bluuebunny/crossref_metadata_2025 huggingface dataset:
Grain Boundary Precipitate Density as a Function of Time and Misorientation in an Al-5 WT% Cu Alloy
Authors: Cantrell M. A. , Shiflet G. J.
Published: March 2011

The variation of θ (CuAI2) precipitate density as a function of heat treatment time and grain boundary misorientation was investigated in an Al-5 wt % Cu alloy. In this study, precipitate densities have been quantitatively linked to grain boundary structure. It was found that, for a given heat treatment time, the precipitate density varied in a reproducible manner (108 to 1010 ppts/cm2) as a function of misorientation between the grains (20 to 60 degrees). Additionally, misorientation was found to be the most important factor governing the precipitate density at a given grain boundary. The grain boundary plane orientation played a secondary role in determining the precipitate density. Bollmann O-lattice modeling allows comparison of the relative effects of grain boundary plane orientation relative to the misorientation between grains.
Computer Simulation of Vacancy Segregation at Antiphase Domain Boundaries During Coarsening
Authors: Chen Long-Qing
Published: March 2011

A computer simulation technique based on the Master Equation Method (MEM) is developed for modeling the spatial distribution of vacancies during ordering and subsequent domain coalescence and coarsening. A vacancy mechanism is assumed for the atomic diffusion and the single-site approximation is employed. It is demonstrated that vacancies strongly segregate into the antiphase domain boundaries (APBs) during coarsening, resulting in the vacancy concentration at APBs more than an order of magnitude higher than that inside the ordered domains. As the antiphase domains coarsen, the vacancy concentration at the APBs continues to increase and its spatial s segregation profile moves accompanying the APB migration. The effect of vacancy concentration on the antiphase domain coarsening kinetics is discussed.
Interface Controlled Amorphization of Crystalline Ni/Ti Multilayers
Authors: Jankowski Alan F. , Hayes Jeffrey P. , Ramsey Philip B.
Published: March 2011

Solid-State Amorphization (SSA) of crystalline interfaces is observed in the Ni/Ti multilayer system. The amorphization reaction nucleates at location(s) of crystallographic disorder, i.e. the multilayer interfaces. Microstructural analyses reveal the sputter-deposited growth structure to be epitaxial with semi-coherent interfaces. Strain energy originating from interface lattice distortions varies as a function of the multilayer repeat spacing. Therefore, the interfacial energy effects the onset conditions for SSA. Differential thermal analysis is used to measure the critical temperature Tc, to the nucleation of the SSA, which is found to vary with the Ni/Ti multilayer pair spacing.
Evidence by Hrem of Interfacial Modification Induced by Equilibrium Segregation in GE(S) Bicrystal
Authors: Charai A. , Rouviere J.L.
Published: March 2011

The structure of a [001] tilt grain boundary Σ=25(710) in germanium was investigated by HREM as prepared and after sulfur segregation. The characterisation of Ge(S) system was performed using techniques such as Auger Electron Spectroscopy, Electron Energy Loss Spectroscopy and radioactive tracer analysis. Before the sulfur introduction, this grain boundary is usually sensitive to the preferential etching during the TEM sample preparation which makes the observation difficult. This effect, depending on experimental conditions, can either partially or completely disappear. The interfacial structure as prepared and after annealing under argon or under hydrogen atmosphere can be described as a stacking of several structures with different energies. After sulfur segregation no preferential etching is observed because the lowest energy structure of the as prepared grain boundary is stabilized. The tilt angle and the plane of the grain boundary remain constant.
Correlation of Electric Parameters Change and Structural Changes Induced in Silicon Systems by Pulsed Magnetic Field Treatment
Authors: Levin Mark N. , Maslovsky Vladimir M.
Published: March 2011

The results of comparison investigations of structural and electric parameters changes in silicon systems induced by pulsed magnetic field (MF) treatment (PMFT) are presented for the first time. The characteristics of (PMFT) that can induce considerable parameters changes of the silicon system were determined. Amplitudc of thc magnctic impulscs is 0.1-0.3 MA/m and duration of thc impulscs is 10-30 ms. The investigations were carried out by means of scanning electron microscope (SEM), X-ray diffraction analysis, C-V and DLTS spectroscopies. The PMFT induces the generation of A-centers in the near-surface region of silicon, the changes of the crystal lattice parameter and the concentration of free electrons and results in emergence of an extent structural microdefects in subsurface. The obtained experimental data testifies that PMFT is possible to increase the vacancy concentration at subsurface region of silicon.
Effect of Interface on Capture and Emission Processes Via Deep Centers
Authors: Yassievich I. N. , Pakhomov A. A.
Published: March 2011

It is shown that the carrier capture and emission rates by deep centers situated near an interface (crystal surface or heterojunction) dramatically increase.
Transition From Inclined to In-Plane 60° Misfit Dislocations in a Diffuse Interface
Authors: Ning X. J. , Pirouz P.
Published: March 2011

