bluuebunny/crossref_metadata_2025 huggingface dataset:
Microanalytical Tem Studies of Ceramic Materials
Authors: Bischoff Ewald , Rühle Manfred
Published: March 2011

Frequently, the processing of dense, strong, and tough ceramics requires the addition of sintering aids. The materials are therefore composed of various components. For the understanding of the properties of the ceramics it is essential that the microstructure and the distribution of the different components within the material are known. The possibilities and limitations of microanalytical TEM studies will be described for different structural ceramics, e.g. ZrO2 , NiO, SiC, and SiAlON.
Thermal Annealing of Vacancy and Interstitial Loops in Ion Irradiated Copper
Authors: Larson B. C. , Noggle T. S. , Barhorst J. F.
Published: March 2011

X-ray diffuse scattering has been used to study the thermal annealing of vacancy and interstitial loops in Ni-ion irradiated copper. The diffuse scattering formalism is reviewed and diffuse scattering measurements are reported on liquid-He temperature Ni-ion irradiated copper after annealing to 40, 275, and 300 °C. Size distributions are presented for vacancy and interstitial loops after each anneal and the thermal-induced changes are discussed in terms of loop dissolution and coalescence.
Characterization of Tilt Boundaries by Ultra High Resolution Electron Microscopy
Authors: Krakow W. , Wetzel J. T. , Smith D. A. , Trafas G.
Published: March 2011

A high resolution electron microscope study of grain boundary structures in Au thin films has been undertaken from both a theoretical and experimental point of view. The criteria necessary to interpret images of tilt boundaries at the atomic level, which include electron optical and specimen effects, have been considered for both 200kV and the newer 400kV medium voltage microscopes. So far, the theoretical work has concentrated on two different [001] tilt bounda-ries where a resolution of 2.03Å is required to visualize bulk lattice structures on either side of the interface. Both a high angle boundary, (210) σ=5, and a low angle boundary, (910) σ=41, have been considered. Computational results using multislice dynamical diffraction and image simulations of relaxed bounda-ries viewed edge-on and with small amounts of beam and/or specimen inclina-tion have been obtained. It will be shown that some structural information concerning grain boundary dislocations can be observed at 200kV. However, many difficulties occur in the exact identification of the interface structure viewed experimentally for both [001] and [011] boundaries since the resolution required is near the performance limit of a 200kV microscope. The simulated results at 400kV indicate a considerable improvement will be realized in obtain-ing atomic structure information at the interface.
High Resolution Z-Contrast Imaging and Lattice Location Analysis of Dopants in Ion-Implanted Silicon
Authors: Pennycook S. J. , Narayan J. , Cijlbertson R. J. , Fogarassy E. , Batson P. E.
Published: March 2011

Two new electron microscopy techniques have been developed which greatly extend the capabilities for the micro-characterization of semiconductors. The first is a technique for the direct imaging of dopants in semiconductors, whether or not they are in solution, using Z-contrast, and the second is a technique for determining the substitutional fraction of dopant. Both techniques are capable of nanometer spatial resolution and allow the detailed study of dopant segregation, precipitation, and clustering effects.
The Use of Ion Channeling Axial Scans in the Study of Ion-Implanted Al2O3
Authors: Farlow G. C. , White C. W. , Appleton B. R. , Sklad P. S. , McHargue C. J.
Published: March 2011

Rutherford backscattering and ion channeling-axial scans have been used to study lattice sites for several impurities implanted into A12O3. The case of Ga implanted in A12O3 is discussed and is shown to be substitutional on the Al sublattice. Additionally, the use of this technique in the study of precipitates in A12O3 is discussed with reference to Fe implanted A12O3 which was annealed in either oxygen or hydrogen.
Exafs Studies of Defects in β-Alumina Superionic Conductors
Authors: Roth W. L. , Wong R.
Published: March 2011

The local structures of Zn and Co in solid solution in Na βN–alumina have been investigated using EXAFS spectroscopy. These additives are members of a group that must be incorporated in βN–alumina to stabilize the rhombohedral structure. It is found that they are tetrahedrally coordinated to four oxygen atoms. The N–O bond distances are larger than the Al–O distance in β–alumina and larger than the Al(2)–O distances in disordered crystals determined by x-ray and neutron diffraction. The bonds about individual ions and mixtures of ions are consistent with predictions from bond-strength-bonddistance theory. A preliminary study of Na β–alumina that has been implanted with in and annealed indicates that a substantial fraction of the Mn is tetrahedrally coordinated.
Fim/Iap/Tem Studies of Ion Implanted Nickel Emitters
Authors: Walck S. D. , Hren J. J.
Published: March 2011

Accurate depth profiling of implanted hydrogen and its isotopes in metals is extremely important. Field ion microscopy and atom-probe techniques provide the most accurate depth profiling analytical method of any available. In addition, they are extremely sensitive to hydrogen. This paper reports our early work on hydrogen trapping at defects in metals using the Field Ion Microscope/Imaging Atom Probe (FIM/IAP). Our results deal primarily with the control experiments required to overcome instrumental difficulties associated with in situ implantation and the influence of a high electric field. Transmission Electron Microscopy (TEM) has been used extensively to independently examine the influence of high electric fields on emitters.
Electronic and Chemical Analysis of a Metal-Insulator Interface Utilizing Transmission Electron Energy Loss Spectroscopy at 5Å Spatial Resolution
Authors: Scheinfein Michael , Isaacson Michael
Published: March 2011

