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

Zirconolite (CaZrTi2O7) is the primary actinide-bearing Synroc phase for the immobilisation of high-level nuclear waste. Using X-ray absorption spectroscopy and microanalysis we have investigated the incorporation of cerium, as a simulant for plutonium, on both zirconolite's Ca and Zr sites under a range of redox conditions. The Ce valence state was found to vary between Ce3+ and Ce4+ depending on the both the sintering atmosphere and temperature. The existence of alternative charge compensation schemes, predominantly cation vacancies, in addition to those used in the sample design was inferred in many of the zirconolites and will be discussed in detail.

Three wt% each of Cs and Tc were mixed with the standard Synroc precursor and the ceramic was formed by hot-pressing. Attempts were made to incorporate the Tc as either metal or Tc 4+, using different redox conditions in processing. Volatile losses of Tc during calcination were < 0.1% in all cases. Short-term Tc leach rates when the Tc was present as a metal alloy were in the order of 10−4 g/m /d at 90°C with frequently changed water, and decreased with increasing leaching time. The valence of the Tc was monitored by X-ray absorption spectroscopy at the drying and calcination stages of the production. The general viability of Synroc/glass composites for immobilising the Hanford HLW sludges is further demonstrated by using further refinements of additive schemes for the inactive “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 to two sets of leach tests at 150°C where the leachant was and was not replaced during the test (replacement and non-replacement testing). The leachant was a KH-phthalate buffered solution (pH 4.2). Samples were characterised before and after leach testing using SEM, AEM and SIMS. Elemental concentrations in leachates were measured using ICP-MS. In concert with the findings of i) a dissolution study of perovskite in a flowing leachant and ii) a previous Synroc dissolution study (wherein Synroc containing 10 wt% simulated HLW was subjected to periodic replacement, leach testing in deionised water at 150°C), the results of this study show that when the leachant replacement frequency is varied from 7 d to the duration of the test, there is no effect on leach rate or leaching mechanisms.

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

Preparation and characterization of inductively-melted Synroc containing 20 wt% simulated plant “Mayak” reprocessing waste were performed. The sample bulk composition was as follows, (in wt.%): 55.4 TiO2; 15.8 ZrO2; 7.5 CaO; 7.4 BaO; 4.3 Al2O3 2.0 MnO; 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 by melting in air at 1550–1600 °C under barometric pressure. It is composed of a few crystalline phases and a minor glass phase. Most of the phases (hollandite, zirconolite, perovskite and rutile) are similar to the analogous phases found in the other Synroc formulations. An additional phase with 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 was found. Some elements (Ba, Si, Ni, K, Na, Fe) were present in the phase in negligible quantities. Its formula (Ca2.65U0.3Ce0.2)(Ti7.3Mn0.6Zr0.4Al0.3)O20.0 is rather close to a rare mineral uhligite - Ca3(Ti,Zr,Al)9O20. Another possible counterpart of the phase is murataite-like mineral previously described in tailored ceramic designed for Savannah River Plant wastes fixation. This phase as well as zirconolite are the major host for U in the sample Preliminary data on the material leachability in water at 350 °C and 50 MPa have been obtained Uranium contents in the solution were about 1 ppb and close to the uranium dioxide solubility in deionized water under the same P-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 at Radon, Russia. One of the melted samples was prepared using the Australian sol-gel precursor and the second one was obtained from an oxide-nitrate mixture. It was established that the specimens have closely similar mineral compositions, with major hollandite, perovskite, zirconolite, and rutile. Small amounts of hibonite were also found. Unlike the hot-pressed Synroc containing metallic alloy particles, melted Synrocs contain molybdates. An investigation of mineral compositions and elemental distribution in the samples was carried out. Features of hot-pressed and melted ceramics were compared. Unit cell parameters of the Synroc phases were determined and preliminary results on durability of the melted Synroc are presented.

Syntheses of novel titanate-based waste forms, potentially suitable for immobilizing Pu(III) and Am(III), have been performed. The dititanate compounds An2Ti2O7 (An=Pu or Am) were found to have a monoclinic structure. A series of pyrochlore-type solid solutions between An2Ti2O7 and the cubic pyrochlore Ln2Ti2O7 (Ln=Gd, Er, or Lu) has been established, and limits of solid solubility for An2Ti2O7 were determined. These limits were found to increase as the ionic radius of the lanthanide in the host decreased. Strontium-containing titanate compounds SrAn2Ti4O12 were also prepared and were characterized as exhibiting a perovskite-type structure. Such compounds could serve as waste hosts for the immobilization of not only the actinide but also 90 Sr. A 2 month leaching study of Er1.78Pu0.22Ti2O7 and SrPu2Ti4O12 by WIPP “A” brine was performed. Very low amounts (< 1 ppm) of Pu(III) were leached. Although the 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 systems investigated.

