Haut PDF Zirconium(IV) electrochemical behavior and electrorefining in molten fluoride salts

Zirconium(IV) electrochemical behavior and electrorefining in molten fluoride salts

Zirconium(IV) electrochemical behavior and electrorefining in molten fluoride salts

(black). Inset. Variation of the peak current density (◊) and the peak potential (□) versus the square root of the potential scan rate. Working electrode: Ag (S = 0.36 cm 2 ); auxiliary electrode: glassy carbon; comparison electrode: Pt Reduction of Zr(IV) occurs in one step Quasi-reversible system

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Zirconium(IV) electrochemical behavior and electrorefining in molten fluoride salts

Zirconium(IV) electrochemical behavior and electrorefining in molten fluoride salts

A first step consists in investigating the feasibility of the electrochemical recovery of Zr metal in fluoride media. Thus, the present work focused on the electrochemical behavior study of zirconium in molten fluoride using transient electroanalytical techniques, e.g. cyclic voltammetry, square wave voltammetry, and chronopotentiometry. These different technics allowed to understand the zirconium reduction mechanism by determining the number of exchanged electrons and assessed the thermochemical properties of Zr in the salt (diffusion coefficient, etc.). Zirconium electrocrystallisation process was also investigated by chronoamperometry and cyclic voltammetry. This set of data is of first importance in order to estimate the further feasibility of the process.
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Zirconium(IV) electrochemical behavior in molten LiF-NaF

Zirconium(IV) electrochemical behavior in molten LiF-NaF

mixture on silver electrode using chronoamperometry. In these conditions, zirconium nucleation was demonstrated to be pro gressive, which means that the nuclei are formed continuously. In addition, the shape of zirconium nuclei is hemispherical. Their growth is achieved in the three dimensions and is limited by the diffusion of Zr(IV) ions. Finally, the temperature and the over voltage do not have any in fluence on the zirconium nucleation mode, which remains progressive. However, the overvoltage in fluences the nucleation rate: the increase in the overvoltage gen erates an increase in the number of nuclei formed.
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Electrochemical study of the Eu(III)/Eu(II) system in molten
fluoride media

Electrochemical study of the Eu(III)/Eu(II) system in molten fluoride media

reliable data on the electrochemical behaviour and thermodynamic data of both An and fission products dissolved in molten salt solvents. Among the fission products, the lanthanides (Ln) such as Ce, Sm, Nd, Eu, Gd, Dy have very comparable chemical properties with the An, which makes the An-Ln separation very difficult. Furthermore, in order to ensure a longer lifetime of the solvents used in these processes, the decontamination of the fluoride salt from the Ln is also a key issue. On top of these P&T concepts, the Molten Salt Reactor (one of the six nuclear reactors concept evaluated in the frame of the Generation IV Forum), which should operate with fluoride molten salts, requires an online reprocessing in order to remove fission products and particularly Ln [3].
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Zirconium(IV) electrochemical behavior in molten LiF-NaF

Zirconium(IV) electrochemical behavior in molten LiF-NaF

at 750  C. Fig. 13 a shows a photograph of the obtained deposit, where it can be noted that the coating is covered by cooled salt. However, some parts of the deposit are clearly visible and indicate a metallic dendritic structure. In order to con firm that metallic Zr was obtained, a cross section of the cathode was polished and charac terized by SEM EDS analysis (see Fig. 13 b). On this micrograph, the Zr coating is adherent and made of a dense layer (~10 m m of thickness) with some dendrites. The feasibility of metallic Zr electrodeposition is thus demonstrated.

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Feasibility of the electrochemical way in molten fluorides for separating Thorium and Lanthanides and extracting Lanthanides from the solvent.

Feasibility of the electrochemical way in molten fluorides for separating Thorium and Lanthanides and extracting Lanthanides from the solvent.

