bluuebunny/crossref_metadata_2025_split huggingface dataset:
Nanocomposites — the Multifunctional Family of Materials
Authors: Roy Rustum 
Published: March 2011

AbstractThis paper starts by discussing the concept of multifunctional materials. Multifunctionality can be achieved in a single phase in a composite or a mixture. While a few exceptional single phase materials such as diamond, Al2O3, CaSrZr4P6O24 exhibit peak values of more than one function, it will be argued that the “invention” of the nanocomposite class of materials made by solution sol-gel processing has opened up the potential for designing several multifunctional materials.Examples will be drawn from our recent results—both successes and failures—in trying to design and synthesize nanocomposites of two or more ceramic phases for various chemical, thermal, and electrical functions.
Multifunctional Ferroic Nanocomposites
Authors: Newnham R. E. , McKinstry S. E. , Ikawa H. 
Published: March 2011

AbstractAs trends towards miniaturized components and systems continue in many fields, there has been a rapid development in similarly scaled-down composites. In the electronics industry, these nanocomposites (and especially active nanocomposites based on ferroic elements) form a basis for many of the recent advances in both information and charge storage. While the overall properties of some of these composites can be explained as straightforward extrapolations from the bulk properties, in other instances the small size of the ferroic phase has important consequences on the macroscopic behavior of the composite. This paper reviews some of the recent developments in small-scale ferroic nanocomposites and details the relation between component size and the resultant properties.
Designing Pore-Size in Silica Gels: [H2O] – [TMOS] System
Authors: Kido H. , Malla P. B. , Komarneni S. 
Published: March 2011

AbstractA series of microporous silica gels were synthesized by hydrolyzing tetramethylorthosilicate (TMOS) with varying amounts of water. No catalyst or alcohal was added. The gels were characterized by water sorption isotherms, the BET nitrogen surface area measurements, and differential thermal analysis. Although the gelation time for all H2O/TMOS molar ratios (R) was approximately the same (about 3 hr), the pore-size (pore diameter) and sorption capacity were found to be minimum in the R range 4.96–5.37 but increased with both increasing and decreasing R. Water sorption isotherms measured at 25°C after degassing at 200°C of gels prepared with R = 4.96–5.37 exhibited moderate Type I (Brunauer classification) isotherm which changed to intermediate between Type I and Type IV with decreasing or increasing R. These results indicated that the amount of organic or water which was not utilized in the reaction and the degree of cross-linking apparently played a role for the observed variations in pore-size or shape of water isotherms.
Synthesis and Properties of Comb-Like Liquid Crystalline Polymers with Electrooptically Active Mesogenic Side Groups
Authors: Maeda M. , Kumar R.S. , Blumstein A. , Tripathy S.K. , Sixou P. , Faubert F. 
Published: March 2011

AbstractIn an attempt to synthesize side chain liquid crystalline polymers containing NLO active moieties, a number of psubstituted aromatic azomethine phenols with donor-acceptor substituents such as N≡C and NO2 were prepared and incorporated into a polymer chain as side groups.Liquid crystalline properties of intermediate compounds composed of such moieties were explored separately and in mixtures. Phase diagrams were established. These intermediate compounds of M2C8 and M5C8 show interesting mesogenic properties.Methacrylic esters with such moieties were also synthesized. Homopolymer and copolymer systems with donor-acceptor substituents were obtained. Such copolymers also display liquid crystalline properties.
Silicone Containing Liquid Crystalline Polyurethanes and Polyesters
Authors: Kosfeld Robert , Braun F. , Willner L. , Heb M. 
Published: March 2011

AbstractSeries of poly(urethanes) were prepared in which the spacer groups between the rigid segments contained siloxanes. The conditions for the appearence of liquid crystallinity and a fair thermal stability were investigated and compared to those of corresponding esters.
Study on Diopside Whiskers Precipitation Type Hydroxyapatite
Authors: Nonami Toru 
Published: March 2011

