An analysis of sintered uranium dioxide has been conducted using a hyperspectral camera sensitive to short-wave infrared wavelengths in the range 949–2472 nm. Three groups of sintered UO2 nuclear fuel pellets were prepared and analysed, with stable sub-group surrogates introduced at the preparation stage to emulate the presence of fission product elements. Results show a clear, consistent, and reproducible spectral response across the pellet groups for pure UO2. Furthermore, the addition of fission product elements is observed to affect the shortwave infrared response, causing an overall flattening of the spectra. We have shown that this spectral change is correlated significantly with the presence of lanthanides in the fuel matrix. This result could have important potential in post-irradiation examination for quantifying nuclear fuel burn-up and radiotoxicity at discharge, as the hyperspectral imaging setup allows multiple (> 20) samples to be analysed in a single image, captured in under 30 s.

Measurements of refractory and volatile components in soils from Northern England confirm 244Pu/239Pu consistent with the global average but refractory 240Pu/239Pu significantly higher but consistent with Chernobyl fallout, i.e., 0.390 ± 0.006. Refractory formation suggests temperatures > 3000°C, consistent with atmospheric injection by nuclear-driven explosions rather than by steam or hydrogen. Volatile 240Pu/239Pu is consistent with the global average, 0.181 ± 0.002, and hence lower-temperature formation in the chemical explosions and subsequent fire. This hypothesis is supported by 239Pu fission and capture cross-section enhancement due to the S-wave resonance at 0.3 eV. A further calibration with a 240Pu standard is recommended to substantiate these observations.

Disposal MOX (dMOX), a sintered mixed U, Pu, oxide containing a neutron poison, such as Gd, is currently under consideration for immobilisation of the UK’s civil plutonium inventory. Prior to deployment, it is important to understand how material homogeneity impacts overall behavior of the wasteform. Therefore, while using Ce as a non-active Pu surrogate, this study outlines a methodology for the production of a homogenous dMOX precursor powder to be subsequently sintered to a homogenous dMOX pellet. The powder was produced via reverse-strike oxalate co-precipitation of a Ce(III), Gd(III), and photochemically conditioned U(IV) precursor solution, which was then calcined to a mixed oxide in a 5%H 2 /95% N 2 reaction atmosphere. SEM–EDX and PXRD analysis revealed the oxalate and mixed oxide to be fully homogeneous, with no evidence of segregated phases. TGA decomposition found the oxalate to decompose in several distinct stages, agreeing with established literature, resulting in a fully homogenous mixed oxide powder. Graphical abstract

Recycling nuclear fuel to recover materials such as uranium and plutonium involves high temperature processes to treat the resulting highly active (HA) waste. This waste contains ruthenium, an important fission product due to its tendency to form volatile compounds and the presence of the radioactive isotope 106Ru. These properties, combined with the elevated temperatures during HA waste treatment, necessitate a deeper understanding of ruthenium volatilisation mechanisms. Key ruthenium species present or potentially formed in HA waste include nitro and nitrosyl complexes, and ruthenium dioxide (RuO2). This study focuses on the behaviour of RuO2 at high temperatures, due to its relevance to the vitrification process used to immobilise HA waste. Preliminary thermogravimetric analysis of commercially available RuO2 revealed a dehydration process under N2, O2, air atmospheres, and volatilisation under oxidising conditions. Post-heating analyses using scanning electron microscopy and X-ray diffraction identified a concurrent, thermally induced recrystallisation process, most prominent under non-oxidising conditions.