Using a 0.5 nm diameter probe of 100 keV electrons, we have been able to detect significant changes in the transmission electron energy loss spectra in the region of valence shell and L23 shell excitation within a spatial extent of 0.4 nm of an Al-AlF3 interface. The spectra have been recorded with a dose significantly less than the critical dose for destruction of the AlF3.
Experimental Analysis of Fracture in High Strength Cementitious Composites
Authors: Ansari Farhad
Published: March 2011

Characterization of cementitious composites by their fracture properties has been difficult due to controversial results reported in the technical literature. Existing studies on the fracture behavior of plain concrete reveal some fracture characteristics that differ from those normally observed in metallic materials. Among these characteristics is the existence of a micro-cracking zone or process zone at the tip of an advancing crack. The determination of the fracture process zone in concrete is a difficult experimental problem, because the resulting deformation is strongly localized.

In the present study, in-plane displacements in front of notched high strength concrete have been monitored. Laser speckle metrology, which is a special technique utilizing the speckle patterns of laser light for measurement of in-plane displacements, is employed. Experimental results indicate that a precise description of the fracture process zone is possible by speckle metrology.
Porosity and its Relationship to the Strength of Hydraulic Cement Pastes
Authors: Kendall K , Birchall J D
Published: March 2011

The concept of ‘intrinsic strength’ of cement was addressed by testing the bending strength propertes of MDF cement pastes[1–5], containing no organic material, over a range of pore volumes and flaw lengths. The MDF pastes, made only of cement and water, exhibited interesting behaviour, particularly the well packed compositions which gave remarkably low porosity, about 1%, and high modulus, 75 GPa. It was demonstrated that the bending strength of these materials depended on flaw size in a brittle fashion, and that the toughness increased as the volume porosity fell. Fracture mechanics described the results satisfactorily. Therefore, theories based on intrinsic strength, or on the need for polymers to provide high strength [7, 10] are unnecessary.
Fracture Behavior of Lightweight Aggregate Concrete Under Multiaxial Compressive Stress
Authors: Grübl Peter
Published: March 2011

Lightweight aggregate concrete under triaxial compressive load shows a different failure behavior compared with normalweight concrete. The increase of strength due to confining pressure is not as large as it is with normalweight concrete. It depends on the type of aggregate. The larger the stiffness of the aggregate the larger the increase of strenoth. Under hydrostatic pressure the low strength lightweight aggregate concrete (Bc ∼ 15 N/mm2 ) showed a confining strength which was about 2 to 2,5 times higher than the urfiaxial strength. Concrete with aggregates with high stiffness (E ∼ 14,000 N/mm2 ) and a uniaxial strength between 33 and 48 N/mm2 reached at a confining factor of 25% a strength increasing factor of about 4. The volumetric change curves show two remarkable facts. With confining factors of 0,05 and 0,10 lightweight aggregate concrete has at a stress level of 80% to 90% of the confining strength a volumetric increase which changes into a volumetric decrease shortly before reaching the ultimate load. With a confininn factor of 0.25 the volume decreases constantly until the fracture happens. The temporary volumetric increase is caused by a widening of the cracks. The fracture of lightweight aggregate concrete under confining pressure is mainly caused by an internal collapse of the aggregate grains.
High Stpength Fiber Reinforced Cement Composites
Authors: Naamian Antoine E.
Published: March 2011

The effects of fiber reinforcenent; on the response of high strength fiber reinforced concrete in conmpression, tension and flexure are reviewed. Tradeoffs between the use of high fiber contents to achieve comiposite strength and ductility, and a reduction in matrix porosity to achieve high matrix strength are discussed. Methods used to increase composite strength include application of pressure after casting, addition of fly-ash, silica fumes and or sulerplasticizers to the cement matrix, and mixing high volurite contents of high strength stiff fibers.
Reactions in the Setting of High Strength Cement Pastes
Authors: Rodger S. A. , Sinclair W. , Groves G. W. , Brooks S. A. , Double D. D.
Published: March 2011

Materials of very high flexural strengths (> IOOMPa) can be made by high shear mixing of cement pastes with small amounts of water soluble polymer. Two such systems - high alumina cement/partially hydrolysed polyvinyl acetate and OPC/polyacrylamide - have been examined using a variety of techniques including isothermal calorimetry, infra-red spectroscopy and X-ray diffraction. In both systems the polymer addition appears to become cross-linked by ions released by the cement powder leading to an increase in stiffness of the material. In particular the reactions involved in the OPC system are analogous to those occurring in certain dental cements. The importance of removing excess water from the pastes is demonstrated.
Moving Boundary Model for Leaching of Nuclear Waste Glass
Authors: Banba T. , Murakami T. , Kimura H.
Published: March 2011

