The setting of the technical subsystem within the overall socio-political-economic-technical radwaste system will be described and the highly interactive nature of the larger setting emphasized. It will be shown that because of the dominance of the socio-political subsystem, the importance of the technical subsystem is overshadowed. Moreover the key issue in the technical subsystem, whether to put more reliance on the immobilization (via the waste package or engineeered barriers) or the isolation (via the geology) has stayed tilted toward the latter since 1978, when we organized the first symposium on radwaste science (at the Materials Research Society meeting). Now that the isolation strategy is stymied, the opportunity arises again for the materials community to make a compelling case for the waste package's true significance.

This review is made mainly from the perspective of one laboratory in one country, the U.S. What is remarkable about the state of research in the technical subsystem is how little the big picture of the science or the technology has changed after some billions spent on R/D. The borosilicate glass (almost unchanged) is still the establishment's choice of reference waste form for HLW Cost, however, is finally forcing cement encapsulated forms to be given a second look. Mineral-modeled ceramics have received a great deal of scientific attention but remain esoteric to managers. It will be shown that in the author's opinion an enormous amount of detailed science has been done but most of it is unlikely to prove to be of any relevance or use. The policy implications for future R/D are discussed.

Crystallization kinetics of a simulated high-level waste (HLW) glass were measured and modelled. Kinetics of acmite growth in the standard HW39–4 glass were measured using the isothermal method. A time-temperature-transformation (TTT) diagram was generated from these data. Classical glass-crystal transformation kinetic models were empirically applied to the crystallization data. These models adequately describe the kinetics of crystallization in complex HLW glasses (i.e., RSquared = 0.908). An approach to measurement, fitting, and use of TTT diagrams for prediction of crystallinity in a HLW glass canister is proposed.

A single-step plasma arc-vitreous ceramic (PAVC) process is described for converting spent nuclear fuel (SNF), SNF sludges, and associated wastes into a vitreous ceramic waste form. This proposed technology is built on extensive experience of nuclear waste form development and nuclear waste treatment using the commercially available plasma arc centrifugal (PAC) system. SNF elements will be loaded directly into a PAC furnace with minimum additives and converted into vitreous ceramics with up to 90 wt% waste loading. The vitreous ceramic waste form should meet the functional requirements for borosilicate glasses for permanent disposal in a geologic repository and for interim storage. Criticality safety would be ensured through the use of “batch” modes, and controlling the amount of fuel processed in one batch. The minimum requirements on SNF characterization and pretreatment, the one-step process, and minimum secondary waste generation may reduce treatment duration, radiation exposure, and treatment cost.

This paper describes a new concept for a high-temperature, electrodeless melter for vitrifying radioactive wastes. Based on the principles of induction heating, it circumvents a number of difficulties associated with existing technology. The melter can operate at higher temperatures (1500–2000°C vs 1150°C), allowing for a higher quality, more durable glass which reduces the long-term leaching rate. Higher processing temperatures also enable conversion from borosilicate to high-silica glass which can accommodate 2 to 3 times as much radioactive waste, potentially halving the ultimate required long-term disposal space. Finally, with high temperatures, conversion of nuclear waste into ceramics can also be considered. This too leads to higher waste loading and the reduction of repository space. The melter is toroidal, linked by an iron core transformer that allows efficient electrical operation even at 60 Hz. One-dimensional electrical and thermal analyses are presented.

We have pursued some of the fundamental chemistry and materials science of americium in three glass matrices, two being high-temperature (850° and 1400°C melting points) silicate-based glasses and the third a sol-gel glass. Optical spectroscopy was the principal investigating tool in the studies. One aspect of this work was to determine the oxidation state exhibited by americium in these matrices, as well as factors that control and/or may alter this state. We have noted a correlation between the oxidation state of the f-elements in the two high-temperature glasses with their high-temperature oxide chemistries. One exception was americium: although americium dioxide is the stable oxide encountered in air, when this dioxide was incorporated into the high-temperature glasses, only trivalent americium was found in the products. When trivalent americium was used to prepare the sol-gel glasses at ambient temperature, and after these products were heated in air to 800°C, again only trivalent americium was observed. Potential explanations for the unexpected behavior of americium is offered in the context of its basic chemistry. Experimental spectra, spectroscopie assignments and other pertinent data obtained in the studies are discussed.

