The incorporation of Pu and Np in zirconolite (CaZrTi2O7) has been investigated over a range ofredox conditions. Zirconolite formulations designed to favour eithertrivalent or tetravalent Pu and Np were prepared by limiting the amount ofcharge compensating additives available to maintain electroneutrality. Fromnear-edge X-ray absorption spectroscopy the Pu valence state was found tovary with the processing atmosphere, from completely tetravalent when firedin air, and located on either the Ca or Zr sites, to trivalent, whensubstituted on the Ca site after annealing in 3.5% H2/N2. Np was predominantly tetravalent over therange of redox conditions examined and was readily incorporated on either ofzirconolite's Ca or Zr sites. The charge compensation mechanisms at work indifferent zirconolites are also discussed.

Zirconolite (CaZrTi2O7) is the primaryactinide-bearing Synroc phase for the immobilisation of high-level nuclearwaste. Using X-ray absorption spectroscopy and microanalysis we haveinvestigated the incorporation of cerium, as a simulant for plutonium, onboth zirconolite's Ca and Zr sites under a range of redox conditions. The Cevalence state was found to vary between Ce3+ and Ce4+depending on the both the sintering atmosphere and temperature. Theexistence of alternative charge compensation schemes, predominantly cationvacancies, in addition to those used in the sample design was inferred inmany of the zirconolites and will be discussed in detail.

Three wt% each of Cs and Tc were mixed with the standard Synroc precursorand the ceramic was formed by hot-pressing. Attempts were made toincorporate the Tc as either metal or Tc 4+, using differentredox conditions in processing. Volatile losses of Tc during calcinationwere < 0.1% in all cases. Short-term Tc leach rates when the Tc waspresent as a metal alloy were in the order of 10−4 g/m /d at 90°Cwith frequently changed water, and decreased with increasing leaching time.The valence of the Tc was monitored by X-ray absorption spectroscopy at thedrying and calcination stages of the production. The general viability ofSynroc/glass composites for immobilising the Hanford HLW sludges is furtherdemonstrated by using further refinements of additive schemes for theinactive “All-blend” formulation and initial studies using the U-containing“All-blend” waste formulation.

Synroc containing 20 wt% simulated high level waste (HLW) was subjected totwo sets of leach tests at 150°C where the leachant was and was not replacedduring the test (replacement and non-replacement testing). The leachant wasa KH-phthalate buffered solution (pH 4.2). Samples were characterised beforeand after leach testing using SEM, AEM and SIMS. Elemental concentrations inleachates were measured using ICP-MS. In concert with the findings of i) adissolution study of perovskite in a flowing leachant and ii) a previousSynroc dissolution study (wherein Synroc containing 10 wt% simulated HLW wassubjected to periodic replacement, leach testing in deionised water at150°C), the results of this study show that when the leachant replacementfrequency is varied from 7 d to the duration of the test, there is no effecton leach rate or leaching mechanisms.

Synroc ceramics were synthesized in an induction-heated cold crucible atlaboratory scale (1 kg) from an oxide mixture, and at industrial prototypescale (45 kg) from Synroc previously produced by sintering under load athigh temperature. After melting, both materials contained the major phasesof Synroc-C. The chemical durability of both melted materials, as determinedby static leaching of powder samples in initially pure water at 90°C with anSA/V ratio of 20000m−1, was equivalent to that of conventionalhot-pressed Synroc-C. Cerium, used in this investigation to simulate thepresence of tri-and tetravalent actinides, was found in steady-stateconcentrations on the order of 1 ppb (i.e. NL(Ce) ≤ 10−6 g·m−2). The concentration in the leachates was independent ofthe initial CeO2 content of the Synroc (at least up to 10 wt%);moreover, it is similar to the results obtained with hot-pressed Synroc-Cspecifically formulated for conditioning long-lived actinides.

