Pyrochemical Separation Research Articles

Thermodynamics and the Separation Factor of Dy/U Couple on Ga and Ga–Al Electrodes in Fused LiCl–KCl Eutectic

The radiological danger of spent nuclear fuel (SNF) mainly depends on the presence of minor actinides (Np, Am, Cm) and certain long-lived fission products (Cs, I, Tc). If these elements are effectively separated from SNF and converted into short-lived or stable isotopes, the radiological danger of SNF storage can be significantly reduced. Pyrochemical reprocessing of spent nuclear fuel represents a promising alternative to the existing hydrometallurgical technology (solvent extraction PUREX-process) with subsequent transmutation of long-lived radionuclides. Pyrochemical separation methods in molten salts have some potential advantages over hydrometallurgical processes for separating actinides from fission products present in irradiated fuels. High radiation stability of molten salts and liquid metals as well as absence on neutron moderators make possible reprocessing highly active fuel after a short cooling time.In this study the principal thermodynamic properties and the separation factor of U/Dy couple were determined on gallium and gallium-aluminum electrodes in molten LiCl–KCl eutectic at 723–823 K using open-circuit potentiometry.The equilibrium electrode potentials of metallic dysprosium, uranium and their alloys with Ga and Ga-Al eutectic in molten LiCl–KCl–MeCl3 solutions were determined by open-circuit potentiometry. The potential-time dependencies of Me–Ga–Al and Me–Ga alloys (Me = Dy, U) after cathodic polarization are presented in Fig. 1.The electrochemical behavior of the alloys is characterized by the apparent standard potential of the alloys, . Temperature dependencies of the apparent standard potentials of Me(Ga–Al) and Me(Ga) alloys were fitted to the following expressions using the Origin Pro software (ver. 7.5):E** Dy(Ga-Al) = –(2.705±0.009) + (3.5±0.1) ·10–4∙T ±0.005 V [1]E** U(Ga-Al) = –(2.688±0.013) + (6.4±0.1) ·10–4∙T ±0.005 V [2]E** Dy(Ga) = –(2.726±0.007) + (3.4±0.2) ·10–4∙T ±0.009 V [3]E** U(Ga) = –(2.565±0.004) + (4.2±0.1) ·10–4∙T ±0.003 V [4]Activity coefficients of Dy and U in liquid Ga and Ga–Al alloys were calculated from the following equation:log(γMe(alloy)) = [nF(E* Me(III)/Me – E** Me(alloy))]/2.303RT [5]Values of Dy(III)/Dy and U(III)/U equilibrium electrode potentials were taken from our previous works (1, 2).E* Dy(III)/Dy = –(3.401±0.009) + (6.2±0.1) ·10–4∙T ±0.007 V [6]E* U(III)/U = –(3.051±0.006) + (7.6±0.1) ·10–4∙T ±0.003 V [7]Activity coefficients of solid U and Dy in liquid Ga and Ga–Al alloys as a function of temperature were fitted to the following equations:log(γDy(Ga)) = 4.24 – 10224/T ±0.12 [8]log(γDy(Ga-Al)) = 4.09 – 10542/T ±0.12 [9]log(γU(Ga)) = 5.15 – 7361/T ±0.14 [10]log(γU(Ga-Al)) = 1.16 – 4962/T ±0.17 [11]Partial excess Gibbs free energy change of Dy and U in liquid Ga and Ga–Al alloys was calculated according to eq. [13] and previously obtained expressions [8–11]:ΔGex Me(alloy) = ΔHMe(alloy) – TΔSex Me(alloy) [12]ΔGex Me(alloy) = 2.303RTlog(γMe(alloy)) [13]ΔGex Dy(Ga) = –195.4 + 0.081· T ±2.8 kJ/mol [14]ΔGex Dy(Ga-Al) = –201.5 + 0.078· T ±2.8 kJ/mol [15]ΔGex U(Ga) = –140.7 + 0.098· T ±3.1 kJ/mol [16]ΔGex U(Ga) = –140.7 + 0.098· T ±3.1 kJ/mol [17]When the separation of lanthanides and actinides in the spent nuclear fuel is studied, the effectiveness of electrochemical separation methods is usually described by the value of the distribution or separation factor. The separation factor for uranium and dysprosium on liquid gallium and gallium–aluminum based alloys was calculated using the following equation:log(θ) = [(n–m)FE + mFE** Dy(alloy) – nFE** U(alloy)]/2.303RT [18]The temperature dependence of the apparent standard potentials of dysprosium [1, 3] and uranium [2, 4] alloys were used to derive the following expressions for the separation factor of Dy/U couple:log(θ(Ga)) = 1.21 + 2438/T ±0.04 [19]log(θ(Ga-Al)) = 4.39 + 275/T ±0.6 [20]The separation factor of Dy/U couple in molten LiCl–KCl eutectic calculated according to the above formula indicates that dysprosium will be concentrated in the molten salt phase and uranium in the liquid metal phase. Analysis of experimental data shows that these systems are interesting for application in pyrochemical methods of nuclear waste disposal.The study was carried out with the financial support of the RFBR in the framework of the scientific project # 20-03-00743.References А. Novoselova, V. Smolenski, and V. A. Volkovich, J. Electrochem. Soc. 167, 112510 (2020). A. Novoselova and V. Smolenski, J. Radioanal. Nucl. Chem., 326, 621 (2020). Fig. 1. The potential-time dependencies of Me–Ga and Me–Ga–Al alloys (Me = Dy, U) after cathodic polarization in LiCl–KCl–MeCl3 solutions at 726 K. Polarization current 50–150 mA, duration 20–40 s. [U(III)] = 4.7·10-2 mol·kg-1. [Dy(III)] = 5.8·10-2 mol·kg–1.1 – The equilibrium potential of U(Ga-Al) alloy. 2 – The equilibrium potential of U(Ga) alloy.3 – The equilibrium potential of Dy(Ga-Al) alloy. 4 – The equilibrium potential of Dy(Ga) alloy. Figure 1

