bluuebunny/crossref_metadata_2025_split huggingface dataset:
Use of Dilute, Aqueous HF for Selective Etching
Authors: Gow Catherine J. , Martin Dale W. , Albaugh Kevin B. 
Published: March 2011

ABSTRACTHot, dilute aqueous solutions of hydrofluoric acid (HF) are used to etch silicon nitride films selectively relative to thermally grown silicon dioxide films. Etching is performed in a batch, immersion reactor. The wafers are then rinsed in deionized water and dried with isopropanol vapor. A maximum selectivity of 10.9 to 1 for a 1000:1 HF:H2O dilution at 70 °C is observed. Wafers are etched in hot, dilute HF and cleaned using standard techniques [1]. A thin (12 nm) oxide is then grown and dielectric properties are measured to determine optimum process conditions. For very dilute solutions at lower temperatures, relatively low dielectric strengths (5-6 MV/cm) for the grown oxide are observed, which indicates roughening of the silicon surface. Although the parameter space for exploiting the selective etch process is constrained, a significant range of process conditions exist where selectivity is present and no electrical properties are degraded.
Room-Temperature HF Vapour-Phase Cleaning for LPCVD EPI of Si and SiGe
Authors: Kuiper A.E.T. , Lathouwers E.G.C. 
Published: March 2011

ABSTRACTSilicon dioxide films can be etched in a mixture of HF and H2O vapour. This process can be performed at reduced pressure, which has allowed us to develop a highvacuum compatible etch facility. Channelling-RBS analysis demonstrates the feasibility of epitaxial LPCVD of Si and SiGe layers at 600°C on Si(100) substrates that previously have been vapour-etched in situ at room temperature. With RBS also kinetic information about these deposition reactions can be derived.
Selectivity Mechanisms in Low Temperature (<950°C) Selective Silicon Epitaxy
Authors: Fitch J.T. , Denning D.J. 
Published: March 2011

ABSTRACTLow temperature selective silicon epitaxy was studied over a range of process pressures and HCI flows using a SiH2Cl2/HCl/H2based chemistry. Thermodynamic modelling was carried out with the aid of the SOLGAS program to investigate the effect of process pressure, HCI flow rate, and leaks on the distribution of gas phase species. Selectivity results are interpreted in terms of the defect microchemistry on SiO2surfaces.
A New Two-Step Plasma-Assisted Surface Cleaningoxidation and Film-Deposition Process Sequence for The Formation of Si(100)/SiO2 Interfaces with Low Densities of Interfacial Traps
Authors: Ma Y. , Yasuda T. , Habermehl S. , He S.S. , Stephens D.J. , Lucovsky G. 
Published: March 2011

ABSTRACTDevice quality SiO2 films have been deposited by Remote Plasma Enhanced Chemical Vapor Deposition (RPECVD). The substrate surface cleaning, both ex-situ and insitu, influence the density of trapping states at the Si/SiO2 interface. Pre-deposition exposure to plasma-generated atomic-H significantly reduces hydrocarbon and 0 contamination, but damages the Si surface leading to high interface trap densities (Dit) in the Si band-gap. Pre-deposition exposure to plasma generated O-atoms, removes the residual hydrocarbons, and oxidizes the Si substrate while maintaining an atomically smooth interface. Metal Oxide Semiconductor (MOS) structure fabricated by RPECVD have midgap Dit values of ∼1-3×1010cm−2eV−1 for 0-plasma cleaned surface. Final rinse after RCA cleaning in HF solutions with different pH values affects surface morphology; however surface roughening by atomic-H is always the determinant contributor to high values of Dit.
Effects of Hydrogen Coverage on Silicon Surface Reactivity
Authors: Dillon A.C. , Robinson M.B. , George S.M. , Gupta P. 
Published: March 2011