Abstract: An alternative way of reprocessing nuclear fuel by hydrometallurgy could be using treatment with molten salts, particularly fluoride melts. Moreover, one of the six concepts chosen for GEN IV nuclear reactors (Technology Roadmap – http://gif.inel.gov/roadmap/) is the Molten Salt Reactor (MSR). The originality of the concept is the use of molten salts as liquid fuel and coolant. During the running of the reactor, fission products, particularly lanthanides, accumulate in the melt and have to be eliminated to optimise reactor operation.
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On the use of electrochemical techniques to monitor free oxide content in molten fluoride media

On the use of electrochemical techniques to monitor free oxide content in molten fluoride media

The method consists in measuring the reduction peak intensity obtained on the SWV voltammograms which was proved to be linear with the content of each species. Then, the content of these species in a given glass solution was determined using a calibration method. The accuracy of the results obtained by this methodology leaded us to build a probe allowing an on-line titration of the Fe (III) /Fe (II) ratio in industrial glasses by measuring the ratio of SWV peaks of these species [7]. Similar results were obtained by our Laboratory for the titration of Niobium species in molten fluoride (Nb V and Nb IV ) during the electrodeposition process of Nb metal [8]. In addition, the method was shown to be sensitive to determine the oxide ions content in the molten salts media [9].
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Direct electrochemical reduction of solid uranium oxide in molten fluoride salts

Direct electrochemical reduction of solid uranium oxide in molten fluoride salts

1. Introduction Advanced nuclear fuel cycles are under development worldwide in order to minimise the amount of high radiotoxic waste gener- ated by nuclear power plants operation. New technologies have to be economically competitive, environmentally safe and resistant to proliferation. Within most of the developed future fuel cycles, recycling of actinides (Cm, Pu, Am, Np) from spent nuclear fuel is required due to their significant impact on its radiotoxicity and pyrochemical methods represent one of the promising options to fulfil this task.

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Lanthanides extraction processes in molten fluoride media. Application to nuclear spent fuel reprocessing

Lanthanides extraction processes in molten fluoride media. Application to nuclear spent fuel reprocessing

A B S T R A C T This paper describes four techniques of extraction of lanthanide (Ln) elements from molten salts in the general frame of reprocessing nuclear wastes; one of them is chemical: the precipitation of Ln ions in insoluble compounds (oxides or oxyfluorides); the others use electrochemical methodology in molten fluorides for extraction and measurement of the progress of the processes: first electrodeposition of pure Ln metals on an inert cathode material was proved to be incomplete and cause problems for recovering the metal; electrodeposition of Ln in the form of alloys seems to be far more promising because on one hand the low activity of Ln shifts the electrodeposition potential in a more anodic range avoiding any overlapping with the solvent reduction and furthermore exhibit rapid process kinetics; two ways were examined: (i) obtention of alloys by reaction of the electroreducing Ln and the cathode in Ni or preferably in Cu, because in this case we obtain easily liquid compounds, that enhances sensibly the process kinetics; (ii) codeposition of Ln ions with aluminium ions on an inert cathode giving a well defined composition of the alloy. Each way was proved to give extraction efficiency close to unity in a moderate time.
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Electrochemical reduction of Gd(III) and Nd(III) on reactive cathode material in molten fluoride media

Electrochemical reduction of Gd(III) and Nd(III) on reactive cathode material in molten fluoride media

extracting lanthanides exhibits many advantages, in particular that of avoiding the addition of any chemicals to the melt to be purified. The molten salts used for this study were alkaline fluorides, chosen for their high chemical stability and their favourable neutronic properties. Electrodeposition can occur in two different ways based on two kinds of cathode material: (i) The cathode material is inert, which means that the Ln ions lose their charge to yield pure Ln metal at more cathodic potentials than the Ln equilibrium potential in the melt, given by the Nernst equation, with the activity of the Ln(0) phase equal to one. In earlier works [2, 3], we demonstrated that only incomplete extraction can be expected. The first reason for this is that the equilibrium potential of Ln ions is too negative, i.e. too close to the solvent discharge potential, to reach the efficiency predicted by thermodynamics. Secondly, electrodeposited Ln metal is dendritic and falls to the bottom of the bath in the form of fine particles, difficult to recover.
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Electrochemical extraction of europium from molten fluoride media

