Glass Waste Form Research Articles

Volcanic Glass as a Natural Analog for Borosilicate Waste Glass

ABSTRACTObsidian and basaltic glass are opposite end-members of natural volcanic glass compositions. Syngenetic and diagenetic tensile failure in basaltic glass (low silica glass) is pervasive and provides abundant alteration fronts deep into the glass structure. Perlitic fracturing in obsidian (high silica glass) limits the alteration zones to an “onion skin” geometry. Borosilicate waste glass behaves similarly to the natural analog of basaltic glass (sideromelane).During geologic time, established and tensile fracture networks form glass cells (a three-dimensional reticulated pattern) where the production of new fracture surfaces increases through time by geometric progression. This suggests that borosilicate glass monoliths will eventually become rubble. Rates of reaction appear to double for every 12C° of temperature increase. Published leach rates suggest that the entire inventory of certain radionuclides may be released during the 10,000 year regulatory time period. Steam alteration prior to liquid attack combined with pervasive deep tensile failure behavior may suggest that the glass waste form is not license defensible without a metallic- and/or ceramic-type composite barrier as an overpack.

Leaching of Al2O3-Based Nuclear Waste Forms.

Abstract The chemical durability of LaAlO3 and CaAl 12 O19 containing simulated high-level waste (HLW) elements (Nd, Ce and Sr) was examined at 90 and 150°C in HCl solution and deionized water. Because of the amphotericity of the major aluminium ions and the high solubility of the other matrix constituents, both waste forms dissolved congruently in HCl solution of pH 1 at 90°C, resulting in high leach rates of more than 10−3g. m−2.d−1 for all constituents. In deionized water, they dissolved incongruently, and leach rates of Al and Nd were less than 10−4g. m−2g.d−1. The leach rates in deionized water were lower than those of borosilicate glass waste forms by about 3 orders of magnitude. CaAl12O19 was superior to LaAlO3 in the durability in HCl solution and deionized water. It was found that the leach rates of Nd and Ce from both waste forms decreased with increasing temperature in deionized water. This phenomenon was attributed to the increase of pH resulting from the dissolution of Ca or Sr.

Nuclear Waste Analytical Round Robins 1-6 Summary Report

ABSTRACTThe Materials Characterization Center (MCC) at Pacific Northwest Laboratory(a) conducts Nuclear Waste Analytical Round Robins on simulated waste glasses with and without radionuclides. The analytical round robins are used to provide analytical laboratories with a forum for improving the accuracy and precision of analytical techniques that will be used for characterizing nuclear waste glasses. Chemical analysis data on glass waste forms produced by the nuclear waste community may need to withstand legal challenges for waste material storage qualification and for licensing repositories. Determining the quality of analytical data is important for meeting these challenges, making process operation decisions, and predicting long-term waste behavior.The MCC has conducted six round robins for the waste management, research, and development community from 1987 to the present time. The laboratories that have participated regularly are Ames, Argonne, Catholic University, Lawrence Livermore, Pacific Northwest Laboratory, Savannah River, and West Valley Nuclear. The glass types analyzed in these round robins all have been simulated nuclear waste compositions expected from the vitrification of high-level nuclear waste. A wide range of analytical procedures have been used by the participating laboratories, including Atomic Absorption spectroscopy, inductively coupled plasma-atomic emission spectro-scopy, direct current plasma-emission spectroscopy, and inductively coupled plasma-mass spectroscopy techniques. The consensus average relative error for Round Robins 1 through 6 is 5.4%, with values ranging from 9.4 to 1.1%. The trend on the average improved with each round robin. When the laboratories analyzed samples over longer periods of time, the intralaboratory variability increased. Lab-to-lab variation accounts for most of the total variability found in all the round robins. To date, participation in the radiochemistry portion has been minimal and the analytical results poor when compared to the nonradiochemistry portion of the round robins. Additional radiochemical work is needed in future round robins.