Despite tremendous activity during the last few decades in the study of strain relaxation in thin films grown on substrates of a dissimilar material, there are still a number of problems which are unresolved. One of these is the nature of misfit dislocations forming at the film/substrate interface: depending on the misfit, the dislocations constituting the interfacial network have predominantly either in-plane or inclined Burgers vectors. While, the mechanisms of formation of misfit dislocations with inclined Burgers vectors are reasonably well understood, this is not the case for in-plane misfit dislocations whose formation mechanism is still controversial. In this paper, misfit dislocations generated to relax the strains caused by diffusion of boron into silicon have been investigated by plan-view and crosssectional transmission electron microscopy. The study of different stages of boron diffusion shows that, as in the classical model of Matthews, dislocation loops are initially generated at the epilayer surface. Subsequently the threading segments expand laterally and lay down a segment of misfit dislocation at the diffuse interface. The Burgers vector of the dislocation loop is inclined with respect to the interface and thus the initial misfit dislocations are not very efficient. However, as the diffusion proceeds, non-parallel dislocations interact and give rise to product segments that have parallel Burgers vectors. Based on the observations, a model is presented to elucidate the details of these interactions and the formation of more efficient misfit dislocations from the less-efficient inclined ones.
Transmission Electron Microscopy Studies of a Complex Oxide System
Authors: Chen H. , Sung C. , Wei G.C.
Published: March 2011

A glass ceramic based on the CaO-A12O3-BaO-MgO-B2O3 system was investigated using transmission electron microscopy in order to understand interactions among multiple oxide phases. The microstructure consisted of various crystalline and amorphous phases with stacking faults and dislocation loops in some of the crystalline phases. The microstructural development and phase assemblage of the complex system appeared to be related to defect-interface interactions during heating and cooling processes.
Characteristics of Pdai Schottky Contacts to n-GaAs
Authors: Huang T.S. , Pang J.G.
Published: March 2011

Metallurgical and electrical properties of β-phase PdAl Schottky metallizations on n-GaAs after rapid thermal annealing for 20 s in the temperature range 500-1000°C have been investigated using x-ray diffraction, transmission electron microscopy, Auger depth profiling and current-voltage measurement. The Al-rich contacts were stable up to 900°C, whereas the Pdrich contacts were less stable. The thermal stability of Pd-rich contacts decreased with increasing Pd composition, and interfacial reaction after high temperature annealing resulted in the formation of PdGa compound. The interface between Al-rich PdAI and GaAs substrate was quite sharp even after 900°C anneal. The Schottky barrier heights of Al-rich PdAl contacts increased with annealing temperature. The barrier height enhancement in the annealed Al-rich contacts can be attributed to the thin AlxGal−xAs layer formed at the interface between PdAl and GaAs.
The Development of Solid Phase Regrowth on GaAs and its applications
Authors: Wang L. C.
Published: March 2011

A solid phase regrowth process on GaAs has been observed in Pd- and Ni- based bi-layer structures, e.g. the Si/Ni, the Ge/Pd, the In/Pd, and the Sb/Pd structures. Due to the regrowth, uniform epitaxial layers of Ge, GaAs, InxGa1-xAs, and GaSbl-xAsx on GaAs substrates by solid phase reactions can achieved. The model of this regrowth process will be presented. Based on this regrowth mechanism, a series of non-spiking planar ohmic contacts on n and p type GaAs have been developed. Low contact resistivity in the range of mid 10−7 Ω-cm2 was obtained. The ohmic contact formation mechanism of these contacts will also be discussed. All the studies suggest that the ohmic behavior is a result of the formation of an n+ or p+ surface layer via solid phase reactions. The regrowth process has also been utilized to achieve compositional disordering of GaAs/AlGaAs superlattices, and low loss AlGaAs/GaAs waveguide has been obtained.
Radionuclide Decay Effects on Waste Glass Corrosion
Authors: Wronkiewicz D. J.
Published: March 2011

The release of glass components into solution, including radionuclides, may be influenced by the presence of radiolytically produced nitric acid, carboxylic acid, and transient water dissociation products such as ·OH and O2-. Under batch test conditions, glass corrosion has been shown to increase up to a maximum of three-to five-fold in irradiated tests relative to nonirradiated tests, while in other studies the presence of radiolytic products has actually decreased glass corrosion rates. Bicarbonate groundwaters will buffer against pH decreases and changes in corrosion rates. Under high surface area-to-solution volume (S/V) conditions, the bicarbonate buffering reservoir may be quickly overwhelmed by radiolytic acids that are concentrated in the thin films of water contacting the samples. Glass reaction rates have been shown to increase up to 10-to-15-fold due to radiation exposure under high S/V conditions.

Radiation damage to solid glass materials results in bond damage and atomic displacements. This type of damage has been shown to increase the release rates of glass components up to four-fold during subsequent corrosion tests, although under actual disposal conditions, glass annealing processes may negate the solid radiation damage effects.
Waste Glass Alteration Processes, Surface Layer Evolution and Rate Limiting Steps
Authors: Clark D. E. , Schulz R. L. , Wicks G. G. , Loading A. R.
Published: March 2011

This paper provides an overview of waste glass alteration. Emphasis is on the evolution of surface layers and factors affecting the alteration rate when glass is subjected to an aqueous environment. The extent, type and rate of alteration is determined by a variety of parameters including time, temperature, glass composition, pH, Eh, composition of leachate/geology, the presence of other engineered barriers, flow conditions and surface area/volume ratio. Models (based on kinetic and thermodynamic considerations) developed by other researchers that are consistent with the experimental observations from the laboratory and field experiments are discussed. The morphology of the surface layers as described by interpretation of secondary ion mass spectrometry (SIMS) analyses on waste glasses and the role the various layers play in the alteration process is presented as well.
In Situ Testing of Waste Glass in Clay
Authors: Van Iseghem Pierre Ph.
Published: March 2011

