Reprocessing Of Spent Nuclear Fuel Research Articles

Plutonium and Cerium Perrhenate/Pertechnetate Coordination Polymers and Frameworks.

Spent nuclear fuel (SNF) contains transuranic and lanthanide species, which are sometimes recovered and repurposed. One particularly problematic fission product, 99TcO4-, hampers this recovery via coextraction with high valence metals, perhaps by complexation during aqueous reprocessing of SNF. There is limited molecular-level knowledge concerning the coordination chemistry between TcO4- or its well-known surrogate ReO4- and transuranic/lanthanide species. In the current study, we investigated the coordination of ReO4-/TcO4- with plutonium and cerium cations by structural and chemical characterization of a series of isolated extended solids. In this study, Ce represents both trivalent lanthanides and is considered a surrogate for Pu, respectively, in its common trivalent and tetravalent oxidation states. The structural elucidation of the seven isolated crystalline solids revealed that ReO4-/TcO4- directly connects to PuIV, PuVIO22+, CeIII, and CeIV in the terminal and bridging coordination modes, leading to 1-, 2-, and 3-dimensional frameworks. For example, ReO4- coordination to Pu(IV) formed a 1D chain or 2D framework, isostructural with previously isolated Th(IV) compounds. However, PuVIO22+ alternating with ReO4- led to a unique 1D chain, different from the prior-reported U(VI)/Np(VI)-ReO4-/TcO4- structures. Coordination of ReO4-/TcO4- with Ce(III) promotes the assembly of 3D frameworks. Finally, attempted synthesis of a Ce(IV)-ReO4- compound resulted in a 2D framework with a mixed-valence CeIII/IV. The highly acidic reaction conditions supported the reduction of both CeIV and TcVII, challenging isolation of compounds featuring these species. Only one TcO4-containing structure was obtained in this study (CeIII-TcO4 3D framework), vs the six total Ce/Pu-ReO4 compounds. Our three Pu-ReO4 crystal structures are the first reported and translated to atomic-level information about Pu-TcO4 coordination in nuclear fuel reprocessing scenarios, in addition to broadening our knowledge of bonding trends in the early, high-valence actinides.

A Review of Opportunities and Methods for Recovery of Rhodium from Spent Nuclear Fuel during Reprocessing

Rhodium is one of the scarcest, most valuable, and useful platinum group metals, a strategically important material relied on heavily by automotive and electronics industries. The limited finite natural sources of Rh and exponentially increasing demands on these supplies mean that new sources are being sought to stabilise supplies and prices. Spent nuclear fuel (SNF) contains a significant quantity of Rh, though methods to recover this are purely conceptual at this point, due to the differing chemistry between SNF reprocessing and the methods used to recycle natural Rh. During SNF reprocessing, Rh partitions between aqueous nitric acid streams, where its speciation is complex, and insoluble fission product waste streams. Various techniques have been investigated for Rh recovery during SNF reprocessing for over 50 years, including solvent extraction, ion exchange, precipitation, and electrochemical methods, with tuneable approaches such as impregnated composites and ionic liquids receiving the most attention recently, assisted by more the comprehensive understanding of Rh speciation in nitric acid developed recently. The quantitative recovery of Rh within the SNF reprocessing ecosystem has remained elusive thus far, and as such, this review discusses the recent developments within the field, and strategies that could be applied to maximise the recovery of Rh from SNF.

Development and Tasks of Radiochemical Technologies: History and Modern Challenges

The main prerequisites for creating and developing industrial radiochemical production in the USSR and the Russian Federation is reviewed. The history of spent nuclear fuel (SNF) management in relation to the RT-1 and RT-2 plants, the features and tasks of the plants are considered from the position of today. Current specifics of the platforms of the Mayak Production Association, the Mining and Chemical Complex and the Siberian Chemical Combine of the closed nuclear fuel cycle of the Russian Federation was analyzed. Topical issues are formulated on SNF reprocessing and radioactive waste management technologies.

