic calculations indicate that the brines approach saturation with all major minerals in their host horizon. The brines therefore, appear to be in equilibrium with their geologic environment and do not have the capacity for substantially degrading the integrity of the site. When analyzed through solute versus bromide plots, the data indicate that the most likely origin for the brines is Permian seawater. As a result, the reservoirs appear to have been derived from waters no longer being introduced into the rock (i.e., they have a finite source).
Partitioning of Cesium in Phases Produced by Grouted Waste Injection
Authors: Stinton D. P. , Mcdaniel E. W. , Weerent H. O.
Published: March 2011

Phases present in injected grouts were characterized by use of optical microscopy, scanning electron microscopy, x-ray diffraction, and β-γ autoradiography. A laboratoryproduced sample containing 1 wt % stable cesium and an actual grout sheet obtained by core drilling were examined. The phases present in these samples were identified, and cesium was found to be absorbed almost entirely by illite clay agglomerates. These clay agglomerates were tightly bound within the grout structure by hydrated calcium silicates. The β-γ autoradiography of the core-drilled sample verified that cesium and other radionuclides were trapped within the 20-year-old grout and had not migrated into trapped shale fragments.
Corrosion Behaviour of Truw Base and Reference Glasses
Authors: Van Iseghem P. , Timmermans W. , De Batist R.
Published: March 2011

The corrosion behaviour of five TRUW base and reference glasses in distilled water has been investigated for times up to 8 months. The glasses were chosen to evaluate some of the main features of the basaltic slag as produced by the prototype FLK incinerator, i.e. the iron valence state and the relative influence of Al2O3 vs FeOx. Diffusion processes control the corrosion for times below 80 d at 90°C at the SA.V-1 value currently considered (1 cm-1), resulting in the formation of complex films enriched in Al, Fe and Mg over 200 nm thick. The partial replacement of 10 mole 7 Fe2Ox by Al2O3 improves the chemical resistance and strongly accelerates saturation of the leachate for Si, and apparently also for Na and K. The iron valence

state in the glass does not seem to influence the corrosion behaviour. When leaching at temperature up to 200°C, the surfaces of all glasses crystallize, while the corrosion rates do not decrease with time. At SA.V−1 ≠ 0.1 cm−1 , 90°C, the weight losses increase by a factor up to 10 relative to SA.V−1 ≠ 1 cm−1 , 90°C, and the initial diffusion controlled stage might be followed by matrix dissolution reactions.
Metal Matrix Integrity and Related Technology Development in the Canadian Nuclear Fuel Waste Management Program
Authors: Mani Mathew P. , Krueger Paul A.
Published: March 2011

One of the concepts under development as a nuclear fuel waste isolation container is a thin-wall corrosion-resistant shell, supported internally by a cast metal matrix in which intact used fuel bundles are invested. The integrity of the metal matrix can be influenced by metallurgical factors and by process parameters. Finite element solidification modelling and laboratory experiments with lead as an investment material have shown the influence of heat transfer parameters on matrix integrity. Controlled cooling of the container walls, for example, can be used to reduce the interaction time between the molten matrix, the fuel sheathing and the container wall, and achieve a void-free matrix. The results of the computer simulations have been used to design an improved casting system, based on controlled wall cooling, for investing nuclear fuel waste containers. Ultrasonic testing of bonds between some candidate container and metal matrix materials, in combination with the metallurgical characterization of the interface region, has allowed differentiation between bonded and unbonded regions. Matrix cracking near bonded interfaces was identified as a potential problem, which could limit the use of the ultrasonic scanning technique for matrix inspection. To produce a high quality interface with good chemical bonding, induction skin melting looks promising and is being further evaluated.
Modeling Chemical Interactions in the Hydrated Layers of Nuclear Waste Glasses.
Authors: Sullivan Terrence M. , Machiels Albert J.
Published: March 2011

The distribution of glass constituents within the hydrated layer is investigated through a mathematical representation of the fundamental processes that influence mass transport. A brief description of the models that are developed is presented. The resulting equations are implemented in a computer code named GELOH. The importance of the processes of network hydration and layer growth, alteration product formation, and dissolution is studied through the use of GELOH. The results of these studies are presented and interpreted in the light of the experimental evidence available on the aqueous corrosion of nuclear waste glass.
Dissolution of Technetium From Nuclear Waste Forms
Authors: Yasser Khalil M. , White William B.
Published: March 2011