ABSTRACTHydrogen passivation of silicon surfaces plays an important role in silicon surface cleaning and preparation. To measure the effect of hydrogen passivation on silicon surface reactivity, Fourier transform infrared (FTIR) transmission spectroscopy was used to monitor the oxidation of silicon surfaces versus hydrogen coverage. Experiments were performed insitu in an ultrahigh vacuum (UHV) chamber using high surface area poroussilicon samples. Si-H stretching and bending vibrations and Si-O-Si stretching vibrations were employed to monitor the silicon surface species. Oxidation studies with O2 conducted versus various initial hydrogen coverages revealed that oxidation rates and apparent oxygen saturation levels on porous silicon decreased as a function of initial surface hydrogen coverage. Exceptional surface stability was observed when the porous silicon surface was passivated by both monohydride and dihydride surface species. In addition, new blue-shifted Si-H stretching and bending features were observed following the oxidation of partially hydrogen-passivated porous silicon which indicated the presence of Ox SiH species. Thermal annealing studies revealed that the thermal stability of these OxSiH species increased with increasing oxidation of the silicon surface. These results have important implications for silicon growth and surface cleaning because they indicate that hydrogen removal is more difficult when the silicon surface is contaminated with oxygen. These FTIR results have also been compared with earlier results of oxidation versus hydrogen coverage on Si(111) 7×7.
Transport of 14C and Uranium in the Carrizo Aquifer of South Texas, a Natural Analog of Radionuclide Migration
Authors: Andrews Robert W. , Pearson F. J. 
Published: March 2011

ABSTRACTGround-water flow and transport were modeled along a one-dimensional, vertical cross-section of the Carrizo aquifer from its outcrop down gradient for 100 km. Using a constant flux of 14C at the recharge boundary, the 14C mass fraction distribution downgradient was simulated until steady-state was attained. The match between the observed data and simulated results is excellent for a porosity of 35% and a KD of zero. The transport of both 234U and 238U was simulated with varying retardation coefficients. The best fit occurred with a RD of about 30, which corresponds to a KD of 6. The results indicate that modeling may be reliably conducted over time periods and distances of interest in radionuclide transport performance assessments and that regionally averaged hydrologic properties can provide appropriate values for predicting ground-water flow and nuclide transport.
The Behavior of 99Tc in Doped-Glass/Basalt Hydrothermal Interaction Tests
Authors: Coles D.G. , Apted M. J. 
Published: March 2011

ABSTRACTThe release of polyvalent radionuclides from a nuclear waste repository located in basalt may be sensitively related to the redox potential (Eh) imposed by the basalt. A series of tests are reported here evaluating the effect of basalt on the concentrations of 99Tc released into solution from a borosilicate glass waste form. Crushed PNL 76-68 glass, doped with 0.7 mg 9 99Tc/g glass, was reacted with reference basalt groundwater under oxic hydrothermal conditions in a sampling autoclave, both alone and in the presence of crushed basalt. The steady state fluid concentrations of 99Tc and various stable species were determined from samples obtained at the test conditions of 200 °C, 30 MPa, and a initial solution to solid mass ratio of 10 for both tests.In the glass + groundwater test, the 99Tc concentration rose rapidly to about 50 mg/L after only 200 hr of run time and remained at a value between 50 and 60 mg/L throughout the duration of the test. For the basalt + glass + groundwater test, the 99Tc concentration rose to an initial value of about 2.5 mg/L. At about 700 hr, the 99Tc concentration began to drop rapidly until a value near the analytical detection limit (approximately 0.005 mg/L) was reached after a test duration of 1,400 hr. It is concluded that the presence of basalt in these hydrothermal experimtents reduces the concentration of 99Tc in solution by nearly four orders of magnitude, probably by control of solution Eh and subsequent precipitation of a solid containing a reduced form of technetium. Reaction mechanisms are discussed that can account for these observations.
The Source Term for the Release of Nuclides From a Radioactive Waste Repository – 1. Vitrified waste in granite
Authors: Ewart F.T. , Morris J.B. , Severn J. , Sharpe B.M. , Thomason H.P. 
Published: March 2011

ABSTRACTSamples of simulated vitrified radioactive waste have been prepared containing, separately, neptunium, plutonium and technetium. These glasses have been leached in static granite groundwater for 150 days. The resulting solutions have been characterised with respect to both the ionic and colloidal species: these solutions have then been contacted with concrete and granite to produce initial measurements of the source term from a typical repository. The results show that (a) the colloidal species arise from detached particles of the so-called gel layer on the surface of the glass, (b) e solution activities were 590 Bq.ml-1 in the case of 99Tc and 20 Bq.ml-1 for 237Np, and (c) the sorption coefficients wIre 1.4 and > 300 ml.g for concrete and 0.4 and 8.3 ml.g-1 for granite respectively. The technetium was only weakly sorbed and the neptunium strongly sorbed in each case.
Uranium Migration at Some Hydrothermal Veins Near Marysvale, Utah: A Natural Analog For Waste Isolation.
Authors: Shea Micheal 
Published: March 2011

