Grouped actinide extraction (GANEX) is a concept developed in Europe for the "homogeneous" separation of actinides for recycling into Generation IV reactors. The first cycle (GANEX-1) is a selective separation of uranium from fission products and transuranic actinides typically using a monoamide extractant such as N,N-di-(2-ethylhexyl)isobutyramide (DEHiBA). This is followed by a 2(nd) cycle (GANEX-2) that selectively separates the transuranic actinides. Solvent extraction with DEHiBA has other applications including the recovery of uranium from fuel residues or reprocessing high assay low enriched uranium (HALEU) for use in advanced modular reactors. A flowsheet test of the GANEX-1 cycle using a simulated feed has been carried out in laboratory-scale centrifugal contactors at a higher solvent phase loading than previously demonstrated. The test was monitored using on-line UV-vis spectrophotometry on the uranium products as well as off-line analysis of products and profiles (for acidity, uranium, neptunium, americium and technetium). Good results were obtained, particularly with regard to neptunium decontamination and uranium recovery. Technetium behaviour remained problematic with similar to 25% routed with the uranium product. Batch distribution experiments indicate that this is due to co-extraction of technetium with uranium, whereas unlike the PUREX process, co-extraction with zirconium does not occur to any measurable degree. Good decontamination factors were also obtained for a range of other important fission product elements.

A comprehensive understanding of solvent physicochemical properties is essential to developing advanced solvent extraction processes, such as Advanced PUREX (Plutonium Uranium Reduction Extraction) and GANEX (Group Actinide Extraction), that can safely and efficiently recycle spent nuclear fuel. The densities, viscosities and surface tensions of binary mixtures of tributyl phosphate (TBP) (0.92-3.66 M) in n-dodecane and N,N-di (2ethyl hexyl)isobutyramide (DEHiBA) (0.69-2.78 M) in n-dodecane were measured at atmospheric pressure from 278.15 to 333.15 K. Through these measurements, empirical expressions to predict the density and viscosity at any molar composition of these solvents within the studied temperature range were constructed and shown to be in good correlation with all experimental data across the studied variables. The relationship between temperature and viscosity of pure TBP and DEHiBA can be described by the Arrhenius equation for viscosity in the temperature range studied. Surface tension measurements (air-liquid interface) were obtained for varying extractant concentrations between 278.15 and 333.15 K, which showed that DEHiBA acts as a surfactant when mixed with n-dodecane whilst TBP exhibits a higher surface tension than n-dodecane.

Many countries, including the United Kingdom, have committed to reaching “net zero” emissions by 2050. To meet this challenge requires urgent deployment of low-carbon energy-generating technologies, not just for electricity generation but also other sectors, including transportation and heating. However, this will only be successful if the other two pillars of sustainability (social and economic impacts) are balanced with the environmental drivers. All energy-generation technologies have benefits and drawbacks, and these must be objectively and fairly assessed using a “level playing field” approach. Nuclear energy has benefits that are complementary to renewables and thus can play a valuable role in delivering large amounts of low-carbon energy globally. However, critics of nuclear energy raise concerns related to safety (and security), radioactive waste management, and economics that have challenged its acceptance as a sustainable energy source in some quarters. Nevertheless, objective consideration of sustainability in global energy needs and the different generating technologies clearly indicate a valuable role for nuclear energy in a sustainable and low-carbon future. It is concluded that nuclear energy should be recognised as “sustainable”, and the analysis shows that energy portfolios incorporating nuclear provide the most sustainable system overall.

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.

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.

We report the results of a "CHON-compliant" i-SANEX flowsheet test using a surrogate feed with process con-centrations of 241Am (>1 g/L) and major lanthanide and fission product species in a 32-stage counter-current centrifugal contactor (CC) rig. By combining flowsheet modelling with experience from the European GEN-IORS project, a flowsheet was designed that included the lanthanide/actinide extraction, fission product scrubbing, actinide stripping and finally, lanthanide stripping; therefore, covering the entire flowsheet except solvent treatment. The extraction system was 0.2 M N,N,N'N'-tetraoctyldiglycolamide (TODGA) in Isane-185 with 5 v/v.% 1-octanol as a phase modifier. Actinide stripping was afforded by the aqueous soluble pyridine-2,6-bis(1H-1,2,3-triazol-4-yl) (PTD), whilst the lanthanide stripping was performed by N,N,N',N'-tetraethyldi-glycolamide (TEDGA). This represents the first test of PTD and TEDGA across the full flowsheet using counter-current CC and with process concentrations of americium. This CHON-compliant i-SANEX flowsheet offers clear improvements to the reference process. The use of a full PUREX raffinate simulant as the feed for the process has also permitted the measurement of concentration profiles of many species that are not normally followed in such detail. This has highlighted that under the conditions used - 8 stages of extraction, 8 stages of scrubbing (0.7 M HNO3) - some of the lanthanides recycle and do not reach steady-state in the timeframe of the experiment. Overall, the flowsheet demonstrated good control of 241Am with 98.6 % being routed to the actinide product although further work is needed to fully control some lanthanides and fission products, including; La, Ce, Pr, Gd, Tc, Ru and Sr.

