bluuebunny/crossref_metadata_2025_split huggingface dataset:
Quantification of Water-Content of Simulated Nuclear Waste Glasses Using Nuclear Reaction Measurements
Authors: Sundaram S. K. , Thevuthasan S. , McCready D. E. 
Published: April 2011

AbstractWater content of glasses is an important factor in glass manufacture. In the case of slurryfed processing of waste glass melters, the water content of waste glasses may change the waste form properties significantly. A reliable, non-contact and quantitative measurement technique, with a spatial resolution of about 2-mm was used to determine the concentration of dissolved water molecules in simulated nuclear waste glasses. The 19F resonant reaction was used. The 1H (19F, αγ)16O reaction requires a resonance energy that is available by use of a 6.471 MeV19F beam. A series of ten segments of slurry-fed, joule-heated, research-scale melter tests were conducted to determine the redox effects on processing of various feeds. Representative glass samples were collected at the middle of each segment. Their hydrogen contents were measured. The hydrogen content was correlated to the feed processing and redox of the glasses.
Low Temperature Vitrification of Radioiodine Using AgI-Ag2O-P2O5 Glass System
Authors: Fujihara H. , Murase T. , Nisli T. , Noshita K. , Yoshida T. , Matsuda M. 
Published: April 2011

AbstractA new vitrification process for radioiodine has been developed for safe disposal of the spent iodine adsorbent generated from reprocessing off-gas systems. The proposed process consists of the following two steps: (1) separation of radioiodine as AgI from spent iodine adsorbent and (2) vitrification of the separated AgI with silver phosphate. An AgI-Ag2O-P2O5 glass system can homogeneously contain up to 60mol%AgI, which leads to high volume reduction efficiency (approximately 1/25). It also can vitrify the AgI without volatilization of iodine because of its low melting temperature (below 400°C). The leachabilities of iodine from simulated vitrified waste forms were evaluated by the MCC-1 static leach test in an anaerobically controlled glove box, which was purged by nitrogen gas with 3% hydrogen. The leaching behavior of the Agl-Ag2O-P2O5 glass system was influenced by the composition of the actual glass, type of leachant, and redox conditions. When the leach test was carried out using simulated ground wateroriginating from rainfall, the leach rate of iodine from 3AgI-Ag4P2O7 glass was 6×10−8g/cm2/d in the early period of the leach test at 35°C under the reducing condition, after which it decreased to 3×10−9g/cm2/d. This glass showed a lower leach rate in the simulated ground water originating from sea water or cement saturated water, since a precipitation layer was formed on the surface.
Thermochemical Models for Nuclear Waste Glass Subsystems – MgO-CaO and MgO-Al2O3
Authors: Besmann T. M. , Spear K. E. , Beahm E. C. 
Published: April 2011

AbstractA relatively simple model, the associate species model, is being applied to nuclear waste glass compositions in order to accurately predict behavior and thermodynamic activities in the material. In the model, the glass is treated as a supercooled liquid, with the liquid species allowed to exist below their melting point. The approach requires an initial assembly of binary and ternary oxide liquid solution data that sufficiently reproduce the equilibrium phase diagrams. Two binary oxide subsystems, MgO-CaO and MgO-Al2O3, have been modeled and results compared to published phase diagrams. Computed activities of the glass constituent species are plotted as a function of composition at 1200°C.
Leaching of Spent Fuel in Cladding Segments and After Removal From Cladding
Authors: Oversby V. M. 
Published: April 2011

