Polycrystalline pyrochlore oxides consisting of selected f elements (lanthanides and actinides) and Zr and Hf have been prepared and characterized. Characterization to date has been primarily by X-ray diffraction, both at room and at elevated temperatures. Initial studies concentrated on selected lanthanides and the Np, Pu and Am analogs (reported here) but have been extended to the other actinide elements through Cf. Data from these studies have been used to establish a systematic correlation regarding the fundamental materials science of these particular pyrochlores and structurally related fluorite-type dioxides. In addition to pursuing their materials science, we have addressed some potential technological applications for these materials. Some of the latter concern: (1) immobilization matrices; (2) materials for transmutation concepts; and (3) special nuclear fuel forms that can minimize the generation of nuclear wastes. For f elements that display both a III and IV oxidation state in oxide matrices, the synthetic path required for producing the desired pyrochlore oxide is dictated by their pseudo-oxidation potential the stability of the compound towards oxygen uptake. For the f elements that display an oxidationreduction cycle for pyrochlore-dioxide solid solution, X-ray diffraction can be used to identify the composition in the oxidation-reduction cycle, the oxygen stoichiometry and/or the composition. This paper concentrates on the Np, Pu and Am systems, and addresses the above aspects, the role of the crystal matrix in controlling the ceramic products as well as discussingsome custom-tailored materials.

The formation enthalpy, - 3752.3 ± 4.7 kJ·mol−1, of Hf-zirconolite, CaHfTi2O7, was obtained using high temperature oxide-melt solution calorimetry. Combined with heat capacity data obtained using low temperature adiabatic calorimetry we report the heat capacity (Cp) and the standard molar formation energetics (ΔH°f. elements, Δ S°T, and ΔG°f. elements)for Hf-zirconolite from T = 298.15 K to T = 1500 K. Comparison of Hf-zirconolite with Zr-zirconolite is made.

Titanate-based ceramic waste forms are currently under development for the immobilization of excess weapons plutonium. Both Hf and Gd are added to the ceramic formulation as neutron absorbers in order to satisfy a defense-in-depth concept for the waste form. The introduction of significant amounts of hafnium may be responsible for the presence of zirconolite-2M crystals in pyrochlore-based ceramics and the formation of zirconolite lamellae within pyrochlore. The zirconolite grows epitaxially on { 111 }planes of pyrochlore. Although the zirconolite lamellae within pyrochlore are non-cubic, any volume expansion due to radiation damage in the pyrochlore should still be isotropic; in addition, the presence of these intergrowths may allow some stress relief in the ceramic.

Pyrochlore is a complex oxide with the nominal formula A2B2X6Y, where A and B are cations in VIII and VI-fold co-ordination, X and Y are anions. Its structure is derived from the cubic fluorite structure. In natural pyrochlores A = Na, Mg, K, Ca, Mn, Fe, Sr, Sb, Cs, Ba, REEs, Pb, Bi, Th, and U; B = Nb, Ta, Ti, Zr, Sn, W, Fe, and Al; X = O; Y = O, OH, or F. Synthetic pyrochlores have been repeatedly described as matrices designed for actinide-bearing waste immobilization. In synthetic pyrochlores site "A" is mainly occupied by Ca, U, An, and REEs; B = Ti and Zr; X and Y = O. In this work we have studied pyrochlores in crystalline titanate-based waste forms. The ceramics were fabricated in the system: Ca-Mn-U-REE-Zr-Ti-Al-O by cold pressing and sintering, melting in a high-temperature furnace, and inductive melting in a cold crucible. All specimens were studied by XRD, SEM/EDS and TEM methods. The amount of pyrochlore in the samples varied from 10 to 70%. Other phases in these ceramics were brannerite, perovskite, zirconolite, murataite, hibonite, loverengite, pseudobrookite, and rutile. Compositions of the pyrochlores correspond to stoichiometry: A2B2O7-x, 0.1<x<0.4, where A = Ca, Mn, REEs, U, Zr; B = Ti, Zr, Al, Mn. The positions and intensities of the peaks of pyrochlores from various ceramics were: d222=2.89-2.93A, 1=100; d400=2.51, 1=10-25; d440=1.779-1.809, 1=20-60; d622=1.512-1.540, 1=20-35; d444=1.451-1.477, I=10-15; d662=1.158-1.173, I=10-15. These data allowed the determination of the unit-cell dimensions of the pyrochlores as 1.00-1.02 nm. Results obtained from TEM research agree well with these values. Distribution of U and REEs among all phases of the ceramics was characterized. The main substitutions which have influenced the pyrochlore compositions are discussed.