Dissolution of subglacial volcanic glasses from Iceland: laboratory study and modelling

Abstract Subglacial hyaloclastites from Iceland with ages ranging from 2 ka to 2.2 Ma have been studied from a mineralogical and geochemical standpoint. The chemical composition of palagonite (alteration crust formed on the surface of the glass) is almost identical with that of the clayey material filling the intergranular spaces of the rock. The clayey material is made up of two particle populations: the first is Si-, Mg- and Ca-rich with a smectite structure, while the second is amorphous, Fe−, Ti− and Al-rich, and has a smectite-like morphology. It is suggested that these two types of particles can be formed simultaneously, in the same solution, such that it is not necessary to explain their existence by local or temporary equilibria. The mineralogical sequences observed in natural samples were reproduced using the geochemical computer code DISSOL. The geochemical mass balances calculated with DISSOL also fit quite well with those calculated from Icelandic samples, illustrating the predictive capability of such a calculation code and gives us confidence in applying a similar approach to nuclear waste-form glass problems.

Dissolution Model for a Glass Having an Adherent Insoluble Surface Layer

Waste form glasses that contain substantial quantities of iron, manganese, and aluminum oxides, such as the Savannah River SRL TDS-131 glass, form a thick, hydrated surface layer when placed in contact with water. The dissolution of such a glass has been modeled with the Savannah River Model. The authors showed previously that the equations of the Savannah River Model could be fitted to published experimental data if a time-dependent diffusion coefficient was assumed for species of diffusing through the surface layer. The Savannah River Model assumes that all of the material dissolved from the glass enters solution, whereas it was observed that substantial quantities of material were retained in the surface layer. An alternative model, presented contains a mass balance equation that allows material either to enter solution or to be retained in the surface layer. It is shown that the equations derived using this model can be fitted to the published experimental data assuming a constant diffusion coefficient for species diffusing through the surface layer.

Mineralogical, Textural and Compositional Data on the Alteration of Basaltic Glass From Kilauea, Hawaii to 300°C: Insights to the Corrosion of a Borosilicate Glass Waste-Form

ABSTRACTMineralogical, textural and compositional data accompanying greenschist facies metamorphism (to 300°C) of basalts of the East Rift Zone (ERZ), Kilauea, Hawaii may be evaluated relative to published and experimental results for the surface corrosion of borosilicate glass. The ERZ alteration sequence is dominated by intermittent palagonite, interlayered smectite-chlorite, chlorite, and actinolite-epidote-anhydrite. Alteration is best developed in fractures and vesicles where surface reaction layers root on the glass matrix forming rinds in excess of 100 microns thick. Fractures control fluid circulation and the alteration sequence. Proximal to the glass surface, palagonite, Fe-Ti oxides and clays replace fresh glass as the surface reaction layer migrates inwards; away from the surface, amphibole, anhydrite, quartz and calcite crystallize from hydrothermal fluids in contact with the glass. The texture and composition of basaltic glass surfaces are similar to those of a SRL-165 glass leached statically for sixty days at 150°C. While the ERZ reservoir is a complex open system, conservative comparisons between the alteration of ERZ and synthetic borosilicate glass are warranted.

Leach Rates and Thermal Properties of Lead-Iron Phosphate Glass Waste Forms

Lead-iron phosphate (LIP) glasses loaded with a simulated high-level nuclear waste were studied on their leach rates and thermal properties. The obtained results showed that the phosphate glass matrix consisting of lead monoxide, phosphorus pentoxide and ferric oxide of 56:35:9w/0 is able to vitrify the waste, pretreated with formic acid to remove Zr, to about 15 w/0 at 950°C. The leach rate of the vitrified waste glass was in the order of 10−7 g/cm2.d at 110°C, which is low compared with that of the borosilicate glass waste form. Increasing the phosphorus pentoxide content of the matrix to higher than 35% enabled it to produce the glass form with the waste near 20 w/0 at 950°C, but this increase rendered the glass waste form more soluble than the former. Thermal properties such as thermal expansion coefficient, critical cooling rate for vitrification and temperatures of glass transition, softening and maximum rate of crystallization were measured and discussed. Removing Na ions from wastes improves consi...