The leaching experiments of Soxhlet type have been carried out[l, 2] and on the basis of the results we developed the mathematical leaching model which used one dimensional diffusion and could treat the growth of surface layers. The model adopted the following assumptions: 1) The velocity of the bulk glass-surfacé layers boundary depends on time alone. 2) Some of the diffusing substances are immobilized in the surface layers by an irreversible first-order reaction. 3) A fictitious film exists at the solution-surface layers interface. The fundamental equations were established based on these assumptions and the numerical solutions were obtained by the Crank-Nicholson implicit method. The values of the diffusion coefficient, the reaction rate and the film mass transfer coefficient were obtained by a trial-and-error method. The applicability of the model was confirmed by the fact that the leaching mechanisms deduced from the calculated results were consistent with those mechanisms deduced from the experimental results. The present study showed the proposed model was valid for calculation of the leach rates of waste glasses when surface layers grew during leaching, and the study also indicated which parameters should be measured experimentally to predict the leach rates.
The Effect of Phase Formation on Glass Leaching
Authors: Haaker Richard , Malow Gonter , Offermann Peter
Published: March 2011

The leaching of nuclear waste glasses in aqueous solution is still considered to be the most important process when discussing a potential release of radionuclides. However, in the last few years it has become clear that the release mechanism under static leaching conditions is very complex involving several processes [1]. The kinetics gave no evidence for pure leaching (i.e. the alkali-ion exchange) which should result in a square root time law, or for pure corrosion (i.e. congruent dissolution of the glass) which should result in a linear time law. Moreover, fits using a combination of these two processes gave unsatisfactory results. This is certainly due to additional processes which are not considered by ion exchange and congruent dissolution. Some explanations may be the build up of alteration layers, saturation effects of some elements with phase formation due to solubility limits and disappearance of metastable phases. To describe such a complicated system thermodynamic data are needed and measurements of mass losses as well as analyses of leachates and surfaces layers have to be performed.
Detailed Processes of Surface Layer Formation in Borosilicate Waste Glass Dissolution
Authors: Flintoff John F. , Harker Alan B.
Published: March 2011

The morphology and chemical structure of surface layers formed during the aqueous leaching of the SRP-type borosilicate waste glass in simulated ground waters have been studied by SEM/EDXS, XRD, and surface spectroscopy. Comparison of layers formed in deionized, silicate, and brine waters shows the processes of both corrosion and precipitation to be highly localized. The corrosion process proceeds on the glass surface preferentially at points of stress and inhomogeneity. Crystalline and noncrystalline precipitates form in well-defined regions within the surface layers indicating the layer development process to be dominated by species migration and local saturation effects.
A General Rate Equation for Nuclear Waste Glass Corrosion
Authors: Grambow B.
Published: March 2011

A general rate equation was developed to describe the reaction of nuclear waste glasses with aqueous solutions as a function of pH, the ratio of sample surface area to solution volume, temperature, time, glass- and solution composition. Thermodynamic and kinetic models have been combined in a new model. Glass network dissolution, pH variation, precipitation of stable or metastable solid reaction products, silica saturation at the glass and solution interface (reaction zone) and a residual affinity for the long term reaction under near saturation conditions have been addressed. The model has been successfully used to interpret a great number of experimental data reported in literature.
A Kinetic Model of Nuclear Waste Glass Dissolution in Flowing Water Environments
Authors: Barkatt Aaron , Gibson Barbara C. , Brandys Marek
Published: March 2011

A simple kinetic model for the description of the interaction of nuclear waste borosilicate glass with water has been developed. In the case of SRL TDS-131 glass leached in water at 70°C over a broad range of flow rates this model was found useful both in describing the evolution of leachate composition as a function of exposure time at a particular flow rate and in describing the dependence of the steady-state concentrations of the leached elements in solution on flow rate in a series of flow experiments.
Solubility Limits. on Radionuclide Dissolution
Authors: Kerrisk Jerry F.
Published: March 2011

This paper examines the effects of solubility in limiting dissolution rates of a number of important radionuclides from spent fuel and high-level waste. Two simple dissolution models were used for calculations that would be characteristic of a Yucca Mountain repository. A saturation-limited dissolution model, in which the water flowing through the repository is assumed to be saturated with each waste element, is very conservative in that it overestimates dissolution rates. A diffusion-limited dissolution model, in which element-dissolution rates are limited by diffusion of waste elements into water flowing past the waste, is more realistic, but it is subject to some uncertainty at this time. Dissolution rates of some elements (Pu, Am, Sn, Th, Zr, Sm) are always limited by solubility. Dissolution rates of other elements (Cs, Tc, Np, Sr, C, I) are never solubility limited; their release would be limited by dissolution of the bulk waste form. Still other elements (U, Cm, Ni, Ra) show solubility-limited dissolution under some conditions.
Modelling of Waste Form Performance and System Release
Authors: Barkatt Aaron , Macedo Pedro B. , Gibson Barbara C. , Montrose Charles J.
Published: March 2011

Kinetic and thermodynamic approaches to the derivation of long-term release rates of species from defense waste glass are reviewed. It is concluded that at high flow rates kinetic factors are pre-dominant, while at low flow rates saturation of the aqueous medium with respect to major matrix elements, particularly with respect to silica present in the glass and in its alteration products, becomes a controlling factor. Quantitative calculations indicate that under likely repository conditions the release rates can be expected, in general, to fall below the NRC criterion of 10−5 yr−1.
Investigation of the Hydrothermal Interaction of 99Tc-Doped Glass with Basalt Repository Nuclear Waste Package Components
Authors: Coles David G. , Simonson S. A. , Thomas L. E. , Schranke J. A. , McKinley S. G.
Published: March 2011

Hydrothermal experiments using basalt-repository nuclear waste package components have been carried out at 200°C and 30 MPa using 99Tc-doped PNL 76–68 waste glass. This work was conducted in support of the Basalt Waste Isolation Project. The experiments were carried out in rocking autoclaves that allowed for periodic solution sampling. Preliminary results that illustrated the effect of basalt on 99Tc solution behavior were discussed previously[l]. In this paper, we continue those observations by discussing the additional experiments that investigated the effects of steel. We also include the post-experiment solid phase analyses from all test configurations.