Laser-Induced Breakdown Spectroscopy (LIBS) in a Novel Molten Salt Aerosol System.

In the pyrochemical separation of used nuclear fuel (UNF), fission product, rare earth, and actinide chlorides accumulate in the molten salt electrolyte over time. Measuring this salt composition in near real-time is advantageous for operational efficiency, material accountability, and nuclear safeguards. Laser-induced breakdown spectroscopy (LIBS) has been proposed and demonstrated as a potential analytical approach for molten LiCl-KCl salts. However, all the studies conducted to date have used a static surface approach which can lead to issues with splashing, low repeatability, and poor sample homogeneity. In this initial study, a novel molten salt aerosol approach has been developed and explored to measure the composition of the salt via LIBS. The functionality of the system has been demonstrated as well as a basic optimization of the laser energy and nebulizer gas pressure used. Initial results have shown that this molten salt aerosol-LIBS system has a great potential as an analytical technique for measuring the molten salt electrolyte used in this UNF reprocessing technology.

High resolution ICP-OES analysis of neptunium-237 in samples from pyrochemical treatment of spent nuclear fuel

A straightforward and reliable ICP-OES procedure for 237Np analysis of samples potentially originating from spent nuclear fuel has been developed. In contrast to most other analytical techniques employed for the determination of 237Np in radioactive samples, 237Np was analyzed without prior separation from the residual elements present in the analyte solution. External calibration of the ICP-OES response was accomplished using a certified 237Np stock solution that was additionally characterized in-house using ICP-MS as well as γ-spectrometry. Thirteen prominent 237Np emission wavelengths were tested for their suitability in terms of sensitivity, peak shape, occurrence of spectral interferences and stability of spectral background. The investigated 237Np emission wavelengths revealed complex ICP-OES spectra with either broad peaks having widths of up to 30pm or splitting of the 237Np signal into multiple peaks, i.e. multiplets. Compared to previous ICP-OES investigations, 237Np detection limits (LOD) were lowered by at least one order of magnitude in the current study. The most intensive 237Np emission line at λ=382.92nm providing the lowest LOD of 3.1μgkg−1, however, was of limited practical use because severe spectral interference from Nd (a representative fission product with a significant content in spent fuel) at λ=382.914nm overlapped with the 237Np signal. Additional detailed interference studies helped identify potential complications during 237Np analysis caused by concomitant elements (Nd, Pu, Th, U) in the analyte solution. Ultimately, emission signals at λ=410.84nm, λ=429.09nm, and λ=456.04nm were used to analyze selected salt and deposit samples from pyrochemical separation experiments. ICP-OES Np results were cross-validated using sector field ICP-MS, largely confirming each other, and thus underpinning the validity of the developed ICP-OES procedure. The successful use of two independent analytical techniques for all analyses helped to prove the quality of the achieved results.

