A single borosilicate glass composition has previously been shown to dissolve 10 and 25 mass% PuO2 under oxidizing and reducing conditions, respectively. A simplified version of this glass has been thoroughly investigated to determine the effect of increasing the alkali:aluminum ratio on the HfO2 solubility in borosilicate glasses. We are investigating HfO2 solubility because specific Pu wastes are being considered for disposal in glass, and Hf(IV) serves as a structural surrogate for Pu(IV) and as a neutron absorber in glass. Three series of base glasses were produced using the same initial composition, but varying the oxides B2O3, Al2O3, or Na2O one at a time. In a fourth series of the same initial composition, both Na2O and A12O3 were varied. Hafnia was added to these glasses and the mixture equilibrated for 2 hours: 1 hour at 1450°C after 1 hour at 1560°C. A wide range of HfO2 additions were made to the base glasses, and the solubility of HfO2 determined to within ±1 mass%. The highest solubility determined was 14 mol% (35 mass%) HfO2 in a low-Al glass. We conclude that increasing Na2O/Al2O3 increases the HfO2 solubility, and increasing the B2O3 content apparently has little effect on HfO2 solubility in the borosilicate glasses studied.

Mixtures of rare earth and actinide oxalates will be vitrified into boro-aluminosilicate-based glasses for intermediate term stabilization according to current plans. The reaction chemistry involved with converting these oxalate feed stocks into glass products determines the potential for foaming, redox, and other melt and off gas related phenomena associated with this process. We've undertaken a detailed study of this conversion process using a variety of complementary techniques. A closed quartz crucible contained in a vertical furnace equipped with a quartz window and video camera was used to study volume expansion of the feed/melt during heating while monitoring the off-gas using a gas chromatograph-mass spectrometer. Simultaneous thermogravimetric and differential thermal analyses were conducted on small samples of feed and frit mixtures. Samples containing Ce were analyzed using established wet chemical techniques to determine Ce3+/Ce4+ ratio (redox) as a function of temperature. We evaluate the results and provide a description of the reaction chemistry of these oxalate feeds during vitrification.

Gadolinium and lanthanum solubility limits in a sodium-alumino-borosilicate glass system were studied. As melting temperature increased from 1400°C to 1450°C, 1500°C and 1550°C, the solubility of gadolinium in the baseline glass 15B2O3-5A12O3-20Na2O-60SiO2 (in molar composition) increased from 10.1 to 11.3, 12.2 and 13.1 (in mole percent of Gd2O3). The enthalpy change of Gd2O3 dissolution in this baseline glass is about 43.6 kJ/mol. Boron effect on lanthanum solubility was studied using the following baseline glasses: xB2O3-20Na2O- 5Al2O3-60SiO2, where x equals to 5, 10, 15, and 20, respectively. It was found that lanthanum solubility limit increased from 8.4 to 10.3, 12.5 and 14.9 (in mole percent of La2O3) as B2O3 increased from 5.1 to 9.5, 13.1 and 16.2 mol%. Gd2O. and La2O3 have similar solubility limits. Solubility limits of mixtures containing different ratios of Gd2O3 to La2O3 in the baseline glass 15B2O3-20Na2O-5A12O3-60SiO2 were found insensitive to the ratio of La/Gd. As far as gadolinium is concerned, its solubility limit will decrease when other lanthanides are introduced.

Magic Angle Spinning Nuclear Magnetic Resonance (MAS-NMR) spectroscopy has been used to characterize the structural and chemical environments of B, Al, and Si in model Hanford low-activity waste glasses. The average 29Si NMR peak position was found to systematically change with changing glass composition and structure. From an understanding of the structural roles of Al and B obtained from MAS-NMR experiments, we first developed a model that reliably predicts the distribution of structural units and the average 29Si chemical shift value, δ, based purely on glass composition. A product consistency test (PCT) was used to determine the normalized elemental release (NL) from the prepared glasses. Comparison of the NMR and PCT data obtained from sodium boro-aluminosilicate glasses indicates that a rudimentary exponential relationship exists between the 29Si chemical shift value, and the boron NL value.