The actual status of an in-situ test programme exposing different waste glass samples directly to Boom clay is reviewed. Corrosion test tubes have been retrieved after residence for 5 years at 16° C, 2 years at 90° C, and 5 years at 170° C. The corrosion is interpreted in terms of mass loss, surface analysis by SEM and profiling by EPMA and SIMS. At 16° C, glasses dissolve about 0.02 – 0.08 µm per year. At higher temperature dissolution is more than two orders of magnitude larger. A good agreement is obtained between the mass losses and the surface analyses. The advantages and limitations of the Belgian in-situ tests are compared with the conclusions of an international expert group.
The Role of Natural Glasses as Analogues in Projecting the Long-Term Alteration of High-Level Nuclear Waste Glasses: Part 1
Authors: Mazer James J.
Published: March 2011

The common observation of glasses persisting in natural environments for long periods of time (up to tens of millions of years) provides compelling evidence that these materials can be kinetically stable in a variety of subsurface environments. This paper reviews how natural and historical synthesized glasses can be employed as natural analogues for understanding and projecting the long-term alteration of high-level nuclear waste glasses. The corrosion of basaltic glass results in many of the same alteration features found in laboratory testing of the corrosion of high-level radioactive waste glasses. Evidence has also been found indicating similarities in the rate controlling processes, such as the effects of silica concentration on corrosion in groundwater and in laboratory leachates. Naturally altered rhyolitic glasses and tektites provide additional evidence that can be used to constrain estimates of long-term waste glass alteration. When reacted under conditions where water is plentiful, the corrosion for these glasses is dominated by network hydrolysis, while the corrosion is dominated by molecular water diffusion and secondary mineral formation under conditions where water contact is intermittent or where water is relatively scarce. Synthesized glasses that have been naturally altered result in alkali-depleted alteration features that are similar to those found for natural glasses and for nuclear waste glasses. The characteristics of these alteration features appear to be dependent on the alteration conditions which affect the dominant reaction processes during weathering. In all cases, care must be taken to ensure that the information being provided by natural analogues is related to nuclear waste glass corrosion in a clear and meaningful way.
Remaining Uncertainties in Predicting Long-Term Performance of Nuclear Waste Glass From Experiments
Authors: Grambow B. , Karlsruhe Kernforschungszentrum
Published: March 2011

The current knowledge on the glass dissolution mechanism and the representation of glass dissolution concepts within overall repository performance assessment models are briefly summarized and uncertainties related to mechanism, radionuclide chemistry and parameters are discussed. Understanding of the major glass dissolution processes has been significantly increased in recent years. Long-term glass stability is related to the long-term maintenance of silica saturated conditions. The behavior of individual radionuclides in the presence of a dissolving glass has not been sufficiently and results do not yet allow meaningful predictions. Conservative long-term predictions of glass matrix dissolution as upper limit for radionuclide release can be made with sufficient confidence, however these estimations generally result in a situation were the barrier function of the glass is masked by the efficiency of the geologic barrier. Realistic long-term predictions may show that the borosilicate waste glass contributes to overall repository safety to a much larger extent than indicated by overconservatism. Today realistic predictions remain highly uncertain and much more research work is necessary. In particular the long-term rate under silica saturated conditions needs to be understood and the behavior of individual radionuclides in the presence of a dissolving glass deserves more systematic investigations.
The Lack of Technical Basis for Requiring a Ten Thousand Year Prediction for Nuclear Waste Management
Authors: Ramspott Lawrence D.
Published: March 2011

There is no technical basis for setting a time limit of 10,000 years on the regulated performance of a nuclear waste repository. First, accurate prediction of releases for such periods is not possible. Second, there is nothing unique about 10,000 years. Third, equally toxic materials, which never transform to non-toxic substances by radioactive decay, have less stringent requirements. And fourth, during a 10,000 year time frame, natural disasters will dwarf the worst possible outcomes of repository placement.

Analyses could be required to extend as long as doses above current radiation protection guidelines are possible (perhaps several million years), but these results should be recognized as qualitative information rather than evidence of quantitative compliance with exact numerical limits. Concern for what will happen over long times can be addressed for the next several hundred years by maintaining waste retrievability. At that time, uncertainty about future performance should have been reduced significantly.
Objectives and Limitations of Scientific Studies with Reference to the Swedish RD&D Programme 1992 for Handling and Final Disposal of Nuclear Waste
Authors: Sjöblom Rolf , Dverstorp Björn , Wingefors Stig
Published: March 2011

The Swedish Nuclear Power Inspectorate (SKI) has recently concluded its evaluation of the Swedish programme for the development of a system for the management of nuclear waste. The programme was compiled and issued by the Swedish Nuclear Fuel and Waste Management Company (SKB). In this process of programme formulation and review, considerable attention has been paid to the question of how scientific studies should be directed and performed in order to provide the support needed in the programme.

