The setting of the technical subsystem within the overallsocio-political-economic-technical radwaste system will be described and thehighly interactive nature of the larger setting emphasized. It will be shownthat because of the dominance of the socio-political subsystem, theimportance of the technical subsystem is overshadowed. Moreover the keyissue in the technical subsystem, whether to put more reliance on theimmobilization (via the waste package or engineeered barriers) or theisolation (via the geology) has stayed tilted toward the latter since 1978,when we organized the first symposium on radwaste science (at the MaterialsResearch Society meeting). Now that the isolation strategy is stymied, theopportunity arises again for the materials community to make a compellingcase for the waste package's true significance.

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

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

A single-step plasma arc-vitreous ceramic (PAVC) process is described forconverting spent nuclear fuel (SNF), SNF sludges, and associated wastes intoa vitreous ceramic waste form. This proposed technology is built onextensive experience of nuclear waste form development and nuclear wastetreatment using the commercially available plasma arc centrifugal (PAC)system. SNF elements will be loaded directly into a PAC furnace with minimumadditives and converted into vitreous ceramics with up to 90 wt% wasteloading. The vitreous ceramic waste form should meet the functionalrequirements for borosilicate glasses for permanent disposal in a geologicrepository and for interim storage. Criticality safety would be ensuredthrough the use of “batch” modes, and controlling the amount of fuelprocessed in one batch. The minimum requirements on SNF characterization andpretreatment, the one-step process, and minimum secondary waste generationmay reduce treatment duration, radiation exposure, and treatment cost.

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

We have pursued some of the fundamental chemistry and materials science ofamericium in three glass matrices, two being high-temperature (850° and1400°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 byamericium in these matrices, as well as factors that control and/or mayalter this state. We have noted a correlation between the oxidation state ofthe f-elements in the two high-temperature glasses with theirhigh-temperature oxide chemistries. One exception was americium: althoughamericium dioxide is the stable oxide encountered in air, when this dioxidewas incorporated into the high-temperature glasses, only trivalent americiumwas found in the products. When trivalent americium was used to prepare thesol-gel glasses at ambient temperature, and after these products were heatedin air to 800°C, again only trivalent americium was observed. Potentialexplanations for the unexpected behavior of americium is offered in thecontext of its basic chemistry. Experimental spectra, spectroscopieassignments and other pertinent data obtained in the studies arediscussed.

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 asol-gel technology. The oxidation state of these actinides in the sol-gelproduct was examined by absorption spectroscopy after solidification, aging,and thermal treatment. The oxidation state of the actinides in the startingsolutions was essentially maintained through the solidification process ofthe silica matrix. However, during densification and removal of residualsolvents at elevated temperatures, both actinides converted eventually totheir tetra valent states while in the different sol-gel products. Thisfinding is in accord with reports that tetravalent states of plutonium andneptunium are acquired in glass products prepared by dissolution of theactinide in molten glasses. Comparisons between room temperature spectraobtained from neptunium and plutonium in heated sol-gel products and frommolten glass products showed subtle differences that can be related to themetal ion's environments.

CCIM technology allows syntheses of various materials over a wide range ofcompositions. This provides a means of applying the CCIM to thesolidification of wastes obtained by different technologies of the spentnuclear fuel reprocessing, including toxic wastes containing heavy metalsand those arising from fractionation. Solid compounds produced by thismethod meet both the requirements for their subsequent disposal in deepgeological formations and spent fuel standards.

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

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

The liquidus temperature (TL) often limits the loading ofhigh-level waste in glass through the constraint that TL must beat 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 phasewere measured as a function of glass composition. The test glasses werebased on high-iron Hanford Site tank wastes. All studied glassesprecipitated spinel (Ni,Fe,Mn)(Cr,Fe)2O4 as theprimary crystalline phase. TL was increased by additions of Cr2O3, NiO, Al2O3, Fe2O3, MgO, and MnO; while Li2O, Na2O, B2O3, and SiO2 had anegative effect. Empirical mixture models were fitted to data.

A study was conducted on glasses based on a simulated transuranic waste withhigh concentrations of ZrO2and Bi2O3 todetermine the compositional dependence of primary crystalline phases andliquidus temperature (TL). Starting from a baseline composition, glasses were formulated bychanging mass fractions of Al2O3, B2O3, Bi2O3, CeO2, Li2O, Na2O, P2O5, SiO2, and ZrO2, one at a time, while keeping theremaining components in the same relative proportions as in the baselineglass. Liquidus temperature was measured by heat treating glass samples for24 h in a uniform temperature furnace. The primary crystalline phase in thebaseline 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 phaseto baddeleyite (ZrO2), while cerium oxide (CeO2)precipitated from glasses with more than 3 wt% CeO2. ZirconTL 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 T L as a function of composition for zircon primary crystalline phaseglass.

