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Photoluminescent and electroluminescent compositions are provided which comprise a matrix comprising aromatic repeat units covalently coordinated to a phosphorescent or luminescent metal ion or metal ion complexes. Methods for producing such compositions, and the electroluminescent devices formed therefrom, are also disclosed.
Deformation behavior of continuous W and B fibers reinforced Al alloy matrix composites, monitoring stress-strain characteristics by X ray diffraction
The cores and reflectors in high-temperature gas-cooled reactors (HTGRs) are made of graphite materials, with the graphite acting as a moderator, a fuel host matrix, or the foundation for various structural components. This study aims to survey the models in the literature for graphite materials being used as host matrices in pebble/fuel compacts and to implement those surveyed models into Bison to conduct an early assessment of graphite's thermo-mechanical response under various reactor conditions. In this study, thermal (e.g., thermal conductivity, and specific heat capacity) and mechanical (e.g., elastic properties, thermal expansion, irradiation-induced dimensional changes, and irradiation-induced creep) material models for various graphite grades (e.g., H-451, IG-110, G-348, 2020, A3-3, and A3-27) are incorporated into Bison. Two benchmark problems are then exercised utilizing these new graphite-related capabilities: (1) modeling an Advanced Gas Reactor (AGR)-2 fuel compact, and (2) modeling the debonding of a particle-matrix interface.
Compositions and methods for adjustable lenses are provided. In some embodiments, the lenses contain a lens matrix material, a masking compound, and a prepolymer. The lens matrix material provides structure to the lens. The masking compound is capable of blocking polymerization or crosslinking of the prepolymer, until photoisomerization of the compound is triggered, and the compound is converted from a first isomer to a second isomer having a different absorption profile. The prepolymer is a composition that can undergo a polymerization or crosslinking reaction upon photoinitiation to alter one or more of the properties of the lenses.
Compositions and methods for adjustable lenses are provided. In some embodiments, the lenses contain a lens matrix material, a masking compound, and a prepolymer. The lens matrix material provides structure to the lens. The masking compound is capable of blocking polymerization or crosslinking of the prepolymer, until photoisomerization of the compound is triggered, and the compound is converted from a first isomer to a second isomer having a different absorption profile. The prepolymer is a composition that can undergo a polymerization or crosslinking reaction upon photoinitiation to alter one or more of the properties of the lenses.
The combined third and fourth irradiation in the Advanced Gas Reactor (AGR) program (AGR-3/4) contained tristructural isotropic (TRISO)-coated particle fuel and designed-to-fail (DTF) fuel particles. The DTF particles were only coated with highly anisotropic pyrocarbon (PyC) so they would purposely fail during the AGR-3/4 irradiation and provide a source of fission products for measurement. To observe the post-irradiation morphology of these DTF particles and the TRISO-coated particles, three AGR-3/4 compacts were mounted in epoxy, sectioned above their centerlines, and ground/polished. Three rounds of grinding/polishing and optical microscopy were performed so that the particles could be observed at multiple planes. Each compact contained approximately 1,872 TRISO particles and exactly 20 DTF particles. The three AGR-3/4 compacts examined covered a wide temperature range and featured both the hottest average irradiation temperature (1375°C) and the coldest/lowest burnup (872°C and 5.5% fissions per initial metal atom) of any compacts to undergo post-irradiation examination (PIE) in the AGR program to date. A total of 29 DTF particles were located and observed via microscopy across the three compacts. All the DTF particles observed via microscopy had completely failed PyC coatings. Some different DTF kernel and PyC morphologies were observed that have been attributed to the differences in irradiation temperature and/or burnup. At low and medium irradiation temperatures (roughly 850 to 1050°C), it appears irradiation-induced dimensional changes in the DTF PyC caused it to fracture and fold in on itself, and the kernel deformed to accommodate this PyC deformation. The extent of DTF kernel deformation to accommodate DTF PyC buckling tended to be greater for the medium-temperature fuel compared to the cold fuel. In the high-irradiation-temperature compact (1375°C), the DTF PyC layer appears to have completely reacted chemically with the kernel material. The only material surrounding the DTF particles