Possible routes for pyrochemical separation: Focus on the reductive extraction in fluoride media

Molten salts (MSs) such as fluoride or chloride salts at high temperature (400–800 °C) are solvents known for their high solvation power and electroactivity range. Rare earths, lanthanides, actinides, and refractory metals can be dissolved, treated, and purified in MSs. The properties of these solvents are particularly interesting for nuclear spent-fuel reprocessing. The pyrochemical separation and extraction of solutes can be performed using several methods taking into account the effects of redox and/or acidity. This paper is focused on the reductive extraction method performed by contacting a liquid metal (LM) containing reductive species and an MS. The analytical model developed to calculate the efficiency of such a method is detailed in this paper. To apply this model, one essential point is the establishment of a database related to the redox and solvation properties of solutes in MSs. The approach retained to propose a database based on the analysis of both thermochemical data of pure compounds and experimental measurements reported in the literature is described in this paper in the case of lanthanides in fluoride MSs. The use of the database to calculate efficiency as a function of process parameters is given in this paper as well as the comparison between two reducing agents considered.

A study on the electrochemical deposition behavior of uranium ion in a LiCl–KCl molten salt on solid and liquid electrode

Electrochemical behaviors of uranium ion (U3+) from a molten salt electrolyte to metal solid and liquid cathodes were studied for an electrorefinning recovery in pyrochemical separation. Cyclic voltammetry (CV) and electrochemical impedance spectroscopy (EIS) were carried out to investigate the interfacial processes related with cathodic reduction of uranium ion. Electrodeposition at solid Mo and liquid Cd electrodes are discussed as well as mass transfer in liquid Cd metal and LiCl–KCl molten salt phases. The reduction of U3+/U at both of the electrodes preceded a (quasi-)reversible reaction with exchange of three electrons. The difference of the standard formal potential at both solid and liquid electrodes was found which may be attributable to lowering activity of U in liquid Cd phase. The impedance data measured at different potentials were fitted to the equivalent circuits for each electrode. In case of solid electrode, the impedance spectrum was related with surface diffusion followed by lattice formation as well as charge transfer at low cathodic potential. At high cathodic potential, the effect of surface relaxation induced by U deposition on the surface of Mo electrode was found. In case of liquid electrode, the atomic diffusion of reduced U metal in liquid Cd phase appeared before the saturation of Cd by U. Over the solubility limit, the formation of U dendrite on the Cd surface made the impedance magnitude decreased by increasing the electrode surface area.

Assessment of the Complete Core of the Reference Pyrochemical Process, Developed by the CEA

Pyrochemical separation technology using high-temperature molten salts presents a potential interest for an overall separation strategy. Pyrochemical R&D program, launched at CEA, aims at demonstrating the feasibility of a grouped separation of actinides with sufficient decontamination. The process is based on two liquid/liquid extraction steps: a selective extraction in liquid aluminium from molten fluoride and the actinide back-extraction in molten chloride. The present work focuses on a complete study of this core of process. The reductive extraction has been assessed as function of the salt composition wile the back-extraction step focused on the behaviour of the actinides.

Electrochemistry of Selected Lanthanides in FLiBe and Possibilities of their Recovery on Reactive Electrode

Among other applications, electrochemical based separation of actinides and lanthanides from molten salt media seems to be suitable method for reprocessing of spent nuclear fuel in proposed future types of nuclear reactors such as Molten Salt Reactor. One of the most interesting features of the MSR concept is a circulation of liquid molten fluoride fuel mixture. It allows the fuel to be continuously (“on-line”) reprocessed as it is necessary for continuous reactor operation. The pyrochemical separation processes seem suitable for the “on-line” reprocessing technology and electrochemical separations are one of the most promising methods for separation of fissile material and fission products. This work is focused on electrochemical behaviour of several lanthanides (Sm, Gd) in FLiBe on inert (Mo, Mo) and reactive (Ni) electrodes. Electrochemical behaviour was studied by cyclic voltammetry and in the case of gadolinium, electrolytic product was analysed by SEM-EDX analysis. Specific interactions between rare earth elements and polymer-like structure of Be-based melt were taken into account and experimental results are interpreted. Evaluation of possibilities of lanthanides recovery on reactive electrode (Ni) is presented.