Uranium mononitride (UN) is a potential versatile fuel for use in both thermal and fast spectrum reactors. Knowledge of the thermodynamic properties of actinides and fission products in spent UN fuel is required to understand their properties such as phase stability and retention during long term storage and disposal. The present study reviews thermodynamic data and calculates the free energies of formation (Delta(f)G(m)) of the nitrides and other components formed in the spent fuel to predict the actinide and fission product behaviour. An End of Life (EoL) spent fuel inventory was calculated for a high burnup UN fuel (60 MWd kg(-1)) using the FISPIN fuel inventory code. The spent fuel consisted predominantly of a solid solution of nitrides (U, An, Ln, Y, Zr, Nb)N forming a single homogeneous and stable phase as expected from the Delta(f)G(m) variation with T of its components or of the fuel as calculated for an ideal solid solution. Since delta Delta(f)G(m)/delta T > 0 for the reduction: MN double left right arrow M + 1/2 N-2, dissociation is consequently more likely at high temperature. Other fission products are expected to be divalent (Ba, Sr), monovalent (Cs, Rb) or non-valent (Tc, Ru, Rh, Pd) as well as noble gases (Xe, Kr), and halides (I, Br) which may form nano-precipitates (e.g. with metal ions). The behaviour of Mo is more complex. At low fuel temperature (<1100 K) it may form nitride precipitates while at higher temperature (>1100 K) MoN and MoN0.5 decompose to Mo metal. The stoichiometry of the spent fuel is related to the burn-up and the temperature of the fuel during operation. It is also shown to be dependent on the molybdenum species generated in pile (metal or nitride precipitate type). Thermodynamic calculations for potential Pellet Clad Interaction (PCI) showed that with Zr alloys interactions are expected while with stainless steel clad no reaction between the steel components (Fe, Ni, and Cr) and UN is expected. Finally, the free energy evaluation for UN hydrolysis shows that UN reaction with water is spontaneous.

Tritium (3H) is one of the hardest isotopes to detect by most traditional radiometric means due to the low energy of the & beta;- emission, (& beta;-MEAN 5.67 keV, & beta;-MAX 18.59 keV). The high mobility of the isotope in groundwater environments and subsequent entry into the food chain constitutes a radiation safety risk justifying assessment. Accordingly, there is a need to measure 3H accurately and efficiently, often in low concentrations, both in laboratory settings and on-line flow-cells for potential in situ measurement requirements. This review covers technologies developed to assess aqueous tritium-containing samples. Of the techniques reviewed, liquid scintillation counting (LSC) is the best performing means of aqueous 3H detection with a minimal detectable activity of 6 x 10-4 Bq L-1 for a 195-min counting time. LSC is also established as the industry standard and is the basis of the first, commercially-available, real-time 3H detection system. This review also covers the other means described in literature for the detection of tritium in aqueous samples, including the use of plastic and inorganic scintillators, imaging plates, both in off-line and on-line modes of operation. Whilst most of these techniques lag LSC in terms of technological maturity, several offer detection sensitivities that could rival LSC, without the need for the sample preparation and waste generation associated with LSC, and providing real-time in situ measurements.

In fuel reprocessing, product finishing is the conversion of aqueous metal nitrates into solid forms that can either be re-used in new fuels or safely interim stored and so is the key step at the interface between reprocessing and fuel manufacturing processes. Conversion processes were originally developed for the fabrication of oxide fuels and have typically involved the generation of UO3 (or U3O8) and PuO2 powders as separate products. However, whilst this is an industrial proven process, research and development of mixed oxide (MOX) fuels and minor actinide targets by advanced reprocessing routes is also underway. It is envisaged that advanced recycle processes will allow the multi-recycling of plutonium and the transmutation of minor actinides into shorter lived isotopes to enhance sustainability of the nuclear fuel cycle and fully utilise fissionable material recoverable from spent fuels whilst improving a number of key features over current conversion processes including: product conversion efficiency, higher throughput, flexibility in product specification, proliferation resistance and a reduction in the number of waste streams produced. Internationally, research programmes to examine future options for advanced fuel cycles are focusing on the development of advanced reprocessing flowsheets for future actinide recycling. It is anticipated that these separation processes will produce a range of mixed transuranic (TRU) actinide nitrate products rather than the separated pure plutonium stream produced in current reprocessing plants. The easiest assumption is that these nitrate products will be converted to oxides by the oxalate co-precipitation route. However, this has certain limitations. The focus of this review is to identify any alternative nitrate to oxide conversion processes which have been applied to mixed oxides and evaluate their suitability for MOX production. A variety of factors including process complexity, technical maturity, effluent treatment/recycling and scale up into an industrial process will also be considered.