AbstractA biomedical material for use as artificial bone and dental root is provided which takes the form of a sintered composite body comprising a Hydroxyapatite(HAP) and diopside whisker. The mixed powder of HAP and diopside was heated to make this composite material. The sintering of HAP and the precipitation of diopside whisker were examined, to find a method of making them occur simultaneously. Also, diopside was added to HAP and sintering at 1200°C for 2hr. was done. As a result of these investigations, diopside whisker of aspect ratio 101˜15 was formed. Moreover the intermediate layer of CaO-SiO2-P2O5-MgO system was generated at the interface of HAP and whisker. The bending strength of this sintered body was 300MPa and fracture toughness was 3.2MPa ·m1/2. These values were higher by about 2times or 3times than those of matrix HAP respectively. The increase in bending strength was attributed to an increase in fracture toughness, caused by an increase in fracture surface energy.
Assessment of the Performance of Cement-Based Composite Material for Radioactive Waste Immobilization
Authors: Atkins M. , Cowie J. , Glasser F.P. , Jappy T. , Kindness A. , Pointer C. 
Published: March 2011

ABSTRACTThe problem of predicting the future performance of cement-based systems is complicated by a poor understanding of the behaviour of cement systems at long ages, as well as of the complex interactions which can occur between cement and waste components - active as well as inactive - and with cement blending agents including fly ash, glassy slags and natural pozzolans. The progress achieved in developing a predictive capability is reviewed. Considerable success has been achieved in modelling the chemically-related features of cement based systems in terms of pH, Eh, and of element partition between solid and aqueous phases. The behaviour of model radwaste elements - iodine and uranium - has been studied in depth and indicates that both sorption and precipitation occur. U(VI), in particular, forms solubilitylimiting compounds, e.g. uranophane. But in general, presently-available data are inadequate to predict many cement-waste stream teractions; future progress in modelling is likely to rely heavily on additions to the data base.The repository environment will also condition chemical exchanges in cement-based systems. Progress is being made in predicting the impact of CO2, a common ground water component, on the performance of cement systems.
Encapsulation of Radioiodine in Cementitious Waste Forms
Authors: Atkins M. , Glasser F.P. 
Published: March 2011

ABSTRACTRisk assessment models applied to radioactive waste repository design disclose that iodine is one of the nuclides causing most concern. Computer calculations for these scenario studies assume that iodine, in the form of I−, is poorly sorbed on most geological materials. Therefore it is important that iodine be retained at source, ie. within the vault, for as long as is practicable.In the UK context, cements are likely to form a major part of the waste package for low and medium active wastes, and of engineered vault structures. These cements are likely to be blends of one form or another, including Portland cement blended with blast furnace slag, or fly ash. It is therefore important to assess the effect of Portland cement and blending agent on iodine speciation and uptake by the constituent solid phases. Data are presented on the uptake of I− on specific phases: Ca(OH)2, calcium aluminate sulphate hydrates, hydrotalcite and calcium silicate hydrogel (C-S-H). Combining this with a model for predicting phase assemblages in well-aged slag cements, yields an optimum blend for immobilising radioiodide, on ageing.Precipitation of I as AgI during cementation and radiolytic effects on I are also discussed. The production of gaseous radioiodine (I2) is potentially a serious problem.
Durability of Blended Cements in Contact with Sulphate-Bearing Ground Water
Authors: Duerden S L , Majumdar A J , Walton P L 
Published: March 2011

ABSTRACTIn the concept of radioactive waste disposal developed in the UK, OPC blended with pulverised fuel ash or ground granulated blastfurnace slag is being considered for encapsulation of waste forms, as a material for backfilling and sealing a repository, and for concrete in repository construction.This paper describes a laboratory study of the long term durability of such cements in contact with sulphate-bearing ground water under accelerated exposure conditions. Mineralogical analysis of the cements over the exposure period, carried out with the aid of scanning electron microscope observations and x-ray diffraction studies, provides an indication of the stability of cementitious phases exposed to an aggressive environment.It is shown that for intact cement blocks there is minimal interaction between cement and sulphate-bearing ground water. Sulphate minerals produced by the reaction are accommodated in voids in the cement with no adverse effect on the cement structure. However, crystallisation of C-S-H and sulphate minerals along cracks in hardened cement specimens causes expansion of fracture surfaces resulting in a more accessible route for ground water intrusion and radionucleide migration.The reaction of cement with ground water is greatly accelerated by the use of powdered material. Ettringite formed in the reaction is found to be unstable under these conditions. The mineralogical assemblage after exposure for 1 year is calcite, hydrotalcite, C-S-H and quartz.
A Comparison of the Behaviour of Vitrified HLW in Repositories in Salt, Clay and Granite. I: Experimental
Authors: Mertens L. A. , Lutze W. , Marples J. A. C. , van Iseghem P. , Vernaz E. 
Published: March 2011