When the objectives and limitations of such scientific studies are to be analysed, it is vitally important that the implications of the long timescales and the lack of feedback of post-closure experience are fully understood and adequately dealt with. It must be realised, that simulations will have to be utilized for the systems development work as well as for the assessments of safety. Such performance simulations have to be based on a thorough understanding of the pertinent phenomena as well as on a comprehensive base of experimental data. They must also be based on a good overall perspective that can only be obtained through full performance assessments.

Thus, a sound scientific base is needed. The components of this base must be traceable. Moreover, the links must be clearly identified between the knowledge base and the development of the disposal system as well as the performance assessment of such a system. Performance simulations and performance assessments should be utilized, not only as tools to develop a system and to evalute its safety but also to identify areas where further research is required - or unnecessary.

The knowledge base will, however, always be limited, and a lack of understanding must also be recognised and adequately dealt with, e.g. by accounting for uncertainties in the performance simulations, by relying on bounding (conservative) assumptions or by robust repository design.

Thus, the success of a programme for the handling and disposal of nuclear waste is highly dependent on the strategy applied for the utilization of the scientific knowledge base. The requirements on such a strategy increase considerably when substantial commitments are to be made.
Radiation Defects and Energy Storage in Natural Polycrystalline Rock Salt. Results of an In-Situ Test in the Permian Rock Salt of the Asse
Authors: Gies H. , García Celma A. , de Haas J. B. M. , Pederson L. , Rothfuchs T.
Published: March 2011

Radiation damage development and the corresponding energy storage in pure undeformed single crystals have frequently been studied in laboratory experiments, however little is known of irradiation experiments on natural rock salt (polycrystalline, deformed and impure) under geological conditions. The relevance of these parameters to the defect formation was revealed by a joint United States/Federal Republic of Germany in-situ test in the Asse Mine. Natural rock salt was heated and irradiated using Co-60 sources.

Calculations of the amount of halite expected to be decomposed by radiolysis during the experiment were performed using the 1985 version of the Jain-Lidiard model. Qualitative agreement between theory and analyses was found for all the performed analyses. Quantitative and qualitative deviations of the natural samples behaviour from that of single undeformed crystals were observed and attributed to the influence of sulfatic admixtures, ‘polycrystallinity’ and strain on radiation damage development and anneal.

Special chemical methods, such as iodometric titration and uv-visible spectroscopy were applied in order to measure the hypochlorite ion, which forms in irradiated salt. Composite samples located closest to the Co-60 source averaged 0,4 micromoles neutral chlorine atoms per gram salt, a factor of two more than from other positions. Uv-vis analyses revealed more than a factor of ten greater neutral chlorine concentrations in coloured halite. Similarly, optical absorption measurements indicated a factor of ten difference in sodium metal colloid concentrations.
Radiation Stability of Ceramic Waste Forms Determined by In Situ Electron Microscopy and He Ion Irradiation
Authors: White T.J. , Mitamura H. , Hojou K. , Furuno S.
Published: March 2011

The radiation stability of polyphase titanate ceramic waste forms was studied using analytical transmission electron microscopy, in combination with in situ irradiation by 30 keV He ions, followed by staged annealing. Two experiments were conducted. In the first, a reconnaissance investigation was made of the stabilities of the synthetic minerals hollandite, zirconolite, and perovskite when subjected to a total dose of 1.8 × 1017 He+ cm-2. It was found that all phases amorphized at approximately the same rate, but perovskite recovered its structure more rapidly and at lower temperatures than the other phases. In particular, annealing for 10 minutes at 1000°C was sufficient for perovskite to completely regain its crystallinity, while zirconolite and hollandite were only partially restored by these conditions. In the second experiment, the response of a thin hollandite crystal to irradiation was examined by selected area electron diffraction. At a dose of 1.5 × 1015 He+ cm-2its incommensurate superstructure was disrupted, but even at a dose of 3 × 1016 He cm-2 the hollandite subcell was largely intact. For this dose, total recovery was achieved by annealing for 1 minute at 1000°C.
Effects of Radiation Exposure on SRL 131 Composition Glass in a Steam Environment
Authors: Wronkiewicz D. J. , Bradley C. R. , Bates J. K. , Wang L. M.
Published: March 2011

Monoliths of SRL 131 borosilicate glass were irradiated in a saturated air-steam environment, at temperatures of 150°C, to examine the effects of radiation on nuclear waste glass behavior. Half of the tests used actinide and Tc-99 doped glass and were exposed to an external ionizing gamma source, while the remaining glass samples were doped only with uranium and were reacted without any external radiation exposure. The effects of radiation exposure on glass alteration and secondary phase formation were determined by comparing the reaction rates and mineral paragenesis of the two sets of samples.