This article presents a review of the behaviour of uranium nitride (UN) fuels during thermal reactor fuel-based reprocessing. UN is one of the leading candidate materials for use in Advanced Technology Fuels (ATFs) due to it having several superior properties over UO2 or MOx based fuels, including: good thermal conductivity; high melting point; low thermal expansion; high fissile density; and good compatibility with metallic or gaseous coolants currently under consideration for Generation IV reactors. However, it also has a number of drawbacks in fuel-based applications including expected higher fabrication costs and oxidative instability in water. The former is mainly due to the expected need to enrich the nitrogen component in 15N to increase the neutron economy and avoid formation of 14C from the n,p reaction of 14N. The latter property may be advantageous for UN's post-irradiation reprocessing but must be addressed if UN is to be deployed safely in the near term in existing and under-construction LWR reactors. One means by which UN's hydrolytic instability may be addressed is by the introduction of suitable protective dopants such as Al, Cr or Zr to greatly increase oxidative resistance. However, this may also impact on the reprocessability of spent UN post-irradiation. Thus, in this review, we will focus on the management and recycle options for UN fuels after irradiation. Whilst 15N enrichment is an important potential economic driver of the reprocessing of spent UN, it is not the primary concern of this review. Rather, we focus on issues that may arise during the dissolution and head-end treatment of UN fuels, and the subsequent expected behaviour of the dissolved fuel in existing and proposed solvent extraction processes. Where they exist, similarities with the reprocessing of thermal UO2 fuels will be highlighted, as will the effect of protective dopant materials on the chemistry of UN reprocessing.

An Advanced PUREX process for the recycling of spent nuclear fuel is currently under active development in the UK. Its key aims are to avoid pure separated plutonium at all stages of the process to enhance the level of proliferation resistance, and to achieve a single cycle flowsheet that has a smaller plant footprint with consequent decreases in the capital cost and secondary wastes generated. Addressing these aims, a significant feature of the process is the co-treatment of U and Pu and thus the in situ co-conversion of mixed actinide metal nitrate solutions into oxide powders, suitable for the fabrication of new mixed metal oxide (MOx) fuel. The baseline industrial process for plutonium recovery is by oxalate precipitation; however, in order to quantitatively recover both U and Pu the uranium must be in the U(IV) oxidation state due to the high solubility of U(VI) oxalate. The first stage of this co-conversion, the development of which is reported on here, is the rapid, clean photochemical co-reduction of a mixed U(VI)/Pu(IV) nitrate stream to U(IV)/Pu(III). Here we describe a study of the reduction of U(VI) in preparation for mixed U(VI)/Pu(IV) reduction trials. Exploiting the photochemistry of U, we demonstrate the convenient and efficient photo-excitation and chemical reduction of U(VI) upon exposure to 407 nm wavelength light in the presence of alcohol-based reductants. Using a purpose built laboratory-scale photochemical reactor, U(VI) solutions of up to process-relevant concentrations of 630 mmol/dm(3) (150 g/l) U have been successfully converted to a U(IV) product, achieving a conversion efficiency of similar to 98% within 1.5-15 min when using propan-2-ol as a sacrificial reductant. Modelling of the dependence of the rate of U(IV) generation on initial U(VI) concentration reveals the importance of light penetration depth and effective solution mixing in determining the efficiency of the photochemical process at high U-loadings. It also reveals that the photoreduction of U(VI) to U(IV) occurs by two sequential 1-electron reductions: (i) the photochemically driven reduction of UO22+ to UO2+ by propan-2-ol, which itself is oxidised to form an alpha-hydroxyalkyl radical, immediately followed by (ii) a second chemical reduction of UO22+ to UO2+ and/or UO2+ to U-4(+) by the so-formed radical. With the addition of a nitrous acid scavenger to prevent re-oxidation of the photochemically generated U(IV), a stable product is maintained indefinitely, and the solution is suitable for subsequent oxalate co-precipitation as part of a MOx fuel fabrication process.