Zirconolite and pyrochlore are considered as promising host phases for high level waste (HLW). However, correct information on substitution mechanisms, forms of dopants incorporation in their structures and distortions in coordination polyhedra is presently unavailable. To clarify these points we use the electron paramagnetic resonance (EPR). Pyrochlore and three of zirconolite polytypes: zirconolite-2M, zirconolite-3T, and zirconolite-30 are considered. Pyrochlore is the “parent” structure for zirconolite since any zirconolite variety is produced by means of distortion of the initial pyrochlore structure. Space groups of pyrochlore and basic polymorphous zirconolite varieties found from XRD and TEM data, as well as interatomic distances and angles, were taken from reference data. This allows the determination of the most probable sites for impurities, substitution mechanisms, and local symmetry of coordination polyhedra (initial). Ions chosen for EPR were Gd (III) as an analog of trivalent rare earth and actinide elements which are also occurred in HLW and Fe (III) as a typical corrosion product which occurs in all HLW. For Gd (III) a strong ligand field approximation is suggested, theoretical computation using perturbation theory in this approximation has been carried out. All the non-diagonal members plus magnetic field were chosen as perturbation and formulae for transition frequencies, estimations of fine structure and g-factors parameters in the given approximation have been obtained.

Halides can be retained in bellows hot-pressed Synroc-C. Loadings of 0.2 wt% of fluoride or 2 wt% of sulphate did not appear to exert deleterious effects on the aqueous durability of the Synroc, but 0.2 wt% of chloride or 0.5 wt% of fluoride increased the Cs leach rate by a factor of ˜10. Fluoride was present principally as CaF2 in both calcined and hot-pressed Synrocs containing (a) ˜2 wt% of F as the only waste or (b) ˜2 wt% of F plus 20 wt% of simulated PW-4b Purex reprocessing waste. The speciation of Cl- in Synroc is not clear, though there was some evidence of CsCl from nmr and a compound giving a similar X-ray pattern to that of TiCI3. Iodine can retained in the hot-pressed Synroc containing the full complement of fission products, probably as insoluble Ag, Cr or Pd group iodides.

Scanning electron microscopy (SEM), transmission electron microscopy (TEM), and analytical electron microscopy (AEM) studies have been conducted on samples crystallized from a melt with a composition of zirconolite {(Ca0.9Gd0.1)Zr(Ti1.9Al0.1)2, O7}. The formationof a whole suite of Synroc minerals (zirconia, ZrTiO4, zirconolite, perovskite, and rutile) has been observed. The formation of these minerals follows the crystallization sequence of Ti-bearing zirconia → ZrTiO4 phase → Zr-rich zirconolite → Zr-poor zirconolite →rutile/perovskite. This sequence is induced by a fractional crystallization process, in which Zr-rich mineral phases tend to crystallize first, resulting in continuous depletion of Zr in melt. Consistent with this melt compositional evolution, Zr content in the zirconolite decreases from the area next to ZrTiO4 phase to the area next to rutile or perovskite. High-resolution TEM images show that there are no glassy phases at the grain boundary between zirconolite and perovskite. The fractional crystallization-induced textural heterogeneity may have a significant impact on the incorporation of radionuclides into crystalline phases and the resistance of radioniclides to leaching processes. Exsolution lamellae and multiple twinning result from the phase transition from tetragonal zirconia to monoclinic zirconia may decrease durability of the Synroc. Fast cooling of melt may produce more zirconolite phase andrelatively uniform texture. In general, however, a Synroc prepared by a through-melt method is less uniform in texture than that prepared by a through-sol-gel method. The reaction path calculation for the alteration of U-bearing zirconolite in an oxidizing fluid shows that zirconolite is first altered into a perovskite-like phase (CaZrO3), followed by rutile (TiO2 ), and U6+ -bearing phases of soddyite [(UO2)2SiO42H20] and haiweeite [Ca(U02)2Si6O15·5H2O].