The immobilization of 99Tc in both ceramic and borosilicate glass waste forms and the mechanism of dissolution of technetium from these waste forms are complicated by the range of technetium valence states. Metallic Tc,ionic Tc4+ oxides, and volatile molecular species containing Tc7+ all occur within the stability field of water. Tc4+ will substitute readily for Ti4+ in titanatebased ceramics in such phases as perovskite (CaTiO3) and spinel (Mg2TiO4). Tc4+ can be homogeneously distributed in borosilicate glass by re-melting the glass under closed system conditions to prevent the re-oxidation of technetium. The dissolution of technetium from both glass and ceramic waste forms follow similar kinetic behavior although the dissolution rate from the titanate is two orders of magnitude lower. Dissolution under highly reducing conditions is too slow to measure on the time scale of the experiments.
Influence of Redox Condition in Iron, Silicon and Hydrogen Contents of Leached Glass Surface
Authors: Manara A. , Lanza F. , Della Mea G. , Rossi C. , Salvagno G.
Published: March 2011

Surface analysis has been conducted on samples leached in a Sochlet apparatus at 100°C in the presence and in the absence of air. The XPS technique was applied to analyse the content of iron and silicon while the nuclear reaction method was utilized to analyse the content of hydrogen. Samples leached in argon atmosphere have shown a smaller content of iron and silicon with respect to the samples leached in air atmosphere. The H concentration has shown the same behaviour. The results are discussed in terms of possible formation of iron compounds in the different redox condition and of their different stabilities and in terms of their efficiency in reducing exchange between Na+ and H+ ions.
Time Dependent Leaching in Two-Phase Composite Glasses
Authors: Montrose C. J. , Barkatt Aaron , Macedo P. B.
Published: March 2011

A model has been developed for the elemental leach rates as a function of time in the high-dilution limit for a two-phase composite glass material in which one phase exists as unconnected inclusions embedded within a major continuous phase. For those elements that are preferentially concentrated in this minor and presumed less durable phase, the leach rates are strongly time-dependent, decreasing significantly from the initial to the final steady-state value.
New Data on Ion-Induced Modifications of Aqueous Dissolution of Silicates
Authors: Della Mea G. , Dran J.-C. , Petit J.-C. , Bezzon G. , Rossi-Alvarez C.
Published: March 2011

The ion-induced modifications of aqueous dissolution have been thoroughly investigated in several silicates (amorphous silica, quartz, soda-lime glass and albite) by applying depth profiling techniques (RBS and resonant nuclear reactions) for the characterization of leached surfaces. We show in particular that the H penetration after implantation and leaching is markedly increased when the ion dose exceeds a critical value comprised between 1013 and 1014 Pb.cm-2. This feature supports the radiation damage origin of the observed effects. Such experiments provide interesting clues about the aqueous dissolution of silicates and show that no important effect of α-decay is expected on the hydration of HLW glasses during disposal.
Laboratory and Field Data Needs for Site-Specific Repository Modeling
Authors: Apted M.J.
Published: March 2011

The relative importance of material properties, geochemical processes, and environmental parameters affecting containment and release of radionuclides vary strongly as a function of time and spatial position within a nuclear waste repository located in basalt (NWRB). A simple matrix of spatial regions (Engineered-Barrier System, Disturbed-Rock Zone, Site System of the Controlled Access Zone) and time periods (Pre-emplacement, Containment, Isolation/Slow-release)can be used to identify the dominant processes that affect the containment, release, and migration of radionuclides. From this analysis, a directed and efficient program of field and laboratory studies can be conducted to obtain the data required to assess the feasibility of an NWRB and compliance with Federal regulatory criteria.
Predicting Amounts of Radiolytically Produced Species in Brine Solutions
Authors: Simonson S. A. , Kuhn W. L.
Published: March 2011

The corrosion of iron overpack materials for nuclear waste forms in salt is highly dependent upon the species produced by the radiolysis of brines which may contact the metal. A computer code has been developed to predict the concentrations of the radiolytically produced species as a function of time using published values for radiation yields of species, chemical reaction rate constants, and parameters derived from experimental evidence.
Solubility Constraint: An Important Consideration in Safety Assessment of Nuclear Waste Disposal
Authors: Rai Dhanpat , Ryan Jack L.
Published: March 2011

