665 Matching Results

Search Results

Advanced search parameters have been applied.

Dissolution studies of plutonium oxide in LaBS glass

Description: As part of international agreement between the United States and Russia, a significant amount of plutonium requires disposition. One of the disposition paths is to immobilize it and dispose of it in a geological repository. The two favored immobilization forms are glass and ceramic. The plutonium, as an oxide, would be reacted with the glass or ceramic to form a homogeneousmaterial. The resulting solid product would then be encased in High-Level Waste (1-ILW)glass for the can-in-canister option. The HLW glass gives a radiation barrier to increase proliferation resistance. The glass canister would then be disposed of by geological emplacement. This paper discusses how glass meets two criteria: the condition of significant actinide volubility, and That the PuO{sub 2} feed should be incorporated into the matrix without significant amount of unreacted material.
Date: May 5, 1997
Creator: Riley, D.; Bourcier, W.; Vienna, J.; Meaker, T.; Peeler, D. & Maffa, J.
Partner: UNT Libraries Government Documents Department

Coffee Maker

Description: The view of the glass coffee maker is from an angle showing the lid. The two screws holding the wooden handle are also visible in this view.
Access: This item is restricted to UNT Community Members. Login required if off-campus.
Date: 1941
Creator: Schlumbohm, Peter
Partner: UNT College of Visual Arts + Design

Comparison of the results of short-term static tests and single-pass flow-through tests with LRM glass.

Description: Static dissolution tests were conducted to measure the forward dissolution rate of LRM glass at 70 C and pH(RT) 11.7 {+-} 0.1 for comparison with the rate measured with single-pass flow-through (SPFT) tests in an interlaboratory study (ILS). The static tests were conducted with monolithic specimens having known geometric surface areas, whereas the SPFT tests were conducted with crushed glass that had an uncertain specific surface area. The error in the specific surface area of the crushed glass used in the SPFT tests, which was calculated by modeling the particles as spheres, was assessed based on the difference in the forward dissolution rates measured with the two test methods. Three series of static tests were conducted at 70 C following ASTM standard test method C1220 using specimens with surfaces polished to 600, 800, and 1200 grit and a leachant solution having the same composition as that used in the ILS. Regression of the combined results of the static tests to the affinity-based glass dissolution model gives a forward rate of 1.67 g/(m{sup 2}d). The mean value of the forward rate from the SPFT tests was 1.64 g/(m{sup 2}d) with an extended uncertainty of 1.90 g/(m{sup 2}d). This indicates that the calculated surface area for the crushed glass used in the SPFT tests is less than 2% higher than the actual surface area, which is well within the experimental uncertainties of measuring the forward dissolution rate using each test method. These results indicate that the geometric surface area of crushed glass calculated based on the size of the sieves used to isolate the fraction used in a test is reliable. In addition, the C1220 test method provides a means for measuring the forward dissolution rate of borosilicate glasses that is faster, easier, and more economical than the SPFT test method.
Date: January 29, 2007
Creator: Ebert, W. L. & Engineering, Chemical
Partner: UNT Libraries Government Documents Department

Reactivity of high plutonium-containing glasses for the immobilization of surplus fissile materials

Description: Experiments have been performed on glasses doped with 2 and 7 wt % plutonium to evaluate factors that may be important in the performance of these high-Pu-loaded glasses for repository storage. The high Pu loadings result from the need to dispose of excess Pu from weapons dismantling. The glasses were reacted in water vapor to simulate aging that may occur under unsaturated storage conditions prior to contact with liquid water. They were also reacted with liquid water under standard static leach test conditions. The results were compared with similar tests of a reference glass (202 glass) containing only 0.01 wt % Pu. In vapor hydration testing to date, at 2 wt % loading, the Pu was incorporated into the glass without phase separation, and reaction in water vapor proceeded at a rate comparable with that of the 202 glass. At wt % loading, a Pu phase separated and was not uniformly incorporated into the glass. The vapor reaction of this glass proceeded at a more rapid rate. This phase separation was manifested in the static leach tests, where colloidal phases of Pu-rich material remained suspended in solution, thereby increasing the absolute Pu release when compared to the 202 glass.
Date: June 1, 1995
Creator: Bates, J.K.; Hoh, J.C.; Emery, J.W.; Buck, E.C.; Fortner, J.A.; Wolf, S.F. et al.
Partner: UNT Libraries Government Documents Department

Direct conversion of plutonium-containing materials to borosilicate glass for storage or disposal

