Over 60 samples of plutonium dioxide (PuO2) powders of varying provenance have been exposed to humid atmospheres and the hydrogen (H2) generation rates measured by gas chromatography. The effects of relative humidity (RH), specific surface area (SSA), plutonium isotopic composition (absorbed dose), and overlying atmosphere have been investigated for ‘as received’ PuO2 powders from the United Kingdom’s Thorp and Magnox reprocessing plants, high surface area powder produced in the laboratory and Magnox PuO2 that was recalcined at 800, 900 or 950°C. Hydrogen generation was shown to be susceptible to subtle influences with the most consistent results observed at 95% RH. However, it was shown that the measured (net) hydrogen decreases with decreasing RH and with atmosphere in the order: Air > argon > nitrogen. There was no clear effect of SSA, apart from with the highest SSA samples (∼40 m2.g−1) but these powders also have very different morphology, porosity, and carbon content to the rest. The results presented substantially enhance the growing body of literature on the factors that determine hydrogen and gas generation from PuO2 that has significant implications for long term safe storage of plutonium globally.

Over 80 heat treatment experiments have been made on samples of chloride-contaminated plutonium dioxide retrieved from two packages in storage at Sellafield. These packages dated from 1974 and 1980 and were produced in a batch process by conversion of plutonium oxalate in a furnace at around 550 °C. The storage package contained a poly(vinyl chloride) (PVC) bag between the screw top inner and outer metal cans. Degradation of the PVC has led to adsorption of hydrogen chloride together with other atmospheric gases onto the PuO2 surface. Analysis by caustic leaching and ion chromatography gave chloride contents of ∼2000 to >5000 ppm Cl (i.e., μgCl g-1 of the original sample). Although there are some subtle differences, in general, there is surprisingly good agreement in results from heat treatment experiments for all the samples from both cans. Mass loss on heating (LOH) plateaus at nearly 3 wt % above 700 °C, although samples that were long stored under an air atmosphere or preexposed to 95% relative humidity atmospheres, gave higher LOH up to ∼4 wt %. The majority of the mass loss is due to adsorbed water and other atmospheric gases rather than chloride. Heating volatilizes chloride only above ∼400 °C implying that simple physisorption of HCl is not the main cause of contamination. Interestingly, above 700 °C, >100% of the initial leachable chloride can be volatilized. Surface (leachable) chloride decreases quickly with heat treatment temperatures up to ∼600 °C but only slowly above this temperature. Storage in air atmosphere post-heat treatment apparently leads to a reequilibration as leachable chloride increases. The presence of a "nonleachable" form of chloride was thus inferred and subsequently confirmed in PuO2 samples (pre- and post-heat treatment) that were fully dissolved and analyzed for the total chloride inventory. Reheating samples in either air or argon at temperatures up to the first heat treatment temperature did not volatilize significant amounts of additional chloride. With regard to a thermal stabilization process, heat treatment in flowing air at 800 °C with cooling and packaging under dry argon appears optimal, particularly, if thinner powder beds can be maintained. From electron microscopy, heat treatment appeared to have the most effect on degrading the square platelet particles compared to those with the trapezoidal morphology.

The yield of molecular hydrogen, as a function of nitric acid concentration, from the α-radiolysis of aerated nitric acid and its mixtures with sulfuric acid containing plutonium or americium has been investigated. Comparison of experimental measurements with predictions of a Monte Carlo radiation track chemistry model shows that, in addition to scavenging of the hydrated electron, its precursor, and the hydrogen atom, the quenching of excited state water is important in controlling the yield of molecular hydrogen. In addition, increases in solution acidity cause a significant change in the track reactions, which can be explained as resulting from scavenging of eaq- by Haq+ to form H•. Although plutonium has been shown to be an effective scavenger of precursors of molecular hydrogen below 0.1 mol dm-3 nitrate, previously reported effects of plutonium on G(H2)α between 1 and 10 mol dm-3 nitric acid were not reproduced. Modeling results suggest that plutonium is unlikely to effectively compete with nitrate ions in scavenging the precursors of molecular hydrogen at higher nitric acid concentrations, and this was confirmed by comparing molecular hydrogen yields from plutonium solutions with those from americium solutions. Finally, comparison between radionuclide, ion accelerator experiments, and model predictions leads to the conclusion that the high dose rate of accelerator studies does not significantly affect the measured molecular hydrogen yield. These reactions provide insight into the important processes for liquors common in the reprocessing of spent nuclear fuel and the storage of highly radioactive liquid waste prior to vitrification.