AbstractIn the early 1980s, tests of the leaching behavior of spent light water reactor fuel were conducted in Sweden by SKB and in the USA by the NNWSI Project. Both organizations used fuels with similar burnup, leaching solutions with similar chemical compositions, and conducted the tests at ambient hot cell temperature. Most of the test results were closely similar. The exception was in the recovery of actinide elements at the end of leaching cycles. In the NNWSI tests, the test vessels were stripped with nitric acid at the end of the leaching cycle. When the actinide inventories recovered in the original leaching cycle plus vessel rinse solutions were added to the amount recovered from the acid stripping, the relative abundances of uranium, plutonium, and other actinides were approximately the same as their inventories in the fuel samples. In the SKB tests, the materials recovered from stripping the leaching vessels were low in amount and interpreted to be fine fuel fragments. Only a few percent of the plutonium inventory that would correspond to the uranium recovered in solution was accounted for in the solution samples. To investigate the reasons for this difference in recovery of the actinides, a new test was undertaken using a fuel sample that had been leached for several years using the SKB methods. The fuel was removed from the cladding after a total of a bit more than two years of leaching inside the cladding. The bare fuel was then leached for several cycles using a geometry that simulated the NNWSI tests and a test procedure that was similar to that used in the NNWSI tests. The cladding from which the fuel was removed was also leached in a separate vessel in parallel with the bare fuel leaching. This paper presents the results of the test series and discusses the effects of specimen geometry on the mobility of actinides during leaching. During leaching of spent fuel inside the original cladding, a secondary phase containing Pu seems to form. This phase appears to be less easily dissolved than spent fuel itself.
Microstructure Evolution and Weathering Reactions of Synroc Samples Crystallized from CaZrTi2O7 melts: TEM/AEM investigation and geochemical modeling
Authors: Xu Huifang , Wang Yifeng 
Published: April 2011

AbstractScanning electron microscopy (SEM), transmission electron microscopy (TEM), and analytical electron microscopy (AEM) studies have been conducted on samples crystallized from a melt with a composition of zirconolite {(Ca0.9Gd0.1)Zr(Ti1.9Al0.1)2, O7}. The formationof a whole suite of Synroc minerals (zirconia, ZrTiO4, zirconolite, perovskite, and rutile) has been observed. The formation of these minerals follows the crystallization sequence of Ti-bearing zirconia → ZrTiO4 phase → Zr-rich zirconolite → Zr-poor zirconolite →
rutile/perovskite. This sequence is induced by a fractional crystallization process, in which Zr-rich mineral phases tend to crystallize first, resulting in continuous depletion of Zr in melt. Consistent with this melt compositional evolution, Zr content in the zirconolite decreases from the area next to ZrTiO4 phase to the area next to rutile or perovskite. High-resolution TEM images show that there are no glassy phases at the grain boundary between zirconolite and perovskite. The fractional crystallization-induced textural heterogeneity may have a significant impact on the incorporation of radionuclides into crystalline phases and the resistance of radioniclides to leaching processes. Exsolution lamellae and multiple twinning result from the phase transition from tetragonal zirconia to monoclinic zirconia may decrease durability of the Synroc. Fast cooling of melt may produce more zirconolite phase and
relatively uniform texture. In general, however, a Synroc prepared by a through-melt method is less uniform in texture than that prepared by a through-sol-gel method. The reaction path calculation for the alteration of U-bearing zirconolite in an oxidizing fluid shows that zirconolite is first altered into a perovskite-like phase (CaZrO3), followed by rutile (TiO2 ), and U6+ -bearing phases of soddyite [(UO2)2SiO42H20] and haiweeite [Ca(U02)2Si6O15·5H2O].
The Effect f Waste Package and Engineered Barrier System Design:Options on Long-Term Waste Package Degradation Characteristics
Authors: Lee J. H. , Mon K. G. , Bullard B. E. , Longsine D. E. 
Published: April 2011

AbstractSensitivity analyses were conducted to evaluate the impact of a number of waste package and engineered barrier system (EBS) design options on long-term waste package degradation in the potential repository at Yucca Mountain, Nevada. The evaluations were performed by analyzing the waste package degradation in terms of the first breach and first pit-breach profiles with time. Results for each design option were assessed by comparing to results for the reference-case design. An important finding from the analyses is that the waste package lifetime could be extended substantially by limiting and delaying the contact of dripping water with the Alloy 22 corrosion resistant material (CRM) barrier. In the current model, this would keep the CRM barrier outside the conditions that makes the alloy susceptible to localized corrosion. Uncertainty in waste package degradation analysis would be reduced by improved understanding and characterization of the processes that create the local exposure condition on waste package.
Principal Component Analyses of Potential Repository Groundwaters
Authors: Sasamoto H. , Salter P. , Apted M. , Yuiv M. 
Published: April 2011