During the plutonium uranium reduction extraction (PUREX) process, nitric acid facilitates the extraction of actinides from the aqueous phase into the organic phase by forming neutral, organic soluble complexes with trin-butyl phosphate (TBP). The concentration of nitric acid is generally measured by titration; however, titration is a time-consuming method that generates significant volumes of additional waste. Optical spectroscopic techniques can be used to perform fast, automated measurements off-line or on-line, without generating any waste. In this work, the effectiveness of Raman and mid-infrared (MIR) spectroscopy has been compared for the first time as an alternative to titration for the quantification of nitric acid in PUREX-relevant mixtures. Samples of 0-12 M nitric acid in the aqueous phase and 0-1.10 M nitric acid in the organic phase (TBP/odourless kerosene (OK)- H2O-HNO3 model system) were analysed and partial least squares (PLS) regression models were built to predict nitric acid concentration. MIR spectra required less pre-processing than Raman spectra and more accurate predictions of nitric acid concentration were obtained for MIR spectroscopy than for Raman spectroscopy, with root mean square error of prediction (RMSEP) values of 0.099 M versus 0.148 M obtained for the aqueous phase and root mean square error of cross validation (RMSECV) values of 0.006 M versus 0.013 M obtained for the organic phase. To investigate the ability to predict nitric acid concentration in the presence of uranyl nitrate, samples containing uranium (0-100 g/L) and nitric acid (0.15-0.64 M) in the organic phase (U-TBP/OKH2O-HNO3 model system) were analysed by Raman and MIR spectroscopy. The RMSECV was 0.027 M and 0.066 M for MIR and Raman spectroscopy, respectively; these values are higher than those obtained in the absence of uranyl nitrate owing to differences in the experimental approaches employed. Therefore, the results obtained demonstrate that MIR or Raman spectroscopy could be used to measure the concentration of nitric acid in the organic and aqueous phases in the PUREX process.

Controlling the routing of technetium to keep it away from the uranium product streams is essential in order to deliver efficient and safe recycling of spent nuclear fuel in future processes such as Advanced PUREX. In this paper, a flowsheet simulation code is developed for the simulation of technetium and zirconium routing in an Advanced PUREX flowsheet. The essence of this code is its consideration of the slow conversion of inextractable forms of zirconium to extractable forms in the modelling of mass transfer rates. The code is applied to model a flowsheet tested experimentally and is validated by experimental results. Sensitivity of the flowsheet to operating parameters, including temperature, organic-to-aqueous (O/A) flow ratio, and scrub nitric acid concentration is investigated; an optimized set of operating conditions is suggested that could result in lower amounts of tech-netium being routed to the uranium product.

EURO-GANEX is an innovative hydrometallurgical process for the group separation of transuranic actinides from future spent nuclear fuels. A flowsheet test of a process upset (or ‘maloperation’) has been carried out involving the reduction in the concentration of the scrub acid feed to the extract-scrub contactors. The experimental test was completed in two stages. First, the extract-scrub contactors were run under normal flowsheet conditions before initiating the process upset by reducing the scrub acidity from 0.5 to 0.05 mol/L HNO3. At low acidity, a change in the online UV-Vis spectra of the solvent phase indicated the formation of hydrolyzed plutonium species or plutonium colloid formation. Surprisingly, however, this did not lead to the recycle and accumulation of plutonium in the extract-scrub contactors as expected indicating that the EURO-GANEX process is quite robust towards this process upset. Further spectroscopic investigations of the organic-phase species were also performed to characterize the conditions under which the hydrolyzed plutonium species form in the EURO-GANEX solvent as well as separate TODGA and DMDOHEMA phases. The spectroscopic studies supported the view that Pu(IV) hydrolyses at low acidity but the hydrolysis is limited by the organic ligands and is consequently reversible on raising the acidity again. The hydrolysis was also observed in separate TODGA and DMDOHEMA phases.