Preparation and characterization of inductively-melted Synroc containing 20wt% simulated plant “Mayak” reprocessing waste were performed. The samplebulk composition was as follows, (in wt.%): 55.4 TiO2; 15.8 ZrO2; 7.5 CaO; 7.4 BaO; 4.3 Al2O3 2.0MnO; 1.8 SiO2; 0.7 Na2O; 1.9 K2O, 0.5 Ce2O3; 1.0 UO2; 0.9 NiO; 0.6 Cr2O3, and 0.2 FeO. The sample was produced bymelting in air at 1550–1600 °C under barometric pressure. It is composed ofa few crystalline phases and a minor glass phase. Most of the phases(hollandite, zirconolite, perovskite and rutile) are similar to theanalogous phases found in the other Synroc formulations. An additional phasewith average composition, wt.%: 59.8 TiO2; 15.6 CaO; 7.0 UO2; 5.6 ZrO2; 4.7 MnO; 4.1 Ce2O3, and 1.8 Al2O3 wasfound. Some elements (Ba, Si, Ni, K, Na, Fe) were present in the phase innegligible quantities. Its formula (Ca2.65U0.3Ce0.2)(Ti7.3Mn0.6Zr0.4Al0.3)O20.0is rather close to a rare mineral uhligite - Ca3(Ti,Zr,Al)9O20. Another possiblecounterpart of the phase is murataite-like mineral previously described intailored ceramic designed for Savannah River Plant wastes fixation. Thisphase as well as zirconolite are the major host for U in the samplePreliminary data on the material leachability in water at 350 °C and 50 MPahave been obtained Uranium contents in the solution were about 1 ppb andclose to the uranium dioxide solubility in deionized water under the sameP-T conditions.

Three Synroc-C samples, containing simulated high level waste were studied.One was produced by the conventional hot-pressing method at ANSTO,Australia, and the others were obtained using cold crucible technology atRadon, Russia. One of the melted samples was prepared using the Australiansol-gel precursor and the second one was obtained from an oxide-nitratemixture. It was established that the specimens have closely similar mineralcompositions, with major hollandite, perovskite, zirconolite, and rutile.Small amounts of hibonite were also found. Unlike the hot-pressed Synroccontaining metallic alloy particles, melted Synrocs contain molybdates. Aninvestigation of mineral compositions and elemental distribution in thesamples was carried out. Features of hot-pressed and melted ceramics werecompared. Unit cell parameters of the Synroc phases were determined andpreliminary results on durability of the melted Synroc are presented.

Syntheses of novel titanate-based waste forms, potentially suitable forimmobilizing Pu(III) and Am(III), have been performed. The dititanatecompounds An2Ti2O7 (An=Pu or Am) were foundto have a monoclinic structure. A series of pyrochlore-type solid solutionsbetween An2Ti2O7 and the cubic pyrochlore Ln2Ti2O7 (Ln=Gd, Er, or Lu) has beenestablished, and limits of solid solubility for An2Ti2O7 were determined. These limitswere found to increase as the ionic radius of the lanthanide in the hostdecreased. Strontium-containing titanate compounds SrAn2Ti4O12 were also prepared and werecharacterized as exhibiting a perovskite-type structure. Such compoundscould serve as waste hosts for the immobilization of not only the actinidebut also 90 Sr. A 2 month leaching study of Er1.78Pu0.22Ti2O7 and SrPu2Ti4O12 by WIPP “A” brine wasperformed. Very low amounts (< 1 ppm) of Pu(III) were leached. Althoughthe ionic radius of Np(III) is similar to that of Pu(III) and Am(III),analogous Np compounds could not be prepared in any of the titanate systemsinvestigated.

The compositional flexibility of the sodium zirconium phosphate (NaZr2(PO4)3) structure has beenexploited in the design of monophasic radiophases capable of immobilizingthe most common cations associated with reprocessed high-level commercialwaste streams. Highly crystalline, monophasic members of the NaZr2(PO4)3 structural family ([NZP])have been prepared with conventional processing methods and equipment. Theseradiophases were tailored to accommodate 10–20 wt % modified PW-4b simulatedcalcine as single phases isostructural with NaZr2(PO4)3. To meet the challenge ofdesigning monophasic materials capable of accommodating the chemicalcomplexity of PW-4b, an ionic substitution scheme based on crystal chemicalprinciples was developed. The radiophases were prepared with inexpensive,inorganic precursors and a solution sol-gel method; these materials wereheat treated and/or sintered under a variety of conditions to determine theoptimum conditions for single phase [NZP] formation. X-ray powderdiffraction provided valuable information that was used to assess thesuitability of the ionic substitution model developed in this investigation.The results of this investigation suggest that monophasic [NZP] radiophasescapable of accommodating 10–20 wt % modified PW-4b simulated calcine may becontinuously processed with conventional ceramic processing methods andequipment. Moreover, the relatively low temperatures involved and thereproducibility of the process make [NZP] radiophases economicallyattractive hosts for radioactive and heavy metal industrial wastes.