AbstractThe chemical composition of ambient groundwater for a geological, high level radioactive waste repository is of crucial significance to issues such as radioelement solubility limits, sorption, corrosion of the overpack, behavior of compacted clay buffers, and many other factors involved in repository safety assessment. At this time, there are no candidate repository sites established in Japan for the geological disposal of high-level radioactive waste, and only generic rock formations are under consideration. It is important that a small, but representative set of groundwater types be identified so that defensible models and data for generic repository performance assessment can be established. Over 15,000 separate analyses of Japanese groundwaters have been compiled into a data set for the purpose of evaluating the range of geochemical conditions for waste repositories in Japan. This paper demonstrates the use of a multivariate statistical analysis technique, principal component analysis (PCA), to derive a set of statistically based, representative groundwater categories from the multiple chemical components and temperature that characterize the deep Japanese groundwater analyses. PCA also can be used to guide the selection of groundwaters that could be used in scenario analyses of future geological events in Japan.
Radionuclide Sorption at Yucca Mountain, Nevada - A Demonstration of an Alternative Approach for Performance Assessment
Authors: Turner David R. , Pabalan Roberto T. , Prikryl James D. , Paul Bertetti F. 
Published: April 2011

AbstractAn approach is developed for including aspects of mechanistic models of radionuclide sorption into performance assessment (PA) calculations. Water chemistry data from the vicinity of Yucca Mountain (YM), Nevada are screened and used to calculate the ranges in key parameters that could exert control on radionuclide sorption behavior. Using a diffuse-layer surface complexation model, sorption parameters for Np(V) and U(VI) are calculated based on the chemistry of each water sample. Model results suggest that lognormal probability distribution functions (PDFs) of sorption parameters are appropriate for most of the samples; the total calculated range is almost five orders of magnitude for Np(V) sorption and nine orders of magnitude for U(VI) sorption, but most samples fall in a narrower range. Finally, statistical correlation between the calculated Np(V) and U(VI) sorption parameters can be included as input into PA sampling routines, so that the value selected for one radionuclide sorption parameter is conditioned by its statistical relationship to the others. The approaches outlined here can be adapted readily to current PA efforts, using site-specific information to provide geochemical constraints on PDFs for radionuclide transport parameters.
Kinetics of MgO Dissolution and Buffering of Fluids in the Waste Isolation Pilot Plant (Wipp) Repository
Authors: Monastra Valerie , Grandstaff D. E. 
Published: April 2011

AbstractDOE has proposed to add MgO to backfill in the WIPP repository to stabilize pH and reduce gas pressures and spallation. Experiments show that MgO dissolution rates are proportional to surface area, nearly independent of pH, and decrease with increasing dissolved NaCl. The probable repository MgO dissolution rate will be ≤ 10 to 20 μmol m2 min−1. Simulations indicate that MgO dissolution will buffer pH and provide stabilizing cement; however, under highly reducing conditions MgO will not control total gas pressures. Adding oxidizing agents or materials containing oxidized species, for example anhydrite, to the backfill, might control gas pressures.
Modeling the Durability of Silica Additives Proposed to Protect Nuclear Waste Glass Packages in A Geological Repository
Authors: Minet Yves , Vernaz Étienne 
Published: April 2011

AbstractSilica additives have been proposed to protect nuclear waste glass packages in deep geological repositories by presaturating silicon adsorption sites* in the media surrounding the package (engineered barrier, canister and overpack corrosion products). The durability of such additives was estimated using a one-dimensional dissolution-transport model developed at the CEA's Rhône Valley Research Center for the French reference glass in typical disposal sites. The silica additive would be placed between the wasteform and the overpack and thus dissolved before the waste glass. Its lifetime is proportional to Madd2(where Madd is the added silica mass) up to a critical value Mcrit at which it increases considerably (from 103 or 104 to 108 years) due to saturation of the adsorption sites. Excess silica consumption is then limited by the typically very low (3 × 10−3 l·m−2y−1 ) groundwater flow in the geological site. After consumption of the added silica, the evolution of the altered waste glass fraction is comparable to the evolution without added silica if Madd « Mcrit, but much lower if Madd ≈ Mcrit or Madd > Mcrit. These findings were confirmed by some experimental evidence, and establish that a few kilograms of added silica for a 400 kg package can considerably enhance the waste glass durability.
Modeled Near-Field Environment Porosity Modification due to Coupled Thermohydrologic and Geochemical Processes
Authors: Nitao John J. , Glassley William E. 
Published: April 2011