The effect of steel on glass dissolution was observed to be minimal when basalt was not present, i.e., there were no discernible differences in the amount of glass dissolved (based on boron release), with or without steel present.

The 99Tc solution concentration results showed that basalt, steel, and a combination of basalt and steel have an ability to dramatically lower the concentration of 99Tc in the solution, probably through a redox mechanism[2]. Solid run product analyses showed that without basalt present, a gel-like secondary phase consisting of an iron, zinc clay with an apparent smectite structure formed. When basalt was present clinoptilolite formed as the major secondary phase, due primarily to the presence of aluminum in the basalt mesostasis. Separation of the various solid run products and residual initial solid phases has not yet been achieved. Such a separation would facilitate the identification of the phase or phases with which 99Tc was associated. The solution results indicated that 99Tc, a potentially mobile radionuclide, may be incorporated in a relatively insoluble phase in the environment of a basalt repository. In addition, no synergistic effects between waste package components were observed that would increase the concentration of 99Tc in solution.
237Np aND 239Pu Solution Behavior During Hydrothermal Testing of Simulated Nuclear Waste Glass with Basalt and Steel
Authors: Schramke Janet A. , Simonson Scott A. , Coles David G.
Published: March 2011

A series of hydrothermal experiments were carried out on 237Np- and 239Pu-doped PNL 76–68 glass, synthetic basalt groundwater, basalt, and cast steel. These hydrothermal experiments are part of the Basalt Waste Isolation Project investigation of the interactions of waste package components in a basalt repository. Experiments of three months duration were conducted in Dickson-type rocking autoclaves at 200°C and 30 MPa, with an initial fluid to solid weight ratio of 10:1. All solid materials were ground and sieved to a narrow size fraction. The experiments carried out were: glass and groundwater; glass, basalt, and groundwater; glass, steel, and groundwater; and glass, steel, basalt, and groundwater. Unfiltered, 4000 Â filtered, and 18 Å filtered solutions were analyzed to determine the concentrations of radionuclides in solution and those associated with colloids.

The quantities of 237Np and 239Pu in solution were very small. Worstcase calculations indicate that 0.01% or less of the total radionuclide inventories were present in solution. The highest solution concentrations of the actinide dopants were observed in the experiments with basalt, even though smaller amounts of glass were dissolved than in the experiments without basalt. The observed differences in the solution concentrations of 237Np and 239Pu were probably controlled by differences in the reaction products, which were clinoptilolite in the experiments with basalt, and an Fe-Zn smectite clay in the experiments without basalt.
Solubility and Colloid Generation of Plutonium from Leaching of a HLW Glass in Salt Solutions
Authors: Kim J. I. , Treiber W. , Lierse Ch. , Offermann P.
Published: March 2011

Solubilities of Pu from leaching of a simulated HLW glass (C31–3-EC) spiked with 5 wt % Pu and from PuO2 are determined in salt solutions of different NaCl concentration and in groundwaters which are representative for the Gorleben area. Th.e leaching experiment is carried out either in an autoclave at 200 °C or under normal conditions (1 atm, 25°C). Solubilities of Pu in all investigated solutions are observed to be considerably higher than the values known from theoretical estimates. The predominant Pu species in solutions are found to be microcolloids, which are characterized by means of ultrafiltration and pulsed laser-induced photoacoustic spectroscopy.
Sorption-Capacity Limited Retardation of Radionuclides Transport in Water-Saturated Packing Materials
Authors: Pescatore C. , Sullivan T.
Published: March 2011

Radionuclides breakthrough times as calculated through constant retardation factors obtained in dilute solutions are non-conservative. The constant retardation approach regards the solid as having infinite sorption capacity throughout the solid. However, as the solid becomes locally saturated, such as in the proximity of the waste form-packing materials interface, it will exhibit no retardation properties, and transport will take place as if the radionuclides were locally non-reactive. The magnitude of the effect of finite sorption capacity of the packing materials on radionuclide transport is discussed with reference to high-level waste package performance. An example based on literature sorption data indicates that the breakthrough time may be overpredicted by orders of magnitude using a constant retardation factor as compared to using the entire sorption isotherm to obtain a concentrationdependent retardation factor.
Ceramic Containers for Spent Nuclear Fuel. I. Homogeneous Sealing of Rutile Containers at Low Temperatures
Authors: Forberg S , Bergman B
Published: March 2011

Some ceramic materials may withstand granitic ground water for millions of years, or, ideally be thermodynamically stable under these conditions. Some conceptual design criteria and materials selection guidelines are discussed with reference to ceramic containers for spent fuel.