in this hot compact resembles the compact matrix material. Observations of the TRISO-coated driver fuel particles and fuel compact graphitic matrix were also made. The TRISO particle kernels in the medium-temperature compact (1047°C), and many in the high-temperature compact (1375°C), showed morphologies consistent with what was seen in AGR-1 and AGR-2. In other TRISO particles from the high-temperature compact, a spatial gradient in the kernel appearance was evident. Here the cool side of the kernel (away from the center of the compact) had a much darker appearance (like that of the buffer and pyrocarbon). This is the first observation of this kind of kernel spatial gradient in AGR fuel. It is believed that the high irradiation temperature of that compact (an average of 1375°C) activates or accelerates an unidentified chemical reaction, and the temperature gradient within the fuel causes a sharp spatial gradient. The low-temperature/low-burnup compact also showed unique kernel morphologies where the TRISO-coated particles had clearly distinguishable oxide and carbide phases at the center of the kernel, an oxide rind surrounding this, and remnants of a carbide skin surrounding the oxide rind. These are features observed in the as-fabricated fuel that are no longer present in higher burnup fuel. Occasional gaps between the outer pyrolytic carbon (OPyC) and graphitic matrix material were observed in all compacts. Gaps were most often found in the small, matrix-filled spaces between adjacent particles. Finally, buffer morphologies in AGR-3/4 TRISO particles were like those observed in AGR-1 and AGR-2, and the buffer fracture frequencies in AGR-3/4 and AGR-2 TRISO particles were plotted versus fast neutron fluence and irradiation temperature. In AGR-3/4 the buffer fracture rate was highest (23%) for an irradiation temperature of 1047°C and a fast fluence of 5.18E25 n/m 2 . Increasing the irradiation temperature to 1375°C reduced the buffer fracture rate to 14%. Reducing the irradiation temperature to 872°C and the fast fluence by a factor of 3 reduced the buffer fracture rate to 7%. The temperature and fluence dependencies observed for AGR-3/4 buffer fracture were consistent with those observed in AGR-2. Lower fluences and/or higher temperatures significantly reduced the buffer fracture frequencies. Lower buffer fracture frequencies were also observed at lower temperatures as long as the fluence was also significantly reduced. This is believed to be due to lower fluence resulting in less irradiation-induced dimensional change (shrinkage) of the buffer, and a higher temperature promoting enhanced creep relaxation of stresses within the buffer, leading to less buffer fracture.
A method for evaluating the pH of an aqueous solution by utilizing the optical properties of a pH sensing material comprised of plurality of optically active nanoparticles dispersed in matrix material. The optically active nanoparticles have an electronic conductivity greater than about 10−1 S/cm and generally have an average nanoparticle diameter of less that about 500 nanometers, and the matrix material is a material which experiences a change in surface charge density over a pH range from 2.0 to 12.0 of at least 1%. The method comprises contacting the pH sensing material and the aqueous solution, illuminating the pH sensing material, and monitoring an optical signal generated through comparison of incident light and exiting light to determine the optical transmission, absorption, reflection, and/or scattering of the pH sensitive material. The optical signal of the pH sensitive material varies in response to the pH of the aqueous solution.
Resin systems for matrix material in ablative composite for liquid rocket combustion chambers using fluorine-containing propellants
Results are presented of a combined chemical, isotopic, and mineralogical study of fine-grained opaque matrix material from a type-3 ordinary chondrite Allan Hills A77299. It was found that the matrix of this chondrite (in contrast to chondrules) shows no siderophile element depletions and is remarkably unfractionated relative to CI chondrites. It was also found that the matrix lump has a unique oxygen isotopic composition which lies below the terrestrial fractionation line and differs markedly from the composition of fine-grained matrix lumps and chondrule rims from Semarkona chondrules (Grossman and Wasson, 1987; Grossman et al., 1987); the mineralogy and textures of the matrix lump (which consists of rare angular grains of pyroxene and olivine embedded within a ground-mass of rounded fine-grained olivines), as well as its major element composition, are typical of matrix material of other type-3 chondrites studied. It is concluded that the matrix lump in these chondrites consists largely of material that formed by annealing of amorphous presolar dust or nebular condensates.