Simultaneous quantification of Zr, Mo, U, Np and Pu in technological solutions of spent nuclear fuel reprocessing with a potentiometric multisensor system

The chemical analysis of spent nuclear fuel (SNF) reprocessing is a very challenging research field due to the high radioactivity of process streams and their complex composition. The process managing requires thorough chemical analysis to ensure safe and efficient performance. Traditional ways of chemical control of SNF reprocessing are based on sampling methods and involve long and tedious procedures, failing to provide immediate information on the process status. Recent literature suggests that potentiometric multisensor systems can be employed for on-line control. However, up to now these systems were only studied in very simple model media. This work reports on the potentiometric multisensor array applied for the first time for the simultaneous quantification of the content of actinides (uranium, plutonium, neptunium) and several other analytes (zirconium, molybdenum and nitric acid) in the complex samples obtained from the pilot extraction unit for SNF reprocessing. The same samples were also analyzed by UV-Vis spectrometry (another prospective method for on-line control) and the analytical performance of these two methods was compared using established analytical figures of merit (sensitivity and analytical sensitivity metrics) for multivariate calibration. In spite of the extremely challenging analytical task, both simple methods provide the characteristics suitable for real industrial applications.

Justification of the economic and radiological advantages of a closed nuclear cycle with fast neutron reactors in view of IAEA radiation safety standards and equalization of carcinogenicity of the radioactive waste and natural uranium raw materials

Economic and radiological advantages of the closed nuclear fuel cycle with fast neutron reactors in view of radiological equivalence, i.e. the equalization of carcinogenicity of radioactive waste and the natural uranium ore materials are considered in the paper. Potential cancer risks of the spent nuclear fuel (SNF) from the WWER-1000 reactor and radioactive waste from the BR-1200 reactor generating 1 GW of electric power per year have been estimated. It is assumed that the SNF from WWER-1000 reactor will be sent to 10 or 30-year storage. In the closed nuclear fuel cycle radioactive wastes, the products of SNF reprocessing, consisting of 0.1% of Sr, Cs, Tc, I, U, Np, Pu, Am, Cm of their content in SNF and all other radionuclides are disposed. Over a period of 10,000 years, the total radiation detriment from SNF from the WWER-1000 reactor is more than 132 times higher than the radiation detriment from radioactive waste from the BR-1200 reactor. The time of radiological equivalence achievement was evaluated. It is assumed that if WWER-1000 and BR-1000 generate 1 GW of electricity per year, radiological equivalence of carcinogenicity of radioactive waste (BR-1200 reactor) and natural uranium ore material will be achieved in 100 years of radioactive waste storage. The equivalence of carcinogenicity of SNF storage and natural uranium ore materials will be achieved after 15,600 years storage. When disposing of spent fuel from a WWER-1000 reactor without achieving radiological equivalence, socio-economic losses due to excess mortality from cancer are estimated at 129 billion rubles/GW year.

Insights into the Complexation Mechanism of a Promising Lipophilic PyTri Ligand for Actinide Partitioning from Spent Nuclear Fuel.

The challenging issue of spent nuclear fuel (SNF) management is being tackled by developing advanced technologies that point to reduce environmental footprint, long-term radiotoxicity, volumes and residual heat of the final waste, and to increase the proliferation resistance. The advanced recycling strategy provides several promising processes for a safer reprocessing of SNF. Advanced hydrometallurgical processes can extract minor actinides directly from Plutonium and Uranium Reduction Extraction raffinate by using selective hydrophilic and lipophilic ligands. This research is focused on a recently developed N-heterocyclic selective lipophilic ligand for actinides separation to be exploited in advanced Selective ActiNide EXtraction (SANEX)-like processes: 2,6-bis(1-(2-ethylhexyl)-1H-1,2,3-triazol-4-yl)pyridine (PyTri-Ethyl-Hexyl-PTEH). The formation and stability of metal-ligand complexes have been investigated by different techniques. Preliminary studies carried out by electrospray ionization mass spectrometry (ESI-MS) analysis enabled to qualitatively explore the PTEH complexes with La(III) and Eu(III) ions as representatives of lanthanides. Time-resolved laser fluorescence spectroscopy (TRLFS) experiments have been carried out to determine the ligand stability constants with Cm(III) and Eu(III) and to better investigate the ligand complexes involved in the extraction process. The contribution of a 1:3 M/L complex, barely identified by ESI-MS analyses, was confirmed as the dominant species by TRLFS experiments. To shed light on ligand selectivity toward actinides over lanthanides, NMR investigations have been performed on PTEH complexes with Lu(III) and Am(III) ions, thereby showing significant differences in chemical shifts of the coordinating nitrogen atoms providing proof of a different bond nature between actinides and lanthanides. These scientific achievements encourage consideration of this PyTri ligand for a potential large-scale implementation.