The use of garnet/perovskite-based ceramic, with formula type (Y, Gd,..) 3(AI, Ga,..) 5O12 12/(Y, Gd,..)(A1, Ga,..)O3, was tested for immobilizing plutonium residue wastes. Pu residue wastes originate from nuclear weapons production and can contain more than 50% of impurities including such elements as Am, Al, Mg, Ga, Fe, K, La, Na, Mo, Nd, Si, Ta, Ce, Ba, B, W, Zn, Zr, C and Cl. While for some of these residues, direct conversion to typical glass or ceramic forms may be difficult, ceramic forms based on durable actinide host-phases are preferred for Pu, Am and other actinides immobilization. Garnet/perovskite crystalline host-phases are chemically and mechanically durable and desirable for the incorporation of Pu and most of the impurity elements in the Pu residue wastes in the lattices of host-phases in the form of solid solutions. Experiments on the synthesis of garnet/perovskite ceramic samples were carried out using melting in air at temperatures from 1300°C (for samples doped with 10 wt.% Pu residue waste simulant) to 2000°C (for samples doped with 10 wt.% Ce or U). Samples were studied by XRD, SEM and cathodoluminescence techniques. It was found that the garnet phase can incorporate upto 6 wt.% Ce and up to 4.0-5.5 wt.% U, which is correlated with the increase of Ga content and decrease of Al content in the melt. In one of the features of the melt, the perovskite phase formation substitutes for the formation of garnet. The capacity of the perovskite lattice to accommodate Ce and U is higher than the capacity of garnet, reaching about 8 and 7 wt.%, respectively. It was shown that cathodoluminescence can be effectively used to determine the valence state of Ce and U, an important step to optimize the starting precursor preparation. Incase of U4+ in the melt, the charge-compensating elements (Sn2+, Ca2+...) are needed to successfully incorporate U in the garnet lattice.

We synthesized a ceramic containing simulated excess weapons plutonium waste in solidsolution with zirconia (ZrO2)ss. ZrO2 has a large solubility for other metal oxides. More thantwenty binary systems AxOy- ZrO2 have been reported in the literature, including PuG2, rare earth elements, and oxides of metals contained in weapons plutonium wastes. We show that significant amounts of gadolinium (neutron absorber) and yttrium (stabilizer of the cubic modification) can be dissolved in ZrO2, together with plutonium (simulated by Th4+, Ce4+, or U4+) and impurities (e.g., Ca Mg, Fe, Si). Sol-gel and powder methods were applied to make homogeneous, single phase zirconia solid solutions. Pu waste impurities were completely dissolved in the solid solutions. In contrast to other phases, e.g., zirconolite and pyrochlore, yttrium stabilized cubic zirconia does not undergo amorphization upon irradiation.

The radioactivity of the Hanford site waste tanks is primarily from 137Cs and 90Sr, of which can both be selectively removed from solution using a crystalline silicotitanate (CST) ion exchanger. We are currently seeking waste forms alternative to borosilicate glass for Cs-CSTs. In order to obtain a fundamental basis for the development of an alternative waste form, we are investigating synthesis and characterization of CST component phases, namely Cs-Si-Ti-O phases. Two novel Cs-Ti-Si-O phases (one porous, one condensed) have been hydrothermally synthesized, characterized and evaluated as waste form candidates based on chemical and thermal stability, leachability, and ion exchange capabilities.