ABSTRACTThe fission track method was used to map 235Udistribution in thin sections made from twenty-one rock samples. Uranium locations within the studied samples include: veins and veinlets, microcracks, secondary minerals sealing microcracks, rock matrix, grain boundaries and cleavages, and accessory minerals. Two traverses show evidence for nuclide movement by diffusion (−5 and −1 cm, respectively) away from a vein and a veinlet and two other traverses show limited movement along a few microcracks. Three traverses show evidence for no uranium movement. One shows a complex combination of perhaps both bulk diffusion and microcrack transport. Apparent diffusivity (Da) values derived for the rock matrix (10-19to 10-16[m2/sec]) are several orders of magnitude less, and distribution constant (Kd) values (1 to 102[m3/kg]) are several orders of magnitude greater than predicted by laboratory experiments. Values for chlorite phases within the rock matrix were Da= 10-22to 10-19[m2/sec] and Kd= 101to 104[m3/kg].
The Corrosion Resistance of Alloys to Salt Brine Attack
Authors: Mirschinka V. , Odoj R. 
Published: March 2011

ABSTRACTSix metal alloys were subjected to surface and creeprupture stress corrosion tests using NaCl- and Q-brines. Low-allou materials do not seem to be suitable as a barrier against salt brine attack due to the high surface corrosion rate. Creep-rupture strength corrosion was insignificant for the tested materials. Hastelloy C 4 showed no corrosion attack by salt brines in autoclaves at 200°C and 100 bar during 8 months of testing. Bentonite was shown to reduce the corrosion attack considerably.
Investigations of Hydraulic Properties in Crystalline Rock.
Authors: Carlssn Leif , Winberg Anders , Rosander Björn 
Published: March 2011

ABSTRACTHydraulic properties of crystalline rock from four potential repository sites in Sweden were analysed. The hydraulic conductivity of the bedrock was established by means of transient water-injection tests with constant head conducted in 25 m sections. The bedrock at the sites was divided into different hydraulic units. An effective hydraulic conductivity was calculated for the rock mass at each site. This was done on the basis of the frequency distribution of all measured values within this unit. A log-nornal distribution was found to fit the data reasonably well. Regression analysis of hydraulic conductivity as a function of depth indicated similar relationships between the four sites. At a depth of 500 m the effective hydraulic conductivity for three-dimensional flow was about 5.10-11 m/s.The fracture frequency of the sites was established from existing core-logs. At depths of about 500 m the mean fracture frequency of the rock mass at the four sites was 1.8-2.5 fractures per meter. Of this total fracture frequency only a a certain proportion is considered to be hydraulically conductive. This proportion was established from results of hydraulic tests perforned in 2 or 3 m sections. Results obtained indicated a frequency of hydraulically conductive fractures of 0.1-0.3 fractures per meter in the rock mass at depths below 300 m.
NNWSI Performance Assessment Considerations
Authors: Tyler L. D. , Peters R. R. , Hayden N. K. , Johnstone J. K. , Sinnock S. 
Published: March 2011

ABSTRACTThe Nevada Nuclear Waste Storage Investigations (NNWSI) project includes a Performance Assessment task to evaluate the containment and isolation potential for a nuclear waste repository at Yucca Mountain in southern Nevada. This task includes calculations of the rates and concentrations at which radionuclides might be released and transported from the repository and will predict their consequences if they enter the human environment. Among the major tasks required for these calculations will be the development of models for water flow and nuclide transport under unsaturated conditions and in fractured hard rock. The program must also quantify the uncertainties associated with the results of the calculations. The performance assessment will provide evaluations needed for making major decisions as the U. S. Department of Energy seeks a site for a repository. An evaluation will be part of the environmental assessments prepared to accompany the potential nomination of the site. If the Yucca mountain site is selected for characterization and development as a repository, the assessments will be required for an environmental impact statement, a safety analysis report, and other documents.This program has been divided into five tasks. Collectively they will provide the performance assessments needed for the NNWSI Project.
Migration of Radionuclides: Experiments Within the Site Investigation Program at Gorleben
Authors: Warnecke Ernst , Hollmann Almut , Stier-Friedland Gerhard 
Published: March 2011