The Effect of Temperature on the Redox Constraints for the Processing of High-Level Nuclear Waste into a Glass Waste Form

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.

Comparison of Surface Layers Formed on Synthetic Basaltic Glass, French R7T7 and HMI Borosilicate Nuclear Waste form Glasses - Materials Interface Interactions Tests, Waste Isolation Pilot Plant

ABSTRACTMaterials Interface Interaction Test results for up to 2 years indicate that surface layers formed on synthetic basaltic glass in brine are predominantly Mg-chloride or oxychloride. Layers formed on HMI and R7T7 borosilicate glass are Mg-silicates. Layer thickness increases from 1 year to 2 years for basaltic glass and HMI borosilicate glass. Mg-silicate contains Ni and Zr in amounts that are generally related to the composition of associated glass, but layers formed on basaltic glass also contain Ni and Zr, elements not present in the glass. Ni and Zr are, in part externally derived, however, Ti appears to be completely immobile, and is not found in layers formed on Ti-free glass (R7T7). Some chemical components of the surface layers are derived, in part, from alteration of glass but major components of the layers are from the brine.

Leach rates and thermal properties of lead-iron phosphate glass waste forms.

Leach rates and thermal properties of lead-iron phosphate glass waste forms.

Leach Rates of Lead-Iron Phosphate Glass Waste Forms

Lead-iron phosphate glasses loaded with simulated high-level nuclear wastes at temperatures between 900 and 1,100°C were studied on their soaking behavior in distilled water by means of leachate solution analysis. The obtained results showed that the leach rates of the glass waste forms were at least 10 to 100 times lower than that of the currently investigated borosilicate glass, even though the selective release of Na ion from the forms was observed. Zirconium of the waste led the glass to partial crystallization at 900°C, but was able to be incorporated in the glass at near 1,100°C. The liquid chromatographic analysis of poly-phosphate ions in the leachate solution revealed that the low leachability of the glass forms was brought about by a certain degree of depoly-merization of long poly-phosphate chains of lead metaphosphate caused by the addition of ferric oxide.

Preferential leaching and natural annealing of alpha-recoil tracks in metamict betaflte and samarskite

Leaching experiments on naturally occurring, metamict betafite and samarskite minerals in a bicarbonate-carbonate solution show strongly enhanced release to the solution of short-lived 228Th relative to its parent isotope 232Th (by a factor of 3 to 6), but only slightly enhanced dissolution (by a factor of 1.2 to 2) of long-lived 234U and 230Th relative to 238U and 232Th, respectively. The betafite (a complex Ca–U–Ti–Nb oxide of the pyrochlore group, A2−mB2X6Y0−1.H2O) and samarskite (a complex Y–Fe–U–Nb oxide with varying stoichiometry between AB O4 and AB2O6) are x-ray and electron diffraction amorphous having experienced doses of 2.2–2.8 × 1017 alpha-decay events/mg (24–31 dpa) and 3.8–4.4 × 1017 alpha-decay events/mg (39–45 dpa), respectively. The isotopic fractionation is attributed to the radiation damage created by the alpha-decay events. Individual alpha-recoil tracks are preserved for some time as disordered regions of higher chemical reactivity in already fully damaged, aperiodic areas that result from the alpha-decay events. The annealing time of the alpha-recoil track within the aperiodic atomic array of the metamict state is calculated to be 29 300 ± 8100 years for samarskite and 42 700 ± 13 700 years for the betafite. These data plus data on an earlier studied betafite sample (annealing time = 2000 ± 1300 years) give an average annealing time of 25 000 ± 21 000 years. The variation in calculated annealing time is due, in part, to post-metamict alteration of the sample, particularly for the betafite. The wider range of values for the betafite is attributed to its greater degree of alteration. These results demonstrate that recoil nuclei of alpha-decay events may be selectively leached from damaged, aperiodic phases, but that these tracks are subject to low-temperature annealing in short periods of time relative to the age of the samples (∼ 500 m.y.). The same phenomena are expected in actinide-containing nuclear waste form glasses.