Solubilities of key solid compounds that are either present in the waste form or can readily precipitate from waste package leachates under repository conditions can be used to set maximum limits on radionuclide concentrations expected in groundwater. This is because the solubility limited concentrations are independent of the release scenarios, hydrologic transport characteristics, and adsorption/desorption reactions. Some of the important factors that control solubilities are pH, pe, type of solid phase, and nature of complexing ligands in the ground waters. Most of the above factors are affected by radiolysis due to the inherent radiation field of the waste form. Experimental results pertaining to the solubilities of selected Am, U, Np, and Pu compounds and the effects of radiolysis are discussed. These results show that: 1) at expected repository pH and reducing conditions, solubility controlled concentrations of several actinides are low and near acceptable limits, 2) the redox conditions at the waste form-water interface may be very oxidizing due to radiolyticeffects, despite the fact that normal repository conditions are assumed to be reducing, 3) additional data on solubility limits and key thermodynamic parameters are needed.
Strontium Isotopic Study of Fracture Filling Minerals in the Grande Ronde Basalt, Washington
Authors: Brookins D. G. , Murphy M. T. , Wollenberg H. A.
Published: March 2011

An important issue for assessing the feasibility of a nuclear waste repository located in basalt at the Hanford Site, Washington, is whether or not waters can percolate vertically through the rocks at the BWIP site. Flow in well defined horizontal, interbed units is well known, but vertical flow has not been fully investigated. Because vertical fractures are common, it is possible that some waters could penetrate the basalts and move normal to the horizontal flow layers. To investigate this, we have chosen to examine the chemistry and isotopics of minerals found in the fractures. These minerals may have originated earlier in the basalt history, i.e. deuteric, in which case their chemistry and isotopic composition should reflect that of the basalt. Alternately, they may have formed late in the basalt's history, in which case their chemistry may be a reflection of material deposited from percolating waters. Thirdly, they may represent a mixing of the basalt and groundwater reservoirs. We examined this issue by determining the strontium isotopic composition of the various reservoirs, supported by other chemical studies. In this fashion we have carefully investigated the origin of the strontium and other trace elements in the fracture filling minerals. To date, waters from surface or subsurface reservoirs at Hanford have not been provided for isotopic analysis, although we plan to run these as soon as they are available. Our conclusions on the strontium isotopic work are therefore qualitative to semi-quantitative at best, but the results can be directly applied to the source of the strontium in the fracture filling minerals.
Generation and Transport Properties of Colloidal Tri-And Tetravalent Actinide Species in Geologic Environments
Authors: Olofsson U. , Bengtsson M. , Allard B.
Published: March 2011

The formation of colloidal particle fractions of triand tetravalent actinides (Am(III), Th(IV), Pu) has been studied in aqueous solutions under various conditions. (Variations of nuclide concentration 10-7-10-11 M; pH 3-12; salt medium 0.01-1.0 M NaCl04; storage time 6 h-6 months; temperature 5°C, 25°C, 70°C). A large fraction of the nuclides is sorbed on the vessel walls, but there is also a formation of centrifugable particles. For americium this fraction has two maxima, at pH 5-8 and at pH above 12. For thorium and plutonium this fraction is very small except at pH about 12. In the pH-range 7-9, which is expected in underground waste repositories, the particle fractions do not seem to be highly mobile. Plutonium behaves very similar to thorium in all studied systems, so it is probably tetravalent to a large extent.
The Behavior of Americium in Aqueous Carbonate Systems
Authors: Silva R. J.
Published: March 2011

The solubilities of crystalline Am(OH)3and AmOHCO3were measured at 25° C in aqueous solutions of 0.1 M NaCl04by determination of the solution concentrations of Am. Prior to use in the measurements, the solid materials were characterized by their x-ray powder diffraction patterns. The solubility product quotients were calculated from the experimental data. The hydrolysis quotients of Am3+were also estimated from the hydroxide solubility data. Using the thermodynamic data derived from these experiments and the recently reported formation constants for the Am3+carbonate complexes, the solid phases and concentrations of solution species of americium in several aqueous carbonate systems were calculated using the computer code MINEQL.
Cemical Durability of a Multicomponent Glass in a Simulated Carnallite/Rock Salt Environment
Authors: Conradt R. , Roggendorf H. , Scholze H.
Published: March 2011