Description: A new process, the Glass Material Oxidation and Dissolution System (GMODS), has been invented for the direct conversion of plutonium metal, scrap, and residue into borosilicate glass. The glass should be acceptable for either the long-term storage or disposition of plutonium. Conversion of plutonium from complex chemical mixtures and variable geometries into homogeneous glass (1) simplifies safeguards and security; (2) creates a stable chemical form that meets health, safety, and environmental concerns; (3) provides an easy storage form; (4) may lower storage costs; and (5) allows for future disposition options. In the GMODS process, mixtures of metals, ceramics, organics, and amorphous solids containing plutonium are fed directly into a glass melter where they are directly converted to glass. Conventional glass melters can accept materials only in oxide form; thus, it is its ability to accept materials in multiple chemical forms that makes GMODS a unique glass making process. Initial proof-of-principle experiments have converted cerium (plutonium surrogate), uranium, stainless steel, aluminum, and other materials to glass. Significant technical uncertainties remain because of the early nature of process development.
Date: June 27, 1995
Creator: Forsberg, C.W. & Beahm, E.C.
Partner: UNT Libraries Government Documents Department

Conversion of plutonium scrap and residue to boroilicate glass using the GMODS process

Description: Plutonium scrap and residue represent major national and international concerns because (1) significant environmental, safety, and health (ES&H) problems have been identified with their storage; (2) all plutonium recovered from the black market in Europe has been from this category; (3) storage costs are high; and (4) safeguards are difficult. It is proposed to address these problems by conversion of plutonium scrap and residue to a CRACHIP (CRiticality, Aerosol, and CHemically Inert Plutonium) glass using the Glass Material Oxidation and Dissolution System (GMODS). CRACHIP refers to a set of requirements for plutonium storage forms that minimize ES&H concerns. The concept is several decades old. Conversion of plutonium from complex chemical mixtures and variable geometries into a certified, qualified, homogeneous CRACHIP glass creates a stable chemical form that minimizes ES&H risks, simplifies safeguards and security, provides an easy-to-store form, decreases storage costs, and allows for future disposition options. GMODS is a new process to directly convert metals, ceramics, and amorphous solids to glass; oxidize organics with the residue converted to glass; and convert chlorides to borosilicate glass and a secondary sodium chloride stream. Laboratory work has demonstrated the conversion of cerium (a plutonium surrogate), uranium (a plutonium surrogate), Zircaloy, stainless steel, and other materials to glass. GMODS is an enabling technology that creates new options. Conventional glassmaking processes require conversion of feeds to oxide-like forms before final conversion to glass. Such chemical conversion and separation processes are often complex and expensive.
Date: November 28, 1995
Creator: Forsberg, C.W.; Beahm, E.C.; Parker, G.W.; Rudolph, J.; Elam, K.R. & Ferrada, J.J.
Partner: UNT Libraries Government Documents Department

[Localized fracture damage effects in toughened ceramics]. Final report

Description: The primary research goal was to investigate localized fracture damage due to single point cutting of ceramic materials and then to compare this to multipoint cutting during precision grinding of the same materials. Two test systems were designed and constructed for the single-point cutting tests. The first system used a PZT actuator for closed-loop load control. An acoustic emission data acquisition system was used for crack initiation detection. The second test system employed a high-precision diamond-turning machine for closed-loop position (cutting depth) control. A high stiffness load cell and data acquisition system were used for crack initiation detection. Microcutting tests were carried out on silicon, borosilicate glass and CVD silicon carbide. The crack initiation thresholds and the fracture damage distribution were determined as a function of the loading conditions using a Vickers diamond as the cutting tool. The grinding tests were done using a plunge-grinding technique with metal-bonded diamond wheels. Optical microscopy, surface roughness and specific cutting energy were measured in order to characterize the fracture damage as a function of the grinding infeed rate. Simulation models were developed in order to estimate the average grain-depth of cut in grinding so that the response could be compared to the single-point microcutting tests.
Date: December 31, 1997
Partner: UNT Libraries Government Documents Department

