A program of work has been undertaken to treat plutonium-residues wastes at Sellafield. These have arisen from past fuel development work and are highly variable in both physical and chemical composition. The principal radiological elements present are U and Pu, with small amounts of Th. The waste packages contain Pu in amounts that are too low to be economically recycled as fuel and too high to be disposed of as lower level Pu contaminated material. NNL and ANSTO have developed full-ceramic and glass-ceramic waste forms in which hot-isostatic pressing is used as the consolidation step to safely immobilize the waste into a form suitable for long-term disposition. We discuss development work on the glass-ceramic developed for impure waste streams, in particular the effect of variations in the waste feed chemistry glass-ceramic. The waste chemistry was categorized into actinides, impurity cations, glass formers and anions. Variations of the relative amounts of these on the properties and chemistry of the waste form were investigated and the waste form was found to be largely unaffected by these changes. This work mainly discusses the initial trials with Th and U. Later trials with larger variations and work with Pu-doped samples further confirmed the flexibilitymore » of the glass-ceramic. (authors)« less

The structural behavior of zirconolite (CaZrTi2O7) under reducing conditions at high temperature has been studied, mainly by scanning electron microscopy (SEM) and x-ray diffraction (XRD), but also with x-ray absorption spectroscopy, thermogravimetry, and electron paramagnetic resonance. The partial reduction of Ti4+ to Ti3+, associated with a reducing atmosphere heat treatment, led to the initial formation of perovskite (CaTiO3) as a second phase. As the concentration of Ti3+ in the zirconolite increased, so did the amount of perovskite until the zirconolite was totally transformed into a fluorite structured phase. Analysis of the reduced zirconolites showed them to be consistently deficient in Ca and enriched in Zr, in proportion to the concentration of Ti3+. To determine how electroneutrality was preserved in these reduced zirconolites, a series of zirconolites were prepared in air using In3+ and Ga3+ as models for Ti3+. These samples were then investigated by neutron and x-ray diffraction, SEM, solid state nuclear magnetic resonance (NMR), and nuclear quadrupole resonance (NQR). 71Ga MAS NMR studies of the Ga substituted zirconolite exhibited a narrow resonance at ˜13 ppm which was attributed to six-coordinate Ga incorporated in a trace perovskite phase. Broadline 71Ga NMR and 69/71Ga NQR were required to characterize the Ga incorporated in the zirconolite. The resultant quadrupolar parameters of CQ = 30.0 ± 0.05 MHz and η = 1.0 ± 0.03 indicate that the Ga site is in a highly distorted environment which would suggest that it is located on the five-coordinate Ti site within the zirconolite lattice. These results were complemented by Rietveld refinement of the neutron diffraction data from the In-doped zirconolite sample, which was optimal when all the In was located on the five-coordinate Ti site with the excess Zr located on the Ca site. It would therefore appear that charge compensation for the presence of Ti3+ in zirconolite is effected via the substitution of an appropriate amount of Zr on the Ca site. The Ti3+-stabilized fluorite structure was readily oxidized back to a single phase zirconolite upon heating in air.

Solid solubility limits of U, Pu, and the neutron absorbers Hf and Gd have been measured for zircon (ZrSiO4), monazite (CePO4), titanite (CaTiSiO5), perovskite (CaTiO3), apatite (Ca10(PO4)6O), in almost all cases where these limits were not known beforehand. The method used was to oversaturate the host phase with the dopant, using a nominated substitutional scheme, and then establish the dopant content of the host phase by microanalysis/scanning electron microscopy. Tetravalent U has limited solid solubilities in titanite, perovskite and apatite. X-ray absorption near-edge and diffuse reflectance spectroscopies were used to show that U was tetravalent in U-doped perovskite prepared in both argon and hydrogen-nitrogen atmospheres, with different charge compensating schemes. Tetravalent Pu has solubilities of 0.13 and 0.02 formula units (f.u.) in perovskite and titanite respectively. Trivalent Pu has a solubility of 0.05 f.u. in titanite. Pu3+ dominates tetravalent Pu in monazite fired in air at 1400°C. At least 0.5 and < 0.1 f.u. of Hf are soluble in titanite and monazite respectively.Hf solubility in apatite is estimated as < 0.1 f.u. Approximately 0.3 and < 0.1 f.u. of Gd are soluble in titanite and zircon respectively

New X-ray diffraction and scanning electron microscopy data are given for the incorporation of Np and Pu in zirconolite, at levels of tens of percent. The actinide valences and the cations they replace are deduced from the microanalysis of the zirconolite compositions, and X-ray absorption data are used to obtain more direct information on the valences of Ce and Nd, which are used as simulants of Pu and trivalent actinides respectively. Trivalent rare earths and actinides have extensive solid solubility in zirconolite, mainly but not exclusively in the Ca site. Tetravalent rare earths and actinides have considerable solid solubility in the Zr site of zirconolite, and some solubility in the Ca site, but the strong tendency of zirconolite with ions substituted in the Zr site to undergo phase separation complicates structural interpretation. In zirconolite-rich Synroc-type ceramics designed to immobilise waste actinides, the target actinide waste loading has been set at 20 wt% and early leach results indicate the durability is at least as good as that of Synroc-C.