A model is presented based upon calculated bridging oxygens which allows theprediction of the region of acceptable glass compositions for alime-soda-silica glass-forming system containing mixed waste. The model canbe used to guide glass formulation studies (e.g., treatability studies) orassess the applicability of vitrification to candidate waste streams.

The Savannah River Site has the majority of the United States' supply ofneptunium currently stored in an acid solution in one of their canyonfacilities. A program is being developed that could be utilized to ship thismaterial, as glass, to Oak Ridge National Laboratory where the Np could beleached from the glass, purified by ion exchange and made into targetmaterial for the production of Pu-238. Ion exchange purification dictates nomaterial be in the leachate making the isolation of the Np difficult. Wehave developed a process using thorium as a surrogate for Np that couldimmobilize the Np into a soda borosilicate glass for shipment. To achieverecovery of the Np, the glass can be phase separated prior to leaching withnitric acid. Phase separation would produces a Np-rich sodium-borate phaseand a Si-rich phase similar to a Vycor® glass. The nitric acid selectivelyattacks the sodium-borate phase allowing high Np recovery in a solution thatcontains only sodium and boron. These can be easily separated from Np by ionexchange. Essentially all of the silicon which would interfere with ionexchange by precipitation is retained in the Vycor®-type phase. Thistechnology may also be applied to other actinides stored in relatively puresolutions.

This paper will report the optimization of variables for maximizing Th (a Npsurrogate) 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 heattreatment.

Immobilization by vitrification is one potential disposition option for aportion of the United States' excess plutonium inventory. Research has beenperformed at the Savannah River Site (SRS) to determine the optimumcomposition of a lanthanide borosilicate frit for the vitrification ofplutonium using a Plackett-Burman design and simplex algorithm as astatistical tool. This technique uses various response variables to rank andoptimize a composition. The variables used in this study correspond tohomogeneity, durability, actinide solubility and devitrification afterheat-treatment.

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

Batches containing various loadings of ThO2 were melted todetermine if actinide solubility was improved in the optimized compositioncompared to that of a similar lanthanide borosilicate glass. No noticeableimprovement in ThO2 solubility was realized as a result of usingthis optimization technique.

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

A recovery flowsheet was demonstrated using a curium-containing glass toextract the Am/Cm from the glass matrix. The procedure involved grinding theglass to less than 200 mesh and dissolving in concentrated nitric acid at110°C. Under these conditions, the dissolution was essentially 100% after 2hours except for the insoluble silicon. Using a nonradioactive surrogate,the expected glass dissolution rate during Am/Cm recovery was bracketed byusing both static and agitated conditions. The measured rates, 0.0082 and0.040 g/hrcm2, were used to develop a predictive model for thetime required to dissolve a spherical glass particle in terms of the glassdensity, particle size, and measured rate. The calculated dissolution timewas in agreement with the experimental observation that the curium glassdissolution was complete in less than 2 hr.

Induction melting by using a cold crucible is a suitable technology for theimmobilisation of ash residues after incineration of solid radioactivewaste. We investigated the possibility of using glass composites produced bystirring the ash into meltedglass. Glass composites containing 15 -40 wt. %of ash were obtained in both laboratory and bench scale units. Infraredspectroscopy, electronic paramagnetic resonance, X-ray diffractometry, TEMand SEM analyses were applied in order to characterise the structure of theglass composites obtained. The glass composites consisted of a relativelyhomogeneous glass matrix with embedded polycrystalline aggregates. Thefraction of aggregates increases when the fraction of ash rises. Theisothermal curing of composites at 1100 °C leads to dissolution of the ashcomponents into the melt as well as to their inclusion into the glassstructure, according to the analysis of the spectra obtained.