To mimic the reactions that have occurred between plutonium dioxide (PuO2) and degraded polyvinyl chloride during prolonged storage, batches of PuO(2 )powders with varying specific surface area (4-38 m(2) g(-1)) were exposed to dry hydrogen chloride gas forming a surface-sorbed exchangeable chloride species that was readily leached by dilute caustic solution. One batch was also humidified under a 95% relative humidity atmosphere. Heat treatment at temperatures between 100 and 950 degrees C steadily reduced the leachable chloride content at least when the initial leachable chloride surface coverage was calculated to be > 0.5 monolayer (ML). Variations between samples are attributed to differences in surface area as well as production route and period of interim storage in air or humidified atmospheres. The data are consistent with surface species and reactions proposed in the literature for HCl adsorption on to TiO2. This indicates that it is interactions between hydrogen chloride/chloride and water/hydroxyl species on the PuO2 surface that are key to understanding the behaviour of these materials when heated in a furnace. These data will inform the design of a future thermal stabilisation process for chloride-contaminated PuO2. (C) 2018 Elsevier B.V. All rights reserved.

One of the most successful uses of nuclear energy, other than nuclear fission reactors, is the use of radioisotopes to provide a simple source for heat and electrical power in space applications. The alpha decay heat from suitable radioisotopes can be harnessed to heat instrumentation and generate electricity through thermoelectric generators. Within Europe, the feasibility of using americium (241Am) in such applications is being considered. Part of this study is to develop a way to access a stock of suitable material and the UK stockpile of separated civil plutonium dioxide, generated through reprocessing operations, is a rich source of americium. This paper summarises 5 years of development work performed to establish a way to separate americium from the plutonium dioxide in a safe and cost effective way, generating a high purity product with high recovery efficiency.

The yield of HNO2, as a function of absorbed dose and HNO3 concentration, from the α-radiolysis of aerated HNO3 solutions containing plutonium or americium has been investigated. There are significant differences in the yields measured from solutions of the two different radionuclides. For 0.1 mol dm-3 HNO3 solutions, the radiolytic yield of HNO2 produced by americium α-decay is below the detection limit, whereas for plutonium α-decay the yield is considerably greater than that found previously for γ-radiolysis. The differences between the solutions of the two radionuclides are a consequence of redox reactions involving plutonium and the products of aqueous HNO3 radiolysis, in particular H2O2 and HNO2 and its precursors. This radiation chemical behavior is HNO3 concentration dependent with the differences between plutonium and americium α-radiolysis decreasing with increasing HNO3 concentration. This change may be interpreted as a combination of α-radiolysis direct effects and acidity influencing the plutonium oxidation state distribution, which in turn affects the radiation chemistry of the system.

Electrical power sources used in outer planet missions are a key enabling technology for data acquisition and communications. State–of-the-art power sources generate electricity from alpha decay of 238Pu via thermoelectric conversion. However, production of 238Pu requires specialist facilities including a nuclear reactor, a source of 237Np for target irradiation and hotcells to chemically separate neptunium and plutonium within the irradiated targets. These specialist facilities are expensive to build and operate, so naturally, a more economical alternative is attractive to the space industry. Within Europe 241Am is considered a promising alternative heat source for radioisotope thermoelectric generators (RTGs) and radioisotope heating units (RHUs) since, as a daughter product of 241Pu decay, 241Am exists in 1000 kgs quantities within the UK civil plutonium stockpile. A chemical separation process is required to extract the 241Am in a pure form and this paper describes the AMPPEX process (Americium and Plutonium Purification by Extraction), successfully developed over the past five years to isolate 241Am in high yield (> 99%) and to a high purity (> 99%). The process starts by dissolving plutonium dioxide in nitric acid with the aid of a silver(II) catalyst, which is generated electrochemically. The solution is then conditioned and fed to a PUREX type solvent extraction process, where the plutonium is separated from the americium and silver. The plutonium is converted back to plutonium dioxide and the americium is fed forward to a second solvent extraction step. Here the americium is selectively extracted leaving the silver in the aqueous phase. The americium is stripped from the solvent and recovered from solution as americium oxalate, which is calcined to give americium dioxide as the final product. This paper describes the development of the AMPPEX process over a series of six solvent extraction separation trials using centrifugal contactors. The technical underpinning reported here has allowed the design of a facility capable of producing 25-50 g of americium per day.