AbstractHeat generated by waste packages in nuclear waste repositories can modify rock properties by instigating mineral dissolution and precipitation along hydrothermal flow pathways. Modeling this reactive transport requires coupling fluid flow to permeability changes resulting from dissolution and precipitation. Modification of the NUFT thermohydrologic (TH) code package to account for this coupling in a simplified geochemical system has been used to model the timedependent change in porosity, permeability, matrix and fracture saturation, and temperature in the vicinity of waste-emplacement drifts, using conditions anticipated for the potential Yucca Mountain repository. The results show dramatic porosity reduction approximately 10 m above emplacement drifts within a few hundred years. Most of this reduction is attributed to deposition of solute load at the boiling front, although some of it also results from decreasing temperature along the flow path. The actual distribution of the nearly sealed region is sensitive to the time-dependent characteristics of the thermal load imposed on the environment and suggests that the geometry of the sealed region can be engineered by managing the waste-emplacement strategy and schedule.
Determination of the Porosity, Permeability and Diffusivity of Rock in the Excavation-Disturbed Zone Around Full-Scale Deposition Holes Using the He-Gas Method
Authors: Autio J. , Timonen J. , Aaltonen T. , Laajalahti M. , Kuoppamäki K. , Maaranen J. 
Published: April 2011

AbstractIn a nuclear waste repository, rock in the excavation-disturbed zone adjacent to the walls of deposition holes for waste canisters is a potential pathway for the transport of corrosive agents and radionuclides. Three experimental holes the size of deposition holes in a KBS-3 type repository (depth 7.5 m and diameter 1.5 m) were bored in hard granitic rock in the Research Tunnel at Olkiluoto and the porosities, effective diffusivities and permeabilities of rock in the excavation-disturbed zone were determined in a direction parallel to the disturbed surface using He-gas methods. Permeability and diffusivity in a direction parallel to the rock schistosity was found to be clearly larger than in a direction perpendicular to it.
Corrosion Protection of Metallic Waste Packages Using Thermal Sprayed Ceramic Coatings
Authors: Wilfinger K. R. , Farmer J. C. , Hopper R. W. , Shell T. E. 
Published: April 2011

AbstractCeramic coated carbon steel coupons were corrosion tested in water with dissolved salts to simulate exposure to evaporation concentrated groundwater in an underground nuclear repository. Metallography revealed no corrosion at the ceramic metal interface of dense coatings, even though electrical measurements demonstrated that the coatings were slightly porous. Experimental results and a model to predict corrosion rates influenced by a porous ceramic coating and coating lifetimes are presented.
Zircaloy Corrosion in a Repository Environment
Authors: Christian Jerry D. 
Published: April 2011

AbstractAssessments are made of the corrosion characteristics of spent nuclear fuel Zircaloy cladding in a Yucca mountain repository environment and the potential for the cladding to provide protection against radionuclide release following waste package failure. Considerations and assumptions includes a waste package life near 10,000 years and air-saturated water contacted with waste package corrosion product goethite, based on the near-field geochemical environment evaluated in the Yucca Mountain Viability Assessment [3]. Literature corrosion data (general, pitting, and localized crevice attack) are evaluated on the basis of these conditions and the expected chemical environments that can result on the surface of the fuel. General corrosion of Zircaloy is expected to be negligible and result in a lifetime of the SNF cladding of several hundred thousand years, approaching a million years. General surface pitting is not expected. Effects of crevice localized corrosion for periods beyond 10,000 years are uncertain and require modeling development and experimental characterization. Details of the evaluations that provide the basis for the conclusions are presented.
Sectrophotometric Investigation of U(VI) Chloride Complexation in the Naci/NaClO4 system
Authors: Paviet-Hartmann Patricia , Lin Mavis R. 
Published: April 2011