Vitrification is the most developed method for the immobilisation of highlyradioactive waste obtained from reprocessing irradiated nuclear fuel. Highlyactive wastes continue to be vitrified at Sellafield in the WasteVitrification Plant (WVP) and the product glass has previously beensubjected to rigorous characterisation in order to ensure waste formintegrity and stability during interim storage. However, the vitrified wastemay eventually be placed in an underground repository and so further studiesmay be needed to confirm the chemical stability of vitrified wastes indifferent disposal scenarios. One area of uncertainty is in knowing how thealteration products and surface layers on a corroded glass effect thedissolution process and the solution concentrations of long-livedradionuclides. This study shows how the technique of laser microprobeinductively coupled plasma mass spectrometry (LM-ICP-MS) can be used toprovide a direct chemical analysis of vitrified samples, both in theiras-cast state and after leach testing. LM-ICP-MS, solutions' nebulisationICP-MS (to provide chemical analyses of the leachants), light microscopy,scanning electron microscopy and confocal laser microscopy were used toexamine the effect of leaching on the surfaces of vitrified samples. Datawere obtained for a number of simulated WVP-type product glasses that weresubjected to soxhlet leach testing for varying lengths of time. TheLM-ICP-MS data clearly show that the technique offers a rapid and relativelyaccurate method for detecting and monitoring minute changes in the surfacelayers of glass and in the build-up of alteration products. Hence, LM-ICP-MScould be used to obtain valuable mechanistic information on the degradationof glass by examining radionuclide behaviour in the surface layers of glassas 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 beconverted to glass because halogens (chlorides, fluorides, etc.) formpoor-quality waste glasses. Furthermore, halides in glass melters often formsecond phases that create operating problems. A new waste vitrificationprocess, the Glass Material Oxidation and Dissolution System (GMODS),removes these limitations by converting halogen-containing wastes intoborosilicate glass and a secondary, clean, sodium-halide stream.

Vitrification studies of actual Savannah River M-Area mixed wastes haveshown that the limiting factor for high waste loading of this waste streamis its chemical durability as defined by the toxicity characteristicsleaching procedure (TCLP). As part of the optimization study of SavannahRiver M-Area wastes, a number of additives were examined including Na2O, Li2O, B2O3, ZrO2, and TiO2. This paper reports on the effect ofvarying the boron to total alkali ratio and on the effect of substitutionssuch as ZrO2 for waste and TiO2 for SiO2 onthe chemical durability and processability of M-Area waste glasses.

We have characterized the corrosion behavior of several Defense WasteProcessing Facility (DWPF) reference waste glasses by conducting staticdissolution tests with crushed glasses. Glass dissolution rates werecalculated from measured B concentrations in tests conducted for up to fiveyears. The dissolution rates of all glasses increased significantly aftercertain alteration phases precipitated. Calculation of the dissolution rateswas complicated by the decrease in the available surface area as the glassdissolves. We took the loss of surface area into account by modeling theparticles to be spheres, then extracting from the short-term test resultsthe dissolution rate corresponding to a linear decrease in the radius ofspherical particles. The measured extent of dissolution in tests conductedfor longer times was less than predicted with this linear dissolution model.This indicates that advanced stages of corrosion are affected by anotherprocess besides dissolution, which we believe to be associated with adecrease in the precipitation rate of the alteration phases. These resultsshow that the dissolution rate measured soon after the formation of certainalteration phases provides an upper limit for the long-term dissolutionrate, and can be used to determine a bounding value for the source term forradionuclide release from waste glasses. The long-term dissolution ratesmeasured in tests at 20,000 m−1 at 90°C in tuff groundwater at pHvalues near 12 are about 0.2,0.07, and 0.04 g/(m2•d) for theEnvironmental Assessment glass and glasses made with SRL 131 and SRL 202frits, respectively.

We have investigated the alteration behavior of synthetic basalt and SRL 165borosilicate waste glasses that had been reacted in water vapor at 70°C fortime periods up to seven years. The nature and extent of corrosion ofglasses have been determined by characterizing the reacted glass surfacewith optical microscopy, scanning electron microscopy (SEM), transmissionelectron microscopy (SEM), and energy dispersive x-ray spectroscopy (EDS).Alteration in 70°C laboratory tests was compared to that which occurs at150–200°C and also with Hawaiian basaltic glasses of 480 to 750 year oldsubaerially altered in nature. Synthetic basalt and waste glasses, bothcontaining about 50 wt % SiO2, were found to react with watervapor to form an amorphous hydrated gel that contained small amounts ofclay, nearly identical to palagonite layers formed on naturally alteredbasaltic glass. This result implies that the corrosion reaction in naturecan be simulated with a vapor hydration test. These tests also provide ameans for measuring the corrosion kinetics, which are difficult to determineby studying natural samples because alteration layers have often spalled offthe samples and we have only limited knowledge of the conditions under whichalteration occurred.