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Studtite is known to exist at the back-end of the nuclear fuel cycle as an intermediate phase formed in the reprocessing of spent nuclear fuel. In the thermal decomposition of studtite, an amorphous phase is obtained at calcination temperatures between 200 and 500 °C. This amorphous compound, referred to elsewhere in the literature as U2O7, has been characterised by analytical spectroscopic methods. The local structure of the amorphous compound has been found to contain uranyl bonding by X-ray absorption near edge (XANES), Fourier transform infrared and Raman spectroscopy. Changes in bond distances in the uranyl group are discussed with respect to studtite calcination temperature. The reaction of the amorphous compound with water to form metaschoepite is also discussed and compared with the structure of schoepite and metaschoepite by X-ray diffraction. A novel schematic reaction mechanism for the thermal decomposition of studtite is proposed.

Uranyl oxalate (UO2C2O4·xH2O) may exist at the back-end of the nuclear fuel cycle (NFC) as an intermediate in spent fuel reprocessing. The conditions used in aqueous reprocessing and thermal treatment can affect the physical and chemical properties of the material. Furthermore, trace impurities, such as Fe, may incorporate into the structure of these materials. In nuclear forensics, understanding relationships between processing variables aids in determination of provenance and processing history. In this study, the thermal decomposition of UO2C2O4·3H2O and phase analysis of its thermal products are examined. Their morphologies are discussed with respect to a matrix of solution processing conditions.

A digital colour image may be composed of hundreds of thousands of pixels, every pixel exhibiting a single colour. Each colour can be described as a combination of red, green and blue (RGB) components, of discrete values between 0-255. The RGB data contained within the pixels of an image could, therefore, be used to quantitatively establish the colour of nuclear material powders from digital images, particularly for use in nuclear forensics applications, where there is a need for consistent, objective analysis. This paper sets out a standard method for the photography and analysis of digital images of uranium oxide powder, for the objective quantification of colour by mean RGB values. Eight heat treated (up to 550°C) powder samples of studtite ([(UO2)(O2)(H2O)2]·2H2O) were photographed at room temperature and analysed by the RGB method. Hue, saturation and value of the coloured samples were obtained alongside mean RGB values, both of which were used to successfully determine the heating temperatures of unknown specimens of studtite.

Actinide chlorides, such as those arising from pyrochemical reprocessing operations can be problematic to immobilise, as the high chloride content often makes their solubilities in melts very low, and even in small quantities can seriously affect the properties of the waste-form. Rather than attempt to immobilise the chlorine, one potential approach is to utilise the chloride salt as a reaction medium from which the An(III) cations can be extracted and immobilised. To this end, lanthanide titanate pyrochlores have been prepared by molten salt synthesis in CaCl 2 :MgCl 2 , CaCl 2 :NaCl and MgCl 2 :NaCl eutectics. Single-phase pyrochlore is found to be formed at temperatures as low as 650 °C in the CaCl 2 :NaCl system, whereas in the CaCl 2 :MgCl 2 and MgCl 2 :NaCl eutectics reaction with Mg produces a magnesium titanate secondary phase. Compositions of Yb 2 Ti 2 O 7 doped with Sm 3+ as an actinide surrogate have been synthesised, and cold-pressing and sintering at 1500 °C yields fully dense pellets.

At the back-end of the nuclear fuel cycle, spent nuclear fuel (SNF) may be processed in solution to form a precipitate of uranyl peroxide tetrahydrate, also known as studtite ([(UO2)(O2)(H2O)2]•2H2O). The physical and chemical properties of studtite are influenced by the solution processing conditions employed, thus, the processing history of the precipitate may be elucidated from these characteristics. This is useful in the field of nuclear forensics, where the provenance of intercepted illicit nuclear material is of interest. Studtite is often calcined to ≥500 °C to form alpha uranium trioxide (α-UO3), an oxide prevalent in waste storage. In this study, the crystal structure of UO3 is analysed, with a discussion on the α/α’-UO3 C2mm symmetry. The morphological and structural signatures of both studtite and UO3 are investigated, with respect to a matrix of solution processing conditions and the thermal treatment of studtite.

The structure of amorphous U2O7 has been examined by extended x-ray absorption near edge spectroscopy (EXAFS) and modelled using density functional theory (DFT) simulations. A hybridised metastudtite-UO3 structure is proposed, consisting of peroxide bonds (O-Operoxo), uranyl units (U=Oyl) and U-O bonding. Experimental and simulated X-ray diffraction (XRD) is used to confirm the proposed structure.

The UK has a considerable stockpile of separated plutonium; a legacy of over 50 years of civilian nuclear programmes. This material has been considered both as an asset for future energy generation and a liability due to the proliferation threat. A proportion of the PuO2 stocks may be consumed by nuclear fission, in mixed oxide (MOx) or inert matrix (IMF) fuels but a quantity of waste PuO2 will remain which is unsuitable for fuel manufacture and will require immobilisation. A research program is currently underway to investigate the potential of various single phase ceramic formulations for the immobilisation of this waste PuO2 fraction. In this work a number of synthetic mineral systems have been considered including titanate, zirconate, phosphate and silicate based matrices. Although a wealth of information on plutonium disposition in some of the systems exists in the literature, the data is not always directly comparable which hinders comparison between different ceramic hosts. The crux of this research has been to compile a database of information on the proposed hosts to allow impartial comparison of the relative merits and shortcomings in each system.

Ceramics are promising potential hosts for the immobilization of actinide containing wastes. Work has been reported in the literature on multiphase systems, such as SYNROC [1], and on single phase systems such as pyrochlores [2] and zirconia [3], but assessment of the different waste-forms by direct comparison of literature data is not always easy due to the different processing and fabrication routes employed. In this study a potential range of different ceramic systems were investigated for plutonium disposition using the same processing scheme. Durable actinide containing minerals exist in nature and provided excellent target phases for the titanate, zirconate, silicate and phosphate based formulations examined here [4]. The Ce solid solution limits for each particular substitution mechanism were established and the processing parameters required to produce high quality ceramic specimens were optimised. Importantly, this was achieved within the constraints of a generic processing route suitable for fabrication of Pu bearing samples. (authors)

Substituting Pu(IV) onto the Ca site and using Fe(III) as a charge balancing species, a series of 6 compositions of 239Pu doped zirconolite (of general formula (Ca1−xPux)Zr(Ti2-2xFe2x)O7) have been fabricated, where x = 0.1, 0.2, 0.3, 0.4, 0.6 and 0.7. Sintered pellets from each of these compositions have been characterised by XRD, SEM, EDX and Raman techniques in order to analyse the change in microstructure and determine the polytype boundaries as the doping level increases, showing that the zirconolite is stable in the natural 2M polytype up until x = 0.3, then adopting the 3T polytype from x = 0.3 to x = 0.4, but from there onwards an increasing degree of phase separation occurs.