As the level of Pu4+ substituted on the Zr-site in CaZr1-xPuxTi2O7 zirconolite increased, from x=0.1 to 0.6, a series of structural transitions occurred from zirconolite-2M to zirconolite-4M and subsequently from zirconolite-4M to pyrochlore. The solid-solution limit for Pu4+ substituted on the Zr-site in zirconolite-2M was ~ 0.15 formula units. Zirconolite-4M was only stable over a narrow compositional range, centered about CaZr0.59Pu0.41Ti2O7, whilst the pyrochlore structure was stabilized with CaZr0.4Pu0.6Ti2O7 stoichiometry. The stability of the zirconolite polytypes is therefore sensitive to the average effective ionic size of the ions occupying the seven-coordinated Zr-site. The reduction in Pu from Pu4+ to Pu3+ destabilized the zirconolite-4M, producing a mixture of perovskite and possibly zirconolite-3T. The CaZr0.4Pu0.6Ti2O7 pyrochlore was also predominantly transformed to perovskite as a result of this reduction of Pu.

A calorimetric investigation of the enthalpies of formation of Gd 2 (Ti 2-x Zr x )O 7 , where 0≤ × ≤ 2 is underway. All samples exhibit pyrochlore (Fd3m) peaks in their XRD patterns. However, where x=2 significant local disorder is observed in the Raman spectra. Preliminary data for the enthalpies of formation from the oxides in kJ/mol are: x=0, ΔH f = -113.4±2.7; x=0.5, ΔH f = -94.0±3.0; x=1.0, ΔH f = -74.2±4.9; x=1.5, ΔH f = -64.5±2.0; x=2, ΔH f = -52.2±4.8. Two additional samples, Gd 1.80 Zr 2.15 O 7.00 (pyrochlore) and Gd 2.15 Zr 1.87 O 7.00 (fluorite), were also studied. Their enthalpies of formation from the oxides in kJ/mole are -50.9±3.3 and -46.4±3.4 respectively. Replacing Ti with Zr, i.e. when x=2, destabilizes the pyrochlore in enthalpy by approximately 60 kJ/mol. The ΔH mix for the Gd 2 (Ti 2-x Zr x )O 7 solid-solution series is positive and can be described by a regular solution formalism with an estimated interaction parameter, ŝ = +20 kJ/mol. The results of this study suggest that the pyrochlore to fluorite transition enthalpy in Gd 2 Zr 2 O 7 is small, of the order of the configurational entropy contribution due to cation disorder at the transition temperature, TΔS conf. ≍ 10 kJ/mol.

A brief history of the technical development of immobilisation strategies for high-level nuclear wastes (HLWs) is given. The desirable performance characteristics of the waste-immobilising matrices (waste forms) are outlined. The pros and cons of different classes of waste forms, as well as spent fuel itself, are discussed, together with common production methods. While borosilicate glass is the baseline waste form to deal with the majority of HLW, ceramic and glass-ceramic waste forms can offer performance and economic benefits for the immobilisation of HLWs that are difficult to incorporate in borosilicate glass, due to limited solid solubility of HLW ions or the presence of volatile species.

Trapping experiments have been performed at the Idaho National Laboratory to assess the performance of AgX sorbent media in capturing volatile iodine during the oxidation of irradiated oxide fuel. The demonstration of iodine release and capture from the used fuel has been accomplished with laboratory-scale equipment in a hot cell environment. Iodine loadings as high as 6 ug/g media have been achieved via chemical adsorption with filter efficiencies in excess of 90%. In addition to iodine, significant quantities of tritium have also been collected on the AgX filter media. Filter media loaded with radioactive iodine has been sequestered in a tin matrix by hot isostatic pressing at 200°C. The placement and encapsulation of the sorbent media was examined by neutron radiography, thus confirming the sequestration of radioactive iodine.