The compositional flexibility of the sodium zirconium phosphate (NaZr2(PO4)3) structure has been exploited in the design of monophasic radiophases capable of immobilizing the most common cations associated with reprocessed high-level commercial waste streams. Highly crystalline, monophasic members of the NaZr2(PO4)3 structural family ([NZP]) have been prepared with conventional processing methods and equipment. These radiophases were tailored to accommodate 10–20 wt % modified PW-4b simulated calcine as single phases isostructural with NaZr2(PO4)3. To meet the challenge of designing monophasic materials capable of accommodating the chemical complexity of PW-4b, an ionic substitution scheme based on crystal chemical principles was developed. The radiophases were prepared with inexpensive, inorganic precursors and a solution sol-gel method; these materials were heat treated and/or sintered under a variety of conditions to determine the optimum conditions for single phase [NZP] formation. X-ray powder diffraction provided valuable information that was used to assess the suitability of the ionic substitution model developed in this investigation. The results of this investigation suggest that monophasic [NZP] radiophases capable of accommodating 10–20 wt % modified PW-4b simulated calcine may be continuously processed with conventional ceramic processing methods and equipment. Moreover, the relatively low temperatures involved and the reproducibility of the process make [NZP] radiophases economically attractive hosts for radioactive and heavy metal industrial wastes.

Neutron diffraction studies of salt-occluded zeolite and zeolite/glass composite samples, simulating nuclear waste forms loaded with fission products, have revealed complex structures, with cations assuming the dual roles of charge compensation and occlusion (cluster formation). These clusters roughly fill the 6–8 Å diameter pores of the zeolites. Samples are prepared by equilibrating zeolite-A with complex molten Li, K, Cs, Sr, Ba, Y chloride salts, with compositions representative of anticipated waste systems. Samples prepared using zeolite 4A (which contains exclusively sodium cations) as starting material are observed to transform to sodalite, a denser alumi-nosilicate framework structure, while those prepared using zeolite 5A (sodium and calcium ions) more readily retain the zeolite-A structure. Because the sodalite framework pores are much smaller than those of zeolite-A, clusters are smaller and more rigorously confined, with a correspondingly lower capacity for waste containment. Details of the sodalite structures resulting from transformation of zeolite-A depend upon the precise composition of the original mixture. The enhanced resistance of salt-occluded zeolites prepared from zeolite 5A to sodalite transformation is thought to be related to differences in the complex chloride clusters present in these zeolite mixtures. Data relating processing conditions to resulting zeolite composition and structure can be used in the selection of processing parameters which lead to optimal waste forms.

Zirconolite-rich ceramics were produced by the cold crucible melting technique in an air atmosphere, at 1550±50 °C and 1 atm. Four samples with overall 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 been studied. Total phases in the ceramics consist of major zirconolite and minor rutile, perovskite, zirconia, aluminium titanate, and glass. The Gd2O3 content in zirconolite reaches up to 31.4 wt.% corresponding to the formula: (Cao4,Gd0.7)Zri.0(Tii4,Alo 5)070. The data on the phase composition agree well with coupled Gd incorporation into the 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 are 25.4 % for zirconia, 12.6 % in glass, 8.8 % in perovskite, and 1.4 % for rutile. The rest of the elements' distribution in the samples are analyzed.

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 isostatic pressing in metal bellows containers. The short-term chemical durability of the phosphate-based material containing Purex waste was within an order of magnitude ofthat for Synroc-C, as measured by 7-day MCC-1 tests at 90°C. The dissolution behaviour showed evidence of re-precipitation phenomena, even after times as short as 28 days. Potential for improvement of NZP-based ceramics for HLW management is discussed.

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

A technique for solidification of high-level liquid wastes (HLLW) to obtain mineral-like forms through the use of water-cooled melters with direct induction heating of a melt (CCIM) has been developed at the State Scientific 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 D were synthesized by the CCIM method at temperatures of 1250 – 1550°C.

The preliminary X-ray diffraction studies of these materials indicate that the structures synthesized by the CCIM method have compositions that are identical to those of the corresponding natural analogs. The chemical durabilities of the synthesized materials upon their exposure to distilled water at 20°C have been determined.