All glass samples readily reacted with the water that condensed on their surfaces, producing two types of smectite clay within the first three days of testing. Additional crystalline phases precipitated on the altered glass surface with increasing reaction times, including zeolites, smectite, calcium and sodium silicates, phosphates, evaporitic salts, and uranyl silicates. Similar phases were produced on both the nonirradiated and irradiated samples; however, the quantity of precipitates was increased and the rate of paragenetic sequence development was accelerated in the latter. After 56 days of testing, the composite smectite layer developed at an average rate of ~0.16 and 0.63 µm/day for the nonirradiated and irradiated samples, respectively. These comparisons indicate that layer development is accelerated approximately four-fold due to the radiation exposure at high glass surface area/liquid volume (SA/V) conditions. This increase apparently occurs in response to the rapid concentration of radiolytic products, including nitric acid, in the thin films of water contacting the sample monoliths.
Immobilization in Ceramic Waste Forms of the Residues From Treatment of Mixed Wastes
Authors: Oversby V. M. , Van Konynenburg R. A. , Glassley W. E. , Curtis P. G.
Published: March 2011

The Environmental Restoration and Waste Management Applied Technology Program at LLNL is developing a Mixed Waste Management Facility to demonstrate treatment technologies that provide an alternative to incineration. As part of that program, we are developing final waste forms using ceramic processing methods for the immobilization of the treatment process residues. The ceramic phase assemblages are based on using Synroc D as a starting point and varying the phase assemblage to accommodate the differences in chemistry between the treatment process residues and the defense waste for which Synroc D was developed. Two basic formulations are used, one for low ash residues resulting from treatment of organic materials contaminated with RCRA metals, and one for high ash residues generated from the treatment of plastics and paper products. Treatment process residues are mixed with ceramic precursor materials, dried, calcined, formed into pellets at room temperature, and sintered at 1150 to 1200°C to produce the final waste form. This paper discusses the chemical composition of the waste streams and waste forms, the phase assemblages that serve as hosts for inorganic waste elements, and the changes in waste form characteristics as a function of variation in process parameters.
Rietveld Analysis of Ceramic Nuclear Waste Forms
Authors: White T.J. , Mitamura H.
Published: March 2011

Powder X-ray diffraction patterns were collected from three titanate waste forms - a calcine powder, a prototype ceramic without waste, and a ceramic containing 10 wt% JW-A simulated waste - and interpreted quantitatively using the Rietveld method. The calcine consisted of fluorite, pyrochlore, rutile, and amorphous material. The prototype waste form contained rutile, hollandite, zirconolite and perovskite. The phase constitution of the JW-A ceramic was freudenbergite, loveringite, hollandite, zirconolite, perovskite and baddeleyite. Procedures for the collection of X-ray data are described, as are assumptions inherent in the Rietveld approach. A selection of refined crystal data are presented.
High-Level Waste Glass Compendium; What it Tells us Concerning the Durability of Borosilicate Waste Glass
Authors: Cunnane J. C. , Allison J. M.
Published: March 2011

Facilities for vitrification of high-level nuclear waste in the United States are scheduled for startup in the next few years. It is, therefore, appropriate to examine the current scientific basis for understanding the corrosion of high-level waste borosilicate glass for the range of service conditions to which the glass products from these facilities may be exposed. To this end, a document has been prepared which compiles worldwide information on borosilicate waste glass corrosion. Based on the content of this document, the acceptability of canistered waste glass for geological disposal is addressed.

Waste glass corrosion in a geologic repository may be due to groundwater and/or water vapor contact. The important processes that determine the glass corrosion kinetics under these conditions are discussed. Testing data together with understanding of the long-term corrosion kinetics are used to estimate radionuclide release rates. These rates are discussed in terms of regulatory performance standards.
Performance of Cementitious Systems in the Repository
Authors: Atkins M. , Damidot D. , Glasser F.P.
Published: March 2011

The behaviour and performance of cement barriers in the repository have been modelled. Since the models employ thermodynamic calculations, an important task is the development of a validated database. This must include phases and aqueous species appropriate to cement and modified cements, containing tly ash and slag. Closed system mineralogy and chemistry are explored as a function of cure duration and thermal excursions. Work has only just begun on open system behaviour: some aspects of reactions with chloride, sulfate and carbonate are described. Future progress depends on database extensions, validation experiments and the coupling of equilibrium and kinetic aspects of reaction.
Ancient Analogues Concerning Stability and Durability of Cementitious Wasteform
Authors: Jiang W. , Roy D.M.
Published: March 2011

The history of cemenütious materials goes back to ancient times. The Greeks and Romans used calcined limestone and later developed pozzolanic cement by grinding together lime and volcanic ash called “pozzolan” which was first found near Port Pozzuoli, Italy. The ancient Chinese used lime-pozzolanic mixes to build the Great Wall. The ancient Egyptians used calcined impure gypsum to build the Great Pyramid of Cheops. The extraordinary stability and durability of these materials has impressed us, when so much dramatically damaged infrastructure restored by using modern portland cement now requires rebuilding. Stability and durability of cementitious materials have attracted intensive research interest and contractors’ concerns, as does immobilization of radioactive and hazardous industrial waste in cementitious materials. Nuclear waste pollution of the environment and an acceptable solution for waste management and disposal constitute among the most important public concerns. The analogy of ancient cementitious materials to modern Portland cement could give us some clues to study their stability and durability. This present study examines selected results of studies of ancient building materials from France, Italy, China, and Egypt, combined with knowledge obtained from the behavior of modern portland cement to evaluate the potential for stability and durability of such materials in nuclear waste forms.
Accelerated Ageing of Blended OPC Cements
Authors: Quilin K.C. , Duerden S.L. , Majumdar A.J.
Published: March 2011