The interactions between stainless steel (SS) and a glass–ceramic designed for immobilisation of the high-level nuclear waste generated at the Idaho chemical processing plant (ICPP) under hot isostatic pressing (HIPing) conditions (100MPa in argon at 1200°C) have been studied and subsequently the effect of such interactions on the chemical durability of the glass–ceramic waste form has been examined. The diffusion of Cr from SS (Cr depletion in SS) through the interaction layer and formation of crystalline Cr/Al oxides in glass dominate the overall interaction process. It appears that the depletion of Cr in SS may reduce the potential of SS as a barrier. However, such interactions have no significant impact on the glass–ceramic and the presence of the interaction layer does not seem to have any detrimental effect on the chemical durability of the glass–ceramic as a waste form.

The mode and energy of simple defect incorporation in SrTiO3 (vacancies and interstitials) is quantified using computer simulation techniques with an empirical partial charge model of interatomic forces, as well as using density functional theory calculations. Oxygen and strontium interstitials form split-interstitial configurations whereas titanium interstitials occupy channel positions. Defect migration energies and paths are also considered; interstitials are more mobile than vacancies, with a low predicted oxygen interstitial migration energy of around 0.3eV. We also calculate the threshold displacement energy (Ed) for each atom type in SrTiO3 perovskite using molecular dynamics simulations, by introducing a primary knock-on atom with a range of energies (20–250eV) in principal crystallographic directions at 300K. We find that all atom types are most easily displaced via direct replacement sequences on their own sublattices, which are extensive for Sr atoms due to focusson processes acting along channels. The weighted average threshold displacement energies (for use in TRIM-type calculations) are 50eV for oxygen, 70eV for strontium and 140eV for titanium atoms. These computed energies for O and Sr are comparable to experimentally-derived values in perovskites, whereas the Ed for Ti is much higher; it is expected that the value reported here is more accurate due to experimental difficulties in distinguishing different types of defects.

Residues and waste streams containing plutonium present unique technical, safety, regulatory, security, and sociopolitical challenges. In the UK these streams range from lightly plutonium contaminated materials (PCM) through to residues resulting directly from Pu processing operations. In addition there are potentially stocks of Pu oxide powders whose future designation may be either a waste or an asset, due to their levels of contamination making their reuse uneconomic, or to changes in nuclear policy. While waste management routes exist for PCM, an immobilisation process is required for streams containing higher levels of Pu. Such a process is being developed by Nexia Solutions and ANSTO to treat and immobilise Pu waste and residues currently stored on the Sellafield site. The characteristics of these Pu waste streams are highly variable. The physical form of the Pu waste ranges from liquids, sludges, powders/granules, to solid components (e.g., test fuels), with the Pu present as an ion in solution, as a salt, metal, oxide or other compound. The chemistry of the Pu waste streams also varies considerably with a variety of impurities present in many waste streams. Furthermore, with fissile isotopes present, criticality is an issue during operations and in the store or repository. Safeguards and security concerns must be assessed and controlled. The process under development, by using a combination of tailored waste form chemistry combined with flexible process technology aims to develop a process line to handle a broad range of Pu waste streams. It aims to be capable of dealing with not only current arisings but those anticipated to arise as a result of future operations or policy changes.

A wide range of plutonium containing wastes and residues are currently stored on the Sellafield site. These wastes and residues arising from early fuel development activities at Sellafield are diverse in nature and in the absence of a long-term disposition strategy, ongoing storage represents a major financial commitment. An immobilisation technology is being developed, producing a glass ceramic wasteform that will be suitable for extended periods of storage and ultimate disposal. Hot isostatic pressing [HIP] has been chosen as the preferred consolidation route. Following demonstration of the chemical feasibility of the approach at ANSTO's Lucas Heights facility, the project will demonstrate the technology by designing and building a fullscale inactive facility at Nexia Solutions' Workington site supported by a small scale active facility in the Technology Centre (BTC) at Sellafield. Delivery of this technology will allow the project sponsors, British Nuclear Group, the option of engineering a full scale plant in order to immobilise the actual waste arisings and discharge the liability to site.

Stainless steel/synroc interactions under HIPing conditions (1280°C/100MPa/3 h) have been studied. The synroc material was based on the zirconolite-rich ceramic targeted for surplus Pu disposition. A ∼300mm-thick complex reaction interface with 8 distinct layers has been identified. Although the Fe diffusion controlled interactions have changed the microstructures of the synroc phases at the interface, they do not affect the integrity of synroc and are unlikely to have any detrimental effect on this synroc derivative.