Electrochemical extraction of europium from molten fluoride media

1. Introduction In France, alternative solutions for radioactive wastes manage- ment are under examination since 1991, in the frame of Partitioning and Transmutation (P&T) concepts aiming at significantly reduc- ing the amount of radiotoxic nuclear waste at the end of the fuel cycle [1] . In these concepts, the most radiotoxic elements (Pu, Am, Cm) are to be burned by transmutation and for this purpose an efficient separation of actinides (An) from lanthanides (Ln) is required. The present hydrometallurgical PUREX process allows extraction of uranium and plutonium for further reuse in MOX fuels. But, concerning the new types of fuels for future genera- tion reactors (i.e. metallic, nitride, carbide, CER-MET) of fuels with inert matrices, a complete recovery of all An is desired, and aque- ous media have not yet proven to be the adequate solvents to dissolve the fuels. Another route could consist of using pyrochem- ical processes involving molten salts, which are currently studied as alternative media due to their good physico-chemical proper- ties (as solvatation for example), but also for some advantages as a faster reprocessing with much shorter cooling times of the fuel, compactness of the reprocessing process yielding the possibility of direct connection of the reprocessing unit with the reactor, etc. Recent progresses have been realised for An extraction in molten salts using extractive reduction [2] and electrorefining [3] . After
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Direct electroreduction of oxides in molten fluoride salts

Direct electroreduction of oxides in molten fluoride salts

according to: M x O y (s) + 2ye − = xM(s) + yO 2− (1) 2O 2− + C(s) = CO 2 (g) + 4e − (2) The FFC process has been intensively studied worldwide, exclu- sively in molten chloride salts (LiCl or CaCl 2 ), for different purposes: high purity Si production [2] , spent nuclear fuel processing (reduc- tion of rare earth oxides [3] , UO 2 [4] , U 3 O 8 [5] , MOX [6] , spent fuel [7] , etc.), pure metal or alloy production (reduction of Nb 2 O 5 , Fe 2 O 3 , NiO–TiO 2 , etc. [8–10] ). No data are available in the literature for the electrochemical reduction of SnO 2 , TiO and Fe 3 O 4 whereas TiO 2
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Electrochemical behavior of neodymium in molten chloride salts

Electrochemical behavior of neodymium in molten chloride salts

HAL Id: cea-02437084 https://hal-cea.archives-ouvertes.fr/cea-02437084 Submitted on 13 Jan 2020 HAL is a multi-disciplinary open access archive for the deposit and dissemination of sci- entific research documents, whether they are pub- lished or not. The documents may come from teaching and research institutions in France or abroad, or from public or private research centers.

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Direct electrochemical reduction of solid uranium oxide in molten fluoride salts

Direct electrochemical reduction of solid uranium oxide in molten fluoride salts

c Institute for Transuranium Elements, Joint Research Centre, P.O. Box 2340, 76125 Karlsruhe, Germany a b s t r a c t The direct electrochemical reduction of UO 2 solid pellets was carried out in LiF–CaF 2 (+2 mass.% Li 2 O) at 850 °C. An inert gold anode was used instead of the usual reactive sacrificial carbon anode. In this case, oxidation of oxide ions present in the melt yields O 2 gas evolution on the anode. Electrochemical charac- terisations of UO 2 pellets were performed by linear sweep voltammetry at 10 mV/s and reduction waves associated to oxide direct reduction were observed at a potential 150 mV more positive in comparison to the solvent reduction. Subsequent, galvanostatic electrolyses runs were carried out and products were characterised by SEM-EDX, EPMA/WDS, XRD and microhardness measurements. In one of the runs, ura- nium oxide was partially reduced and three phases were observed: nonreduced UO 2 in the centre, pure metallic uranium on the external layer and an intermediate phase representing the initial stage of reduc- tion taking place at the grain boundaries. In another run, the UO 2 sample was fully reduced. Due to oxy- gen removal, the U matrix had a typical coral-like structure which is characteristic of the pattern observed after the electroreduction of solid oxides.
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Behavior and Impact of Zirconium in the Soil–Plant System: Plant Uptake and Phytotoxicity