AbstractPost closure radioactive release scenarios from geologic salt formation, such as the WIPP (Waste Isolation Pilot Plant) (USA) include hydrologic transport of radionuclides through a chloride saturated aquifer. Consequently, the understanding of actinide solution chemistry in brines is essential for modeling requiring accurate knowledge of the interaction between AnO22+ and chloride ions. Complexation constants of two U(VI) chloride species UO2Cl+ and UO2Cl20 have been intensively studied for about 40 years using different methods. However, large uncertainties reflect the general difficulty in determining accurate stability constants of weak complexes. In order to model the behavior of U(VI) in brines, we studied the formation of its chloride complexes by UV-Vis spectroscopy as a function of the NaCI concentration at 25 °C. The experiments were performed at constant ionic strength by varying the concentration ratio of NaCI and NaClO4. Deconvolution resulted in single component absorption sepectrum for UO2Cl+ and UO2Cl20. The apparent stability constants of UO2Cl+ and UO2Cl20 are determined at different ionic strengths. The experimental data are used to parameterize using the SIT approach.
He-Dilution to Increase Deposition Rate and Feedstock Utilization During the Growth of a-Si:H and a-SiGe:H Alloys
Authors: Middya A. R. , Wood G. , Lin G. H. , Carlson D. E. 
Published: April 2011

AbstractWe report on the development of helium diluted a-Si:H and a-SiGe:H solar cells with higher deposition rates and better feedstock utilization than devices made with hydrogen dilution. Both the initial and the stabilized efficiencies of the He-diluted single-junction aSi:H and a-SiGe:H cells are similar to those of hydrogen-diluted cells with state-of-the-art intrinsic materials. The total fabrication time for tandem cells has been reduced by 17% by using helium dilution without loss in initial and stabilized efficiency.
ESR Measurements of a-Si:H and a-Si0.5Ge0.5:H Films Under Solid-Phasecrystallization
Authors: Yun I. H. , Roh O. H. , Lee J.-K. 
Published: April 2011

AbstractWe have investigated the solid-phase crystallization of a-Si1-xGex:H (x=0 and 0.5) films by using electron spin resonance and x-ray diffraction. The films were deposited on Coming 1737 glass in a plasma-enhanced chemical vapor deposition system using SiH4 and GeH4 gases. The films were then annealed to be crystallized at 600°C. It was observed that, for the a-Si:H film, both the spin density and the g-value first increased with annealing time, and then rapidly decreased as the film was crystallized. For the a-Si0.5Ge0.5:H film, the Ge dangling bond spin density increased from 3 × 1018 cm-3 to 2 × 1019 cm3 for the first stage of annealing and then decreased to 3 × 1017 cm-3 after being crystallized; the Si dangling bond spin density just increased to about 2 × 1017 cm-3 and remained nearly constant for further annealing. It is thought that exodiffusion of hydrogen resulted in the increase of spin density in the beginning, and then some portions of amorphous components were converted into the crystalline phase by further annealing. And the keen correlation between the dependence of x-ray peak intensity and the dependence of Ge dangling bond spin density on the annealing time suggests that the Ge dangling bonds rather than Si dangling bonds play an important role in the crystallization of the Si0.5Ge0.5 film.
Hydrogen Transport in Phosphorus and Boron Dopedpolycrystalline Silicon
Authors: Nickel N. H. , Kaiser I. 
Published: April 2011

AbstractHydrogen diffusion in phosphorus and boron doped polycrystalline silicon was investigated by deuterium diffusion experiments. The presence of dopants enhances hydrogen diffusion. The effective diffusion coefficient Deff is thermally activated and the activation energy varies between 0.1 and 0.4 eV. This is accompanied by a variation of the diffusion prefactor by 12 orders of magnitude. Using the theoretical diffusion prefactor the actual energy EA was calculated from Deff.EA also depends strongly on the Fermi energy and exhibits a similar dependence as the formation energies of H+ and H- in single crystal silicon.
Characterization of Hydrogen in Hydrogenated Nano-Crystalline Silicon
Authors: Yamamoto K. , Itoh T. , Ushikoshi K. , Nonomura S. , Nitta S. 
Published: April 2011

AbstractThe characterization of hydrogen in hydrogenated nano-crystalline silicon (nc-Si:H) films with various crystallinity has been studied using FTIR absorption spectroscopy and gas effusion spectroscopy. In gas effusion spectrum of a nc-Si:H, three evolution peaks of H2 are found near 400, 500 and 600°C. From the deconvolution of the gas effusion spectra, the hydrogen concentration in amorphous region dose not depend on the volume fraction of crystalline. The surface density of hydrogen at the interface and in the grain boundaries of nano-crystalline grains decreases with the volume fraction of crystalline. The properties of hydrogen in nc-Si:H films are discussed with the results of FTIR absorption spectroscopy.