ACSEPT—Partitioning technologies and actinide science: Towards pilot facilities in Europe

Actinide recycling by separation and transmutation is considered worldwide and particularly in several European countries as one of the most promising strategies to reduce the inventory of radioactive waste and to optimise the use of natural resources. With its multidisciplinary consortium of 34 partners from 12 European countries plus Australia and Japan, the European Research Project ACSEPT (Actinide reCycling by SEParation and Transmutation) aims at contributing to the development of this strategy by studying both hydrometallurgical and pyrochemical partitioning routes. ACSEPT is organised into three technical domains: (i) Considering technically mature aqueous separation processes, ACSEPT works to optimise and select the most promising ones dedicated either to actinide partitioning (for the heterogeneous recycling of actinides in ADS target or specific actinide bearing blanket fuels in fast reactor) or to grouped actinide separation (for the homogeneous recycling of the actinides in fast reactor fuels). In addition, dissolution and conversion studies are underway taking into account the specific requirements of these specific fuels. (ii) Concerning pyrochemical separation processes, ACSEPT focuses on the enhancement of the two reference cores processes selected within FP6-EUROPART. R&D efforts are also devoted to key scientific and technical issues compulsory to set up a complete separation process (head-end steps, salt treatment for recycling and waste management). (iii) By integrating all the experimental results in engineering and system studies, both in hydro and pyro domains, ACSEPT will deliver relevant flowsheets and recommendations to prepare for future demonstrations at a pilot level. After more than two years of work, significant progress was achieved in process development with the demonstration of the SANEX and innovative SANEX flowsheets. Chemical systems were selected for GANEX and are under study. In addition, efforts were made to increase collaborations, mutualise and homogenise procedures and share good practices. Based on these assessments, it is now time to look at the future challenges to overcome. A training and education program is implemented to share the knowledge among the partitioning community, present and future generations of researchers. Specific attention is paid to the funding of post-doctorate fellowships, two having been appointed respectively at the end of 2008 and at the end of 2009. Through this training and education programme, the first ACSEPT International Workshop was organised last March in Lisbon, Portugal. It gave an emphasis to young researchers’ contributions (two thirds of the contributions) and allowed young scientists to meet and exchange with international recognised experts.

An investigation into the electrochemical recovery of rare earth ions in a CsCl-based molten salt

A CsCl-based melt, was used as a supporting electrolyte for a fuel cycle in pyrochemical separation, as it has a high solubility for lanthanide oxide. Cyclic voltammetry and square wave voltammetry were carried out to investigate the cathodic reduction of those rare earth ions. The results prove that the cathodic process of La(III) ions dissolved in a CsCl-based melt, with a one-step reduction La 3+ + 3e − = La, and is similar to those of other reports which have utilised LiCl–KCl or CaCl 2–KCl molten salt systems. However, for the Ce(III) ions that dissolved in a CsCl-based melt, there is a significant difference when compared with published literature as there are two reduction steps instead of the reported single step Ce 3+ + e − = Ce 2+ and Ce 2+ + 2e − = Ce. In order to explain the novel result, a detailed investigation was focused on the cathodic process of Ce(III) in a CsCl-based melt. The identification of the M–O (M = La, Ce) compounds that are stable in the electrolyte, as well as the determination of their solubility products, were carried out by potentiometric titration using an oxide ion sensor. Furthermore, the E-pO 2− (potential-oxide ion) diagram for the M–O stable compound was constructed by combining both theoretical and experimental data.