An accelerated experimental technique using high wateπcement ratios has been developed to study the long term hydration of blended cements that may be used in a repository for the disposal of radioactive waste. This technique has been used to investigate the hydration reactions of Ordinary Portland Cement (OPC) blended with ground granulated blastfurnace slag (ggbs) or pulverised fuel ash (pfa). The effects of high sulphate-bearing and high carbonate-bearing ground waters on the compounds formed on hydration were investigated. Solid/solution compositional data were collected during the course of the hydration process for periods up to 2 years. Thomsonite, thaumasite, afwillite and a tobermorite-like phase were found in addition to the expected cement hydration products. The pH of the aqueous solution in contact with 60 pfa:40 OPC blends hydrated at 90°C fell to below 8. This is lower than the value required to inhibit the corrosion of steel canisters in a repository. The pH of the aqueous solution in contact with OPC and 75 ggbs:25 OPC blends remained above 11, although if the ground waters in contact with the OPC/ggbs blends were periodically replaced the pH eventually fell below 10.
Development of Methodology to Evaluate Microbially Influenced Degradation of Cement-Solidified Low-Level Radioactive Waste Forms
Authors: Rogers Robert D. , Hamilton M.A. , Veeh R.H. , Mcconnell J.W.
Published: March 2011

Because of its apparent structural integrity, cement has been widely used in the United States as a binder to solidify Class B and C low-level radioactive waste (LLW). However, the resulting cement preparations are susceptible to failure due to the actions of stress and environment. An environmentally mediated process that could affect cement stability is the action of naturally occurring microorganisms. The U.S. Nuclear Regulatory Commission (NRC), recognizing this eventuality, stated that the effects of microbial action on waste form integrity must be addressed.

This paper provides present results from an ongoing program that addresses the effects of microbially influenced degradation (MID) on cement-solidified LLW. Data are provided on the development of an evaluation method using acid-producing bacteria. Results are from work with one type of these bacteria, the sulfur-oxidizing Thiobacillus. This work involved the use of a system in which laboratory- and vendor-manufactured, simulated waste forms were exposed on an intermittent basis to media containing thiobacilli. Testing demonstrated that MID has the potential to severely compromise the structural integrity of ion-exchange resin and evaporator-bottoms waste that is solidified with cement. In addition, it was found that a significant percentage of calcium and other elements were leached from the treated waste forms. Also, the surface pH of the treated specimens decreased to below 2. These conditions apparently contributed to the physical deterioration of simulated waste forms after 60 days of exposure to the thiobacilli.
Modeling of UO2 Aqueous Dissolution Over a Wide Range of Conditions
Authors: Steward Steven A. , Weed Homer C.
Published: March 2011

Previously it was not possible to predict reliably the rate at which spent fuel would react with groundwater because of conflicting data in the literature. The dissolution of the UO2 spent fuel matrix is a necessary step for aqueous release of radioactive fission products. Statistical experimental design was used to plan a set of UO2 dissolution experiments to examine systematically the effects of temperature (25-75°C), dissolved oxygen (0.002-0.2 atm overpressure), pH (8-10) and carbonate (2-200x10-4 molar) concentrations on UO2 dissolution. The average dissolution rate was 4.3 mg/m2/day. The regression fit of the data indicate an Arrhenius type activation energy of -8750 cal/mol·K and a half-power dependence on dissolved oxygen in the simulated groundwater.
The Effect of Fuel Burnup and Dispersed Water Intrusion on the Criticality of Spent High-Level Nuclear Fuel in a Geologic Repository
Authors: Culbreth William G. , Zielinski Paige R.
Published: March 2011

Studies of the spent fuel waste package have been conducted through the use of a Monte-Carlo neutron simulation program to determine the ability of the fuel to sustain a chain reaction. These studies have included fuel burnup and the effect of water mists on criticality. Results were compared with previous studies.

In many criticality studies of spent fuel waste packages, fresh fuel with an enrichment as high as 4.5% is used as the conservative (worst) case. The actual spent fuel has a certain amount of “burnup” that decreases the concentration of fissile uranium and increases the amount of radionuclides present. The LWR Radiological Data Base from OCRWM has been used to determine the relative radionuclide ratios and KENO 5.a was used to calculate values of the effective multiplication factor, keff.1

Spent fuel is not capable of sustaining a chain reaction unless a suitable moderator, such as water, is present. A completely flooded container has been treated as the worst case for criticality. Results of a previous report that demonstrated that keff actually peaked at a water-to-mixture ratio of 13% were analyzed for validity. In the present study, these results did not occur in the SCP waste package container.
Effect of Fluoride on Crystallization in High Calcium and Magnesium Glasses
Authors: Wang E. , Kuang H. , Matlack K.S. , Buechele A.C. , Fu Sabrina S.
Published: March 2011

Glasses containing fixed mutual ratios of Al, B, Ca, Fe, Mg, Na, and Si were melted while varying the fluoride content. The effect of fluoride was to suppress the formation of non-fluoride-containing crystals in the composition range from 0 to 11 wt. % F; above 11 wt. % F, the formation of fluoride containing crystals began to define the liquidus temperature. The effect of fluoride on the liquidus temperature was quantified and used to predict the liquidus temperature of melts made from real radioactive wastes. Although the liquidus temperatures of the surrogate and radioactive glasses agreed well, the types of crystals formed differed due to slight compositional differences introduced by the waste stream.
Liquid-Liquid Phase Separation and Phase Diagrams of Simulated Fernald Waste Glasses
Authors: Wang E. , Perez-Cardenas F. , Kuang H. , Buechele A.C.
Published: March 2011