Molecular dynamics simulation techniques were applied to model the incorporation of plutonium in the French nuclear waste glass matrix. Born-Mayer-Huggins analytical potentials were established to characterize short-range interactions between Pu–O and Pu–Pu pairs; the potentials were fitted to the structural properties of plutonium dioxide in the light of a recent experimental study showing that plutonium is found as Pu(IV) in the glass. The transferability of the established potentials to the glass structure is discussed, and the potential parameters are further refined by molecular dynamics simulations in an aluminoborosilicate glass to obtain mean Pu–O interatomic distances and firstneighbor coordination numbers matching the experimental values as closely as possible. Previously published Born-Mayer-Huggins potentials supplemented by Stillinger-Weber three-body terms were used for oxygen-cation and cation-cation interactions. The difficulties encountered in establishing a Pu–O potential that provides satisfactory results in both oxides and glasses are also discussed.

Water content of glasses is an important factor in glass manufacture. In the case of slurryfed processing of waste glass melters, the water content of waste glasses may change the waste form properties significantly. A reliable, non-contact and quantitative measurement technique, with a spatial resolution of about 2-mm was used to determine the concentration of dissolved water molecules in simulated nuclear waste glasses. The 19F resonant reaction was used. The 1H (19F, αγ)16O reaction requires a resonance energy that is available by use of a 6.471 MeV19F beam. A series of ten segments of slurry-fed, joule-heated, research-scale melter tests were conducted to determine the redox effects on processing of various feeds. Representative glass samples were collected at the middle of each segment. Their hydrogen contents were measured. The hydrogen content was correlated to the feed processing and redox of the glasses.

We have been investigating the spectroscopy of actinides in materials having relevance to nuclear waste management. Discussed here are the excitation and emission profiles of curium in high temperature borosilicate glasses, and pure halide compounds. Probing the local coordination environment of Cm(III) in these matrices has been one of the goals of this study. Comparison of the nephelauxetic effects indicate that the silicate matrix provides a more basic environment than either F− or Cl− surroundings found in pure compounds of curium. As a result, f - f transition energies are red shifted significantly in the silicate matrix as compared to that of the neat CmF3 system. Moreover, laser induced emissions have been noted with CmF4 after treatment at elevated temperatures. The higher energy spectral region, consisting of emission bands at 20100 and 16750, is assignable to f - f transitions in Cm(III), which is the expected thermal decomposition product. Energy transfer from Cm(III), followed by a second photon absorption is believed to be responsible for the strong emission bands at 12550, and 13580 cm−1. from presently unidentified emitting species. Results are discussed in relation to the oxidation state and fluorescence dynamics of actinide species in neat compounds, as well as in matrices being considered for waste immobilization.

The successful demonstration of sequentially drying, calcining and vitrifying an oxalate slurry in the Drain Tube Test Stand (DTTS) vessel provided the process basis for testing on a larger scale in a cylindrical induction heated melter. A single processing issue, that of batch volume expansion, was encountered during the initial stages of testing. The increase in batch volume centered on a sintered frit cap and high temperature bubble formation. The formation of a sintered frit cap expansion was eliminated with the use of cullet. Volume expansions due to high temperature bubble formation (oxygen liberation from cerium reduction) were mitigated in the DTTS melter vessel through a vessel temperature profile that effectively separated the softening point of the glass cullet and the evolving oxygen from cerium reduction. An increased processing temperature of 1470°C and a two hour hold time to fine any remaining bubbles successfully reduced bubbles in the poured glass to an acceptable level. The success of the preliminary process demonstrations provided a workable process basis that was directly applicable to the newly installed Cylindrical Induction Melter (CIM) system, making the batch flowsheet the preferred option for vitrification of the americium-curium surrogate feed stream.