This paper reports a detailed chemical and materials characterization study of novel Simulated Spent Nuclear Fuels (SIMFuel) which replicate the chemical and microstructural state of Spent Nuclear Fuel (SNF) discharged from a UK Advanced Gas-cooled Reactor (AGR). Recent advances in the analysis of pure UO2 samples by XPS, micro-Raman spectroscopy and XRD, have been assessed for deployment in the characterisation of SIMFuels generally and these AGR SIMFuels in particular. All three analytical methods reveal the extent that the UO2 bulk matrix is defected in the SIMFuels by the presence of lanthanide dopants. XPS, by inspection of the U4f peaks and their satellites, indicates the presence of U(V) in the UO2 matrix as a means to charge compensate for the incorporation of Ln(III) states and presence of a slight hyperstoichiometry in the UO2 solid solution. This was corroborated by detailed Raman analysis of the UO2 matrix of the SIMFuels - wherein the (Ba,Sr)ZrO3 and metallic particle phases, that simulate the precipitated grey phases and epsilon-particles formed in real spent fuel, were avoided - which indicated the presence of both interstitial oxygens and oxygen vacancies, the latter as a parallel means of charge compensation for the presence of Ln(III). Further confirmation was provided by XRD measurements through observation of a lattice parameter contraction arising from the presence of the smaller U(V) ion in the U(IV)O-2 matrix. XPS and Raman allow for the reporting of O/U ratios in the SIMFuels. These XPS and Raman-derived O/U ratios are in good agreement, indicating: (i) that the SIMFuels are near stoichiometric/slightly hyperstoichiometric, concurring with previous analogous studies on real and simulated light water reactor fuels; and (ii) that either technique may find application in the analysis of real fuel analysis, both pre-and post-irradiation.(c) 2022 The Authors. Published by Elsevier B.V.This is an open access article under the CC BY license (http://creativecommons.org/licenses/by/4.0/)

Electrochemical corrosion of lower-activity spent fuel simulants, SIMFUELs, has been investigated in support of on-going analogous studies of a real AGR spent nuclear fuel electrode. Two electrode coupling methods have been investigated, to study the coupling of a sample of 25 GWd/tU SIMFUEL pellet to a sample of 20/25/Nb AGR cladding to replicate the real AGR fuel electrode setup. Method-1 involves coupling of individual standalone electrodes, while Method-2 involves coupling of the cladding and SIMFUEL within the same electrode. For both coupling methods, open circuit potentials and linear sweep voltammetry measurements were conducted in electrolyte solutions containing 30 μmol/m 3 NaCl, dosed with NaOH to pH 8, 11.4 and 12.5. Resultant data were compared with single-component studies of SIMFUEL and cladding electrodes. Method-1 presents inconsistent electrochemical behaviour which we attribute to pitting/crevice corrosion effects on the cladding, whereas Method-2 shows a consistent behaviour that most resembles that of the SIMFUEL component. Graphical abstract

Neptunium has been previously shown to present challenges within a used nuclear fuel reprocessing scheme due to its tendency to exist in the (IV), (V), and (VI) oxidation states simultaneously. In order to control this neptunium speciation, and informed by relevant work in the literature, we are currently engaged in a study of nitric/nitrous acid redox chemistry with Np(V) and Np(VI). To minimize radiological exposure risks, we are also exploring the validity of using vanadium as an analogue for the study of the kinetics of the Np(VI)/Np(V) reduction by nitrous acid. The kinetics of the reduction of vanadium(V) by nitrous acid in solutions of nitric acid was investigated spectrophotometrically by the method of initial rates. Orders of reaction with respect to V(V), and HNO2 were previously found to be 0.90, and 1.25 respectively, in reasonable agreement with the analogous reaction orders for the reduction of Np(VI) by nitrous acid previously reported by Precek and Paulenova - suggesting that, for this particular reduction, V(V) can serve as a good kinetic analogue for Np(VI). Within this study orders of reaction with respect to [H +], [NO3 -], [SO4 2-], and [ClO4 -] have also been found to be 0.1, -0.2, 0.1, and 0 respectively. Preliminary experiments have also been conducted on the reduction of V(V) by the known reducing agent NO which has hitherto not been considered in the reduction of Np(VI) to Np(V).

Due to radiation induced segregation (RIS) that occurs during fuel use in-reactor, portions of spent fuel cladding are predicted to have enhanced corrosion susceptibility during wet interim storage. The National Nuclear Laboratory UK have produced thermally sensitized analogues (20/25/Nb SS and 304H SS) for RIS-affected cladding to allow study of this effect without radiological risk. Here, we present a study of the corrosion properties of these simulants in order to gain insight into the behavior of real cladding under conditions relevant to wet storage. First, the degree of sensitization of the RIS-affected cladding analogues was verified. In these tests significant susceptibility to localized corrosion was observed. The effect of such vulnerability was analyzed under alkali corrosion conditions. Under open circuit conditions, XPS data indicates that the protective layer formed on the surface of thermally aged SS consists predominately of Fe and Cr oxides/hydroxides, with Fe oxide/hydroxide being the more dominant of the two especially in the case of 304H SS. Despite increased vulnerability and the decrease in Cr oxide/hydroxide in the surface layer, both heat treated samples were found to protected under pond water conditions and it would be expected that real RIS-affected cladding would be similarly passive under the same conditions. Copyright © GLOBAL 2019 - International Nuclear Fuel Cycle Conference and TOP FUEL 2019 - Light Water Reactor Fuel Performance Conference.All rights reserved.

