Image analysis techniques have been applied and shown to be a valuable tool in nuclear forensics analysis. The interlaboratory exercise reported here has tested quantitative and qualitative approaches for characterizing nuclear materials. Particle size, surface features and morphology descriptions were compared by four laboratories on a common image set generated by Scanning Electron Microscopy and Digital Light Microscopy. Quantitative analysis of the image sets through the Morphological Analysis for MAterials software highlighted the strength of image analysis, but also that the application of the software alone can introduce significant bias in the analysis. Qualitative morphology descriptions following the process outlined by Tamasi et al. (J Radioanal Nuclear Chem 307:1611–1619, 2015) were compared with a discussion on the robustness and reproducibility of the results. Future work should continue to focus on proficiency and standardization of image analysis through continued exercises within the extended nuclear forensics community.

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 U 2 O 7 , 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.

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 degrees C) powder samples of studtite ([(UO2)(O-2)(H2O)(2)]center dot 2H(2)O) 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. Crown Copyright (C) 2020 Published by Elsevier B.V. All rights reserved.

X-ray photoelectron spectroscopy (XPS) has applications in many fields ranging from development of thin films for semi-conductors to post failure analysis of organic coatings and structural adhesives. The current work expands on that versatility by applying XPS to the growing field of nuclear forensics. This was achieved by the synthesis and characterisation of several uranium compounds, predominantly in the hexavalent state associated with the nuclear fuel cycle, and by X-ray diffraction and Raman spectroscopy analysis prior to XPS. Spectral characteristics for each compound are discussed, and interpretations made through observations in the binding energy of the U4f region as well as secondary energy loss features such as shake up satellites. The interpretation of such features is related to the stoichiometry, oxidation state and bonding structure of a range of uranium compounds. As XPS is typically insensitive to structural (crystallographic) variations, a rationale is provided for the relationship between structural variations, as measured by Raman and X-ray diffraction and compared to the open literature, and the XPS satellite to parent peak intensity of uranium compounds, providing a novel and useful approach for uranium compound characterisation. In addition to the novel approach described, Wagner chemical state plots have also been generated to provide another comparison tool.

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.

Uranyl oxalate (UO 2 C 2 O 4 ·xH 2 O) 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 UO 2 C 2 O 4 ·3H 2 O and phase analysis of its thermal products are examined. Their morphologies are discussed with respect to a matrix of solution processing conditions.

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.

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 CaCl2:MgCl2, CaCl2:NaCl and MgCl2:NaCl eutectics. Single-phase pyrochlore is found to be formed at temperatures as low as 650°C in the CaCl2:NaCl system, whereas in the CaCl2:MgCl2 and MgCl2:NaCl eutectics reaction with Mg produces a magnesium titanate secondary phase. Compositions of Yb2Ti2O7 doped with Sm3+ as an actinide surrogate have been synthesised, and cold-pressing and sintering at 1500°C yields fully dense pellets.

A series of static lattice calculations were performed to determine the site selectivity of cations of differing size and valence when substituted onto the Ca sites of the calcium chlorosilicate (Ca3(SiO4)Cl2) lattice, a potential host phase for the immobilisation of halide-rich wastes arising from the pyrochemical reprocessing of plutonium. Atomic-scale simulations indicate that divalent cations are preferentially substituted onto the Ca1 site, whilst tri- and tetravalent cations are preferentially hosted on the Ca2 site, with the Ca1 site favoured for forming the vacancies necessary to charge-balance the lattice as a whole. Multi-defect calculations reveal that the site selectivity of the dopant cations is dependent on their ionic radii; as the ionic radii of the divalent cations increase, substitution onto the preferred site becomes more and more strongly favoured, whereas the inverse is true of the trivalent cations.

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.

Tailored ceramics are promising potential hosts for the immobilization of actinides [1]. In this study the potential of a range of different ceramic systems were investigated for the incorporation of waste plutonium, using Ce as a Pu surrogate. Durable actinide containing minerals exist in nature and provided excellent target phases for the titanate, zirconate, silicate and phosphate based formulations examined here [2]. The Ce single phase solid solution limits for each system 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.