[en] This chapter reviews recent research on mixed actinide precipitation using oxalates with a view to obtaining an oxide powder that is homogeneous at the atomic level. These powders can be made into new nuclear fuel that help to avoid hot spots from segregated metal oxide phases and will help in the dissolution of the spent fuel when it comes to reprocessing. By introduction, plutonium oxalate precipitation and calcination is placed in the context of current industrial-scale processes, before discussing the structures of mixed actinide oxalates. Examples of areas where this technology has been demonstrated in the COEX™ process and for minor actinide bearing blanket fuel are also discussed. (author)

[en] In the U.K., High Level Waste from reprocessing operations is vitrified at the Sellafield Waste Vitrification Plant (WVP). A small number of the nuclides present in the waste have the potential to volatilize during vitrification. In order to prevent release of any radionuclides to the environment it is important to understand the mechanisms by which volatilization may occur and to have suitable controls in place. One element of particular concern is ruthenium, formed during the fission of nuclear fuel, which has the potential to form gaseous species such as RuO₄ during the vitrification process and whose behavior must therefore be understood in order to underpin the safe operation of WVP. (authors)

[en] 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 found to be 0.90, and 1.24 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). The value of the rate constant k for the rate law -d[V(V)]/dt = k[V(V)][HNO₂]^1.24/[H⁺] was found to be 7.5 × 10^-4 M^-0.24 s^-1 at 20°C, two orders of magnitude smaller than that for the reduction of Np(VI) by HNO₂, a difference that is attributable to the loss of one of the oxygens during the VO₂⁺ to VO²⁺ reduction reaction. (author)

[en] The processing of spent advanced fuels, such as from Generation IV reactors, will need to manage much higher levels of plutonium, minor actinides and fission products than current experience with LWR UOx fuel reprocessing. It also seems likely that processes will need to be adapted to recover a wider range of products, potentially including the minor actinides (Np, Am, Cm), and to avoid pure separated plutonium at any point within the process. Within European projects such as EUROPART and ACSEPT, hydrometallurgical based processes, including GANEX (Group Actinide Extraction) are being developed to meet these objectives. This paper describes progress towards defining a suitable candidate solvent extraction system for a GANEX process. The extractions of actinides (U, Pu, Np, Am), lanthanides and Tc into a mixture of TOGDA/TBP extractants diluted in a paraffinic diluent have been assessed under process conditions by measuring distribution ratios, third phase boundaries and kinetics. Stripping of actinides from TBP and TOGDA/TBP solutions at low acidities and using complexants has also been investigated. Process models are being developed to design GANEX flowsheets for spiked tests in centrifugal contactor rigs. A preliminary flowsheet using a U, Np, Eu feed to test the hydrodynamic performance of the TOGDA/TBP solvent and routing of U and Np within the flowsheet has been conducted within our miniature centrifugal contactor rig prior to Pu-active glove box trials later in 2009-10. (authors)

[en] The development of advanced separation processes for spent nuclear fuel reprocessing and minor actinide recycling is an essential component of international R and D programmes aimed at closing the nuclear fuel cycle around the middle of this century. While both aqueous and pyrochemical processes are under consideration internationally, neither option will gain broad acceptance without significant advances in process safety, waste minimisation, environmental impact and proliferation resistance; at least when compared to current reprocessing technologies. The UK National Nuclear Laboratory (NNL) is developing flowsheets for innovative aqueous separation processes. These include advanced PUREX options (i.e. processes using tributyl phosphate as the extractant for uranium, plutonium and possibly neptunium recovery) and GANEX (grouped actinide extraction) type processes that use diglycolamide based extractants to co-extract all transuranic actinides. At NNL, development of the flowsheets is closely linked to research on process safety, since this is essential for assessing prospects for future industrialisation and deployment. Within this context, NNL is part of European 7. Framework projects 'ASGARD' and 'SACSESS'. Key topics under investigation include: hydrogen generation from aqueous and solvent phases; decomposition of aqueous phase ligands used in separations prior to product finishing and recycle of nitric acid; dissolution of carbide fuels including management of organics generated. Additionally, there is a strong focus on use of predictive process modelling to assess flowsheet sensitivities as well as engineering design and global hazard assessment of these new processes. (authors)