A new vitrification process for radioiodine has been developed for safe disposal of the spent iodine adsorbent generated from reprocessing off-gas systems. The proposed process consists of the following two steps: (1) separation of radioiodine as AgI from spent iodine adsorbent and (2) vitrification of the separated AgI with silver phosphate. An AgI-Ag2O-P2O5 glass system can homogeneously contain up to 60mol%AgI, which leads to high volume reduction efficiency (approximately 1/25). It also can vitrify the AgI without volatilization of iodine because of its low melting temperature (below 400°C). The leachabilities of iodine from simulated vitrified waste forms were evaluated by the MCC-1 static leach test in an anaerobically controlled glove box, which was purged by nitrogen gas with 3% hydrogen. The leaching behavior of the Agl-Ag2O-P2O5 glass system was influenced by the composition of the actual glass, type of leachant, and redox conditions. When the leach test was carried out using simulated ground wateroriginating from rainfall, the leach rate of iodine from 3AgI-Ag4P2O7 glass was 6×10−8g/cm2/d in the early period of the leach test at 35°C under the reducing condition, after which it decreased to 3×10−9g/cm2/d. This glass showed a lower leach rate in the simulated ground water originating from sea water or cement saturated water, since a precipitation layer was formed on the surface.

A relatively simple model, the associate species model, is being applied to nuclear waste glass compositions in order to accurately predict behavior and thermodynamic activities in the material. In the model, the glass is treated as a supercooled liquid, with the liquid species allowed to exist below their melting point. The approach requires an initial assembly of binary and ternary oxide liquid solution data that sufficiently reproduce the equilibrium phase diagrams. Two binary oxide subsystems, MgO-CaO and MgO-Al2O3, have been modeled and results compared to published phase diagrams. Computed activities of the glass constituent species are plotted as a function of composition at 1200°C.

GLASSFORM-Version 1, a spreadsheet or an algorithm software package, was developed for generating preliminary glass formulations for waste streams. A spreadsheet program on Microsoft Excel 97 for Windows 95 was also developed for predicting key glass properties for laboratory scale vitrification studies of simulated waste streams. The algorithm version based on Modula 2, which can be run on Windows-95 or NT, was developed to plot the glass component ratios versus the waste loading. At this time the algorithm was not developed for predicting the key glass properties. The spreadsheet or the algorithm version is an effective tool for developing preliminary glasses with a potential to be chemically durable, dense for volume reduction, of low viscosity for glass pouring, redox controlled, and resistant to corrosion of melter components such as Inconel 690. For a given solid waste stream oxide composition in wt.% or in grams or in ppm and waste loading in wt.%, the spreadsheet or the algorithm calculates glass component ratios that provide an empirical indication of the quality of a candidate glass. In addition to the component ratios for glass quality indicators, glass property composition relationships for glass durability and processability were incorporated into the spreadsheet version. These spreadsheet or algorithm versions can also be used for studying the effects of such actions as varying waste loadings or of varying individual waste components on glass properties and glass component ratios, provided specific glass property models are incorporated. As an example, the software was applied to candidate phosphate-containing glasses at the Idaho Nuclear Technology and Engineering Center (INTEC, previously Idaho Chemical Processing Plant) and also the borosilicate glasses such as the Savannah River Laboratory Environmental Assessment (EA) glass, the Hanford standard (ARM-1) glass, and the French R7T7 glass. Application of the software to new waste streams such as at Hanford, Fernald, and other countries is the most useful aspect of the GLASSFORM software during the early stages of glass waste form development.

The dissolution rate of a simulated Magnox waste glass has been investigated in single-pass flow-through experiments designed to investigate the role of Al and Si in the dissolution process. The results indicate that both Al and Si species suppress the rate of dissolution. These effects may be modelled using a combined Al/Si affinity term in a conventional glass dissolution rate law. Aluminium species may also play an inhibitory role when present at relatively high solution activities. In Si-rich alkaline media, the concentration of aluminium is controlled to very low levels by the development of secondary aluminosilicate phases. Removal of Al by secondary phase precipitation results in dissolved Al activities below that required to reach ‘saturation’ with respect to the glass.

We have developed a Monte Carlo (simulation) model to describe glass dissolution. This model uses similar parameters to those used in the Grambow model, which has been the main glass dissolution model during the last decade. While the Grambow model is macroscopic, ours is microscopic. This assures that the value of our parameters is time independent and allows changes in the gel structure (not just concentrations) to be described.