The third Advanced Gas Reactor (AGR) irradiation experiment, AGR-3/4, was designed to investigate the migration of fission products in fuel compact graphitic matrix and reactor graphite components. Using destructive methods, radial fission product concentration profiles were measured for gamma-emitting fission products (e.g., Ag-110m, Cs-134, and Eu-154) and beta-emitting Sr-90 in irradiated graphitic and graphite components from six different AGR-3/4 irradiation capsules. These new measured concentration profiles can now be compared to non-destructive measurements and fission product transport simulations and will be used to derive new diffusivities and sorptivities to support refinement of fission product transport models and high-temperature gas-cooled reactor (HTGR) source-term analyses. Each capsule in the AGR-3/4 experiment had four fuel compacts in the middle of two concentric rings of graphitic matrix material, PCEA graphite, or IG-110 graphite. In addition to the approximately 1898 tristructural isotropic (TRISO) coated particles in each compact, there were 20 designed to fail (DTF) particles coated only in pyrocarbon so that they released fission products into the surrounding cylindrical rings of carbonaceous materials. Destructive sampling of the rings involved machining/milling material from around the circumference of the rings, collecting that material, and performing radiochemical analyses on it. Milling operations were performed in multiple steps or segments, and each segment was generally 0.508 mm (0.020 in) thick. Knowing the radial position at which each segment was milled, the volume of the milled material at each segment, and the fission product content in each segment, the radial fission product concentrations were constructed for select isotopes in each ring. Ag-110m profiles had the most variation. Some profiles were peaked at an inner or outer surface. Some were peaked at the middle of the ring wall thickness. Some increased radially outward, and some decreased radially outward. These types of variations and the fact that the measured profiles do not generally compare favorably with the transport model employed for AGR-3/4 may adversely impact the ability to extract reasonable transport parameters for this isotope. In many cases, the Cs-134 profiles decreased somewhat linearly in the outward radial direction, and in cursory comparisons, the shapes of these profiles appeared similar to those from model predictions. The step changes in concentration across the inner-outer ring gap were generally consistent with the model predictions as well. In some cases, there were local maxima in concentration at the outer surface of the rings. This suggests that fission products could have transported in the small gaps between the inner ring and the outer ring and between the outer ring and the sink ring such that some portion of a given fission product can bypass diffusion through the ring itself. The analysis of the small nubs on the outer surfaces of some of the outer rings revealed fission product concentrations in the nubs that were often higher than in the outermost segments of the rings. This further supports the hypothesis that short-circuit, gap transport occurred, causing relatively high surface concentrations on the outer surfaces of the rings. Eu-154 and Sr-90 profiles tend to have very similar shapes, suggesting that they transport via the same mechanisms. The observed profiles were indicative of a transport process where the isotopes are sorbed on the inner surface of the ring, but diffusion into the ring from that surface is quite slow. Some elevated concentrations of Sr-90 (relative to Eu-154) on the outer surface of a ring suggested that rapid, gas-gap transport of gaseous precursor Kr-90 and volatile Rb-90 could have occurred prior to their decaying to Sr-90. Overall, the Eu 154 and Sr-90 profiles were still very similar, which indicates that the transport of short-lived Sr-90 precursors is not a major effect. In some capsules, the qualitative Sr-90 behavior across the ring gaps was consistent with the model (using the available legacy Sr-90 transport parameters), but in other capsules the model was inconsistent with the measurements and seems to underestimate the amount of Sr-90 in the outer rings. The total ring Sr-90 inventories were estimated for all the rings that were subject to physical sampling. These results will be used to adjust the predicted particle and/or compact releases used in the AGR-3/4 fission product transport model. Given the different irradiation temperatures among the capsules and the rings, it was not possible to discern fundamental differences in the transport of isotopes within the different carbon materials, i.e., graphitic matrix, IG-110, or PCEA. It may be possible to do this in the course of determining transport from the concentration profiles in future work.
The material damping in beam-like specimens of aluminum and metal matrix composites was measured. A unique apparatus to determine damping by free decay while the specimens are in free fall in a vacuum was used. The specimens tested include 2024-T3 and 6061-T4 aluminum, and unidirectional graphite/metal matrix specimens with P55 and P100 fibers and 6061 Aluminum and AZ91C Magnesium as matrix materials. Tests were conducted to determine the dependence of damping on frequency and stress level. For the aluminum specimens, the material damping followed the Zener model at very low stress levels. Below the Zener relaxation frequency, a strong dependence of damping on stress was found for even moderate stress levels. Damping for the aluminum matrix materials was slightly above that predicted by the Zener model for a homogeneous bar of the matrix aluminum. For the magnesium matrix specimens, damping significantly above the Zener prediction for the homogeneous matrix material was observed.
The potential for utilizing surface coatings to inhibit the thermal oxidation of polymer matrix composites was studied. Isothermal, inert gas exposures of graphite/PMR-15 composites indicated that after an initial loss of weight, no significant amounts of thermal degradation products are given off during high temperature exposures in the absence of oxygen. As long as a coating remains effective, the composite material should remain stable. It was also found that the glass transition temperature T sub g of the matrix resin could be increased to values in excess of 400 C. This resulted in measured short beam shear strengths of 75.9 MPa (11 Ksi), flexural strengths of 1172 MPa (170 Ksi) and flexural moduli of 141 GPa (20.5 Msi) for the material at a test temperature of 371 C. The treatment that was used caused a decrease in the PMR-15 resin density from 1.31 to 1.29 gm/cc. It was concluded that state-of-the-art composites, protected by oxygen-impervious coatings, can be used as materials of construction with structural integrity to at least 371 C and possibly above.