Electrochemical Determination of Gadolinium and Plutonium Solvation Properties in Liquid Gallium at High Temperature

Pyrochemical separation processes are considered to treat spent nuclear fuel and particularly to separate fission products from actinides. In order to estimate the efficiency and selectivity for various extraction processes based on a molten salt/solvent metal separation technique, we have to know the properties of the elements to be extracted in each solvent, notably their activity coefficients in the two phases. The classical way to measure the activity coefficient of an element in a liquid metal is to use a concentration cell whose the electromotive force is measured. This type of cell involves two electrodes: (a) the element investigated in its pure metallic form and (b) the element solvated in the solvent metal. The electrolyte used for this study is a chloride melt that contains the element under consideration as a solute. In this paper, an effort was made to measure activity coefficients in liquid metals by means of electrochemical techniques rather than a potentiometric technique. The experimental protocol was optimized by measuring the activity coefficient of gadolinium in liquid gallium (solvent metal) γ(Gd/Ga) at 530°C for several amounts of gadolinium in gallium, and log γ(Gd/Ga) was determined to be equal to −10.17 (mole fraction scale). Then, the temperature dependence of the activity coefficient was determined in the range of 535 to 630°C. It appears that log γ(Gd/Ga) varies linearly with the reciprocal value of T, thus following the theoretical variation. The electrochemical method was also performed to determine the activity coefficient of plutonium in liquid gallium at 560°C. The value of log γ(Pu/Ga) so obtained is equal to −8.04 (mole fraction scale). This value was confirmed using electrochemical and potentiometric measurements with a plutonium-saturated gallium electrode.

Proliferation resistance of PEACER system

It is considered that partitioning and transmutation (P&T) could reduce the burden of geological disposal of spent fuel. PEACER (Proliferation-resistant, Environment-friendly, Accident-tolerant, Continuable and Economical Reactor) system has been developed as a conceptual means of P&T of spent fuel in ROK. The PEACER system is based on the combination of pyrochemical separation process, i.e., pyroprocess and Pb–Bi cooled fast reactor. Proliferation resistance is one of key constraints in the deployment of advanced nuclear energy systems. This paper assesses proliferation resistance of the PEACER system.

Activity coefficients of plutonium and cerium in liquid gallium at 1073 K: Application to a molten salt/solvent metal separation concept

Activity coefficients in liquid metal and salt phases are important parameters for predicting the separation efficiency of reductive extraction or electrochemical pyrochemical processes. The electrochemical properties of Ce and Pu in gallium metal and chlorides media – CaCl2 and equimolar NaCl–KCl – have been studied at 1073K. Cyclic voltammetry and chronoamperometry show the thermodynamic feasibility of using gallium as solvent metal for pyrochemical processes involving Pu and Ce. The activity coefficient of Pu in liquid Ga (log(γPu,Ga)=−7.3±0.5) is deduced from the results and is a basis of assessing the potential for using liquid metals in pyrochemical separation of actinides and lanthanides. Evaluation of literature data for Al, Bi and Cd suggests that Ga is most favorable for selective separation of Pu from Ce near 1073K.

Concept of a small-scale electron accelerator driven system for nuclear waste transmutation: Part 2. Investigation of burnup

In our earlier paper the target optimization of an electron–neutron (e–n) converter based molten salt ADS concept (MSENC) was presented. In this paper the burnup in this molten salt ADS is investigated. The system is continuously fueled with TRUs obtained from spent nuclear fuel by means of pyrochemical separation, while fission products are continuously filtered out. Effect of MSENC operation on transuranium consumption, mass reduction and radiotoxicity is considered. It seems that while the amount of plutonium is reduced significantly, the higher actinides will build up, which results that radiotoxicity of nuclear waste will not be reduced considerably.

Role of pyro-chemical processes in advanced fuel cycles

Partitioning and Transmutation (P&T) of Minor Actinides (MAs) and Long-Lived Fission Products (LLFP) arising out of the back-end of the fuel cycle would be one of the key-steps in any future sustainable nuclear fuel cycle. Pyro-chemical separation methods would form a critical stage of P&T by recovering long-lived elements and thus reducing the environmental impact by the back-end of the fuel-cycle. This paper attempts to overview global developments of pyro-chemical process that are envisaged in advanced nuclear fuel cycles. Research and development needs for molten-salt electro-refining as well as molten salt extraction process that are foreseen as partitioning methods for spent nuclear fuels such as oxide, metal and nitride fuels from thermal or fast reactors; high level liquid waste from back-end fuel cycle as well as targets from sub-critical Accelerator Driven Sub-critical reactors would be addressed. The role of high temperature thermodynamic data of minor actinides in defining efficiency of recovery or separation of minor actinides from other fission products such as lanthanides will also be illustrated. In addition, the necessity for determination of accurate high temperature thermodynamic data of minor actinides would be discussed.