Several oxidation states of neptunium and plutonium, Pu(III), Pu(IV), Pu(VI), Np(IV), Np(V), and Np(VI), were studied in glasses prepared by a sol-gel technology. The oxidation state of these actinides in the sol-gel product was examined by absorption spectroscopy after solidification, aging, and thermal treatment. The oxidation state of the actinides in the starting solutions was essentially maintained through the solidification process of the silica matrix. However, during densification and removal of residual solvents at elevated temperatures, both actinides converted eventually to their tetra valent states while in the different sol-gel products. This finding is in accord with reports that tetravalent states of plutonium and neptunium are acquired in glass products prepared by dissolution of the actinide in molten glasses. Comparisons between room temperature spectra obtained from neptunium and plutonium in heated sol-gel products and from molten glass products showed subtle differences that can be related to the metal ion's environments.

CCIM technology allows syntheses of various materials over a wide range of compositions. This provides a means of applying the CCIM to the solidification of wastes obtained by different technologies of the spent nuclear fuel reprocessing, including toxic wastes containing heavy metals and those arising from fractionation. Solid compounds produced by this method meet both the requirements for their subsequent disposal in deep geological formations and spent fuel standards.

On the basis of investigations into the REE and Uranium vitrification, the possibilities of CCIM immobilizing weapons Plutonium in preparation of glass compositions are considered.

The kinetics of spinel crystallization from a molten high-iron simulated high-level nuclear waste glass was studied using isothermal heat treatments. Optical microscopy with image analysis was used to measure volume fraction of spinel as a function of heat treatment time and temperature. The Johnson-Mehl-Avrami equation was fitted to data to determine kinetic coefficients for spinel crystallization. The liquidus temperature and Avrami number are TL = 1337K and n = 1.5.

The liquidus temperature (TL) often limits the loading of high-level waste in glass through the constraint that TL must be at least 100°C below the temperature at which the glass viscosity is 5 Pa-s. In this study, values of TL for spinel primary crystalline phase were measured as a function of glass composition. The test glasses were based on high-iron Hanford Site tank wastes. All studied glasses precipitated spinel (Ni,Fe,Mn)(Cr,Fe)2O4 as the primary crystalline phase. TL was increased by additions of Cr2O3, NiO, Al2O3, Fe2O3, MgO, and MnO; while Li2O, Na2O, B2O3, and SiO2 had a negative effect. Empirical mixture models were fitted to data.

A study was conducted on glasses based on a simulated transuranic waste with high concentrations of ZrO2and Bi2O3 to determine the compositional dependence of primary crystalline phases and liquidus temperature (TL). Starting from a baseline composition, glasses were formulated by changing mass fractions of Al2O3, B2O3, Bi2O3, CeO2, Li2O, Na2O, P2O5, SiO2, and ZrO2, one at a time, while keeping the remaining components in the same relative proportions as in the baseline glass. Liquidus temperature was measured by heat treating glass samples for 24 h in a uniform temperature furnace. The primary crystalline phase in the baseline glass and the majority of the glasses was zircon (ZrSiO4). A change in the concentration of certain components (Al2O3, ZrO2, Li2O, B2O3 and SiO2) changed the primary phase to baddeleyite (ZrO2), while cerium oxide (CeO2) precipitated from glasses with more than 3 wt% CeO2. Zircon TL was strongly increased by Al2O3, Zrb2 and CeO2, and slightly by P2O5 and SiO2; decreased strongly by Li2O and Na2O and moderately by B2O3. A first-order model was constructed for TL as a function of composition for zircon primary crystalline phase glass.

A model is presented based upon calculated bridging oxygens which allows the prediction of the region of acceptable glass compositions for a lime-soda-silica glass-forming system containing mixed waste. The model can be used to guide glass formulation studies (e.g., treatability studies) or assess the applicability of vitrification to candidate waste streams.