Rutile was chosen for practical studies. One reason for this choice was the possibility of joining lid and container below 950°C. TiO2 with 0.1–1.0 w/o B, added as B or B2O3, was densified at 900°C by hot pressing; 0.25 w/o B was used for bonding.

TiO2 from a sol-gel process was, without additions, used for bonding at 900°C and at 800°C. A new TiO2 raw material was, without sintering aids, pressed to high density at 930°C, and also used for joining experiments.

Some bonds are shown in micrographs. The major result is that nearly homogeneous bonding at 900°C seems feasible. The new raw material might possibly be processed to a monolithic ceramic body outside the fuel without any need for sealing.
Ceramic Containers for Spent Nuclear Fuel. II. Reactions Between TiO2 and the Steel Canning During Hot Isostatic Processing
Authors: Bergman B. , Forberg S.
Published: March 2011

Rutile was selected for some practical studies of processing and properties of ceramic containers. Hot isostatic pressing at 1280°C has resulted in reaction zones between the TiO2 powder and the steel canning. The phases ilmenite, pseudobrookite, rutile, and iron have been identified by X-ray diffraction and by microprobe analysis.

The microstructures have been interpreted by classical metallographic methods, and some microstructures obtained by hot pressing and rapid cooling have also been examined for purposes of comparison.

Some implications of the microstructures have been discussed in terms of microcracking and slow crack growth.
Hydrothermal Studies of Simulated Defense Waste Glass Plus Basalt
Authors: Allen Carlton C. , Lane D. L. , Johnston R. G. , Marcy A. D. , Adee R. R.
Published: March 2011

The Basalt Waste Isolation Project (BWIP) is conducted for the U.S. Department of Energy (DOE) by Rockwell Hanford Operations (Rockwell). The BWIP has undertaken an experimental program to test the behavior of defense waste glass under conditions relevant to a repository in basalt. Three autoclave experiments utilizing a nonradioactive simulated waste glass (Savannah River Laboratory (SRL) Type 131/TDS-3A) have been completed. Monolithic discs and crushed glass particles were reacted with crushed basalt and synthetic basalt groundwater for ∼1 yr at temperatures of 90 °C to 150 °C. The crushed qlass proved considerably more reactive than the monoliths at similar temperatures. Raising the temperature from 90 °C to 150 °C produced a striking increase in the extent of glass alteration. At 150 °C, essentially all of the boron and most of the sodium in the glass particles were released. The major reaction products in all three experiments were smectite clay and zeolites. The two tests utilizing crushed glass also yielded analcime. Nickel, cesium, and strontium, important elements in defense waste, were released from the glass in small quantities. The nickel was incorporated into the clay, while the cesium was incorporated into the analcime.
Geoengineering Characterization of Welded Tuffs from Laboratory and Field Investigations
Authors: Zimmermian Roger M. , Nimick Francis B. , Board Mark P.
Published: March 2011

Welded tuff beneath Yucca Mountain adjacent to the Nevada Test Site (NTS) is being considered for development as a high-level radioactive waste repository by the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. Because access into Yucca Mountain has been limited to borehole explorations, early geoengineering materials characterizations have been derived From laboratory tests on cores from Yucca Mountain and from laboratory and field tests on welded tuffs located in G-Tunnel on the NTS. G-Tunnel contains welded tuffs that have similar properties and stress states to those at Yucca Mountain and has been the location for in situ rock mechanics testing. The purpose of this paper is to summarize the geoengineering material property data obtained to date and to compare appropriate laboratory and field data from G-Tunnel to findings from Yucca Mountain.

Geomechanical and thermal data are provided and are augmented by limited geological and hydrological data. A comparison of results of laboratory measurements on tuffs from Yucca Mountain and G-Tunnel indicates good agreement between the bulk densities, saturations, moduli of elasticity, Poisson's ratios, and P-wave velocities. The G-Tunnel tuff has slightly lower thermal conductivity, tensile strength, compressive strength and slightly higher matrix permeability than does the welded tuff near the proposed repository horizon at Yucca Mountain. From a laboratory-to-field scaling perspective, the modulus of deformation shows the most sensitivity to field conditions because of the presence of joints found in the field.
Basalt Glass: An Analogue for the Evaluation of the Long-Term Stability of Nuclear Waste form Borosilicate Glasses
Authors: Byers C. D. , Jercinovic M. J. , Ewing R. C. , Keil K.
Published: March 2011

The long-term stability of nuclear waste form borosilicate glasses can be evaluated by understanding the processes that effect the long-term alteration of glass and by comparing laboratory alteration of synthetic basalt and borosilicate glasses with the observed stability of naturally occurring basaltic glasses in diverse geologic environments. This paper presents detailed electron microprobe analyses of naturally altered basaltic glasses (with maximum ages of 10,000 to 20 million years) from low-temperature environments. These results are compared to laboratory data on the corrosion of a synthetic basaltic glass in MCC-1 tests (90°C, a SA/V of 0.1 cm−1 and time periods up to 182 days), MCC-2 tests (190°C, a SA/V of 0.1 cm−1 and time periods up to 210 days) and hydration tests in saturated water vapor (240°C, an estimated SA/V of ∼ 106 cm−1 and time periods up to 63 days). Additionally, laboratory induced hydration alteration of synthetic basalt and borosilicate glasses is compared. These preliminary experiments provide evidence that the alteration processes observed for natural basalt glasses are relevant to understanding the alteration of nuclear waste glass, as both appear to react via similar processes.
The Importance of Organic Compounds on the Modbilization and Bioassimilation of Thorium in the Morro do Ferro Environment
Authors: Miekeley N. , Dotto R. M. , Kuchler I. L. , Linsalata P.
Published: March 2011