Structural Insight into Complexation Ability and Coordination of Uranyl Nitrate by 1,10-Phenanthroline-2,9-diamides.

Reprocessing of spent nuclear fuel (SNF) is an important task in a frame of ecology and rational use of natural resources. Uranium, as the main component of SNF (>95%), can be recovered for further use as fresh nuclear fuel. To minimize an amount of solid radioactive waste generated during SNF reprocessing, new extractants are under investigation. Diamides of 1,10-phenanthroline-2,9-dicarboxylic acid are perspective tetradentate N-donor ligands that form strong complexes with f-elements, which are soluble in polar organic solvents. As an example of three ligands of this class, we conducted a comparative study and showed how the substituent in the amide functional group affects the extraction ability toward uranyl nitrate from nitric acid media. We have performed a careful study (NMR, FT-IR, XRD, RMC-EXAFS) of the structures of synthesized complexes of new ligands with uranyl nitrate and used quantum mechanical calculations to explain the discovered regularities through.

Evaluation of the corrosion behavior of modern spent nuclear fuels under repository conditions

Abstract. In Germany it is planned to directly dispose spent nuclear fuel (SNF) from nuclear power plants together with other high-level radioactive wastes (HLW) from former SNF reprocessing (e.g., vitrified waste), in a deep geological repository for heat-generating wastes – the siting process for this repository was started in 2017 and is ongoing. Based on several decades of research, development, and demonstration (RD&D) it is generally accepted at the technical and scientific level that direct disposal of HLW and SNF in deep mined geological repositories is the safest and most sustainable option (CEC, 2011; IAEA, 2004). The current efforts to improve the performance and accident tolerance of fuels in nuclear power generation resulted in an increased utilization of a variety of new types of light-water reactor (LWR) fuels such as fuels doped with Cr, Al, and Si. This doping leads to a significant change of the microstructure of the fuel matrix. The corrosion behavior of these types of fuels under conditions relevant to deep geological disposal has hardly been studied so far; however, this is of crucial importance as the development of a robust safety case for deep geological disposal of SNF requires a solid understanding of its dissolution behavior over very long time scales (up to 1 million years). To fill this knowledge gap, additional systematic studies on modern doped UO2 fuels were needed. Corrosion experiments with SNF cannot entirely unravel all of the various concurring effects of the dissolution mechanism due to the chemical and structural complexity of SNF and its high beta and gamma radiation field during the first 1000 years; moreover, technical restrictions only allow a very limited number of experiments. Therefore, within the EU-DisCo project (https://www.disco-h2020.eu, last access: 11 October 2021), a very ambitious programme of corrosion studies on irradiated Cr and Al/Cr doped fuels was carried out, which was complemented by systematic single-effect dissolution studies (e.g., with respect to doping level, grain size and thermodynamic aspects) performed on carefully prepared and characterized, simplified UO2-based model materials. Here, we present recent results on the dissolution behavior of tailor-made UO2 model materials in accelerated static batch experiments using H2O2 as simulant for radiolytic oxidants, present in long-term disposal scenarios for SNF in failed container conditions due to the alpha irradiation of water. In these dissolution experiments pure UO2 reference pellets exhibiting different densities and grain sizes, as well as Cr-doped UO2 pellets with various Cr-doping levels, produced using different doping methods having different grain sizes, were used. In addition, Nd-doped and industrially produced Cr- and Cr/Nd-doped UO2 pellets were used to determine the influence of these parameters on the dissolution rates. The dissolution experiments were performed under strictly controlled conditions with respect to exclusion of oxygen, temperature control, and exclusion of light. This bottom-up approach was followed to understand how the addition of Cr-oxide into the fuel matrix affects SNF dissolution behavior under repository relevant conditions. The results of the dissolution experiments performed with real SNF and the model materials obtained by the DisCo partners build the basis for numerical simulations on the dissolution behavior of modern SNF. First results of the data evaluation indicate that the addition of dopants and the consequential modification of the fuel matrix does not lead to a significant change of the dissolution behavior of these fuels under repository relevant conditions compared to standard SNF (i.e. dissolution rates agree within an order of magnitude).