[en] Evidence to support the Raman assignments of the 1LO (578 cm⁻¹) and 2LO (1158 cm⁻¹) lattice vibrations for PuO₂ material is presented. The T_2g signal is established at 476 ± 2 cm⁻¹ in agreement with literature values. An increase of the 1LO band and an increase of the unit cell lattice parameter with ageing in our samples are found not to be a consequence of PuO_2+x formation but rather a result of simple lattice defects due to radiation damage. The Raman spectrum of AnO₂(OH)₂⋅xH₂O (An = Np, Pu) and laser induced decomposition products suggest that the transition to AnO₂ involves Np₂O₅ for neptunium but no such analogue could be detected for Pu. The presence of a band around 1150 ± 10 cm⁻¹ for a range of MO₂ fluorite structures (CeO₂, ThO₂, UO₂, NpO₂ and PuO₂) suggests that this band is not derived from crystal field electronic f–f transitions as proposed previously and supports recent suggestions that it is the first overtone of the 1LO lattice vibration. The spectrum of PuO₂ is taken across a wide wavenumber range (200–4000 cm⁻¹) and additional signals (2116 and 2611 cm⁻¹) not previously reported have been observed but are not yet assigned with confidence.

[en] The dissolution of uranium or uranium-plutonium carbide fuel in nitric acid leads to ∝50% carbon evolved as carbon dioxide, the remainder remains in the solution as soluble organics. These dissolved organic molecules interfere with the solvent extraction of uranium and plutonium by complexing to the actinide ions and decreasing the efficiency of their extraction. Experiments reported here describe two series of experiments assessing the uranium carbide dissolution liquor treatment by prolonged boiling and electrochemical oxidation. Plutonium losses to aqueous and solvent raffinates are observed for untreated liquors, highlighting that mineralisation of dissolved organics is necessary to reduce the complexing effects of organic acids to an extent that permit efficient operation of a solvent extraction process both in the first solvent use (considered here) and for maintaining solvent quality during industrial solvent reuse in the highly active cycle. Solution carbon analysis and 30% TBP solvent extraction batch tests of uranium and plutonium originating from dissolved uranium carbide liquors untreated and after treatment are compared. These experiments demonstrate the reprocessing of uranium carbides by direct dissolution coupled to a mineralisation process, can achieve near quantitative uranium and high plutonium recoveries (99.9%).

[en] The solvent combination N,N,N'N'- tetraoctyl diglycolamide (TODGA)/tributyl phosphate (TBP)/odourless kerosene (OK) is examined as a potential solvent system for a Grouped Actinide Extraction (GANEX) process to separate all of the actinides from fission products when reprocessing spent nuclear fuel. A series of solvent extraction batch experiments were performed with a range of TODGA/TBP/OK solvent combinations to assess the sensitivity of distribution values for a number of key elements towards [TBP] (0 - 1.1M), [TODGA] (0.1-0.4M), [HNO₃] (0.1-5M) and heavy metal loading ([U] 0-200g/l). There is little impact on D_Am or D_Eu across the solvent range and no influence from U loading. Excellent D_Np values (> 10) are observed, increasing with increasing [TODGA], with [TBP] having little influence. Such high D_Np values may obviate the need for preconditioning of dissolved fuel feeds to control Np routing. High D_Tc values are found even at 5M HNO₃, therefore Tc is expected to remain in the solvent phase. Both Pu(III) and Pu(IV) are readily extracted with D_Pu(III) > D_Pu(IV). Uranium is extracted by both TBP and TODGA and TBP is shown to effectively compete with TODGA for uranium coordination sites. Third phase formation occurs at high [U] loading and [HNO₃] but is suppressed by increasing [TBP].

[en] Highlights: • 5 years of development work is described for ²⁴¹Am production from a plutonium dioxide feedstock. • ²⁴¹Am and plutonium products are produced in high yield (>99%) and high purity (>99%). • A solvent system is defined and optimised to give the best americium recovery and minimise radioactive wastes. • Americium oxalate precipitation and thermal decomposition is used to recover the americium as an oxide. - Abstract: 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.

[en] Graphical abstract: Spent Magnox fuel corroding in-situ in storage ponds forms sludges comprised of brucite and other Mg based phases with uranium oxide particles. Display Omitted Research highlights: → Caracterization study of highly radioactive corroded Magnox sludges. → Unique data from samples of actual corroded nuclear fuel. → Combined electron microscopy and vibrational spectroscopy study. → Analysis of particles from legacy spent fuel storage pond at Sellafield. → Supports major UK decommissioning and nuclear clean up challenge. - Abstract: 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.