Data are presented on the role of naturally occurring and synthetic organic compounds in the mobilization and bicassimilation of Th-232. The study is taking place in the state of Minas Gerais, Brazil, in and around an exposod and deeply weathered are body (the Morro do Ferro) estimated to contain 30,000 tonnes of Th and > 100,000 tonnes of the light rare earth elements. Indigenous plants collected from the are body surface are shown to contain high concentrations in the range of 20–1,000 μg/g ash of Th-232. Prcliminary estimates of Th-232 concentration factors range between 10−5 and 10−3 and are comparable with those reported for plutonium. Leaching studies using different chelating agents indicate that the release of Th-232 from. weathered soil and are as well as plant uptake are directly related to complex stability. Preliminary results of speciation studies using ultrafiltration techniques indicate that for water which has percolated through the are body, most of the dissolved Th-232 is associated with colloidal humic acids of molecular weight > 300,000.
Corrosion of Simulated Nuclear Waste Glass in a Gamma Radiation Field
Authors: Yokoyama H , Hermansson H-P , Christensen H , Björner I-K , Werme L
Published: March 2011

Two types of simulated nuclear waste glasses were leached and simultaneously gamma-irradiated. The leachate pH was reduced through irradiation, which enhanced glass leaching at room temperature. In contrast, the pH was maintained almost neutral in the experiment at 90°C and the leach rates were about five times lower than those for the glass leached at 90°C but out of the radiation field. Most of the leach rates correlated with the final pH of the leachate. Ground water, granite and bentonite moderates the effect of gammairradiation, probably because of a buffering effect. Simultaneous weathering and irradiation could greatly change the surface morphology of the glass.
Crystal Chemical Constraints on the Formation of Actinide Pyrochlores
Authors: Chakoumakos Bryan C. , Ewing Rodney C.
Published: March 2011

The pyrochlore structure type, A1−2 B2 O6Y0−1 Fd3m Z=8, is a common constituent of polyphase, crystalline waste forms. Naturally occurring minerals with the same structure often occur in the radiation damaged, electron-diffraction amorphous, “metamict” state; and therefore, a principal concern with phases of this structure type is the effect of alpha-recoil damage associated with the presence of actinides. In this paper we outline crystal chemical constraints on the formation of novel actinide pyrochlores.
Natural Pyrochlores: Analogues For Actinide Host Phases in Radioactive Waste Forms
Authors: Lumpkin Gregory R. , Ewing Rodney C.
Published: March 2011

Significant amounts of the pyrochlore structure type, and derivative structures like zirconolite, have been reported to occur in crystalline, polyphase, nuclear waste forms [1,2,3,4]. Pyrochlore has abundant and chemically diverse natural analogues. Many of these contain U and Th and have been transformed to the electron-diffraction amorphous, “metamict” state by alpha-recoil nuclei. A variety of geochemical alteration processes may occur before, during, and after reaching the final saturation level for alpha-recoil damage (∼1026 alpha-events/m3 ). The purpose of this paper is to examine these alteration effects in radiation-damaged natural pyrochlores.
Self-Radiation Damage in Actinide Host Phases of Nuclear Waste Forms
Authors: Weber W. J. , Wald J. W. , Matzke Hi.
Published: March 2011

Three crystalline ceramic materials, which occur as host phases for the long-lived actinides in many nuclear waste formulations, were doped with Cm-244, and the effects of self-radiation damage from alpha decay on microstructure and physical properties were investigated. The irradiation-induced microstructure consisted of individual amorphous tracks from both the alpha-recoil particles and the spontaneous fission fragments. The eventual overlap of the tracks at higher doses leads to a completely amorphous state. This radiation-induced amorphization process results in measured increases in volume, leachability, and stored energy. Thermal recovery of the radiation-induced swelling and amorphization occurs with full recrystallization to the initial structures.
Actinide Microdistributions Within Monazite
Authors: Jalajas E. W. , Kaufman A. , Ku T.-L.
Published: March 2011

The results of an induced fission track mapping study of uranium in 12 pegmatitic monazite samples indicate that as the U concentration in monazite surpasses ca. 300–400 ppm, the microdistribution becomes much less homogeneous resulting in localized areas of very high uranium concentration (hot spots) in a phase undetectable by XRD. Microprobe analysis of three additional pegmatitic monazites shows the occurrence of exsolution of the two other endmembers of the monazite solid-solution series, huttonite and cheralite, as indicated by a normative mineralogy computed from the probe data. The presence of each mineral is confirmed by XRD. Also supporting this exsolution hypothesis is a strong similarity between ratios of the structural elements of monazite (LREE, P) within the Th-rich hot spots as compared to these ratios in the surrounding monazite. Gradational contacts between banded monazite and a Th-rich phase can be observed microscopically as well as through microprobe scans for Th, indicating the close genetic relationship of the two phases. To prevent their enhanced leachabilities, actinide redistribution must be prevented in any single- or multi-phase nuclear waste form.
Chemical Implications of Heat and Radiation Damage to Rock Salt
Authors: Pederson Larry R.
Published: March 2011