Laboratories from 11 countries participated in a ‘Round Robin’ comparative test excercise using the European Commission's Repository Systems Simulation Test [1,5]. The compulsory core part of this exercise involved testing an inactive simulant of the French high-level waste glass (SON68) in containers in which the conditions were similar to those that might occur after disposal of vitrified waste in a future repository. There were three options for the test, representing repositories in clay, granite and salt. Fifteen laboratories, mostly from within the European Community, but including laboratories from USA, Canada and Japan took part in the Round Robin. Most of the laboratories only covered one option: clay, 3, granite, 10, and salt 6. The test procedure was closely specified by the European Communities. This contribution and a companion paper [2] summarize the acquired information.The HLW borosilicate glass SON68 (R7T7), to be returned to and eventually disposed of in various geological repositories by Cogema's reprocessing customers, was tested in a stainless steel autoclave developed and provided by CEA, Marcoule. The glass was prepared and delivered by CEA; together with crushed granite, sand, smectite and groundwater for the granite option. SCK-Mol provided the clay, smectite, sand and the recipe for claywater, HMI-Berlin the recipe for the rock salt composition. The analytical method was the only variable to be chosen by the participants. After termination of the experiments (duration 14-364d, temperature 90°C), the concentrations of the glass constituents Si, B, Mo, Li, in solution had to be determined. The data base was used to evaluate the test procedure and to identify analytical problems.
Nuclear Waste Immobilization in Cement-Based Materials: Overview of French Studies
Authors: Atabek Rosemarie , Bouniol Pascal , Revertegat Eliane , Oliver Jean , Gegout Philippe , Jorda Michel 
Published: March 2011

ABSTRACTAn overview of the different problems to be solved for waste form disposal in a shallow-land burial site is presented. Examples of cementbased materials used to immobilize concentrates are given in relation to cement chemistry. Experimental studies, undertaken to assess concrete durability, obviously indicate that pH is the starting parameter of chemical degradation reactions. It is found that blended cement (CLC) has a better cesium retention ability and a stronger resistance to chloride and sulfate attack than Portland cement.
Growth Rate of Alteration Layer and Elemental Mass Losses During Leaching of Borosilicate Nuclear Waste Glass
Authors: Banba T. , Murakami T. , Isobe H. 
Published: March 2011

ABSTRACTMCC-1 static leaching experiments were carried out for a simulated waste glass in deionized water for up to 364 days at 90°C in order to examine the relationship between alteration layer thickness and elemental mass losses. The thickness of the layer increased linearly at a rate of 0.63,μm/day for the first 91 days, and thereafter it remained constant (about 60 μm) independent of time. The change in normalized elemental mass losses of B, Na, and Si with time shows the same trend as that of the change in the layer thickness. It is also demonstrated that the growth rate of the layer can be calculated from the released boron in the leachates. The shrinkage of the alteration layer during air-drying is not significant in spite of the porous structure of the layer, which results from the depletion of cations and the recrystallization reported previously. Although the apparent release of elements almost ceases after a 91-day leaching experiment, reactions in the layer such as crystal growth continue.
Effects of Surface Area-to-Solution Volume Ratio on Chemical Durability of Nuclear Waste Glasses
Authors: Feng X. , Pegg I. L. , Guo Y. , Barkatt AA. , Macedo P. B. 
Published: March 2011