Crystalline silicotitanate ion exchangers are highly selective for separating Cs from Narich waste streams. However, use of these ion exchangers for removal of Cs from radioactive tank waste will result in large volumes of secondary wastes. Thermal conversion of silicotitanates produces a durable waste form with reduced volumes up to 40%. Leach tests (MCC-l and PCT) have shown that Cs leach rates of IE-91 1-Na (heat treated at 900°C for an hour) are extremely low, ranging from 0.1 to lwt% Cs loss in Cs fraction release, or 10-1 to 10-8g/m2day in normalized Cs mass loss. These are several orders of magnitude lower than that of borosilicate glass. In order to understand the interplay between the structure and high Cs durability, X-ray diffraction, 133Cs NMR, and thermogravimetric analysis have been used to identify phase(s) responsible for trapping Cs in these silicotitanates. Results indicate that Cs is likely to be contained in a crystalline silicate phase.

In the framework of radwaste storage, the thorium phosphate-diphosphate Th4(PO 4)4P 2O7 (TPD) has been synthesized in the ThO2 – P2O5 system. This material is very resistant to aqueous corrosion. In this paper we report the conditions of synthesis of TPD doped with small amounts of trivalent actinides and of solid solutions with uranium (IV), neptunium (IV) and plutonium (IV). Sintered samples of TPD and Th4_xUx(PO4)4P2O7 (TUPD) have been prepared after pressing at 100-800 MPa at room temperature then heating at 1250°C for 10 hours. The corresponding density varies from 87 to 93% of the calculated value from crystal data. Leaching tests of Th4 (PO4)4P2O7 doped with Am3+ or Cm3+ and solid solutions Th4_xUx(PO4)4P2O7 (TUPD) or Th4-xPux(PO4)4P2O7 (TPPD) have been performed in water and in several acidic media at room temperature and at 90'C. The dependence of the leaching rate on the pH has been determined: the partial order related to [H3O+] is in the 0.35 – 0.44 range while the apparent leaching rate is equal to about 1 x 10–5x–4 x 10–5 g/(m2d).

The French Act, dated 30 December 1991, related to “research concerning the elimination of radioactive wastes”, defines the research programs including the study of new conditioning processes and long-term surface storage of these wastes. The research of new conditioning processes mainly concerns separated long-lived radionuclides.

A new method to estimate the thermodynamic stability of crystalline host phases for radwaste is based on the determination of the degree of coordination polyhedra distortion. Analysis of the stability of sodium zirconium phosphate (NZP) structure containing Fe and Cr has been completed. The NZP structure can incorporate a limited amount of iron (∼2 mol. % Fe2o3). From the EPR study, Fe3+ and Cr3+ ions occupy two different structural positions substituting for Zr4+and Na+. Moreover, these ions can enter extra phases.

A ceramic waste form is being developed by Argonne National Laboratory* (ANL) as part of the demonstration of the electrometallurgical treatment of spent nuclear fuel [1]. The halide, alkaline earth, alkali, transuranic, and rare earth fission products are stabilized in zeolite which is combined with glass and processed in a hot isostatic press (HIP) to form a ceramic composite. The mineral sodalite is formed in the HIP from the zeolite precursor. The process, from starting materials to final product, is relatively simple. An overview of the processing operations is given. The metrics that have been developed to measure the success or completion of processing operations are developed and discussed. The impact of variability in processing metrics on the durability of the final product is presented. The process is demonstrated to be robust for the type and range of operation metrics considered and the performance metric (PCT durability test) against which the operation metrics are evaluated.