The release of technetium from defense waste processing facility glasses

Description: Laboratory tests are being, conducted using two radionuclide-doped Defense Waste Processing, Facility (DWPF) glasses (referred to as SRL 13IA and SRL 202A) to characterize the effects of the glass surface area/solution volume (SN) ratio on the release and disposition of {Tc} and several actinide elements. Tests are being conducted at 90{degrees}C in a tuff ground water solution at SN ratios of 10, 2000, and 20,000 m{sup {minus}1} and have been completed through 1822 days. The formation of certain alteration phases in tests at 2000 and 20,000 m{sup {minus}1} results in an increase in the dissolution rates of both classes. The release of {Tc} parallels that of B and Na under most test conditions and its release increases when alteration phases form. However, in tests with SRL 202A glass at 20,000 ,{sup {minus}1}, the {Tc} concentration in solution decreases coincidentally with an increase in the nitrite/nitrate ratio that indicates a decrease in the solution Eh. This may have occurred due to radiolysis, glass dissolution, the formation of alteration phases, or vessel interactions. Technetium that was reduced from {Tc}(VII) to {Tc}(IV) may have precipitated, thou-h the amount of {Tc} was too low to detect any {Tc}-bearing phases. These results show the importance of conducting long-term tests with radioactive glasses to characterize the behavior of radionuclides, rather than relying on the observed behavior of nonradioactive surrogates.
Date: December 31, 1995
Creator: Ebert, W.L.; Wolf, S.F. & Bates, J.K.
Partner: UNT Libraries Government Documents Department

Hanford Waste Vitrification Program process development: Melt testing subtask, pilot-scale ceramic melter experiment, run summary

Description: Hanford Waste Vitrification Program (HWVP) activities for FY 1985 have included engineering and pilot-scale melter experiments HWVP-11/HBCM-85-1 and HWVP-12/PSCM-22. Major objectives designated by HWVP fo these tests were to evaluate the processing characteristics of the current HWVP melter feed during actual melter operation and establish the product quality of HW-39 borosilicate glass. The current melter feed, defined during FY 85, consists of reference feed (HWVP-RF) and glass-forming chemicals added as frit.
Date: March 1, 1996
Creator: Nakaoka, R.K.; Bates, S.O.; Elmore, M.R.; Goles, R.W.; Perez, J.M.; Scott, P.A. et al.
Partner: UNT Libraries Government Documents Department

Plutonium dioxide dissolution in glass

Description: In the aftermath of the Cold War, the U.S. Department of Energy`s (DOE) Office of Fissile Materials Disposition (OFMD) is charged with providing technical support for evaluation of disposition options for excess fissile materials manufactured for the nation`s defense. One option being considered for the disposition of excess plutonium (Pu) is immobilization by vitrification. The vitrification option entails immobilizing Pu in a host glass and waste package that are criticality-safe (immune to nuclear criticality), proliferation-resistant, and environmentally acceptable for long-term storage or disposal. To prove the technical and economic feasibility of candidate vitrification options it is necessary to demonstrate that PuO{sub 2} feedstock can be dissolved in glass in sufficient quantity. The OFMD immobilization program has set a Pu solubility goal of 10 wt% in glass. The life cycle cost of the vitrification options are strongly influenced by the rate at which PUO{sub 2} dissolves in glass. The total number of process lines needed for vitrification of 50 t of Pu in 10 years is directly dependent upon the time required for Pu dissolution in glass. The objective of this joint Pacific Northwest National Laboratory (PNNL) - Savannah River Technology Center (SRTC) study was to demonstrate a high Pu solubility in glass and to identify on a rough scale the time required for Pu dissolution in the glass. This study was conducted using a lanthanide borosilicate (LaBS) glass composition designed at the SRTC for the vitrification of actinides.
Date: September 1, 1996
Creator: Vienna, J.D.; Alexander, D.L. & Li, Hong
Partner: UNT Libraries Government Documents Department

Effect of different glasses in glass bonded zeolite

Description: A mineral waste form has been developed for chloride waste salt generated during the pyrochemical treatment of spent nuclear fuel. The waste form consists of salt-occluded zeolite powders bound within a glass matrix. The zeolite contains the salt and immobilizes the fission products. The zeolite powders are hot pressed to form a mechanically stable, durable glass bonded zeolite. Further development of glass bonded zeolite as a waste form requires an understanding of the interaction between the glass and the zeolite. Properties of the glass that enhance binding and durability of the glass bonded zeolite need to be identified. Three types of glass, boroaluminosilicate, soda-lime silicate, and high silica glasses, have a range of properties and are now being investigated. Each glass was hot pressed by itself and with an equal amount of zeolite. MCC-1 leach tests were run on both. Soda-lime silicate and high silica glasses did not give a durable glass bonded zeolite. Boroaluminosilicate glasses rich in alkaline earths did bind the zeolite and gave a durable glass bonded zeolite. Scanning electron micrographs suggest that the boroaluminosilicate glasses wetted the zeolite powders better than the other glasses. Development of the glass bonded zeolite as a waste form for chloride waste salt is continuing.
Date: May 1, 1995
Creator: Lewis, M.A.; Ackerman, J.P. & Verma, S.
Partner: UNT Libraries Government Documents Department