We relate our parameters to those of the Grambow model and test whether the basic assumptions of the Grambow model are consistent with our simulation results. In these simulations, as the solution evolves towards silica saturation, the silica concentration in solution can become higher than the (final) silica saturation concentration. A possible explanation for this behavior, which is observed experimentally as well, is that the silica saturation concentration is not constant as a function of time. Initially, the silica saturation concentration is the glass saturation concentration. The glass saturation concentration is higher than the final silica saturation concentration, which corresponds to saturation of the gel. In the Grambow model, the initial dissolution rate (at the gel/water interface) and the diffusion coefficient of silica in the gel are not constant either. The existence of a final dissolution rate depends on the silica content of the glass and on the protection provided by the gel layer. In the simulations, a protective gel is mainly formed by the adsorption (precipitation) of dissolved silica particles from the solution in contact with the gel. This makes us expect that (1) the gel would be non protective for dynamic tests (with a low silica content in solution) and (2) the gel would be very protective in static tests with high surface to volume ratios.

The pressurized unsaturated flow (PUF) test is a new experimental method for the evaluation of the long-term corrosion behavior of waste forms and other engineered barrier materials. Essentially, the technique provides a means to flow water through a porous bed of test material or materials at elevated temperature and under hydraulically unsaturated conditions. Bulk volumetric content, effluent pH and electrical conductivity are monitored in real time using a computer control and data acquisition system. In previous papers, we have reported on the changes in bulk water content, effluent chemistry, and glass corrosion rates that result from the formation of alteration products during these tests. These measurements are now supplemented through the use of the ultracentrifugation apparatus (UFA) for hydraulic property measurements and highresolution, x-ray microtomography (XMT) to provide 3-D spatial and temporal imaging of water distribution and pore structure alteration during these tests. Quantitative changes in the water retention characteristic were correlated with the onset of zeolite formation in the tests. Extensive alteration of the glass resulted in cementation of the glass grains near the bottom of the column, which was observed in situ using the XMT.

Oxidative dissolution of spent UO2 fuel in vapor and dripping groundwater at 90°C occurs via general corrosion at fragment surfaces. Dissolution along fuel-grain boundaries is also evident in samples contacted by the largest volumes of groundwater, and corroded grain boundaries extend at least 20 or 30 grains deep (> 200 μm), possibly throughout mm-sized fragments. Apparent dissolution of fuel along defects that intersect grain boundaries has produced 50 to 200 nm-diameter dissolution pits that penetrate 1–2 μm into each grain, giving rise to a “worm-like” texture along fuel-grain boundaries. Sub-micrometer-sized fuel shards are common between fuel grains and may contribute to the reactive surface area of fuel exposed to groundwater. Outer surfaces of reacted fuel fragments develop a fmne-grained layer of corrosion products adjacent to the fuel (5–15 μm thick). A more coarsely crystalline layer of corrosion products commonly covers the fine-grained layer, the thickness of which varies considerably among samples (from less than 5 μm to greater than 40 μm). The thickest and most porous corrosion layers develop on fuel fragments exposed to the largest volumes of groundwater. Corrosion-layer compositions depend strongly on water flux, with uranyl oxy-hydroxides predominating in vapor experiments, and alkali and alkaline earth uranyl silicates predominating in high drip-rate experiments. Low drip-rate experiments exhibit a complex assemblage of corrosion products, including phases identified in vapor and high drip-rate experiments.