The Savannah River Site has the majority of the United States' supply of neptunium currently stored in an acid solution in one of their canyon facilities. A program is being developed that could be utilized to ship this material, as glass, to Oak Ridge National Laboratory where the Np could be leached from the glass, purified by ion exchange and made into target material for the production of Pu-238. Ion exchange purification dictates no material be in the leachate making the isolation of the Np difficult. We have developed a process using thorium as a surrogate for Np that could immobilize the Np into a soda borosilicate glass for shipment. To achieve recovery of the Np, the glass can be phase separated prior to leaching with nitric acid. Phase separation would produces a Np-rich sodium-borate phase and a Si-rich phase similar to a Vycor® glass. The nitric acid selectively attacks the sodium-borate phase allowing high Np recovery in a solution that contains only sodium and boron. These can be easily separated from Np by ion exchange. Essentially all of the silicon which would interfere with ion exchange by precipitation is retained in the Vycor®-type phase. This technology may also be applied to other actinides stored in relatively pure solutions.

This paper will report the optimization of variables for maximizing Th (a Np surrogate) recovery while minimizing Si release. Th solubility in glass, heat treatment conditions and leaching parameters will be discussed. Transmission Electron Microscopy (TEM) with energy dispersive spectroscopy (EDS) data will be included to show phase separation after heat treatment.

Immobilization by vitrification is one potential disposition option for a portion of the United States' excess plutonium inventory. Research has been performed at the Savannah River Site (SRS) to determine the optimum composition of a lanthanide borosilicate frit for the vitrification of plutonium using a Plackett-Burman design and simplex algorithm as a statistical tool. This technique uses various response variables to rank and optimize a composition. The variables used in this study correspond to homogeneity, durability, actinide solubility and devitrification after heat-treatment.

The optimized frit composition was determined using a constant ThO2 loading of 20 wt%. No noticeable trends were followed with respect to the individual components which may indicate a relatively robust system able to accommodate variations in the feed.

Batches containing various loadings of ThO2 were melted to determine if actinide solubility was improved in the optimized composition compared to that of a similar lanthanide borosilicate glass. No noticeable improvement in ThO2 solubility was realized as a result of using this optimization technique.

A solution containing kilogram quantities of highly radioactive isotopes of amerícium and curium (Am/Cm) and lanthanide fission products is currently stored in a process tank at the Department of Energy's Savannah River Site (SRS). This tank and its vital support systems are old, subject to deterioration, and prone to possible leakage. For this reason, a program has been initiated to stabilize this material as a lanthanide borosilicate (LBS) glass.1 The Am/Cm has commercial value and is desired for use by the heavy isotope programs at the Oak Ridge National Laboratory (ORNL).

A recovery flowsheet was demonstrated using a curium-containing glass to extract the Am/Cm from the glass matrix. The procedure involved grinding the glass to less than 200 mesh and dissolving in concentrated nitric acid at 110°C. Under these conditions, the dissolution was essentially 100% after 2 hours except for the insoluble silicon. Using a nonradioactive surrogate, the expected glass dissolution rate during Am/Cm recovery was bracketed by using both static and agitated conditions. The measured rates, 0.0082 and 0.040 g/hrcm2, were used to develop a predictive model for the time required to dissolve a spherical glass particle in terms of the glass density, particle size, and measured rate. The calculated dissolution time was in agreement with the experimental observation that the curium glass dissolution was complete in less than 2 hr.

Induction melting by using a cold crucible is a suitable technology for the immobilisation of ash residues after incineration of solid radioactive waste. We investigated the possibility of using glass composites produced by stirring the ash into meltedglass. Glass composites containing 15 -40 wt. % of ash were obtained in both laboratory and bench scale units. Infrared spectroscopy, electronic paramagnetic resonance, X-ray diffractometry, TEM and SEM analyses were applied in order to characterise the structure of the glass composites obtained. The glass composites consisted of a relatively homogeneous glass matrix with embedded polycrystalline aggregates. The fraction of aggregates increases when the fraction of ash rises. The isothermal curing of composites at 1100 °C leads to dissolution of the ash components into the melt as well as to their inclusion into the glass structure, according to the analysis of the spectra obtained.