Hydrogen speciation in hydrated layers on nuclear waste glass

Description: The hydration of an outer layer on nuclear waste glasses is known to occur during leaching, but the actual speciation of hydrogen (as water or hydroxyl groups) in these layers has not been determined. As part of the Nevada Nuclear Waste Storage Investigations Project, we have used infrared spectroscopy to determine hydrogen speciations in three nuclear waste glass compositions (SRL-131 & 165, and PNL 76-68), which were leached at 90{sup 0}C (all glasses) or hydrated in a vapor-saturated atmosphere at 202{sup 0}C (SRL-131 only). Hydroxyl groups were found in the surface layers of all the glasses. Molecular water was found in the surface of SRL-131 and PNL 76-68 glasses that had been leached for several months in deionized water, and in the vapor-hydrated sample. The water/hydroxyl ratio increases with increasing reaction time; molecular water makes up most of the hydrogen in the thick reaction layers on vapor-phase hydrated glass while only hydroxyl occurs in the least reacted samples. Using the known molar absorptivities of water and hydroxyl in silica-rich glass the vapor-phase layer contained 4.8 moles/liter of molecular water, and 0.6 moles water in the form hydroxyl. A 15 {mu}m layer on SRL-131 glass formed by leaching at 90{sup 0}C contained a total of 4.9 moles/liter of water, 2/3 of which was as hydroxyl. The unreacted bulk glass contains about 0.018 moles/liter water, all as hydroxyl. The amount of hydrogen added to the SRL-131 glass was about 70% of the original Na + Li content, not the 300% that would result from alkali=hydronium ion interdiffusion. If all the hydrogen is then assumed to be added as the result of alkali-H{sup +} interdiffusion, the molecular water observed may have formed from condensation of the original hydroxyl groups.
Date: January 15, 1987
Creator: Aines, R. D.; Weed, H. C. & Bates, J. K.
Partner: UNT Libraries Government Documents Department

Large microchannel array fabrication and results for DNA sequencing

Description: We have developed a process for the production of microchannel arrays on bonded glass substrates up to I4 x 58 cm, for DNA sequencing. Arrays of 96 and 384 microchannels, each 46 cm long have been built. This technology offers significant advantages over discrete capillaries or conventional slab-gel approaches. High throughput DNA sequencing with over 550 base pairs resolution has been achieved. With custom fabrication apparatus, microchannels are etched in a borosilicate substrate, and then fusion bonded to a top substrate 1.1 mm thick that has access holes formed in it. SEM examination shows a typical microchannel to be 40 x 180 micrometers by 46 cm Iong; the etch is approximately isotropic, leaving a key undercut, for forming a rounded channel. The surface roughness at the bottom of the 40 micrometer deep channel has been profilometer measured to be as low as 20 nm; the roughness at the top surface was 2 nm. Etch uniformity of about 5% has been obtained using a 22% vol. HF / 78% Acetic acid solution. The simple lithography, etching, and bonding of these substrates enables efficient production of these arrays and extremely precise replication From master masks and precision machining with a mandrel. Keywords: microchannels, microchannel plates, DNA sequencing, electrophoresis, borosilicate glass
Date: January 7, 1999
Creator: Pastrone, R L; Balch, J W; Brewer, L R; Copeland, A C; Davidson , J C; Fitch, J P et al.
Partner: UNT Libraries Government Documents Department

Control of Oxidation Potential for Basalt Repository Simulation Tests

Description: Borosilicate waste glass durability in simulated repository environments can be assessed by use of static tests in leach vessels fabricated of the representative geomedia. Control of the oxidation potential during the test simulates a basalt repository environment. Under very anoxic conditions (i.e., at negative Eh values) the interactions between basalt and SRP waste glass in silica-saturated basaltic groundwaters are the same as those of basalt and groundwater when no waste glass is present. The lack of significant leaching of ions from the waste glass and the lack of any significant changes in either the leached surfaces of glass or basalt under anoxic conditions suggests that the components of this system are at equilibrium when oxygen is absent.
Date: November 13, 1984
Creator: Jantzen, C.M.
Partner: UNT Libraries Government Documents Department