Neutron diffraction studies of salt-occluded zeolite and zeolite/glasscomposite samples, simulating nuclear waste forms loaded with fissionproducts, have revealed complex structures, with cations assuming the dualroles of charge compensation and occlusion (cluster formation). Theseclusters roughly fill the 6–8 Å diameter pores of the zeolites. Samples areprepared by equilibrating zeolite-A with complex molten Li, K, Cs, Sr, Ba, Ychloride salts, with compositions representative of anticipated wastesystems. Samples prepared using zeolite 4A (which contains exclusivelysodium cations) as starting material are observed to transform to sodalite,a denser alumi-nosilicate framework structure, while those prepared usingzeolite 5A (sodium and calcium ions) more readily retain the zeolite-Astructure. Because the sodalite framework pores are much smaller than thoseof zeolite-A, clusters are smaller and more rigorously confined, with acorrespondingly lower capacity for waste containment. Details of thesodalite structures resulting from transformation of zeolite-A depend uponthe precise composition of the original mixture. The enhanced resistance ofsalt-occluded zeolites prepared from zeolite 5A to sodalite transformationis thought to be related to differences in the complex chloride clusterspresent in these zeolite mixtures. Data relating processing conditions toresulting zeolite composition and structure can be used in the selection ofprocessing parameters which lead to optimal waste forms.

Zirconolite-rich ceramics were produced by the cold crucible meltingtechnique in an air atmosphere, at 1550±50 °C and 1 atm. Four samples withoverall composition (in wt.%): 4.9–14.3 CaO; 19.0–41.3 ZrO2;24.1–42.6 TiO2; 1.3–11.3 Al2O3; 6.8–30.0 Gd2O3, and 1.1–8.5 SiO2 have beenstudied. Total phases in the ceramics consist of major zirconolite and minorrutile, perovskite, zirconia, aluminium titanate, and glass. The Gd2O3 content in zirconolite reaches up to 31.4wt.% corresponding to the formula: (Cao4,Gd0.7)Zri.0(Tii4,Alo 5)070. Thedata on the phase composition agree well with coupled Gd incorporation intothe mineral structure. Ca(H) + Ti(IV) = Gd(III) + Al(III), and 2Gd(III) -Ca(II) + Zr(IV). The highest Gd contents observed in the other phases are25.4 % for zirconia, 12.6 % in glass, 8.8 % in perovskite, and 1.4 % forrutile. The rest of the elements' distribution in the samples areanalyzed.

The incorporation of (a) Cs and Sr as; (b) simulated actinides, and (c)simulated Purex waste in sodium zirconium phosphate (NZP) has been studied.The samples were prepared by sintering, by hot pressing and by hot isostaticpressing in metal bellows containers. The short-term chemical durability ofthe phosphate-based material containing Purex waste was within an order ofmagnitude ofthat for Synroc-C, as measured by 7-day MCC-1 tests at 90°C. Thedissolution behaviour showed evidence of re-precipitation phenomena, evenafter times as short as 28 days. Potential for improvement of NZP-basedceramics for HLW management is discussed.

An interaction between Synroc-C samples prepared by inductive melting in acold crucible and deionized water was investigated at room temperature, 100and 200 °C. The leachability of powdered specimens with grainsize 0.10–0.15mm was determined by repeated static tests with regular replacement ofleachant. The experimental data showed that the most leachable elements arethe alkali, alkaline earths and molybdenum. The matrix elements such astitanium, zirconium, and rare earths as well are very leach resistant.Comparison of the calculated leach rates of components with reference dataon leaching of alternative materials showed that cesium leachability fromthe studied specimens are close to borosilicate glass, but leaching behaviorof other components is comparable to leachability from Synroc-C prepared byhot-pressing.

A technique for solidification of high-level liquid wastes (HLLW) to obtainmineral-like forms through the use of water-cooled melters with directinduction heating of a melt (CCIM) has been developed at the StateScientific Center of the Russian Federation VNIINM (SSC RF VNIINM).

Mineral-like materials of different classes such as pyroxenes,pyrosilicates, garnets of the andradite group, titanosilicates and Synroc Dwere synthesized by the CCIM method at temperatures of 1250 – 1550°C.