The long term corrosion behaviour of the German HLW glass SM 58 LW 11, doped with 11 vt.% of inactive waste, was investigated in a simulated carnallite/rock salt environment. The experimental conditions were defined with respect to a hypothetical water entrance to a rock salt repository, i.e. temperatures up to 200°C and a pressure of 130 bar were applied; a specific brine from the quinary system MgCl2-MgSO4-NaCl-KCI-H2O (the so-called Q-brine [1] served as leachant. The effects of the corrosion were determined by weight loss measurements, by analyses of the leachates, and by the application of surface analytical techniques.
Geochemical Studies of Columbia River Basalts
Authors: Brookins D. G. , Murphy M. T. , Wollenberg H. A. , Flexser S.
Published: March 2011

We report our findings for two studies of Columbia River basalts that are relevant to United States radwaste disposal programs. A rhyodacite intrudes the Wanapum basalt and is exposed in a quarry along the Hood River, Oregon. Since the rhyodacite is different geochemically from the basalt, and was intruded at a high temperature, then this site can be used to evaluate the effects of the intrusion on the basalt. In such a study we assume the intrusive rock is a high temperature heat engine, and that cooling may be by convection, conduction, or both. Contact effects of the intrusion are not pronounced in the basalt, and are restricted to minor recrystallization within one meter of the contact, and to possibly some minor fracture filling. Traverses were made from the rhyodacite into the basalt, and samples at closely spaced intervals analyzed for K, U, Th by field and laboratory gamma spectrometry, and by neutron activation analysis for some 35 elements. Separate aliquots were analyzed by isotope dilution to inspect Sr isotopic variations. The chemical studies do not indicate migration of any elements from the rhyodacite into the basalt, nor for any in-basalt or in-rhyodacite migration. Despite a pronounced chemical gradient between the rocks, each has remained a closed, isochemical system. The Sr isotopic studies indicate that the rhyodacite possesses Sr (87/86) = 0.70407 ± 0.00008 while the intruded basalts possess Sr (87/86) = 0.7052 ± 0.0002. Even in samples of the contact, with intermingled rhyodacite and basalt, each rock preserves its own characteristic Sr (87/86) signature, thus again attesting to the closed system conditions. We conclude for this natural experiment that the Wanapum basalt is not readily penetrated by elements from a high temperature, molten source.
Progress on Radiometric Dating of Wolfcamp Brines Using 4He and 40Ar
Authors: Zaikowski A. , Kosanke B. J. , Hubbard N.
Published: March 2011

Ground water samples (brines) from deep wells in the Palo Duro Basin, Texas are being analyzed for noble gases in an attempt to obtain radiometric ages for these brines. The brines contain radiogenic 4He and 40Ar produced from the radioactive decay of U, Th, and K. Consideration of hydrochemical data for the brines, various isotopic, chemical, and mineralogical data for the aquifer rocks and noble gas production rates allow estimating the age of the brines to be about 130 million years at two wells. At a third well interaquifer mixing has occurred and the age is presently indeterminate.
A Proposed Model for the Thermal Conductivity of Dry And Water-Saturated Tuff
Authors: Moss M. , Haseman G. M.
Published: March 2011

The room-temperature thermal conductivities of two kinds of tuff from the Nevada Test Site have been measured on a linear-heat-flow thermal comparator. The results are the basis for an empirical model of the conductivity of these rocks in the dry and water-saturated conditions as a function of porosity. Tuff is one medium being considered for nuclear waste disposal. Devitrified, non-zeolitized tuffs with 6-22% porosity, and non-welded, zeolitized tuffs with 24-35% porosity were examined. The empirical geometric mean model is used to characterize the porosity-dependent thermal conductivity. We consider the rock matrix conductivity and the effective fluid conductivity as adjustable parameters in fitting the model to data for saturated and dehydrated samples, separately. For each rock type, the fitted value of matrix conductivity is the same in both the saturated and dehydrated cases; 2.3 and 1.1 W/(m.K) for the non-zeolitized and zeolitized tuffs, respectively. The fitted values for fluid conductivities are different for the the zeolitized and non-zeolitized tuffs, but the ratio of fitted water conductivity to fitted gas conductivity is very nearly the same for both rock types. This permits the use of a single equation to predict with good accuracy the ratio of saturated to dry rock conductivity.
Comparison of Batch And Autoradiographic Methods in Sorption Studies of Radionuclides In Rock And Mineral Samples
Authors: Pinnioja S. , Jaakkola T. , Miettinen J.K.
Published: March 2011