Leach rates of lead-iron phosphate glass waste forms.

Lead-iron phosphate glasses loaded with simulated high-level nuclear wastes at temperatures between 900 and 1,100°C were studied on their soaking behavior in distilled water by means of leachate solution analysis. The obtained results showed that the leach rates of the glass waste forms were at least 10 to 100 times lower than that of the currently investigated borosilicate glass, even though the selective release of Na ion from the forms was observed. Zirconium of the waste led the glass to partial crystallization at 900°C, but was able to be incorporated in the glass at near 1,100°C. The liquid chromatographic analysis of poly-phosphate ions in the leachate solution revealed that the low leachability of the glass forms was brought about by a certain degree of depoly-merization of long poly-phosphate chains of lead metaphosphate caused by the addition of ferric oxide.

Retrieval and Analysis of Simulated Defense HLW Package Experiments at the WIPP

ABSTRACTIn situ waste package performance experiments involving simulated (non-radioactive) defense high-level waste (DHLW) containers have been in progress since late 1984 at the Waste Isolation Pilot Plant (WIPP) facility. These experiments involve full-size, simulated DHLW containers of several metals and designs emplaced in the WIPP bedded rock salt. These test containers are surrounded by granular backfill (packing) materials, have in many cases been intentionally injected with brines, and are heavily instrumented. A majority of the test packages also contain nonradioactive DHLW borosilicate glass waste form, either within the container and/or outside of it. The primary purpose of these WIPP simulated DHLW experiments is to evaluate the in situ durability and performance of all waste package engineered barrier materials, and to perform package concept validation testing.Twelve of the test DHLW containers, emplaced in WIPP test Room B, have been in heated operation since 1985 and had a maximum surface temperature of about 190°C. These containers were recently retrieved, after about 3 years of heated exposure, for detailed posttest laboratory analyses of: general corrosion and metallurgical degradation, waste form and backfill materials alterations, and other rock salt-brine-barrier materials near-field interactions with the “repository” geochemical environment. Test canisters and overpacks made of ASTM Grade-12 titanium showed essentially no visible degradation in either the base metal or welds; cast mild steel A216/WCA over-packs have suffered some uniform corrosion. Significant degradation of the removed instruments and associated test apparatus has been found: pieces of stainless steel (both 304L and 316) apparatus have undergone extensive stress-corrosion cracking failure and non-uniform attack; Inconel 600-sheathed instruments have undergone both extensive uniform and localized (pitting) attack. Granular backfill materials have been significantly compacted by creep closure to about a density of 2 kg/m. Laboratory analyses are still in progress. Further details on these materials results plus instrumentation data and other in situ WIPP waste package test observations are discussed.

Leach Rates of Composite Waste Forms of Monazite- and Zirconium Phosphate-Type

Leaching behavior of simultaneously synthesized composite waste form of monazite- and zirconium phosphate-type was investigated by making use of simulated high level waste (HLW). The obtained results showed that the leach rates of the composite forms are very low compared to that of the borosilicate glass waste form, indicating that the composite is regarded as being significant to immobilize HLW grossly like vitrification.

The Formation of Surface Layers and Reaction Products in the Leaching of Defense Borosilicate Nuclear Waste Glass

The development of surface alteration layers and solid products in the reaction of borosilicate nuclear waste glass with aqueous solutions has been studied as a function of time, surface morphology, water flow rates, and solution composition. A physical description of the growth of the surface layers has been developed, and several solid reaction products have been identified. The results of the study support a saturation based description of long-term radionuclide release from glass waste forms and demonstrate the complex dependence of short-term leaching experiments upon both surface and solution variables.