The preliminary X-ray diffraction studies of these materials indicate thatthe structures synthesized by the CCIM method have compositions that areidentical to those of the corresponding natural analogs. The chemicaldurabilities of the synthesized materials upon their exposure to distilledwater at 20°C have been determined.

In addition to the investigation of solidification of liquid HLW, the CCIMmethod is examined and applied to the synthesis of mineral-like compositionsinvolving simulated waste components (nonsoluble residues, ashes, and pulps)produced by the chemical and metallurgical manufacturing of fissilematerials.

A ceramic waste form is being developed at Argonne National Laboratory forwaste generated during the electrometallurgical treatment of spent nuclearfuel. The waste is generated when fission products are removed from theelectrolyte, LiCI-KCl eutectic. The ceramic waste form is a composite,fabricated by hot isostatic pressing a mixture of glass frit and zeoliteoccluded with fission products and salt. Past work has shown that thenormalized release rate (NRR) is less than 1 g/m2d for allelements in a Material Characterization Center-Type 1 (MCC-1) leach test runfor 28 days in deionized water at 90°C (363 K). This leach resistance iscomparable to that of early Savannah River glasses. We are investigating howleach resistance is affected by changes in the cationic form of zeolite andin the glass composition. Composites were made with three forms of zeolite Aand six glasses. We used three-day ASTM C1220–92 (formerly MCC-1) leachtests to screen samples for development purposes only. The leach testresults show that the glass composites of zeolites 5A and 4A retain fissionproducts equally well. The loss of cesium is small, varying from 0.1 to 0.5wt%, while the loss of divalent and trivalent fission products is one ormore orders of magnitude smaller. Composites of 5A retain chloride ionbetter in these short-term screens than 4A and 3A. The more leach resistantcomposites were made with durable glasses that were rich in silica and poorin alkaline earth oxides. The x-ray diffraction (XRD) results show that asalt phase was absent in the leach resistant composites of 5A and the betterglasses but was present in the other composites with poorer leachperformance. Thus, the data show that the absence of a salt phase in acomposite's XRD pattern corresponds to improved leach resistance. The dataalso suggest that the interactions between the zeolite and glass depend onthe composition of both.

Glass-bonded zeolite is being developed as a potential ceramic waste formfor the disposition of radionuclides associated with the U.S. Department ofEnergy's (DOE's) spent nuclear fuel conditioning activities. The utility ofseveral standard durability tests [e.g., Materials Characterization CenterTest #1 (MCC-1), Product Consistency Test-B (PCT-B), and Vapor HydrationTest (VHT)] was evaluated as a first step in developing methods and criteriathat can be applied towards the process of qualifying this material foracceptance into the DOE Civilian Radioactive Waste Management System. Theeffects of pH, leachant composition, and samplesurface-area-to-leachant-volume ratios on the durability test results arediscussed, in an attempt to investigate the release mechanisms and otherphysical and chemical parameters that are important for the acceptancecriteria, including the establishment of appropriate test methodologiesrequired for product consistency measurements.

Results from PCT-Bs conducted with 4 μm diameter salt-loaded zeolite powderindicate that a good correlation exists between release rate and ionic sizeand/or charge for the release behavior of the simulated fission products indeionized water (DRV), EJ-13 groundwater, and brine solutions. Simulateddivalent and trivalent fission products [Sr, Ba, and rare earth (RE) ions]were preferentially retained in the zeolite (relative to the singly ionizedcations) after tests with the salt-loaded zeolite in DIW. In general, thepreferential cation release order for salt-loaded zeolite A in DrW is Li> Na ≥ K > Cs > Al > Si > RE > Sr > Ba. Results fromPCT-Bs with the salt-loaded zeolite A immersed in high-ionic-strength brinesat 90°C indicate a significant increase, relative to DIW tests, in therelease rates of the Sr, Ba, and RE ions despite a decrease in the releaseof the Si and Al ions that make up the framework matrix of the zeolite. Anincrease in the Mg and Ca concentrations in the reacted zeolites suggeststhat an ion exchange process may be responsible for this increase.