Chemical changes induced in Palo Duro and Paradox Basin natural rock salts and in synthetic NaCl by heat and gamma radiation were investigated. Heating of unirradiated natural rock salts to 300°C resulted in HCI (most prevalent), SO2, CO2, and H2S evolution, and increased the base content of the remaining salt by not more than 10 microequivalents per gram; whereas, heating of synthetic NaCI gave no product. Gamma irradiation produced sodium colloids and neutral chlorine in amounts similar to the results of Levy and coworkers. When the irradiated salts were heated, three reactions were apparent: (1) radiation-induced defects recombined; (2) neutral chlorine was evolved; and (3) HCl, SO2, CO2, and H2S were evolved, similar to results for unirradiated salts. Because reaction (1) appeared to dominate over reaction (2), it is expected that the influence of radiation damage to salt on the near-field chemical environment will be minor.
Characterization of LWR Spent Fuel Approved Testing Materials for Radionuclide Release Studies
Authors: Barner John O. , Daniel J. L.
Published: March 2011

The Materials Characterization Center (MCC) provides characterized light-water reactor (LWR) spent fuel approved testing materials (ATMs) for use by experimenters in radionuclide release studies in support of repository licensing efforts. The characterization completed to date is described for MCC ATM-101, a Zircaloy-clad UO2 fuel from the H. B. Robinson, Unit 2, Assembly BO-5. ATM-101 is a moderate burnup, low-releasing (fission products in-reactor) spent fuel typical of the fuel that is expected to initially be placed in a repository.
High-Level Radioactive Insoluble Waste Preparation for Vitrification
Authors: Hamm B. A. , Eibling R. E. , Ebra M. A. , Motyka T. , Martin H. D.
Published: March 2011

At the Savannah River Plant (SRP), a process has been developed for immobilizing high-level radioactive waste in a borosilicate glass. The waste is currently stored as soluble salts and insoluble solids. Insoluble waste as stored requires further processing before vitrification is possible. The processes required have been developed and demonstrated with actual waste. They include removal of aluminum in some waste, washing soluble salts out of the insoluble waste, and mercury stripping. Each of the processes and the results with actual SRP waste will be discussed. The benefits of each step will also be included.
Technetium Removal Processes for Soluble Defense High-Level Waste
Authors: Walker Darrel D. , Bibler Jane P. , Wallace Richard M. , Ebra Martha A. , Ryan John P.
Published: March 2011

Two methods for removing technetium from soluble defense high-level waste are described. In the first method, technetium is precipitated as tetraphenylphosphonium pertechnetate and separated from the decontaminated solution using sintered metal crossflow filters. In the second method, pertechnetate is removed from solution using a strong base anion exchange resin and then eluted from the resin with nitric acid. The nitric acid is recovered by sorption of the pertechnetate on a weak base ion exchange resin. The pertechnetate is eluted from the weak base resin with NaOH and recovered by precipitation as the sulfide or oxide.
Determination of Early Stages of Glass Dissolution by pH Titration
Authors: Garland Judith A , White William B.
Published: March 2011

The early stages of glass dissolution are marked by rapid changes in pH for leach solutions with low buffer capacity. Consideration of both dissolution reactions and the electrical neutrality requirement allow the calculation of exchanged cation equivalents from the depletion of the H+-ion reservoir. Comparison was made between sodium trisilicate glass, a commercial waste glass (PNL 76−68), and the defense waste reference glass. The released Na+ calculated from the pH curve of sodium trisilicate glass has the expected parabolic dependence on time. Thenuclear waste glasses react with H+ more slowly and the pH - time plots do not have the shape of a titration curve. The equivalents of cations released follow a power law function, the exponent of which depends on initial pH. There are two distinct rate regimes for all glasses with initial fast rates related to consumption of H+ and later, slower rates nearly independent of H+ activity.
Powder Technological Vitrification of High Level Nuclear Waste by In-Can Hot Pressing Including a Sol-Gel Route
Authors: Caneiro A. , Ondracek G. , E.H. Toscano
Published: March 2011

Powder technology is used to immobilize high level nuclear waste (HLW) in sintered borosilicate glass. By uniaxial in-can hot pressing(temperature 950 K; pressure 1 MPa; heating rate ∼100 K/h; cooling rate 5 K/h), glass products containing simulated HLW(15 wt.%) have been produced in stackable steel cans (≤ 200 mm diameter). High densities, bulk integrities and homogeneities for the waste element distribution are realized. The advantages of powder technology are for example: (i) no segregation due to solid state vitrification, (ii) low evaporation losses and no compatibility problems due to low densification temperatures (sinter temperature ∼0.6 softening temperature of glass), (iii) production in easily arrangeable and interchangeable stacking units at modest pressures adequate especially for the use in hot cells under remote handling conditions, (iv) choosing units with alternatively HLWo r MLWin prescribed sequences to control the heat production of the package. As demonstrated now, the process is appropriate for high level waste (HLW), medium level waste (MLW) and mixed (HLW/ MLW vitrification, and is insensitive to waste modifications.