The environmental legacy of Cold War era nuclear activities presents a major cleanup challenge to the world. Legacy nuclear waste in the US alone is spread across many sites, with the majority of the nuclear waste located at the Department of Energy (DOE) sites at Hanford, Savannah River, and the Idaho National Laboratory. The waste is present in a great many chemical and physical forms and contains a varying mixture of radioactive and hazardous chemical components. The highest cost component of the nuclear waste clean up challenge centres on High-Level Waste (HLW) derived from reprocessing of nuclear fuel to recover plutonium and uranium, and consequently the greatest opportunity for cost and schedule savings lies with optimising the approach to the cleanup of this HLW.

Different modes of charge compensation for trivalent rare earth ions substituted for Ca in perovskite formed by high-temperature sintering are discussed. Cation vacancies were deduced from positron annihilation lifetime spectroscopy (PALS) measurements to exist in Ca((1-x))LaxTiO3 solid solutions (x = 0.001- 0.2) fabricated in air at similar to1500degreesC whether or not they were further heated in reducing atmospheres. Electron paramagnetic resonance measurements have shown the presence of paramagnetic species in Ca((1-x))LaxTiO3 samples fabricated in air, but their identities are not yet clear (other than Mn2+ impurities), although their intensities are not linear in La content, so the species of interest are unlikely to be Ti3+. (139) static nuclear magnetic resonance (NMR) measurements show the La ions in Ca(1-x)La samples fabricated in air to yield very broad featureless line shapes, indicative of low site symmetry and/or interactions with paramagnetic species. However, corresponding broad resonances were also observed in vacancy-compensated samples; this line broadening observed in the compensated and uncompensated systems is attributed to low site point symmetry and short-range disorder in bond angles and bond lengths around each La3+ position. This proposition is supported by the observation of narrow (139) NMR resonances in the vacancy-compensated system Sr((1-3x/2))LaxTiO3 (for x . 0.1), where the point symmetry of the Sr site was cubic rather than orthorhombic, as encountered in the analogous Ca((1-x))LaxTiO3 system. Samples were fabricated in air with Ca((1-x))LaxTi((1-x/4))O-3 stoichiometry to promote Ti vacancies, and the solid solubility limit corresponded to 0.10 f.u. of such vacancies.

Spent nuclear power plant fuel and the waste fission products and actinides from the reprocessing of nuclear fuels for commercial power or weapons production are classed as High-Level Waste (HLW). This chapter presents a brief history of the technical development of immobilization strategies for high-level wastes and the desirable performance characteristics of the waste-immobilizing solids (waste forms). It highlights the pros and cons of different types of waste forms—borosilicate, phosphate, and other glasses; silicate, aluminate, phosphate, and titanate ceramics; glass-ceramics; cements and geopolymers; as well as spent fuel itself, together with common production methods—melting, sintering, and hot uniaxial or isostatic pressing. It also presents some of the issues in the fundamental science of waste form behavior in a repository, discusses geological disposal scenarios, together with some of the political and ethical issues inherent in waste disposal, and presents future developments in waste form science and technology, including the impact of waste form research on nuclear fuel improvements. The chapter further reviews the future of HLW disposition around the world and the material presented in the chapter is focused on activities taking place in the US, but this focus is driven largely by the facts that: (i) the US has had the world's longest nuclear programs in terms of both power production and military applications and (ii) the US therefore has a large fraction of the worldwide HLW inventory.

The structural and electronic properties of Gd2(Ti1−yZry)2O7 (y=0–1) pyrochlores following a 2.0-MeV Au2+ ion-beam irradiation (∼5.0×1014Au2+∕cm2) have been investigated by Ti2p and O1s near-edge x-ray absorption fine structure (NEXAFS). The irradiation of Gd2(Ti1−yZry)2O7 leads to the phase transformation from the ordered pyrochlore structure (Fd3m) to the defect fluorite structure (Fm3m) regardless of Zr concentration. Irradiated Gd2(Ti1−yZry)2O7 with y⩽0.5 are amorphous, although significant short-range order is present. Contrasting to this behavior, compositions with y⩾0.75 retain crystallinity in the defect fluorite structure following irradiation. The local structures of Zr4+ in the irradiated Gd2(Ti1−yZry)2O7 with y⩾0.75 determined by NEXAFS are the same as in the cubic fluorite-structured yttria-stabilized zirconia (Y–ZrO2), thereby providing conclusive evidence for the phase transformation. The TiO6 octahedra present in Gd2(Ti1−yZry)2O7 are completely modified by ion-beam irradiation to TiOx polyhedra, and the Ti coordination is increased to eight with longer Ti–O bond distances. The similarity between cation sites and the degree of disorder in Gd2Zr2O7 facilitate the rearrangement and relaxation of Gd, Zr, and O ions∕defects. This inhibits amorphization during the ion-beam-induced phase transition to the radiation-resistant defect fluorite structure, which is in contrast to the ordered Gd2Ti2O7.