Commercial Ion Exchange Resin Vitrification in Borosilicate Glass

Description: Bench-scale studies were performed to determine the feasibility of vitrification treatment of six resins representative of those used in the commercial nuclear industry. Each resin was successfully immobilized using the same proprietary borosilicate glass formulation. Waste loadings varied from 38 to 70 g of resin/100 g of glass produced depending on the particular resin, with volume reductions of 28 percent to 68 percent. The bench-scale results were used to perform a melter demonstration with one of the resins at the Clemson Environmental Technologies Laboratory (CETL). The resin used was a weakly acidic meth acrylic cation exchange resin. The vitrification process utilized represented a approximately 64 percent volume reduction. Glass characterization, radionuclide retention, offgas analyses, and system compatibility results will be discussed in this paper.
Date: May 1, 1998
Creator: Cicero-Herman, C.A.; Workman, P.; Poole, K.; Erich, D. & Harden, J.
Partner: UNT Libraries Government Documents Department

High-level waste processing at the Savannah River Site: An update

Description: The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) in Aiken, SC mg began immobilizing high-level radioactive waste in borosilicate glass in 1996. Currently, the radioactive glass is being produced as a ``sludge-only`` composition by combining washed high-level waste sludge with glass frit. The glass is poured in stainless steel canisters which will eventually be disposed of in a permanent, geological repository. To date, DWPF has produced about 100 canisters of vitrified waste. Future processing operations will, be based on a ``coupled`` feed of washed high-level waste sludge, precipitated cesium, and glass frit. This paper provides an update of the processing activities completed to date, operational/flowsheet problems encountered, and programs underway to increase production rates.
Date: September 1, 1997
Creator: Marra, J. E.; Bennett, W.M.; Elder, H.H.; Lee, E.D.; Marra, S.L. & Rutland, P.L.
Partner: UNT Libraries Government Documents Department

Impact of Phase Separation on Waste Glass Durability

Description: Phase separation is shown to have an adverse and unpredictable effect on durability of borosilicate nuclear waste glasses. The glass chemistry and thermal history of the waste glass during solidification in a canister can impact the kinetics of phase separation and thus, the long term durability of a the glass. Although waste glasses contain 15-20 components, many of the components are present in minor amounts. Greater than 95 percent of the glass chemistry is dominated by the seven major components, Na2O- K2O-Li2O-SiO2-Al2O3-B2O3-Fe2O3. Although the phase equilibria of this seven component system has never been studied, a compositionally dependent "Phase Separation Discriminator" was developed from a database of 88 High Level Waste (HLW) glasses shown experimentally to be homogeneous and 22 shown experimentally to be phase separated. This discriminator ensures that the HLW glasses produced in the Defense Waste Processing Facility (DWPF) are homogeneous and have predictable long term durability.
Date: April 23, 1999
Creator: Jantzen, C.M.
Partner: UNT Libraries Government Documents Department

Modeling a novel glass immobilization waste treatment process using flow

Description: One option for control and disposal of surplus fissile materials is the Glass Material Oxidation and Dissolution System (GMODS), a process developed at ORNL for directly converting Pu-bearing material into a durable high-quality glass waste form. This paper presents a preliminary assessment of the GMODS process flowsheet using FLOW, a chemical process simulator. The simulation showed that the glass chemistry postulated ion the models has acceptable levels of risks.
Date: January 25, 1996
Creator: Ferrada, J.J.; Nehls, J.W. Jr.; Welch, T.D. & Giardina, J.L.
Partner: UNT Libraries Government Documents Department

Performance of high plutonium-containing glasses for the immobilization of surplus fissile materials

Description: Plutonium from dismantled weapons is being evaluated for geological disposal. While a final waste form has not been chosen, borosilicate glass will be one of the waste forms to be evaluated. The reactivity of the reference blend glass containing the standard amount of Pu ({approximately}0.01 wt %) to be produced by the Defense Waste Processing Facility (DWPF) is compared to that of glasses made from the same nominal frit composition but doped with 2 and 7 wt % Pu, and also equal mole percentages of Gd{sub 2}O{sub 3}. The Gd is added to act as a neutron poison to address criticality concerns. The four different glasses have been reacted using the PCT-B method with a SA/V of 20,000 m{sup {minus}1} and the Argonne Vapor Hydration Test (VHT) method. Both test methods accelerate the reaction of the glass. PCT-B is used to determine the reactivity of the glass by analyzing the solution and reacted test components, while the VHT is used to evaluate the long-term reactivity of the glass and the distribution of Pu to secondary phases that will control the long-term reaction of the glass. The results of the tests with high levels of Pu are compared to those with the nominal levels to be produced in the standard DWPF glass.
Date: July 1, 1995
Creator: Bates, J.K.; Emery, J.W.; Hoh, J.C. & Johnson, T.R.
Partner: UNT Libraries Government Documents Department