In the early 1980s, tests of the leaching behavior of spent light water reactor fuel were conducted in Sweden by SKB and in the USA by the NNWSI Project. Both organizations used fuels with similar burnup, leaching solutions with similar chemical compositions, and conducted the tests at ambient hot cell temperature. Most of the test results were closely similar. The exception was in the recovery of actinide elements at the end of leaching cycles. In the NNWSI tests, the test vessels were stripped with nitric acid at the end of the leaching cycle. When the actinide inventories recovered in the original leaching cycle plus vessel rinse solutions were added to the amount recovered from the acid stripping, the relative abundances of uranium, plutonium, and other actinides were approximately the same as their inventories in the fuel samples. In the SKB tests, the materials recovered from stripping the leaching vessels were low in amount and interpreted to be fine fuel fragments. Only a few percent of the plutonium inventory that would correspond to the uranium recovered in solution was accounted for in the solution samples. To investigate the reasons for this difference in recovery of the actinides, a new test was undertaken using a fuel sample that had been leached for several years using the SKB methods. The fuel was removed from the cladding after a total of a bit more than two years of leaching inside the cladding. The bare fuel was then leached for several cycles using a geometry that simulated the NNWSI tests and a test procedure that was similar to that used in the NNWSI tests. The cladding from which the fuel was removed was also leached in a separate vessel in parallel with the bare fuel leaching. This paper presents the results of the test series and discusses the effects of specimen geometry on the mobility of actinides during leaching. During leaching of spent fuel inside the original cladding, a secondary phase containing Pu seems to form. This phase appears to be less easily dissolved than spent fuel itself.

α-decay will constitute almost entirely the radiation field in and around spent nuclear fuel after a few hundred years in a geological repository. Pellets of UO 2 containing ˜0.1 and ˜10 wt. % 238Pu were fabricated using a sol-gel method and characterized, comparing their properties to those of undoped UO2. The α-radiation fields of different types of commercial LWR spent fuel are of the same order of magnitude as the fuel with the lower Pu-concentration used in this work. The results of static batch leaching tests at room temperature in demineralized water under anoxic atmosphere showed that the amounts of U released during leaching were higher in the case of UO2 containing 238pu than for undoped UO2. Relatively large amounts of Pu were released after the longest leaching times. Lattice parameter measurements using XRD and hardness measurements by Vickers indentation showed a relatively rapid build-up of α-decay damage in the material stored at ambient temperature with the higher concentration of dopant, while for the material with ˜0.1 wt. % Pu no clear variations were detected during the same time intervals.

Further refinements to the oxidation model of Stout et al. have been made. The present model incorporates the burnup dependence of the oxidation rate and an allowance for a distribution of grain sizes. The model was tested by comparing the model results with the oxidation histories of spent-fuel samples oxidized in thermogravimetric analysis (TGA) or oven dry-bath (ODB) experiments. The experimental and model results are remarkably close and confirm the assumption that grain-size distributions and activation energies are the important parameters to predicting oxidation behavior. The burnup dependence of the activation energy was shown to have a greater effect than decreasing the effective grain size in suppressing the rate of the reaction U4O9→U3O8. Model results predict that U3O8 formation of spent fuels exposed to oxygen will be suppressed even for high burnup fuels that have undergone restructuring in the rim region, provided the repository temperature is kept sufficiently low.

An electrochemically based model for predicting the effects of α-radiolysis, the precipitation of U(VI) corrosion products and redox processes with Fe and Fe(II) on the dissolution of UO2 is described. Various aspects of the model are presented, including: the underlying mechanism, the reaction-diffusion equations used to describe the mass transport and homogeneous reactions of the various species considered in the model, the geometrical grid used to simulate both experimental and used fuel/container geometries and the electrochemical boundary conditions used for the numerical solution of the reaction-diffusion equations. The results of preliminary simulations are also discussed.

A mechanistic spent fuel dissolution model has been developed, based on the generalreaction-transport code of AREST-CT. It considers the dissolution of spent fuel under flow conditions. The kinetic reactions of spent fuel dissolution and precipitation of schoepite, uranophane, soddyite, and Na-boltwoodite are included in the model. The results of model prediction are compared against the results of drip-tests that simulate the conditions that may occur in the Yucca Mountain Repository. Comparison shows that the modeling results match the laboratory observations very well and no contradiction has been found. It indicates that the model is a reasonably good representation of the real system.

After validation, the model was used to investigate the release rate of Np from the dissolution of secondary uranyl minerals by examining various degrees of Np incorporation into secondary uranyl minerals. The predicted Np concentration in the aqueous phase is 3 orders of magnitude lower than the upper-bound of the Np solubility range currently used in DOE performance assessment analyses. It suggests that the Np solubility range currently used is too conservative and could be replaced with more realistic values.