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.

Since 2017, the UK Government has invested in research and technology development of advanced nuclear technologies including options for the future recycling of spent nuclear fuels. Whilst the UK has previously reprocessed and recycled uranium and plutonium from spent fuels, industrial scale reprocessing has finished and future fuels are being stored ready for disposal in a deep geological repository around the end of the century (i.e.transition to the open fuel cycle). Nevertheless, it is possible that an expansion of the use of nuclear energy to help meet Net Zero carbon emission targets and changing public perceptions or geopolitical situations over this period may necessitate a return to some form of closed fuel cycle. Consequently, the Advanced Fuel Cycle Programme (AFCP) has been developing advanced recycling options using both aqueous and pyro-chemical methods that would reduce wastes, costs and environmental impacts but with enhanced flexibility and barriers against nuclear proliferation risks. As well as technology development, AFCP has developed approaches for evaluating the benefits and disbenefits of future fuel cycle options under the framework of sustainability. This article provides an overview of the progress made under AFCP on the aqueous recycling of spent fuels, including the management of secondary wastes arising from the reprocessing schemes.

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.

The fission products of technetium and zirconium have historically been problematic in the reprocessing of spent nuclear fuel by solvent extraction using tributyl phosphate (TBP) containing solvents. One of the reasons for this is that the routing of zirconium and technetium becomes difficult to control due to co-extraction mechanisms with other elements/species in the dissolved spent fuel liquors and through alternative extraction pathways that can occur with the presence of solvent degradation products. Consequently, solvent extraction processes based on the PUREX (Plutonium Uranium Redox EXtraction) process incorporate various strategies to ensure these fission products are not present in the final product streams, increasing plant footprints and operational costs. Next generation spent nuclear fuel reprocessing should minimise the need for such scrubbing operations by applying a complete and thorough understanding of the distribution behaviour of technetium and zirconium to optimise the separations chemistry for higher burn up spent fuels from advanced reactor systems that contain higher inventories of fission products. A substantial body of work exists regarding the distribution behaviour of technetium and zirconium with phosphorus based extractants, but studies have tended to be fragmented using various extraction conditions and approaches. This review collates and reviews these data, supporting the development of predictive process models whilst making recommendations for improved control of technetium and zirconium in an advanced PUREX process. Key findings from this review include the significant increase of technetium distribution ratios by coextraction with zirconium. The distribution ratios of zirconium are also seen to increase with increasing technetium concentrations when zirconium is present with technetium through a synergistic effect. To achieve full decontamination of the U/Pu product stream, significant process modifications are required, which can be achieved by the introduction of scrubbing steps or the saturation of the organic phase with uranium. The use of holdback reagents may improve decontamination factors, but further experimental research is required. High acid scrubs to reject technetium are established options and already used at the in-dustrial scale at La Hague (France).

The hydrometallurgical separation concepts for the recycling of irradiated nuclear fuels developed in Europe are presented and discussed. Whilst Part 1 of the review focused on concepts for heterogeneous recycling of minor actinides, this article focuses on group recycling of transuranic actinides, which would support homogeneous recycling scenarios. Most of these concepts were developed within European collaborative projects and involve solvent extraction processes separating all the actinides (U-Cm) in two cycles. The first cycle uses a monoamide extractant to recover uranium leaving all the transuranic actinides in the aqueous raffinate with the fission products. The second cycle aims for a group recovery of the transuranium elements and several strategies have been proposed for this stage. In this review article, the various solvent extraction processes are summarised and the key features of the process schemes are compared.

The production of Pu-238 from Np-237, to the specifications required for space flight, requires the separation of plutonium from neptunium and a range of fission products. The valuable neptunium is then recycled into targets for further irradiation. At Oak Ridge National Laboratory (ORNL), a solvent extraction process has been developed to separate neptunium and plutonium that uses sodium nitrite to control the neptunium oxidation. This results in a neptunium product contaminated with sodium, which needs to be removed prior to recycling. We have developed a separation process that results in a sodium-free neptunium product with very low levels of plutonium contamination, there by reducing the number of purification steps and thus simplifying the process.