In addition to the investigation of solidification of liquid HLW, the CCIM method is examined and applied to the synthesis of mineral-like compositions involving simulated waste components (nonsoluble residues, ashes, and pulps) produced by the chemical and metallurgical manufacturing of fissile materials.

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

Glass-bonded zeolite is being developed as a potential ceramic waste form for the disposition of radionuclides associated with the U.S. Department of Energy's (DOE's) spent nuclear fuel conditioning activities. The utility of several standard durability tests [e.g., Materials Characterization Center Test #1 (MCC-1), Product Consistency Test-B (PCT-B), and Vapor Hydration Test (VHT)] was evaluated as a first step in developing methods and criteria that can be applied towards the process of qualifying this material for acceptance into the DOE Civilian Radioactive Waste Management System. The effects of pH, leachant composition, and sample surface-area-to-leachant-volume ratios on the durability test results are discussed, in an attempt to investigate the release mechanisms and other physical and chemical parameters that are important for the acceptance criteria, including the establishment of appropriate test methodologies required for product consistency measurements.

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

Vapor hydration and MCC-1 tests were performed with ceramic waste form monoliths of glass-bonded zeolite. The VHTs (temperatures at 120,150, and 200°C) provided useful information about the effect of glass composition on corrosion rates and alteration phase formation, and about the overall toughness and structural integrity of the ceramic waste form. The MCC-1 test was investigated as an alternative to the PCT for acceptance criteria measurements. The MCC-1 results indicate that corrosion testing with both DIW and high-ionic-strength leachants (that specifically affect the ion exchange behavior of the fission products) are required to fully assess the durability of the ceramic waste form. These preliminary results establish the utility of the MCC-1 test for providing possible acceptance criteria measurements, including elemental release comparisons between the environmental 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 to the mineral pollucite, CsAlSi2O6, with Ti+4 replacing Al+3. This replacement requires a mechanism for charge compensation. A combination of techniques including neutron diffraction, single crystal x-ray diffraction and x-ray absorption spectroscopy have revealed 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 is five-fold. Pentacoordinate titanium and tetrahedral silicon form a network structure with Cs residing in cages formed by the network. The crystal structure of Cs2TiSi6O15 is unique, with titanium octahedra and silicon tetrahedra forming an open framework structure with the Cs residing in large cavities. The largest covalently bonded ring opening to the Cs cavities in both compounds are smaller than a Cs ion, revealing that the Cs ion has minimal mobility in the structure. Cesium leach rates for both compounds are lower than or comparable to borosilicate glass.

Silicotitanate ion exchangers are potential materials for the removal of radioactive Cs and Sr from tank wastes. In this paper the viability of direct thermal conversion of Cs-loaded silicotitanates to an acceptable high level 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 or above 900 °C for 1 hour further reduces the Cs leach rates to approximately zero (beyond the lppm detection limits). Moreover, Cs volatilization was found to be < 0.8 wt% at temperatures as high as 1000 °C. These results suggest that thermally converted silicotitanate ion exchangers exhibit excellent chemical durability (comparable to or better than borosilicate glass) and thus, have great potential as an alternative waste form.

Aluminosilicates are one of the major class of species controlling the volume of radioactive high-level waste that will be produced from future remediation at Hanford site. Here we present studies of the phases and structures of aluminosilicates as a function of sludge composition using X-ray powder diffraction, solid state 27Al and 29Si NMR, and Raman spectroscopy. The results show that the content of NaNO3 in solution has significant effects on the nature of the insoluble aluminosilicate phases produced. It was found that regardless of the initial Si:Al ratio, nitrate cancrinite was the main phase formed in the solution with pH of 13.5 and 5 MNaNO3. However, at lower NaNO3 concentration with initial Si:Al ratios of 1.1, 2.2, and 11.0 in the solutions, a range of aluminosilicate zeolites was produced with Si: Al ratios of 1.1, 1.3, and 1.5, respectively. Lowering the solution pH appears to promote the formation of amorphous aluminosilicates. The results presented here are important for the prediction of the solubility and dissolution rate of Al in tank wastes.

The effect of the addition of an organic monolayer to the surface of a clay mineral on the speciation of metal ions intercalated into the clay interlayer is probed by X-ray absorption spectroscopy. The presence of the monolayer changes the surface of the clay from hydrophilic to hydrophobic. It inhibits the interlayer ions from exchanging freely into environmental water and reduces the leach rate of cations out of the clay by approximately a factor of 20. Significant changes are observed when these coated samples are treated under hydrothermal and thermal conditions. Reductions of uranium(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 and with the reduced copper species.