Transmutation of minor actinides and innovative fuel cycle concepts

Both physical and chemical properties of actinides show significant variations depending on the element from uranium to curium. The materials database on minor actinides (MA:Np, Am, Cm) is very limited. Even with the scanty database, however, americium and curium did not appear to make viable fuels in the ordinary sense. The solutions may be (1) to tolerate a significantly lower fuel performance whose penalty has to be recovered by an improved overall economy of the MA-fuel cycle, and/or (2) to make a fuel concept whose performance is less restricted by physical and chemical properties of MA. In such a difficult program like actinide burning, the technology components would have to be combined by a modular approach. The system has to be integrated as a whole, but a technological module in the system has to be made as such that can be replaced by the other modules. This type of approach would be only possible if we have a technology which forms a common basis to these “modules”. The advantage of taking the modular approach is to allow the system to evolve with time. The pyrochemical separation can form such a common basis.

Measurement of standard potentials of actinides (U,Np,Pu,Am) in LiCl–KCl eutectic salt and separation of actinides from rare earths by electrorefining

Pyrochemical separation of actinides from rare earths in LiCl–KCl eutectic–liquid metal systems has been studied. The electromotive forces of galvanic cells of the form, Ag|Ag(I), LiCl–KCl‖actinide(III), LiCl–KCl|actinide, were measured and standard potentials were determined for uranium, neptunium and plutonium to be −1.283 V, −1.484 V and −1.593 V (at 450°C vs. Ag/AgCl (1wt%–AgCl)), respectively. A typical cyclic voltammogram of americium chloride has two cathodic peaks, which suggests reduction Am(III)→Am(II) occurs followed by reduction of Am(II) to americium metal. Standard potential of Am(II)/Am(0) was estimated to be −1.642 V. Electrorefining experiments to separate actinides (U, Np, Pu and Am) from rare earths (Y, La, Ce, Nd and Gd) in LiCl–KCl eutectic salt were carried out. It was shown that the actinide metals were recovered on the cathodes and that americium was the most difficult to separate from rare earths. The actinide separation will be achieved by means of the combination of electrorefining with multistage extraction.

Equilibrium Distributions of Actinides and Fission Products in Pyrochemical Separation Systems, (II) LiCl-KCl/Cd System

Equilibrium distributions of actinides and fission products were determined in a LiCl-KCl/Cd system at 800–973 K. The redox potential of the system was controlled with the addition of reductant Li. Different distribution behaviors due to different group elements were observed. The group partitioning in this system was thus supported to be feasible. On reduction, however, the actinide and lanthanide elements were found to be less soluble in the Cd phase and to remain at the interface in the form of their solid intermetallics. A group partitioning process in which reductive extraction is combined with filtration was proposed on the basis of the present observations.

Equilibrium Distributions of Actinides and Fission Products in Pyrochemical Separation Systems, (I)

A thermodynamic model has been described for predicting the equilibrium distributions of solute elements in pyrochemical separation systems of molten salts and liquid metals. The model assumes that the solutes are present in the form of complex compounds in the salt phase and of intermetallic compounds in the metal phase. The thermodynamic data of these compounds are available in the literatures and also can be estimated theoretically. Some predictions have been made by the present model to discuss the feasibility of liquid metal extraction system for the recovery of platinum group elements from spent fuel.

Equilibrium distributions of actinides and fission products in pyrochemical separation systems. (I). Thermodynamic model.

A thermodynamic model has been described for predicting the equilibrium distributions of solute elements in pyrochemical separation systems of molten salts and liquid metals. The model assumes that the solutes are present in the form of complex compounds in the salt phase and of intermetallic compounds in the metal phase. The thermodynamic data of these compounds are available in the literatures and also can be estimated theoretically. Some predictions have been made by the present model to discuss the feasibility of liquid metal extraction system for the recovery of platinum group elements from spent fuel.

Equilibrium distributions of actinides and fission products in pyrochemical separation systems, (II). LiCl-KCl/Cd system.

Equilibrium distributions of actinides and fission products in pyrochemical separation systems, (II). LiCl-KCl/Cd system.