ABSTRACTExperiments concerning the sorption of radionuclides carried out on the site-specific unconsolidated sediments have been in progress in 6 laboratories since 1981 as part of the Gorleben site investigation program for the most relevant radionuclides. Sorption data are being determined for the site-specific conditions, and the influence of parameters, such as e.g. pH, Eh, complexing agents and NaCl concentrations is investicated. The data are necessary for the safety assessment of the repository. As the investigations have not yet been terminated, the results represent a preliminary state.
The Expected Environment for Waste Packages in a Salt Repository
Authors: Pederson L. R. , Clark D. E. , Hodges F. N. , Mcvpy G. L. , Rai D. 
Published: March 2011

ABSTRACTThis paper discusses results of recent efforts to define the very near-field (within approximately 2m) environmental conditions to which waste packages will be exposed in a salt repository. These conditions must be considered in the experimental design for waste package materials testing, which includes corrosion of barrier materials and leaching of waste forms. Site-specific brine compositions have been determined, and “standard” brine compositions have been selected for testing purposes. Actual brine compositions will vary depending on origin, temperature, irradiation history, and contact with irradiated rock salt. Results of irradiating rock salt, synthetic brines, rock salt/brine mixtures, and reactions of irradiated rock salt with brine solutions are reported.
Evaluation of Iron-Base Materials for Waste Package Containers in a Salt Repository
Authors: Westerman R. E. , Nelson J. L. , Pitman S. G. , Kuhn W. L. , Basham S. J. , Moak D. P. 
Published: March 2011

ABSTRACTDesign studies for high-level nuclear waste packages for salt repositories have identified low-carbon steel as a candidate material for containers. Among the requirements are strength, corrosion resistance, and fabricability. The studies of the corrosion resistance and structural stability of iron-base materials (particularly lowcarbon steel) are treated in this paper. The materials have been exposed in brines that are characteristic of the potential sites for salt repositories. The effects of temperature, radiation level, oxygen level and other parameters are under investigation. The initial development of corrosion models for these environments is presented with discussion of the key mechanisms under consideration.
Evaluation of Spent Fuel as a Waste Form In a Salt Repository
Authors: GRAY Walter J. , McVay Gary L. , Barner John. O. , Shade John W. , Cote Roger W. 
Published: March 2011

ABSTRACTLeach tests have been performed on spent fuel in synthetic Permian Basin salt brine at 25 and 75°C. Complementary tests on unirradiated UO2 pellets have been conducted in both salt brine and deionized water in the range 25 to 150°C. Iron and/or oxidized zircaloy coupons were included in some of the tests. Uranium release from spent fuel was more than 100 times greater than from U02. In brine, iron had no significant effect on the total uranium release but substantially reduced the amount in solution by causing the uranium to plate out on the iron coupon and container walls and to precipitate as filterable particles.
Immobilization of HLW in Synroc-E
Authors: Kesson S.E. , Ringwood A.E. 
Published: March 2011

ABSTRACTSYNROC-E is a new TiO2- rich formulation designed to immobilize ≤7% HLW. It comprises hollandite, zirconolite, perovskite, pyrochlore and alloy (totalling ˜20%) in a continuous matrix of fully densified rutile, and utilizes the principle of “synthetic rutile microencapsulation” in its microstructure. Short-term (%100 day) tests at 95°C reveal that the leach-rates for Cs,Ca,Sr and Ba lie within about an order of magnitude of one another and fall by %2 orders of magnitude over this time period. Leach-rates for U,Ce,Nd,Ti and Zr are initially extremely low and rapidly fall below ICP detection capabilities. On longer timescales, when SYNROC phases exposed at the surface of the wasteform have been leached away, the very low solubility of the protective rutile matrix should greatly inhibit further leaching. SYNROC-E is particularly suitable for burial in salt or in deep ocean sediments as a consequence of its thermal output, microstructure and its capacity to achieve public acceptability.
Polyphase Ceramic and Glass-Ceramic forms for Immobilizing ICPP High-Level Nuclear Waste
Authors: Harker Alan B. , Flintoff John F. 
Published: March 2011