ABSTRACTWe report data on the relationship between nuclear waste glass durability, as measured by leachate concentrations and leach rates, and the ratio of glass surface area to solution volume, S/V. The study includes West Valley non-radioactive production glasses (SF6 and SF10), West Valley glasses containing U and Th designed at CUA (WVCM47, WVCM50, and WVCM59), Savannah River SRLTDS-131 glass, and Hanford waste glass HW-39 (for which existing literature data is used). While some of these glasses show departures of leachate concentrations from simple (S/V)t scaling others conform well. The departures are, in all cases, most evident at high values of S/V. It is therefore important, therefore to understand how glass composition determines which corrosion mechanism is dominant in order to assess the region of validity of extrapolations based on (S/V)t scaling. While leach rates show a general tendency to decrease with S/V and time, exceptions are evident for the less durable glasses which show minima and maxima in the S/V-dependence at fixed time.
The Effect of Temperature on the Redox Constraints for the Processing of High-Level Nuclear Waste into a Glass Waste Form
Authors: Schreiber Henry D. , Schreiber Charlotte W. , Riethmiller Margaret W. , Downey J. Sloan 
Published: March 2011

ABSTRACTThe oxidation-reduction equilibria of selected multivalent elements in an alkali borosilicate glass melt (Savannah River Laboratory frit #131) were measured as a function of the imposed oxygen fugacity over the temperature range from 950°C to 1350°C. Redox constraints on the processing of high-level nuclear waste into the glass melt require that the prevailing oxygen fugacity be about 10−5 to 10−12 Zatm at 950°C, about 10−2 to 10−9 atm at 1150°C, and about 100 to 10−7 atm at 1350°C. Such conditions circumvent foaming under oxidizing situations and metal/sulfide precipitation if the system becomes too reducing. The defined oxygen fugacity ranges correspond to the previously prescribed range of 0.1 to 0.5 for the [Fe2+]/[Fe3+] ratio in the resulting glass, independent of the processing temperature from 950°C to 1350°C.
Diffusion of Cesium in Sodium Borosilicate Glasses for Nuclear Waste Immobilisation
Authors: Sengersi E.G.F. , Janssen F.J.J.G. , de Waal H. 
Published: March 2011

ABSTRACTDiffusion of cesium in borosilicate glass for high-level radioactive waste is discussed. For this purpose model glasses with non-radioactive elements are being made, in accordance with the specifications of the reprocessing plants, from which concentration couples are composed. A concentration couple consists of two cylinders of borosilicate glass which contain different amounts of cesium. After heat treatment the couples are studied by means of scanning electron microscopy and X-Ray microanalysis. The model study will provide a basis for predictions of the containment achieved over a longer period of time.
Actinide Speciation in Spent Fuel Leaching Studies
Authors: Choppin Gregory R. 
Published: March 2011

ABSTRACTThe uranium and plutonium chemistry relevant for spent fuel leach studies is discussed. Model calculations are evaluated by comparison with the results of spent UO2fuel leaching obtained in the Swedish SKB program. The thermodynamic data were found to agree sufficiently with the measured solution concentrations that it can be assumed that in oxic natural waters, leachate uranium from spent fuel would be uranyl carbonate complexes in the solution (or hydroxo complexes in the absence of carbonate) while schoepite would be the solubility limiting solid. For plutonium, PuO2+would be the solution species and Pu(OH)4the solubility limiting solid.
Statistical Model for Grain Boundary and Grain Volume Oxidation Kinetics in UO2 Spent Fuel
Authors: Stout R. B. , Shaw H. F. , Einziger R. E. 
Published: March 2011

ABSTRACTThe Yucca Mountain Project of the U.S. Department of Energy is investigating the suitability of a site in the unsaturated zone at Yucca Mountain, NV, for a high-level nuclear waste repository. Most of the waste will consist of UO2 spent fuel in Zircaloy-clad rods from nuclear reactors. If failure of both the waste containers and the cladding occurs within the lifetime of the repository, then the UO2 will be exposed to oxygen in the air and higher oxides of uranium may form. The oxidation state of the spent fuel may affect its dissolution behavior if later contacted by water. A model for the kinetics of spent fuel oxidation under repository-relevant conditions is thus necessary to predict the behavior of the waste form for assessing the performance of the repository with respect to the containment of radionuclides. In spent fuel experiments, the UO2 oxidation front initially propagates along grain boundaries followed by propagation into grain volumes. Thus, the oxidation kinetics is controlled by two processes and the oxidation of spent fuel fragments will depend on the density and physical attributes of grain boundaries. With this in mind, concepts from statistical mechanics are used to define a density function for grain boundaries perunit volume per unit species in a spent fuel fragment. Combining the integral forms of mass conservation and this grain boundary density function, a model for the global rate of oxidation for a spent fuel fragment is obtained. For rapid grain boundary oxidation compared to grain volume oxidation, equations of the model are solved and results compared to existing data.
Measurement of Soluble Nuclide Dissolution Rates from Spent Fuel
Authors: Wilson C. N. , Gray W. J. 
Published: March 2011