Spent Nuclear Fuel Reprocessing and Nuclear Materials Recycling in Two-Component Nuclear Energy

Spent Nuclear Fuel Reprocessing and Nuclear Materials Recycling in Two-Component Nuclear Energy

Application of Aluminium Nitrate for Extraction Reprocessing of Spent Nuclear Fuel

Application of Aluminium Nitrate for Extraction Reprocessing of Spent Nuclear Fuel

Application of Aluminium Nitrate for Extraction Reprocessing of Spent Nuclear Fuel

Application of Aluminium Nitrate for Extraction Reprocessing of Spent Nuclear Fuel

Experimental Modeling of Technetium(VII) Recovery from Raffinates after Extractive Reprocessing of Spent Nuclear Fuel

Experimental justification has been provided for the possibility of extractive isolation of technetium from raffinate of technical process of affinage of uranium–plutonium product resulting from hydrometallurgical reprocessing of spent nuclear fuel (SNF). Technetium (VII) from aqueous solution of Zr(IV) (10 g/L) and 4 mol/L HNO3 with small content of ZrOb+ and $${\text{Z}}{{{\text{r}}}_{{\text{2}}}}{\text{O}}_{{\text{3}}}^{{b + }}$$ oxocations is recovered by 80% in 30% TBP in hydrocarbon diluent over seven stages of counter-current continuous cascade at ratio O : A = 1 : 1. Variation in phase flow ratio for cascade part leads to virtually complete Tc(VII) extraction in organic phase over ten stages. Technetium(VII) is selectively back extracted from loaded organic phase with water virtually completely over eleven stages of counter-current continuous cascade. Technological extraction scheme for technetium isolation from SNF solutions has been proposed.

Corrosion of Ceramic and Carbon-Based Materials in FLiNaK

Molten salt nuclear fast nuclear reactor (MSNFR) is one of the prospective design in frame of Generation IV concept. This technology is also required for reprocessing of spent nuclear fuel (SNF) in molten salts to extract valuable components. Recycling SNF consists of extraction uranium and plutonium with the disposal of minor actinides and fission products. These stages need the materials that retain their corrosion and mechanical properties for long time under the influence of high temperatures, radiation fields and contact with molten salts. Various steels and alloys were widely studied to assess their use under such extreme conditions. Alternative materials include ceramics, composite and carbon materials, which have high corrosion resistance in various media and can be used up to 1000 °C.In the present work the corrosion and mechanical properties of nitride ceramics (Si3N4 and BN) and carbon-based materials (carbon-carbon composite material (C/C) and high-density carbon) were studied in a molten mixture of lithium, sodium, and potassium fluorides (FLiNaK) in the temperature range of 550–750 °C under inert atmosphere. Corrosion tests were performed under static conditions, and the duration of each test was 100 h to enable the comparison of the experimental data.It was found that carbon-containing materials (C/C and high-density carbon) showed high corrosion resistance in the melt. C/C samples had lower corrosion rates (less than 0,1 mm/year in the entire temperature range), while the high-density carbon showed better mechanical properties. The main disadvantage of these materials is their impregnation by molten salt due to relative high porosity. However, C/C composite and high-density carbon are promising structural materials for salt media based on FLiNaK in the selected temperature range.Tested nitride ceramics (BN, Si4N3) demonstrated relatively low corrosion resistance in fluoride systems compare to carbon-containing and metallic materials. The corrosion rates of these materials in FLiNaK exceeded the value of 1 mm/year. The plastic properties of nitride ceramics were also poor. Possible application of such materials for MSNFR and SNF reprocessing is under further investigation.

Simulation of Electrolyzer Processes for the Reprocessing of Spent Nuclear Fuel

The ANSYS software package is used to simulate a functioning electrolyzer for the reprocessing of spent nuclear fuel via electrolysis in LiCl molten salt with the addition of Li2O. The seven considered types of electrolyzer design differ in their mutual arrangement of electrodes and anode thickness. The temperature field, as well as distributions of the current density and electric potential, are calculated. The optimal electrolyzer design is determined in order to achieve the best temperature and electrical characteristics.