Spent Fuel Management at Sellafield includes the reprocessing of spent nuclear fuel from stations across the UK and also the reprocessing of spent fuel from overseas customers. At Sellafield, methods have been developed for the processing of high level wastes, including highly active liquors (HAL), which is a result of reprocessing irradiated nuclear fuel. This Highly Active (HA) raffinate / waste stream is concentrated in evaporators and storage tanks in the Highly Active Liquor Evaporation & Storage (HALES) facility before feeding to the Waste Vitrification Plant (WVP). Here, the resultant HAL feed is calcined and combined with glass before pouring into containers to produce an immobilised HA wasteform. Ruthenium is a fission product possessed of two relatively long lived stable isotopes: Ru-103 (t1/2 = 39.8 days) and Ru-106 (t1/2= 1 year). Both isotopes form part of the inventory of HA waste raffinate during reprocessing of spent fuel. Volatilisation of fission products in nuclear waste generally occurs at high temperature – apart from ruthenium where volatilisation occurs at the lower temperature stages of the vitrification process. Given its volatile nature and high specific radioactivity, ruthenium presents a strong challenge to the nuclear industry in effectively managing its abatement. Part of the challenge is to fully understand the highly complex solution chemistry under conditions relevant to HA waste streams and associated abatement systems. Experimental work within the National Nuclear Laboratory (NNL), UK has demonstrated that the presence of oxidising metal ions in HA waste (e.g. Ce(IV)) can enhance the volatility of ruthenium through a chemical conversion of Ru(III) species to what is assumed to be RuO4. A better understanding of these species, their electrochemical processes and reaction kinetics is required to underpin the empirical evidence gathered to date, in particular to develop gravimetric, electrochemical and spectroscopic analytical methods that will improve the understanding of ruthenium speciation in high nitric acid environments, establish the kinetics of inter-conversion between ruthenium species and establish the mechanism by which metal ions such as Ce(IV) may oxidise ruthenium. We have studied the baseline electrochemical behaviour of ruthenium and present here for the first time the intrinsic thermodynamics of uncomplexed Ru(III) using electrochemical methods to determine as bought RuCl3 to be a mixture of Ru(III) and Ru(IV). We have achieved this via cyclic voltammetry of RuCl3 in HClO4 which yielded two peaks, namely at 0.55 and 0.95 V. A similar peak has been seen by Maya [1] at 0.9 V, reported to be tetramer oxidation. Analysis of the solution using UV-vis shows as bought RuCl3 to be a mixture of Ru(III) and Ru(IV). Subsequently a successful method to electroreduce the mixture to a pure solution of Ru(III) was developed. Cyclic voltammetry of the electroreduced Ru(III) shows an absence of the previously seen peak at 0.55 V, suggesting this peak was associated to the oxidation of Ru(IV), whilst the previously observed peak at 0.95 V had been shifted to ~ 0.8 V suggesting that this peak was not in fact an observation of tetramer oxidation, as proposed by Maya. The latter peak was further interrogated using a scan rate dependence study, which shows the peak potential increasing as scan rate increases showing we have an irreversible, kinetically slow process occurring. A linear relationship between the square root of the scan rate against peak current indicates the oxidation is occurring in solution. The n number of this transition was experimentally determined to be 0.33 and we have confirmed a diffusion coefficient of 4x10-10 m2 s-1 for RuCl3. This shows we have a trimer in solution which is oxidised from: 3 Ru(III) to Ru(III)-Ru(IV)-Ru(III). Future work will focus on the ruthenium nitrosyl species found in spent fuel to understand what effect NO complexation has on the system with the aim of comparing the uncomplexed ruthenium system with the complexed ruthenium system. [1] Maya, L., J. Inorg. Nucl. Chem., 41, 1978, 67

Ruthenium is a fission product possessed of two relatively long lived isotopes, 103Ru and 106Ru, both of which form part of the Highly Active (HA) waste raffinate during spent nuclear fuel reprocessing. During reprocessing ruthenium, which may be in the form of the RuNO3+ complex, encounters temperatures conducive to volatilization. Due rutheium’s high specific radioactivity it is important to understand the mechanism by which volatilisation occurs. Here we use combined CV, RDE and electrochemical microgravimetry experiments in a study of the the RuCl3 system for the first time. We do this in the interest of establishing NO-free Ru(III) baseline behaviour so as to support future studies on NO complexed ruthenium. Using wide aqueous solvent window carbon electrodes we have observed discrete oxidations to a solution phase Ru(III)–Ru(IV)–Ru(III) trimer, to solid RuO2 and volatile RuO4. We have also observed and assigned discrete reductions of solid RuO2 back to Ru(III) and Ru(III) reduction to ruthenium metal.