The potential for utilizing surface coatings to inhibit the thermal oxidation of polymer matrix composites was studied. Isothermal, inert gas exposures of graphite/PMR-15 composites indicated that after an initial loss of weight, no significant amounts of thermal degradation products are given off during high temperature exposures in the absence of oxygen. As long as a coating remains effective, the composite material should remain stable. It was also found that the glass transition temperature T sub g of the matrix resin could be increased to values in excess of 400 C. This resulted in measured short beam shear strengths of 75.9 MPa (11 Ksi), flexural strengths of 1172 MPa (170 Ksi) and flexural moduli of 141 GPa (20.5 Msi) for the material at a test temperature of 371 C. The treatment that was used caused a decrease in the PMR-15 resin density from 1.31 to 1.29 gm/cc. It was concluded that state-of-the-art composites, protected by oxygen-impervious coatings, can be used as materials of construction with structural integrity to at least 371 C and possibly above.
Results are presented on the thermal-vacuum response of a variety of fiber reinforced polymers matrix composites that comprised the UTIAS experiment on the LDEF satellite. Theoretical temperature-time predictions for this experiment are in excellent agreement with test data. Results also show quite clearly the effect of outgassing in the dimensional changes of these materials and the corresponding coefficients of thermal expansion. Finally, comparison with ground-based simulation tests are presented as well. Use of these data for design purposes are also given.
In the High Temperature Gas-Cooled Reactor (HTGR) the TRISO particle fuel serves as the primary fission product containment. However the large number of TRISO particles present in proposed HTGRs dictates that there will be a small fraction (~10 -4 to 10 -5 ) of as manufactured and in-pile particle failures that will lead to some fission product release. The matrix material surrounding the TRISO particles in fuel compacts and the structural graphite holding the TRISO particles in place can also serve as sinks for containing any released fission products. However data on the migration of solid fission products through these materials is lacking. One of the primary goals of the AGR-3/4 experiment is to study fission product migration from failed TRISO particles in prototypic HTGR components such as structural graphite and compact matrix material. In this work, the potential for a Gamma Emission Computed Tomography (GECT) technique to non-destructively examine the fission product distribution in AGR-3/4 components and other irradiation experiments is explored. Specifically, the feasibility of using the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) Precision Gamma Scanner (PGS) system for this GECT application is considered. To test the feasibility, the response of the PGS system to idealized fission product distributions has been simulated using Monte Carlo radiation transport simulations. Previous work that applied similar techniques during the AGR-1 experiment will also be discussed as well as planned uses for the GECT technique during the post irradiation examination of the AGR-2 experiment. The GECT technique has also been applied to other irradiated nuclear fuel systems that were currently available in the HFEF hot cell including oxide fuel pins, metallic fuel pins, and monolithic plate fuel.
To assess whether matrix fracture would result in an unacceptable loss of containment or confinement in TRISO fuel particles, it is crucial to evaluate the micro-tensile strength, fracture toughness, and irradiation effects on matrix materials. Current data must be comprehensive and validated for modeling fractures under various conditions. Relevant material properties surrounding matrix fracture will be discussed during the presentation.
Acousto-ultrasonic nondestructive evaluation of ceramic composite specimens with a lithium-alumino-silicate glass matrix reinforced with unidirectional silicon carbide (NICALON) fibers was conducted to evaluate their reserve of strength. Ceramic composite specimens with different amount of damage were prepared by four-point cyclic fatigue loading of the specimens at 500 C for a different number of cycles. The reserve of strength of the specimens was measured as the maximum bending stress recorded during four-pointed bending test with the load monotonically increased until failure occurs. It was observed that the reserve of strength did not correlate with the number of fatigue cycles. However, it was also observed that higher values of the stress wave factor measurements correspond to higher values of the reserve of strength test data. Therefore, these results show that the acousto-ultrasonic approach has the potential of being used to monitor damage and to estimate the reserve of strength of ceramic composites.
The Y-90 Brachytherapy Seed is a small, cylindrical source of therapeutic radiation for controlled delivery of localized therapeutic radiation to cancerous tissue. The Y-90 Brachytherapy Seed consists of microscopic inorganic particles of highly insoluble yttrium phosphate containing the beta-emitting radionuclide Y-90, gold particles, and an encapsulating bio-absorbable polymer-matrix material. The polymer matrix is formed in a cylindrical shape to contain the yttrium phosphate microspheres and gold x-ray marker particles.