Vapor hydration and MCC-1 tests were performed with ceramic waste formmonoliths of glass-bonded zeolite. The VHTs (temperatures at 120,150, and200°C) provided useful information about the effect of glass composition oncorrosion rates and alteration phase formation, and about the overalltoughness and structural integrity of the ceramic waste form. The MCC-1 testwas investigated as an alternative to the PCT for acceptance criteriameasurements. The MCC-1 results indicate that corrosion testing with bothDIW and high-ionic-strength leachants (that specifically affect the ionexchange behavior of the fission products) are required to fully assess thedurability of the ceramic waste form. These preliminary results establishthe utility of the MCC-1 test for providing possible acceptance criteriameasurements, including elemental release comparisons between theenvironmental assessment benchmark and the ceramic waste form.

Two new zeolitic crystalline phases with stoichiometry, CSTiSi2O6.5 and CS2TiSi6O15, have been discovered. CSTiSi2O6.5 has a crystal structure isomorphous tothe mineral pollucite, CsAlSi2O6, with Ti+4replacing Al+3. This replacement requires a mechanism for chargecompensation. A combination of techniques including neutron diffraction,single crystal x-ray diffraction and x-ray absorption spectroscopy haverevealed that eight extra oxygens are present per unit cell Cs2TiSi2O6.5 as compared to pollucite.As a result of the extra oxygen, the titanium coordination geometry isfive-fold. Pentacoordinate titanium and tetrahedral silicon form a networkstructure with Cs residing in cages formed by the network. The crystalstructure of Cs2TiSi6O15 is unique, withtitanium octahedra and silicon tetrahedra forming an open frameworkstructure with the Cs residing in large cavities. The largest covalentlybonded ring opening to the Cs cavities in both compounds are smaller than aCs ion, revealing that the Cs ion has minimal mobility in the structure.Cesium leach rates for both compounds are lower than or comparable toborosilicate glass.

Silicotitanate ion exchangers are potential materials for the removal ofradioactive Cs and Sr from tank wastes. In this paper the viability ofdirect thermal conversion of Cs-loaded silicotitanates to an acceptable highlevel waste form has been examined. Results show that in aqueous solutions,the Cs leach rates of crystalline silicotitanates (heat treated at 800°C)are 0.04, 0.18, 0.4 g/m2day for Cs loadings of 1, 5, and 20 wt%,respectively. Heating the Cs-loaded (up to 20 wt %) silicotitanates at orabove 900 °C for 1 hour further reduces the Cs leach rates to approximatelyzero (beyond the lppm detection limits). Moreover, Cs volatilization wasfound to be < 0.8 wt% at temperatures as high as 1000 °C. These resultssuggest that thermally converted silicotitanate ion exchangers exhibitexcellent chemical durability (comparable to or better than borosilicateglass) and thus, have great potential as an alternative waste form.

Aluminosilicates are one of the major class of species controlling thevolume of radioactive high-level waste that will be produced from futureremediation at Hanford site. Here we present studies of the phases andstructures of aluminosilicates as a function of sludge composition usingX-ray powder diffraction, solid state 27Al and 29SiNMR, and Raman spectroscopy. The results show that the content of NaNO3 in solution has significant effects on the nature of theinsoluble aluminosilicate phases produced. It was found that regardless ofthe initial Si:Al ratio, nitrate cancrinite was the main phase formed in thesolution with pH of 13.5 and 5 MNaNO3. However, at lower NaNO3 concentration with initial Si:Al ratios of 1.1, 2.2, and11.0 in the solutions, a range of aluminosilicate zeolites was produced withSi: Al ratios of 1.1, 1.3, and 1.5, respectively. Lowering the solution pHappears to promote the formation of amorphous aluminosilicates. The resultspresented here are important for the prediction of the solubility anddissolution rate of Al in tank wastes.

The effect of the addition of an organic monolayer to the surface of a claymineral on the speciation of metal ions intercalated into the clayinterlayer is probed by X-ray absorption spectroscopy. The presence of themonolayer changes the surface of the clay from hydrophilic to hydrophobic.It inhibits the interlayer ions from exchanging freely into environmentalwater and reduces the leach rate of cations out of the clay by approximatelya factor of 20. Significant changes are observed when these coated samplesare treated under hydrothermal and thermal conditions. Reductions ofuranium(VI), in the form of uranyl, and copper(II) ions occur. In addition,the uranium aggregates, forming small particles that appear similar to UO2. Comparable conglomeration occurs with lead cations andwith the reduced copper species.