As the next step the effect of glass modifications was studied. In order to improve the glass leaching resistance and maintain low viscosity, silicon dioxide was doped with titanium dioxide. The mixture (referring to eutectic composition) was produced by a sol gel route and combined finally with 15 w/o HLW-oxides. Hot pressing of the obtained gels at 1273 K and 20 MPa provided highly homogeneous products with high densities.
Iodine Waste Forms: Calcium Aluminate Hydrate Analogues
Authors: Brown Dale R. , Grutzeck Michael W.
Published: March 2011

Phase relations in the system 3CaO·AI2O3-CaSO4-CaI2-H2O in equilibrium with excess water were established by means of room temperature bottle hydration of various bulk chemistries in the system. Starting with end members ettringite (3CaO·Al2O3·3CaSO4·32H2O) and tetracalcium aluminate monosulfate-12-hydrate (3CaO·Al2O3-CaSO4-12H2O), iodine-substituted analogue phases were synthesized which containe increasingly greater percentages of iodine. The iodine-substituted ettringite was found to be unstable whereas the iodine-substituted monosulfate formed readily. SEM, wet chemistry, IR, and x-ray diffraction characterization of the latter phase suggest that its formula is 3CaO·Al2O3·Ca(IO3)2·2H2O. Cement pellets containing this “Afm” iodine-substituted phase were subjected to a modified MCC-1 static leach test. Although the normalized iodine leach rate was relatively high when compared with AgI encapsulated in portland Type III cement, this same leach rate was approximately equal to the rates that have been reported for Ba(IO3)2, Ca(IO3)2, and Hg(IO3)2 in portland cement. The normalized iodine leach rate obtained also was found to be roughly comparable to that given for I-sodalite in cement. Diffusion is indicated as the primary leach mechanism, becoming dominant after the first three days of leaching.
Cementitious Mixtures for Sealing Access Shafts/Boreholes Through Evaporite and Clastic Rocks in a Radioactive-Waste Repository
Authors: Wakeley Lillian D. , Roy Della M. , Grutzeck M. W.
Published: March 2011

Geochemically compatible cement-based mixtures were designed for sealing a potential geologic repository for high-level radioactive waste in evaporite and related rock strata. The slightly expansive cementitious mixtures were based on Class H cement and Class C fly ash, with silica and calcium sulfate additives, and a low ratio of water to cementitious solids (w/s = 0.32). Both a salt-free and a salt-containing version were used in tests with anhydrite/siltstone and halite.

Although it is difficult to achieve strong bonding between anhydrite and cement-based mixtures, tensile bond strengths, interface permeability, and other properties of the composites as a whole indicate achievement of the desired chemical and physical compatibility between these materials Integrity of bonding and low permeability to water (on the order of 110 darcy) suggest the likelihood of attaining durable sealing of repository access shafts/boreholes through the use of such chemically tailored cementitious mixtures.

Rock samples used in ongoing sealing studies were obtained from various areas throughout the United States. The data in the present paper represent results for rocks obtained from the Palo Duro Basin and as such typify bedded-salt strata.
Characterization of Doped Si-TiSi2Bilayers Formed by Ion Beam Mixing and Rapid Thermal Annealing
Authors: Maex K. , de Keersmaecker R.F. , Alkemade P.F.A.
Published: March 2011

The use of rapid thermal processing is reported for simultaneous formation of TiSi2from Ti deposited layers and activation of As or Sb implanted profiles in Si. Properties of the silicide and the doped Si are reported with emphasis on impurity redistribution and defect removal.
Moving Species in Pt2Si, Ni2Si and CrSi2 Formed by Ion Mixing and Thermal Annealing
Authors: Zhao X. A. , Affolter K. , Nicolet M-A.
Published: March 2011

We have determined the dominant moving species during the formation of CrSi2, Ni2Si, and Pt2Si by irradiating with 600 keV Xe ions thin bilayers of evaporated metal on silicon and silicon on metal above room temperature. Two very thin markers were placed at the metal-silicon interface and at the bottom interface with the Si02 substrate. The separation of the two markers as a function of the amount of silicide formed by either ion mixing or thermal annealing was monitored by backscattering spectrometry. This experiment identifies the dominant moving species. The results establish that the moving species for ion mixing and for thermal annealing are the same in CrSi2, but not in Ni2Si and Pt2Si.
Electrical Characterization of Ion Imiplanted, Thermally Annealed TiN Films Acting as Diffusion Barriers on Shallow Junction Silicon Devices
Authors: Armigliato A. , Finetti M. , Gabilli E. , Guerri S. , Ostoja P. , Sabato G. , Scorzoni A.
Published: March 2011

TiN films prepared by implantation onto evaporated Ti layers are tested as diffusion barriers in contact with a thick Al overlayer. Both the TiN/Al and TiSi2/TiN/Al contact structures are evaluated, after thermal treatments up to 600 C, on shallow junction diodes and four terminal resistor test patterns for contact resistance measurements. It is shown that, upon annealing at 600 C, the TiSi2/TiN/Al contact system still exhibits excellent electrical performances. The degradation is found to depend on TiSi2 thickness and contact area.