Vitrification is the most developed method for the immobilisation of highly radioactive waste obtained from reprocessing irradiated nuclear fuel. Highly active wastes continue to be vitrified at Sellafield in the Waste Vitrification Plant (WVP) and the product glass has previously been subjected to rigorous characterisation in order to ensure waste form integrity and stability during interim storage. However, the vitrified waste may eventually be placed in an underground repository and so further studies may be needed to confirm the chemical stability of vitrified wastes in different disposal scenarios. One area of uncertainty is in knowing how the alteration products and surface layers on a corroded glass effect the dissolution process and the solution concentrations of long-lived radionuclides. This study shows how the technique of laser microprobe inductively coupled plasma mass spectrometry (LM-ICP-MS) can be used to provide a direct chemical analysis of vitrified samples, both in their as-cast state and after leach testing. LM-ICP-MS, solutions' nebulisation ICP-MS (to provide chemical analyses of the leachants), light microscopy, scanning electron microscopy and confocal laser microscopy were used to examine the effect of leaching on the surfaces of vitrified samples. Data were obtained for a number of simulated WVP-type product glasses that were subjected to soxhlet leach testing for varying lengths of time. The LM-ICP-MS data clearly show that the technique offers a rapid and relatively accurate method for detecting and monitoring minute changes in the surface layers of glass and in the build-up of alteration products. Hence, LM-ICP-MS could be used to obtain valuable mechanistic information on the degradation of glass by examining radionuclide behaviour in the surface layers of glass as it corrodes over extended time periods.

Glass has become a preferred waste form worldwide for radioactive wastes; however, there are limitations. Halogen-containing wastes can not be converted to glass because halogens (chlorides, fluorides, etc.) form poor-quality waste glasses. Furthermore, halides in glass melters often form second phases that create operating problems. A new waste vitrification process, the Glass Material Oxidation and Dissolution System (GMODS), removes these limitations by converting halogen-containing wastes into borosilicate glass and a secondary, clean, sodium-halide stream.

Vitrification studies of actual Savannah River M-Area mixed wastes have shown that the limiting factor for high waste loading of this waste stream is its chemical durability as defined by the toxicity characteristics leaching procedure (TCLP). As part of the optimization study of Savannah River M-Area wastes, a number of additives were examined including Na2O, Li2O, B2O3, ZrO2, and TiO2. This paper reports on the effect of varying the boron to total alkali ratio and on the effect of substitutions such as ZrO2 for waste and TiO2 for SiO2 on the chemical durability and processability of M-Area waste glasses.

We have characterized the corrosion behavior of several Defense Waste Processing Facility (DWPF) reference waste glasses by conducting static dissolution tests with crushed glasses. Glass dissolution rates were calculated from measured B concentrations in tests conducted for up to five years. The dissolution rates of all glasses increased significantly after certain alteration phases precipitated. Calculation of the dissolution rates was complicated by the decrease in the available surface area as the glass dissolves. We took the loss of surface area into account by modeling the particles to be spheres, then extracting from the short-term test results the dissolution rate corresponding to a linear decrease in the radius of spherical particles. The measured extent of dissolution in tests conducted for longer times was less than predicted with this linear dissolution model. This indicates that advanced stages of corrosion are affected by another process besides dissolution, which we believe to be associated with a decrease in the precipitation rate of the alteration phases. These results show that the dissolution rate measured soon after the formation of certain alteration phases provides an upper limit for the long-term dissolution rate, and can be used to determine a bounding value for the source term for radionuclide release from waste glasses. The long-term dissolution rates measured in tests at 20,000 m−1 at 90°C in tuff groundwater at pH values near 12 are about 0.2,0.07, and 0.04 g/(m2•d) for the Environmental Assessment glass and glasses made with SRL 131 and SRL 202 frits, respectively.

We have investigated the alteration behavior of synthetic basalt and SRL 165 borosilicate waste glasses that had been reacted in water vapor at 70°C for time periods up to seven years. The nature and extent of corrosion of glasses have been determined by characterizing the reacted glass surface with optical microscopy, scanning electron microscopy (SEM), transmission electron microscopy (SEM), and energy dispersive x-ray spectroscopy (EDS). Alteration in 70°C laboratory tests was compared to that which occurs at 150–200°C and also with Hawaiian basaltic glasses of 480 to 750 year old subaerially altered in nature. Synthetic basalt and waste glasses, both containing about 50 wt % SiO2, were found to react with water vapor to form an amorphous hydrated gel that contained small amounts of clay, nearly identical to palagonite layers formed on naturally altered basaltic glass. This result implies that the corrosion reaction in nature can be simulated with a vapor hydration test. These tests also provide a means for measuring the corrosion kinetics, which are difficult to determine by studying natural samples because alteration layers have often spalled off the samples and we have only limited knowledge of the conditions under which alteration occurred.