Hydrodynamic characteristics of 30%TBP/kerosene-HNO3 solution system in an annular centrifugal contactor

Annular centrifugal contactors (ACCs) have many advantages and are recognized as key solvent-extraction equipment for the future reprocessing of spent nuclear fuel (RSNF). To successfully design and operate ACCs for RSNF, it is necessary to understand the hydrodynamic characteristics of the extraction systems in ACCs. The phase ratio (R = Vaq/Vorg, A/O) and liquid holdup volume (V) of the ACC are important hydrodynamic characteristics. In this study, a liquid-fast-separation method was used to systematically investigate the effects of the operational and structural parameters on the V and R (A/O) of a ϕ20 ACC by using a 30%TBP/kerosene-HNO3 solution system. The results showed that the operational and structural parameters had different effects on the V and R (A/O) of the mixing and separating zones of the ACC, respectively. For the most frequently used structural parameters of the ϕ20 ACC, when the rotor speed was 3500 r/min, the total flow rate was 2.0 L/h, and the flow ratio (A/O) was 1, the liquid holdup volumes in the mixing zone and rotor were 8.03 and 14.0 mL, respectively, and the phase ratios (A/O) of the mixing zone and separating zone were 0.96 and 1.43, respectively.

Modern Methods of Radiochemical Reprocessing of Spent Nuclear Fuel

The paper is devoted to the history of development and the current state of technological and scientific advances in radiochemical reprocessing of spent nuclear fuel from water-cooled power reactors. Regarding spent nuclear fuel (SNF) of NPP power reactors, long-term energy security involves adopting a version of its radiochemical treatment, conditioning and recirculation. Recycling SNF is required for the implementation of a closed fuel cycle and the re-use of regeneration products as energy reactor fuels. The basis of modern technological schemes for the reprocessing of the spent nuclear fuel is the “Purex” process, developed since the 60s in the USA. The classic approach to the use of U and Pu nuclides contained in spent nuclear fuel is to separate them from fission products, re-enrich regenerated uranium and use plutonium for the production of mixed-oxide (MOX) fuel with depleted uranium. The modern reprocessing plants are able to deal with fuel with further increase of its main characteristics without significant changes in the initial project. In order to close the fuel cycle, it is needed to add the following technological steps: (1) removal of high-level and long-lived components and minor actinides; (2) return of actinides to the technological cycle; (3) safe disposal of unused components. Each of these areas is under investigation now. Several new promising multi-cycle hydrometallurgical processes based on the joint extraction of trivalent lanthanides and minor actinides with their subsequent separation have been developed. A number of promising materials is suggested to be potential matrices for the immobilization of high-level components of radioactive wastes. To improve the compatibility of fuel processing with the environment, non-aqueous technologies are being developed, for instance, pyro-chemical methods for the reprocessing of various types of highly active fuels based on metals, oxides, carbides, or nitrides. An important scientific and technological task under investigation is transmutation of actinides. The results of international large-scale experiments on the partitioning and transmutation of fuel with various minor actinides and long-lived fission products confirm the real possibility and expediency of closing the nuclear fuel cycle.

Electrochemical Properties of Uranium and Zirconium on Solid and Liquid Electrodes in 3LiCl–2KCl Based Melts

Spent nuclear fuel (SNF) reprocessing for fast neutron reactors can be organized by employing high temperature pyrochemical methods. Fused salts and liquid metals or alloys are the working media in the pyrochemical processes. Electrochemical deposition and electrochemical separation of uranium from fission products is an important stage of pyrochemical SNF reprocessing. Reliable data on electrochemical properties of uranium and fission product elements are required for development and optimization of the pyrochemical processes. Uranium and zirconium have very close electrode potentials and separating uranium (fissile material) from zirconium (fission product) is an important and difficult to solve task in the pyrochemical reprocessing. This work devoted to studying electrochemical properties of zirconium and uranium employing cyclic voltammetry method in 3LiCl–2KCl eutectic based melts containing uranium and zirconium chlorides. Various types of electrodes were tested to determine uranium and zirconium deposition potentials aiming to finding the cathode material with the largest difference of uranium and zirconium potentials. Tungsten was used as a solid cathode, and liquid alloys based on Ga, Al, Zn, Sn and their mixtures were tried as low-melting liquid cathodes. Electrochemical measurements were made in temperature range of 450–800 0C. On the basis of the results of electrochemical measurements the possibility of separating uranium and zirconium employing liquid metal alloys was elucidated. Acknowledgements: This work was supported by the Russian Science Foundation (project No 17–73–20156).