The escape of radionuclides from underground spent nuclear fuel disposal facilities will likely result from anoxic dissolution of spent nuclear fuel by intruding groundwater. Anoxic dissolution of various forms of uranium dioxide (UO2), namely bulk pellet, powder and thin film, has been investigated. Long-duration static batch dissolution experiments were designed to investigate the release of uranium ions in deionized water and any surface chemistry that may occur on the UO2 surface. The dissolved uranium concentration for anoxic dissolution of nearly stoichiometric UO2 was found to be of the order of 10(-9) mol/l for the three different sample types. Further, clusters (similar to 500 nm) of homogenous uranium-containing precipitates of similar to 20-100 nm grains were observed in thin film dissolution experiments. Such a low solubility of UO2 across sample types and the observation of secondary phases in deionized water suggest that anoxic UO2 dissolution does not only occur through a U(IV)((solid)) to U(VI)((aqueous)) process. Thus, we propose that dissolution of uranium under anoxic repository conditions may also proceed via U(IV)((solid)) to U(IV)((aqueous)), with subsequent U(IV) ((precipitates)) in a less defective form. Quantitative analysis of surface-sensitive EBSD diffractograms was conducted to elucidate lattice-mismatch induced cracks observed in UO2 thin film studies. Variable temperature anoxic dissolution was conducted, and no increased uranium concentration was observed in elevated temperatures. (C) 2019 Elsevier B.V. All rights reserved.

Simple hydroxamic acids such as acteohydroxamic acid (AHA) have been identified as suitable reagents for the control of Pu and Np in advanced separation processes for nuclear fuel reprocessing such as the Advanced PUREX or UREX based recycle processes, due to their ability to strip the tetravalent form of Pu and Np from tri-butyl phosphate into nitric acid. However, both free and metal bound hydroxamates are known to undergo acid catalysed hydrolysis at low pH, the kinetics of which must be characterised before implementation of PUREX/UREX based reprocessing flowsheets. In support of this implementation, a comprehensive thermodynamic and kinetic model that describes both the complex speciation and hydrolysis of AHA in the presence of Np(iv) has been developed. The model has two unique features: (i) in the case of the hydrolysis reaction kinetics, the model includes the hydrolysis of not only free AHA but also both the mono- and bishydroxamato-Np(iv) complexes; (ii) for the associated speciation calculations, the model explicitly includes the ionic strength dependence of not only the mono- and bishydroxamato-Np(iv) complexes but also the mono- and bisnitrato neptunium(iv) and monohydroxoneptunium(iv) complexes. For the latter three species, respective SIT coefficients of Δε1,NO3 = -0.13 ± 0.03 kg mol-1, , Δε2,NO3 = -0.37 ± 0.13 kg mol-1, Δε1,OH = -0.36 kg mol-1 and log10 K01,OH = -1.23 were also determined. Using experimental data from a series of kinetic studies on the Np(iv)-AHA system, this model has been used to determine the rate constants for hydrolysis of mono- and bis-acetohydroxamatoneptunium(iv) at 25 °C for the first time. These were found to be 3.5 × 10-5 ± 2.5 × 10-5 dm3 mol-1 s-1 and 1.9 × 10-3 ± 1.3 × 10-3 dm3 mol-1 s-1, respectively. Comparison of these values with the rate constants for hydrolysis of free AHA indicates that complexation of AHA with Np(iv) increases the rate of hydroxamate hydrolysis - an observation that we attribute to the electron withdrawing effect of the metal centre within the Np(iv)-AHA complex increasing the susceptibility of the AHA carbonyl carbon to nucleophilic attack, the accepted first step in its mechanism of hydrolysis.