bluuebunny/crossref_metadata_2025_split huggingface dataset:
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

ABSTRACTVarious 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

ABSTRACTResults 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

ABSTRACTThe 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

ABSTRACTThe 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

ABSTRACTThe 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

ABSTRACTProof 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

ABSTRACTAnalytic 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

ABSTRACTThree 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

ABSTRACTThe 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

ABSTRACTGroundwater 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

ABSTRACTAn 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

ABSTRACTExperiments 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

ABSTRACTA 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

ABSTRACTPreliminary 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

ABSTRACTThe 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

ABSTRACTPressurized 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

ABSTRACTPhases 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

ABSTRACTThe 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 valencestate 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

ABSTRACTOne 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.