An electrometallurgical process is being developed at Argonne National Laboratory to treat spent metallic nuclear fuel. In this process, the spent nuclear fuel is electrorefined in a molten salt to separate uranium from the other constituents of the fuel. The treatment process generates a contaminated chloride salt that is incorporated into a ceramic waste form. The ceramic waste form, a composite of sodalite and glass, contains the fission products (rare earths, alkalis, alkaline earth metals, and halides) and transuranic radionuclides that accumulated in the electrorefiner salt. These radionuclides are incorporated into zeolite A, which can fully accommodate the salt in its crystal structure. The radionuclides are incorporated into the zeolite by hightemperature blending or by ion exchange. In the blending process the salt and zeolite are simply tumbled together at >450°C (723 K), but in the ion exchange process, which yields a product more highly concentrated in fission products, the molten salt is passed through a bed of the zeolite. In either case, the salt-loaded zeolite A is mixed with glass frit and hot isostatically pressed to produce a monolithic leach resistant waste form.Zeolite is converted to sodalite during hot pressing. This paper presents experimental results on the experimental results on the fission product uptake of the zeolite as a function of time and salt composition.

Studying the Synroc, doped with a simulated HLW, we have found, along with conventional Synroc phases (zirconolite, perovskite, hollandite), an extra phase with a stoichiornetry (Ca, Mn, U, TR)4(U, TR, Zr, Ti) 2(AI, Ti)7O22. XRD and TEM study has shown this phase is related to a very rare mineral murataite. In the present work a ceramic based on murataite is studied. The ceramic samples in the system: Ca-Mn-Ti-Zr-U-Ce-AI-Fe-O were produced and examined in details using XRD, SEM/EDS, TEM, and optical microscopy. Total amount of actinide (U) and rare earth (Ce, Gd) elements in the murataite exceeds 20 wt%. Isomorphic substitution schemes in the structures of synthetic and natural murataites are discussed. High isomorphic capacity of the murataite structure towards actinides and REEs, flexibility of its composition, feasibility of synthesis by melting, including a cold crucible melting, and very high chemical durability under hydrothermal conditions make the murataite-based ceramics very promising for actinides and excess weapon Pu fixation.

This paper describes the process of gallium removal from Ga2O3 -doped CeO2−x a surrogate for weapons-grade PuO2−x. Gallium is removed from the surrogate feedstock material using thermal techniques. An Ar-6% H, gas was used in order to reduce the oxide to gaseous Ga2O. Experimental results were shown in the temperature range of 600°C to 1200°C as a function of time and sample geometry. The results to date have shown that CeO2−x is a very good surrogate for PuO2−x.

One of the basic physical parameters which defines: (1) the capacity of the solid matrix for the incorporation (sorption) of the impurity and (2) the stability of the solid matrix -impurity system, is the ion-ion interaction potential, representing the main term of the cohesive energy. Using this parameter determined in the frame of the pseudopotential theory and pseudoatomic approximation we have investigated the systems HAP-Sr and HAP-UO2. In analysis, the hydroxyapatite (HAP, Ca10(PO4)6(OH)2) was selected as a model solid matrix since insoluble phosphates, phosphate ceramic and apatite-like materials are candidates for the immobilization of radionuclides. The substitution of calcium atoms by different impurities in HAP can be presented by the formula: Ca10-xMx(PO4)6(OH)2, where 0 ≤ x ≤ 10 and M = Sr2+, UO2 2+. It has been shown that (1) Sr can be safely immobilized by HAP, natural apatites and phosphate rocks in the whole range of its concentrations, and (2) that HAP is not an appropriate solid-matrix for immobilization of UO2 2+ because its incorporation in the low concentration results in a decrease of the matrix stability, stimulating formation of the inhomogeneous structure containing isolated clusters of the UO2- apatite solid phase.

Our study is concerned on development of a new stabilisation method of ash residue obtained after incineration of solid radioactive waste by means of heat liberating batches - compositions with redox properties. The results of optimisation of heat-liberating batches with the purpose of increasing process safety and obtained product quality of are discussed.