Sorption of 89Sr, 58Co, 134Cs and 241Am in rock and mineral samples was determined by the batch method using crushed sanmples and by autoradiographic method using rock thin sections. The rock samples were tonalite and rapakivi granite, representing the Olkiluoto and Loviisa nuclear power plant sites, respectively. The autoradiographic method was found to be applicable sorption of radionuclides in rock and mineral samles. The 8orption order of nuclides in rock samples was: KaAM> KaCs>KaCo> KaSr the sorption order of the different minerals was biotite> hornblende >plagioclase> potassium feldspar> quartz. Distribution coefficients (Ka) obtained by the batch and autoradiographic methods were of the same order of magnitude. Ka-values of Sr, Co, Cs and Am for tonalite were 5, 20, 190 and 40 × 10-4 m3/m2 by the batch method and 1, 9, 130 and 170 × 10-4 m3/m2 by the autoradiographic method.
Selective Aluminum CVD for Sub-Micron VIA Plug Filling
Authors: Zhu N. , Jo S. K. , Freiler M. B. , Scarmozzino R. , Osgood R. M. , Cacouris T.
Published: March 2011

We present a novel technique to metallize high-aspect-ratio, small-dimension contact holes and via plugs for application to integrated circuits and packaging. The technique uses a laser-assisted process to deposit a thin film of aluminum from DMA1H, which forms a seed layer for subsequent selective CVD. The resistivity of the deposited aluminum is nearly that of the bulk metal, the contact resistivity is good (∼0.03 μΩ-cm2), and the morphology of the deposited film is comparable to that obtained with physical vapor deposition. This process has been used to fill via holes in a SiO2 substrate, and small-diameter (0.7 μm), high-aspect-ratio (3:1), aluminum plugs have been repeatedly formed without the incorporation of voids. A custom-made via chain structure was used to determine the via resistance (plug and contact), which was found to be 0.1 -0.3 Ω. Our technique opens a new process window for void-free high-aspect-ratio via and contact hole filling, and is particularly interesting in that it offers the potential to use aluminum or aluminum-copper in plug metallization.
Self-Aligned Suicides for ULSI
Authors: Maex Karen , Schreutelkamp Rob
Published: March 2011

Since self-aligned suicides are beginning to be used routinely in integral ULSI processes, critical processing characteristics need to be investigated. In this paper, three relevant issues concerning processing are discussed, namely, the formation of ultra-thin suicides, the silicidation of laterally confined areas and the thermal stability of silicide/Si structures. The technological aspects of TiSi2 and COSi2 are discussed and related to their materials properties.
Investigation of an in-situ probe for phase transformations during RTP silicidation
Authors: Schreutelkamp R. J. , Vandenabeele P. , Deweerdt B. , Coppye W. , Vermeiren C. , Lauwers A. , Maex K.
Published: March 2011

In-situ emissivity measurements at a wavelength of 2.4 μα were used to monitor RTP Co silicidation on crystalline and polycrystalline silicon substrates. The influence of various parameters influencing the silicidation reaction was extensively studied. It is shown that particularly the phase transformation from CoSi to the final suicide phase, COSi2, strongly depends on parameters such as background doping level and type of substrate. This is illustrated for As-doped substrates. The method is extremely sensitive for the in-situ detection of the thermal degradation of thin COSi2 films at high temperatures, which is demonstrated for 25 nm COSi2 layers on highly As-doped c-Si substrates.
Electrical Characterization of Phosphorus Doped Ion Beam Synthesised CoSi2/Si Schottky Barrier Diodes.
Authors: Spraggs R. S. , Pananakakis G. , Bauza D. , Reeson K. J. , Gwilliam R. M. , Hunt T. D. , Sealy B. J.
Published: March 2011

The current/voltage characteristics of ion beam synthesised CoSi2/Si (n - type) Schottky barrier diodes implanted with phosphorus to doses between 5 × 1012 and 2 × 1013 ions cm-2are examined after annealing at temperatures in the range 400° - 1000°C. For each dose of implanted phosphorus, the effective barrier height of the CoSi2/Si interface is successively reduced as the anneal temperature increases. The results of Secondary Ion Mass Spectroscopy (SIMS) analysis indicate that these changes are due to an increase in the space charge density at the interface. For lower annealing temperatures the increase in space charge density is attributed to activation of the phosphorus in the tail of the dopant distribution which extends across the CoSi2/Si interface. For higher annealing temperatures larger increases in the space charge density are attributed to a modified dopant distribution resulting from phosphorus diffusion and activation at the interface. For doses of 1 × 1014 P* cm-2and 2×1015P*cm2, ohrnie characteristics are seen after annealing at temperatures of 1000°C and 500°C respectively.