Liquid-liquid extraction is a commonly used technique to separate metals and is a process that has particular relevance to the nuclear industry. There has been a drive to use environmentally friendly ligands composed only of carbon, hydrogen, nitrogen, and oxygen. One example is the i-SANEX process that has been developed to separate minor actinides from spent nuclear fuel. The underlying science of such processes, is, however, both complex and intriguing. Recent research indicates that the liquid phases involved are frequently structured fluids with a hierarchical organization of aggregates. Effective flow-sheet modeling of such processes is likely to benefit from the knowledge of the fundamental properties of these phases. As a stepping stone toward this, we have performed molecular dynamics simulations on a metal free i-SANEX system composed of the ligand N,N,N',N'-tetraoctyl diglycolamide (TODGA), diluent hydrogenated tetrapropylene (TPH), and polar species water and nitric acid. We have also studied the effects of adding n-octanol and swapping TPH for n-dodecane. It would seem sensible to understand this simpler system before introducing metal complexes. Such an understanding would ideally arise from studying the system's properties over a wide range of compositions. The large number of components, however, precludes a comprehensive scan of compositions, so we have chosen to study a fixed concentration of TODGA while varying the concentrations of water and nitric acid over a substantial range. Reverse aggregates are observed, with polar species in the interior in contact with the polar portions of the TODGA molecules and the organic diluent on the exterior in contact with the TODGA alkyl chains. These aggregates are irregular in shape and grow in size as the amount of water and nitric acid increases. At a sufficiently high polar content, a single extended cluster forms corresponding to the third phase formation. No well-defined bonding motifs were observed between the polar species and TODGA. The cluster size distribution fits an isodesmic model, where the Gibbs energy change of adding a TODGA molecule to a cluster ranges between 4.5 and 7.0 kJ mol-1, depending on the system composition. The addition of n-octanol was found to reduce the degree of aggregation, with n-octanol acting as a co-surfactant. Exchanging the diluent TPH for n-dodecane also decreased the aggregation. We present evidence that this is due to the greater penetration of n-dodecane into the reverse aggregates. It is known, however, that the propensity for the third phase formation is greater with n-dodecane as the diluent than is the case with TPH, but we argue that these two results are not contradictory. This research casts light on the driving forces for aggregation, informs process engineers as to what species are present, and indicates that flow-sheet liquid-liquid extraction modeling might benefit by incorporating an isodesmic aggregation approach.

Globally, around half a million tonnes of spent nuclear fuel (SNF) will be in dry or wet storage by around 2050. Continued storage is not sustainable and this SNF must eventually either be disposed (the open nuclear fuel cycle) or recycled (the closed fuel cycle). Many international studies have addressed the advantages and disadvantages of these options which can be considered now in the framework of sustainable development and the three pillars of: economic, environmental and societal impacts. To inform this debate, a detailed survey of the available literature related to economic assessments of closed and open cycles has been undertaken-this complements an earlier review on environmental impacts. Results of economic assessments showing how the management of spent fuels in the open and closed cycles impacts the costs of the nuclear fuel cycle, are usually presented in terms of the levelised cost of electricity (LCOE). It is clear that the costs of the back end of the fuel cycle are a relatively minor component of the LCOE and that there is significant overlap between calculations on open and closed fuel cycles.

Technetium and zirconium are two of the most problematic fission products in the PUREX process due to their complex extraction behaviors. Although there are several published works related to the calculation of Tc and Zr distribution coefficients, the reliability of those methods must be validated carefully, and further knowledge in this area is still needed. This work is, therefore, devoted to developing technetium and zirconium distribution coefficient models for predicting their distribution between nitric acid solution and tributyl phosphate in diluent for solutions containing uranium, technetium, and zirconium. The distribution coefficient model for zirconium is based on the organic nitric acid concentration, rather than the free TBP concentration. An improved technetium distribution coefficient model, in the absence of other metals, is also developed. The coextraction of technetium and zirconium or uranium is calculated by ligand exchange mechanisms. The validation results demonstrate that the distribution coefficient models developed in this work are more accurate than established models.

Plutonium multi-recycling is one of the main aims to gradually achieve a fully closed and sustainable nuclear fuel cycle in the future, particularly once fast reactors will be deployed into the fleet. To reach this objective, optimizations of current processes and technologies and, in some cases, development of innovative concepts are required to meet the scientific, technical and economical challenges that are involved in multi-recycling of spent nuclear fuels with high plutonium contents. Specifically, the issues inherent in the hydrometallurgical treatment of spent mixed oxide (MOX) fuels containing increasing plutonium content must be addressed through new efficient processes for fuel dissolution, plutonium separation, MOX fabrication and waste conditioning. Some of the leading concepts being developed internationally are described in this chapter.