Synthesis and Physical Property Characterisation of Spheroidal and Cuboidal Nuclear Waste Simulant Dispersions.

This study investigated dispersions analogous to highly active nuclear waste, formed from the reprocessing of Spent Nuclear Fuel (SNF). Non-radioactive simulants of spheroidal caesium phosphomolybdate (CPM) and cuboidal zirconium molybdate (ZM-a) were successfully synthesised; confirmed via Scanning Electron Microscopy (SEM), powder X-ray diffraction (PXRD) and Fourier transform infrared (FTIR) spectroscopy. In addition, a supplied ZM (ZM-b) with a rod-like/wheatsheaf morphology was also analysed along with titanium dioxide (TiO2). The simulants underwent thermal gravimetric analysis (TGA) and size analysis, where CPM was found to have a D50 value of 300 nm and a chemical formula of Cs3PMo12O40·13H2O, ZM-a a D50 value of 10 μm and a chemical formula of ZrMo2O7(OH)2·3H2O and ZM-b to have a D50 value of 14 μm and a chemical formula of ZrMo2O7(OH)2·4H2O. The synthesis of CPM was tracked via Ultraviolet-visible (UV-Vis) spectroscopy at both 25 °C and 50 °C, where the reaction was found to be first order with the rate constant highly temperature dependent. The morphology change from spheroidal CPM to cuboidal ZM-a was tracked via SEM, reporting to take 10 days. For the onward processing and immobilisation of these waste dispersions, centrifugal analysis was utilised to understand their settling behaviours, in both aqueous and 2 M nitric acid environments (mimicking current storage conditions). Spheroidal CPM was present in both conditions as agglomerated clusters, with relatively high settling rates. Conversely, the ZM were found to be stable in water, where their settling rate exponents were related to the morphology. In acid, the high effective electrolyte resulted in agglomeration and faster sedimentation.

Applications of Diglycolamide Based Solvent Extraction Processes in Spent Nuclear Fuel Reprocessing, Part 1: TODGA

ABSTRACTOver the last decade there has been much interest in the applications of diglycolamide (DGA) ligands for the extraction of the trivalent lanthanide and actinide ions from PUREX high active raffinates or dissolved spent nuclear fuel. Of the DGAs, the N,N,N’,N’-tetraoctyldiglycolamide (TODGA) is the best known and most widely studied. A number of new actinide separation processes have been proposed based on extraction with TODGA. This review covers TODGA-based processes and extraction data, specifically focusing on how phase modifiers have been used to increase metal loading and thus enhance the operating process envelopes. Effects of third phase formation and the organic phase speciation are reviewed in this context. Relevant aspects of the extraction chemistry of important solvents (TODGA-modifier-diluent combinations) are described and their performances demonstrated by a consideration of the published flowsheet tests. It is seen that modifiers are successfully enabling the use of TODGA in actinide separation processes but to date the identification and testing of suitable modifiers has been rather empirical. There is a growing understanding of the fundamental chemistry occurring in the organic phase and how that affects extractant speciation and metal loading capacity but studies are still needed if TODGA-based flowsheets are to become an industrially deployable option for minor actinide (MA) recovery processes.

Advantages of Production of New Fissionable Nuclides for the Nuclear Power Industry in Hybrid Fusion-Fission Reactors

A concept of a large-scale nuclear power engineering system equipped with fusion and fission reactors is presented. The reactors have a joint fuel cycle, which imposes the lowest risk of the radiation impact on the environment. The formation of such a system is considered within the framework of the evolution of the current nuclear power industry with the dominance of thermal reactors, gradual transition to the thorium fuel cycle, and integration into the system of the hybrid fusion-fission reactors for breeding nuclear fuel for fission reactors. Such evolution of the nuclear power engineering system will allow preservation of the existing structure with the dominance of thermal reactors, enable the reprocessing of the spent nuclear fuel (SNF) with low burnup, and prevent the dangerous accumulation of minor actinides. The proposed structure of the nuclear power engineering system minimizes the risk of radioactive contamination of the environment and the SNF reprocessing facilities, decreasing it by more than one order of magnitude in comparison with the proposed scheme of closing the uranium–plutonium fuel cycle based on the reprocessing of SNF with high burnup from fast reactors.