SRS vitrification studies in support of the U.S. program for disposition of excess plutonium

Description: Many thousands of nuclear weapons are being retired in the U.S. and Russian as a result of nuclear disarmament activities. These efforts are expected to produce a surplus of about 50 MT of weapons grade plutonium (Pu) in each country. In addition to this inventory, the U.S. Department of Energy (DOE) has more than 20 MT of Pu scrap, residue, etc., and Russian is also believed to have at least as much of this type of material. The entire surplus Pu inventories in the U.S. and Russian present a clear and immediate danger to national and international security. It is important that a solution be found to secure and manage this material effectively and that such an effort be implemented as quickly as possible. One option under consideration is vitrification of Pu into a safe, durable, accountable and proliferation-resistant form. As a result of decades to experience within the DOE community involving vitrification of a variety of hazardous and radioactive wastes, this existing technology can now be expanded to include mobilization of large amounts of Pu. This technology can then be implemented rapidly using the many existing resources currently available. An overall strategy to vitrify many different types of Pu will be already developed throughout the waste management community can be used in a staged Pu vitrification effort. This approach uses the flexible vitrification technology already available and can even be made portable so that it may be brought to the source and ultimately, used to produce a consistent and common borosilicate glass composition for the vitrified Pu. The final composition of this product can be made similar to nationally and internationally accepted HLW glasses.
Date: September 1, 1995
Creator: Wicks, G.G.; McKibben, J.M.; Plodinec, M.J. & Ramsey, W.G.
Partner: UNT Libraries Government Documents Department

Formation of novel optical materials by ion implantation

Description: Ion implantation into glasses shows considerable promise as a technique for producing novel materials having unique optical properties. There is also significant commercial interest in the marketing of optical waveguides, high-performance polarizing glasses, and birefringent glasses. In this work, special glass compositions, including both two-phase and anisotropic materials, were prepared by bulk processing techniques and then implanted with Ag, Au, and Cu ions at different temperatures, energies, and doses. The composite materials produced showed strong resonant absorption at the surface plasmon resonance frequency. The plasmon frequency was consistent with the formation of spherical colloids of the implanted metals having diameters on the order of 10 nm. No anisotropic behavior of the implanted materials was observed. Possible applications for these materials, had such anisotropy been developed, would have included optical isolators, Faraday rotators, optical waveguides, and switching devices for use in optical communication and computing.
Date: October 1, 1995
Partner: UNT Libraries Government Documents Department

Chemical durability of simulated nuclear glasses containing water

Description: The chemical durability of simulated nuclear waste glasses having different water contents was studied. Results from the product consistency test (PCT) showed that glass dissolution increased with water content in the glass. This trend was not observed during MCC-1 testing. This difference was attributed to the differences in reactions between glass and water. In the PCT, the glass network dissolution controlled the elemental releases, and water in the glass accelerated the reaction rate. On the other hand, alkali ion exchange with hydronium played an important role in the MCC-1. For the latter, the amount of water introduced into a leached layer from ion-exchange was found to be much greater than that of initially incorporated water in the glass. Hence, the initial water content has no effect on glass dissolution as measured by the MCC-1 test.
Date: April 1, 1995
Creator: Li, H. & Tomozawa, M.
Partner: UNT Libraries Government Documents Department


Description: Porous-walled hollow glass microspheres (PWHGMs) of a modified alkali borosilicate composition have been successfully fabricated by combining the technology of producing hollow glass microspheres (HGMs) with the knowledge associated with porous glasses. HGMs are first formed by a powder glass--flame process, which are then transformed to PWHGMs by heat treatment and subsequent treatment in acid. Pore diameter and pore volume are most influenced by heat treatment temperature. Pore diameter is increased by a factor of 10 when samples are heat treated prior to acid leaching; 100 {angstrom} in non-heat treated samples to 1000 {angstrom} in samples heat treated at 600 C for 8 hours. As heat treatment time is increased from 8 hours to 24 hours there is a slight shift increase in pore diameter and little or no change in pore volume.
Date: April 21, 2008
Creator: Raszewski, F; Erich Hansen, E; Ray Schumacher, R & David Peeler, D
Partner: UNT Libraries Government Documents Department