Water chemistry and dissolution kinetics of waste-form glasses

A simplified water chemistry program was developed for determining the chemical speciation of a solution. This was combined with a linear, first-order kinetic equation to produce a model of the kinetics of dissolution of a nuclear waste glass which includes effects of groundwater flow. Calculations are presented for four waste glasses, giving the rate of glass dissolution in both pure water and a solution intended to simulate basalt groundwater. The calculations are performed for the case of dissolution in a static volume of groundwater and for the case of steady-state dissolution in flowing groundwater. Inclusion of the effects of water chemistry on the solubility of silicon is found to increase the rate of glass dissolution in the case of pure water but to either increase or decrease it in the case of the simulated groundwater, depending upon the glass composition. Implications of the results for prediction of the long-term behavior of a nuclear waste glass are discussed.

Integrated Testing of the Srl-165 Glass Waste Form

AbstractIntegrated testing of the important components of a glass waste form waste package has been performed in order to gain a better understanding of the processes of radionuclide release and transport in the near field environment. Based upon an interpretation of the depth of penetration of hydrogen in reacted SRL-165 glass we have modeled the radionuclide release from the glass as a combined process of (1) the diffusive exchange of alkalis and boron in the glass for hydrogen species in the solution (D=10−16 cm2/s) and (2) surface dissolution. Surface dissolution controls the release of components not exchanged by diffusion and takes place at a rate of 1.5-3.0 μm/yr. Subsequent to release the radionuclides may remain in the leach solution, diffuse into the tuff, or precipitate as secondary phases. Precipitation is particularly important for plutonium and americium. Diffusive transport of radionuclides through the tuff takes place at an extremely slow rate, D=10−16 cm2/s. As such, the mass of radionuclides incorporated in the tuff by diffusion during the tests is inconsequential relative to that in the leach solution (with the exception of plutonium) and can be ignored in mass balance calculations. Mass balance calculations based upon the release of radionuclides by surface dissolution of the glass waste form are in good agreement with observed solution chemistry when allowances are made for a pulse of dissolution early in the tests. This pulse may be due to either the rapid dissolution of high-energy surface features early in the inLegrated tests, or an initially high surface dissolution rate that decreases with time as silica saturation is approached [1], or a combination of the two.

The Effects Of Gamma Radiation on Groundwater Chemistry and Glass Reaction In a Saturated Tuff Environment

AbstractThe Nevada Nuclear Waste Storage Investigations project has completed a series of experiments that provide insight into groundwater chemistry and glass waste form performance in the presence of a gamma radiation field at 90°C. Results from experiments done at 1 × 103 and 0 R/hr are presented and compared to similar experiments done at 2 × 105 and 1 × 104R/hr. The major effect of radiation is to lower the groundwater pH to a value near 6.4. The addition of glass to the system results in slightly more basic final pH, both in the presence and absence of radiation. However, there is essentially no difference in the extent of glass reaction, as measured by elemental release, as a function of dose rate or total dose, for reaction periods up to 278 days.

Natural Analogues: Their Application to the Prediction of the Long-Term Behavior of Nuclear Waste Forms

AbstractOne of the unique and scientifically most difficult aspects of nuclear waste isolation is the extrapolation ofshot-term laboratory data (hours to years) to the long time periods (103-105 years) required by regulatory agencies for performance assessment. The direct verification of these extrapolations is not possible, but methods must be developed to demonstrate compliance with government regulations and to satisfy the lay public that there is a demonstrable and reasonable basis for accepting the long-term extrapolations. Natural analogues of both the repository environment (e.g. radionuclide migration at Oklo) and nuclear waste form behavior (e.g. alteration of basaltic glasses and radiation damage in minerals) have been used to demonstrate the long-term behavior of large scale geologic systems and, on a smaller scale, waste form durability. This paper reviews the use of natural analogues to predict the long-term behavior of nuclear waste form glasses. Particular emphasis is placed on the inherent limitations of any conclusions that are based on “proof” by analogy. An example -- corrosion of borosilicate glass -- is discussed in detail with specific attention to the proper and successful use of natural analogues (basaltic glass) in understanding the long-term corrosion behavior of borosilicate glass.