ABSTRACTPolyphase ceramic and glass-ceramic forms have been consolidated from simulated Idaho Chemical Processing Plant wastes by hot isostatic pressing calcined waste and chemical additives at 1000°C or less. The ceramic forms can contain over 70 wt% waste with densities ranging from 3.5 to 3.85 g/cm3, depending upon the formulation. Major phases are CaF2, CaZrTi2O7, CaTiO3, monoclinic ZrO2, and amorphous intergranular material. The relative fraction of the phases is a function of the chemical additives (TiO2, CaO, and Si02) and consolidation temperature. Zirconolite, the major actinide host, makes the ceramic forms extremely leach resistant for the actinide simulant U238. The amorphous phase controls the leach performance for Sr and Cs which is improved by the addition of SiO2. Glass-ceramic forms were also consolidated by HIP at waste loadings of 30 to 70 wt% with densities of 2.73 to 3.1 g/cm3 using Exxon 127 borosilicate glass frit. The glass-ceramic forms contain crystalline CaF2, Al2O3, and ZrSiO4 (zircon) in a glass matrix. Natural mineral zircon is a stable host for 4+ valent actinides.
Partitioning of High-Level Waste as Pretreatment in Waste Management
Authors: Kubota M. , Yamaguchi I. , Okada K. , Morita Y. , Nakano K. , Nakamura H. 
Published: March 2011

ABSTRACTRemoval of the long-lived radionuclides from high-level waste (HLW) is a potential means not only for making wastes more acceptable in terms of long term hazards, but also for alleviating storage requirement. From these points, the authors are developing a method of partitioning actinides, Sr-90 and Cs-137 from HLW. A chemical flow-sheet has been constructed and experiments with actual HLW were initiated in 1982. Through the partitioning, active elements in HLW can be fractionated into 3 groups. Total volume of the solid materials of the 3 groups was calculated and found to be reduced to less than one-third of the volume of the vitrified material containing 10 wt% of fission products as oxide. Such volume reduction seems to facilitate the long term storage or the deep geological disposal of HLW.
GEL Adsorption Processing for Waste Solidification in 'NZP' Ceramics
Authors: Yang L.J. , Komarneni S. , Roy R. 
Published: March 2011

ABSTRACTSimulated PW-4b waste solution along with Na additive was mixed with Zr-P-O gel, dried and then fired to form the desired sodium zirconium phosphate, NaZr2 (PO4) 3 [NZP] ceramic. NZP and monazite were the only phases produced upon firing at 900°C with 10 to 40% of PW-4b mixed with the gel. CsZr2(PO4) 3 which is isostructural with NZP was also identified when fired under reducing conditions. The -200 mesh powders of these waste forms prepared under reducing conditions showed excellent leach resistance under hydrothermal conditions.Alternatively, PW-4b and Three Mile Island (TMI) wastes were adsorbed on Zr-P-O gel in a column. The gel was dried, pelletized and fired to form the desired [NZP] ceramic. Cesium was found to be selective on the Zr-P-O gel because no breakthrough of Cs was detected up to 38 column volumes of TMI waste. Thus, it is possible to use a tailored gel to sorb Cs and/or Sr from accident waste water and then fired the bed to form the [NZP] ceramic below 1000°C. The main advantages of the gel adsorption process are its simplicity and its enormous compositional flexibility.
Radionuclide Transport Modeling of Diffusion Cell Experiments Specific to a Backfill Barrier in a Salt Repository
Authors: Anderson Harold M. , Pietz John M. , Smith Douglas M. 
Published: March 2011

ABSTRACTExperimental radionuclide migration diffusion cell data have been collected as part of the WIPP Waste Package Performance Program. This data was collected under conditions approximating geologic isolation of a backfill barrier in a salt repository. The experiments are designed to aid in the evaluation of engineered backfill barriers.This paper describes a radionuclide transport model designed to aid interpreting experimental diffusion cell migration data and eventually to simulate the long-term effectiveness of the backfill barrier in a salt repository. The model is designed to test a variety of expressions representative of potential mechanisms for retardation within the backfill for the best-fit with experimental data. From the comparison, the aim is to select the appropriate mechanism from the host of potential mechanisms for retardation. The model employs a novel integral equation approach to the solution of the transport equation with nonlinear retardation terms. The solution technique used in this model is a semi-analytical, iterative method for the general nonlinear problem. It is felt the technique offers improved computational efficiency over comparable finite difference methods.Comparisons between experimental migration diffusion cell data and the model predictions are presented in this paper. Tentative conclusions concerning the importance of the retardation mechanism to radionuclide transport in the backfill barrier will be drawn.