Previous study of crystallization behavior in heat-treated Fernald waste glasses has produced an extensive data base of crystal phases likely to appear in various composition ranges and their corresponding liquidus temperatures. In addition, we have frequently observed amorphous phase separation in these glasses and, occasionally, evidence of crystallization originating from such phase separation. These glasses contain more than 10 components. The composition ranges for the major components are: MgF2 10–26 wt%; CaO 4–27 wt%; Al2O3 3–15 wt%; SiO2 25–40 wt%. The morphology of the phase separation as observed in the Scanning Electron Microscope (SEM) is dark, spherical globules dispersed in a continuous matrix. Globules are depleted in Mg, Ca and F, and enriched in Al and Si compared to the matrix. Phase separation occurs more frequently in melts relatively higher in Si and F. A more systematic study on a simplified and simulated seven component system (Al2O3, B2O3, CaO, Fe2O3, SiO2, Na2O and MgF2) has been undertaken to determine the subliquidus miscibility gap and liquidus curve data. Glasses were formulated by varying the concentrations of MgF2, CaO, Al2O3 and SiO2 within the ranges specified above at fixed levels of Fe2O3, B2O3 and Na2O. The miscibility gap and liquidus curve were obtained by heat-treating the glass samples at different temperatures and observing any phase separation and crystallization in the SEM and the Transmission Electron Microscope (TEM). We report here the results of this study to enhance the understanding of the thermodynamic properties of multi-component silicate systems which are usually the basis of nuclear waste glasses.
Continuous-Melter Vitrification of Fernald Mixed Wastes: Distribution of Contaminants Between Glass, Scrubber Solutions and Off-Gases
Authors: Matlack Keith S. , Hojaji Hamid , Fu Sabrina S. , Pegg Ian L. , Macedo Pedro B.
Published: March 2011

Results are presented from treatability studies on the vitrification of hazardous and low-level radioactive soils and sludges from the Fernald site. The majority of the contaminants of concern are contained in large volumes of sludges accumulated from uranium and thorium production operations that are presently stored in pits at Fernald. These sludges present particular challenges to the vitrification process in that they contain significant amounts of fluorides (mostly magnesium and calcium). In this study, glass formulations used sludges, soils and a minimum amount of additives resulting in waste loadings as high as 94%. Several of these formulations were evaluated by continuous feeding of slurries into a joule-heated vitrification system. Melter off-gases were treated in a two-stage NaOH scrubbing system. HF acid fumes were neutralized, resulting in the formation of a NaF solid precipitate which is removed from the scrubber and recycled to subsequent melter feeds. Elemental and isotope compositions of the glasses, scrubber solutions, NaF precipitate and off-gases show that greater than 99.7% of uranium and thorium, 95% of Pb, Cd, Ni, Sb, Cr, and Hg were retained in the glass. Fluoride and other gaseous emissions from the scrubber system were below regulatory limits, even before HEPA filtration.
Constraints on the Affinity Term for Modeling Long-Term Glass Dissolution Rates
Authors: Bourcier William L. , Carroll Susan A. , Phillips Brian L.
Published: March 2011

Predictions of long-term glass dissolution rates are highly dependent on the form of the affinity term in the rate expression. Analysis of the quantitative effect of saturation state on glass dissolution rate for CSG glass (a simple analog of SRL-165 glass), shows that a simple (1-Q/K) affinity term does not match experimental results. Our data at 100°C show that the data is better fit by an affinity term having the form where σ =10.
Hydrotalcite Formed by Alteration of R7T7 Nuclear Waste Glass and Basaltic Glass in Salt Brine at 190°C
Authors: Abdelouas A. , Crovisier J. L. , Lutze W. , Müller R. , Bernotat W.
Published: March 2011

The R7T7 and synthetic basaltic glasses were submitted to corrosion in a saline MgCl2dominated solution at 190°C. For both glasses, the early alteration product is a hydrotalcite-like compound in which HPO42-, SO4-2and Cl-substitutes to CO32. The measured d003spacing is 7.68 Å for the hydrotalcite formed from R7T7 glass and 7.62 Å for the hydrotalcite formed from basaltic glass which reflect the high aluminium content. Chemical microanalyses show that the hydrotalcite is subsequently covered by a silica-rich gel which evolves into saponite after few months.
Experimental Investigation of Aqueous Corrosion of R7T7 Nuclear Glass at 90°C in the Presence of Humic Acids: a Kinetic Approach
Authors: Gin S. , Godon N. , Mestre J.P. , Vernaz E.Y. , Beaufort D.
Published: March 2011