A flowsheet for a novel GANEX (Grouped ActiNide EXtraction) process has been tested in a spiked flowsheet trial in a 32 stage plutonium-active centrifugal contactor rig with a simulant feed that contained 10 g/L plutonium as well as some fission products and other transuranic actinides. The solvent system used was a combination of 0.2 mol/L N,N,N’,N’-tetraoctyl diglycolamide (TODGA) and 0.5 mol/L N,N’-(dimethyl-N,N’-dioctylhexylethoxy-malonamide (DMDOHEMA) in a kerosene diluent that co-extracted actinides and lanthanides. Actinides were subsequently selectively co-stripped away from the lanthanides using a sulphonated and, therefore, hydrophilic bis-triazinyl pyridine (BTP) complexant in conjunction with acetohydroxamic acid (AHA). Plutonium and americium recoveries were high with decontamination factors across the strip contactors of ˜14,000 and ˜390, respectively. However, approximately 30% of neptunium was lost to the aqueous raffinate which was due to recycling within the first extract-scrub section causing a large build-up of neptunium. Some accumulation of strontium was also observed but in this case it was fully directed to the raffinate stream. In the stripping section, a small fraction of europium (taken as a model lanthanide ion), ca. 7%, was found in the actinide product stream. Modelling of selected data using the PAREX code has shown that even with a relatively simplistic treatment, reasonable agreement between modelling and experiment can be obtained, giving confidence in the use of modelling to refine the GANEX flowsheet design prior to further testing with irradiated fast reactor fuel.

To support the development of an Advanced PUREX process that fully recovers neptunium with uranium and plutonium for recycle, an enhanced understanding of neptunium behaviour in nitric acid and tributyl phosphate phases is essential. The most important reaction affecting neptunium extraction in the PUREX process is the oxidation of Np(V) by nitric acid. This paper reports new data on Np(V) oxidation in nitric acid under conditions relevant to the development of a flowsheet for the primary extraction/scrub contactor in an Advanced PUREX process.

Within the UK there are a number of nuclear legacy fuel storage ponds and silos that contain substantial volumes of corroding spent Magnox fuel pieces. The fuel and wastes within these ponds, including highly radioactive sludges, must be retrieved and processed during decommissioning. Sludges and other intermediate level wastes will then be encapsulated in a wasteform suitable for storage and disposal, whilst residual activity must be removed from pond liquors and process effluents prior to any authorised discharges. Understanding the nature and behaviour of the radionuclides in the ponds, including any potential for activity transfer from solid to solution phases, is critical in the environmental clean up of these nuclear legacy facilities. Plutonium isotopes (with (241)Am) dominate the alpha-activity within these ponds. Herein, the Pu species in samples taken from a UK legacy fuel storage pond and downstream Holding Tank on the Sellafield site are shown to be predominantly associated with suspended solid phases. Analyses of the residual soluble Pu concentrations indicate differences in Pu solubility between different areas of the pond, which speciation studies suggest are related to differing Pu oxidation state distributions. The implication is that Pu redox chemistry varies across the pond and this controls Pu solubilities and, by implication, Pu behaviour during waste processing. Simple treatment methods to suppress soluble Pu-alpha are also suggested.

Samples of filtered particulates and sludges, formed from corroding magnesium alloy clad uranium metal ("Magnox") fuel elements, collected from one of the legacy nuclear fuel storage ponds located at Sellafield (UK) were investigated by Environmental Scanning Electron Microscopy with Energy Dispersive X-Ray analysis (ESEM/EDX), micro-Raman spectroscopy and Fourier transform infra-red spectroscopy (FT-IR). ESEM imaging confirmed the dominant morphology to be clusters of interlocking platelets typical of brucite (Mg(OH)2). EDX analysis was suggestive of some conversion to the related phase, hydrotalcite (Mg6Al2(CO3)(OH)16·4H2O), due to elevated levels of Al associated with Mg. Other apparent morphologies were less commonly observed including flaky sheets, consistent with earlier stages of Magnox alloy corrosion. In a few specific cases, rods were also observed suggestive of some conversion to Mg-hydroxycarbonate phases. Discrete phases rich in U were also identified. Fluorescence in the Raman spectroscopy also indicated surface coatings of organic macromolecules and iron sulphide on hematite containing particles, attributed to microbial activity within the open air pond. Some specific differences in the solid phases between pond areas with differing conditions were apparent.