ABSTRACTGaining a better understanding of the potential release behavior of water-soluble radionuclides is the focus of new laboratory spent fuel dissolution studies being planned in support of the Yucca Mountain Project. Previous studies have suggested that maximum release rates for actinide nuclides, which account for most of the long-term radioactivity in spent fuel, should be solubility-limited and should not depend on the characteristics or durability of the spent fuel waste form. Maximum actinide concentrations should be sufficiently low to meet the NRC annual release limits. Potential release rates for soluble nuclides such as 99Tc, 135Cs, 14C and 129I, which account for about 1-2% of the activity in spent fuel at 1000 years, are less certain and may depend on processes such as oxidation of the fuel in the repository air environment.Dissolution rates for several soluble nuclides have been measured from spent fuel specimens using static and semi-static methods. However, such tests do not provide a direct measurement of fuel matrix dissolution rates that may ultimately control soluble-nuclide release rates. Flow-through tests are being developed as a potential supplemental method for determining the matrix component of soluble-nuclide dissolution. Advantages and disadvantages of both semi-static and flow-through methods are discussed. Tests with fuel specimens representing a range of potential fuel states that may occur in the repository, including oxidized fuel, are proposed. Preliminary results from flow-through tests with unirradiated UO2 suggesting that matrix dissolution rates are very sensitive to water composition are also presented.
A Study on Fabrication Technology of Ceramic Overpack - A Conceptual Design and Fabrication of a Full-Scale Ceramic Overpack -
Authors: Teshima T. , Karita Y. , Katsumoto K. , Ishikawa H. , Sasaki N. 
Published: March 2011

ABSTRACTThe conceptual design and fabrication test of a full-scale ceramic overpack were performed from the viewpoint of structural barriers as a part of program to evaluate their potential use as overpack under conditions of deep geological disposal.Materials investigated were porcelain (used for insulators) and A12O3 with high purity of 99.7 %. The selected design consisted of a cylindrical body with hemispherical heads at each end. The design thickness of overpack is the sum of the structural thickness and corrosion allowance. The thickness required to resist the lithostatic pressure was estimated by the basic cylinder buckling formulas and finite element stress analyses in both case of uniform and non-unifom external pressure conditions. These analyses showed that structural thickness of 119 mm was necessary for overpack of porcelain and 40 - for A12O3 under the predicted maximum uniform pressure. In addition, fracture probability of delayed failure, one of significant degradation mode, was estimated for overpack of porcelain.A full-scale overpack of porcelain, of dimensions 800 mm outer diameter x 2200 mm length x 150mm wall thickness, was fabricated under the ordinary level of fabrication technology.
Leaching/Migration of UO2-Fuel in Compacted Bentonite
Authors: Albinsson Y. , Forsyth R. , Skarnemark G. , Skålberg M. , Torstenfelt B. , Werme L. 
Published: March 2011

ABSTRACTThe release and migration of the fission products cesium, europium and technetium, the actinides plutonium, americium and curium, and the activation product cobalt from spent nuclear fuel pellets in highly compacted bentonite clay has been measured after contact times of 101 and 386 days. Experiments at longer contact times are in progress. In some cases small amounts (0.5–1%) of powdered copper or iron metal, or vivianite (Fe3(PO4)2) have been mixed with the bentonite clay.The results indicate as expected a high mobility of cesium. The actinides have a very low mobility. After 386 days, plutonium has diffused less than 0.5 mm away from the fuel, while americium and curium appear to be somewhat more mobile. The behaviour of europium is similar to that of trivalent actinides. Very little technetium has been leached from all samples. Cobalt shows a strong retention in pure bentonite as well as in the presence of vivianite, while the mobility is much larger when iron or copper is added.