The dissolution kinetics of the French “R7T7” nonradioactive LWR reference glass in solutions containing dissolved humic acids were investigated at 90°C during static tests with imposed or free pH. Experiments conducted in highly dilute media, with a glass-surface-area-to-solution-volume (SA/V) ratio of 5 m-1, showed that the glass dissolution surface reaction is catalyzed by humic acids. With higher degrees of reaction progress (SA/V = 100 m-1 and free pH) the humic acids impose pH modifications on the system compared with inorganic media; moreover, they directly or indirectly enhance the dissolution of certain alkali metals and transition elements, forming aqueous complexes with the latter. During experiments with an imposed pH of 8.5 (SA/V = 1300 and 5300 m-1), the humic acids appear to cause increased silica solubility that cannot be accounted for by the formation of silica complexes. A residual corrosion rate in the humic acid media exceeding the rate measured in inorganic media suggests that, in addition to silica, one or more element complexes formed by humic acids may be a kinetically limiting factor. This hypothesis must be confirmed, however, as the quantity of humic acids per unit glass surface area was too small in this experiment to allow unambiguous characterization of the phenomenon.
Analytical Electron Microscopy Examination of Solid Reaction Products in Long-Term Tests of Srl 200 Waste Glasses
Authors: Buck E. C. , Fortner J. A. , Bates J. K. , Feng X. , L Dietz N. , Bradley C. R. , Tani B. S.
Published: March 2011

Alteration phases, found on the leached surfaces and present as colloids in the leachates of 200-based frit (fully active and simulated) nuclear waste glass, reacted under static test conditions, at a surface area to leachate volume ratio of 20,000 m-1 for 15 days to 728 days, have been examined by analytical electron microscopy. The compositions of the secondary phases were determined using X-ray energy dispersive spectroscopy and electron energy loss spectroscopy, and structural analysis was accomplished by electron diffraction. Long-term samples of simulated glass, which had undergone an acceleration of reaction after 182 days, possessed a number of silicate secondary phases, including; smectite (iron silicate and potassium iron alumino-silicate), weeksite (uranium silicate), zeolite (calcium potassium alumino-silicate), tobermorite (calcium silicate), and a pure silica phase. However, uranium silicates and smectite have also been observed in tests which have not undergone the acceleration of reaction, in both the leachate and leached layer, suggesting that these phases are not responsible for the acceleration of reaction.
Field-Based Tests of Geochemical Modeling Codes: New Zealand Hydrothermal Systems
Authors: Bruton Carol J. , Glassley William E. , Bourcier William L.
Published: March 2011

Hydrothermal systems in the Taupo Volcanic Zone, North Island, New Zealand are being used as field-based modeling exercises for the EQ3/6 geochemical modeling code package. Comparisons of the observed state and evolution of the hydrothermal systems with predictions of fluid-solid equilibria made using geochemical modeling codes will determine how the codes can be used to predict the chemical and mineralogical response of the environment to nuclear waste emplacement. Field-based exercises allow us to test the models on time scales unattainable in the laboratory.

Preliminary predictions of mineral assemblages in equilibrium with fluids sampled from wells in the Wairakei and Kawerau geothermal field suggest that affinity-temperature diagrams must be used in conjunction with EQ6 to minimize the effect of uncertainties in thermodynamic and kinetic data on code predictions.
Formation, Oxidation and Alteration of Ianthinite
Authors: Finch Robert J. , Ewing Rodney C.
Published: March 2011

Ianthinite, UO2.86·1.5H2O, is a weathering product of uraninite, UO2+x, and has been reported as a corrosion product of UO2. Ianthinite alters to schoepite, UO3·2H2O, or studtite, UO4·2-4H2O, in water. The structures of ianthinite, schoepite and studtite are unknown, but their unit cell parameters differ significantly, and transformations are likely reconstructive.

XRD powder data for ianthinite and becquerelite are similar, and they are probably iso-structural; U4+, as U(OH)22+, replaces Ca2+in the interlayer. The proposed formula is

U(OH)22+[(UO2)6O4(OH)6]·6H2O.

Oxidation in air, has been reported to form schoepite; however, in the absence of water, the oxidation of ianthinite,

U(OH)22+[(UO2)6O4(OH)6]·6H2O + 1/2 O2=> (UO2)2+[(UO2)6O4(OH)6]·7H2O ,

shows that epi-ianthinite is structurally distinct from schoepite. The oxidation of fine-grained material is rapid, and the color change — from dark purple to yellow — is dramatic. The acicular habit of ianthinite is preserved during oxidation, and acicular “schoepite” is observed in nature, as well as in leaching experiments on spent fuel. Possible substitution of Pu4+for U4+suggests that Pu-ianthinite may control Pu solubility in oxidizing, U-saturated solutions.
Formation of Secondary Uranium Minerals in the Koongarra Deposit, Australia
Authors: Isobe Hiroshi , Ewing Rodney C. , Murakami Takashi
Published: March 2011

Secondary uranium minerals from the Koongarra deposit, Northern Territory of Australia, were examined in order to understand the formation and alteration processes of the uranium minerals and their relevance to the migration behavior of uranium, lead, calcium and rare earth elements in the weathered zone. In most of the secondary ore zone, the only stable uranium mineral was saléeite (Mg(UO2)2(PO4)2·10H20), occurring as euhedral platy crystals up to 1 mm in length in veins and at surfaces. Apatite (Ca5(PO4)3F), an accessory mineral of the host rock, has saléeite reaction rims, suggesting formation at the expense of apatite. Ca-uranyl phosphates, such as autunite (Ca(UO2)2(PO4)2·10H2O), were not identified, and Ca-rich uranyl silicates are also absent in the primary ore zone. Pb-bearing uranyl phosphates were found only in the graphite layer cross-cutting the secondary ore zone. In the graphite layer, the local low oxidation condition and high hydrocarbonate content of ground water have affected the formation of uranium minerals and the migration behavior of uranium.