Behavior and Impact of Zirconium in the Soil–Plant System: Plant Uptake and Phytotoxicity

Zr presents different oxidation states (varies from +2 to +4), +4 being the form of the predominant stable valence; bonding with oxygen is the prevailing and most common reaction for Zr (Kabata-Pendias 1993 ). The Zr crustal abundance ranges from 20 to 500 mg/kg, and its aqueous chemistry is dominated by the quadrivalent oxidation state (valence electron con fi guration 4 d 2 5 s ¢ ) (Ryzhenko et al. 2008 ) . The lower oxidation states of Zr (0, I, II, and III) occur only in nonaqueous solvents and fused salts ( Cotton and Wilkinson 1980 ). Due to high ionic potential (22.54 e 2 /Å), Zr is the most polarizing among the heavier transition and post-transition quadriva- lent cations. The extent of hydrolysis and polymer stoichiometry depends on the nature of the ionic media (Davydov et al. 2006 ) , with tetramer and trimer forms being the most common stoichiometry. Moreover, hydrolysis and polymerization reactions dominate in the presence of high fi eld strength cations that are capable of rupturing H–O bonds. Similarly, hydrolysis and polymerization is promoted in alka- line solutions and with increasing temperatures.
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Behavior of uranium and its surrogates in molten aluminosilicate glasses in contact with liquid metals

Behavior of uranium and its surrogates in molten aluminosilicate glasses in contact with liquid metals

To achieve good homogenization, glasses are usually milled and melted several times [9,11,20] . A standard remelting procedure was in- troduced with an intermediate milling step. Glass samples were melted once (C0 glass), twice (C5 glass) or three times (C6 glass) at 1250 °C ( Table 2 ). The hafnium concentration gradient has almost vanished after the second melting ( Fig. 3 c). Hafnium concentration in the C5 glass is very close to the targeted value. Nevertheless, few hafnium crys- tals still stay at the bottom. After the third melting, C6 glass is homoge- neous in concentration and only one hafnium crystal is found in the whole sample ( Fig. 4 f). Bubbles are observed in both C5 and C6 glasses because of introduced air by the milling step. The milling step appears to be essential to homogenize the glass. This method (successive milling and remelting steps) is a convenient way of “stirring” the melt, and thus of improving dissolution kinetics, when thermal treatments can only be realized under static conditions [6,24] .
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Comparative study on the chemical stability of Fe 3 O 4 and NiFe 2 O 4 in molten salts

Comparative study on the chemical stability of Fe 3 O 4 and NiFe 2 O 4 in molten salts

L’archive ouverte pluridisciplinaire HAL, est destinée au dépôt et à la diffusion de documents scientifiques de niveau recherche, publiés ou non, émanant des établissements d’enseignemen[r]

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Creep flow and fracture behavior of the oxygen-enriched alpha phase in zirconium alloys

Creep flow and fracture behavior of the oxygen-enriched alpha phase in zirconium alloys

1. Introduction The hexagonal close-packed alpha ( α ) phase of Group 4 elements ‒titanium, zirconium and hafnium‒ can dissolve up to 30 at.% interstitial oxygen [1-3]. It exhibits a high chemical affinity for oxygen and in oxidizing environments, an oxygen-stabilized α phase easily forms at high temperature from the body centered cubic β phase [4], leading to e.g. the so-called “alpha-case” in Ti alloys. Such high amounts of oxygen deeply alter the mechanical behavior of the α phase and lead to a progressive transition of its mechanical behavior from that of a metal to that of an oxide. In Ti alloys, this phase is very detrimental to ductility, fatigue and creep lifetime up to 600°C [5]. The oxygen-enriched α phase of Zr alloys, called α (O), is also brittle at room temperature from 0.5 wt.% oxygen [6]. The high-temperature mechanical behavior of oxygen-enriched Zr alloys has been reported from compression tests and tensile creep tests [7-9]. A significant increase in the compressive steady-state flow stress with increasing oxygen content has been reported, in a linear [7] or an exponential manner [9], for test temperatures between 750°C and 1200°C and for amounts of oxygen up to 2 wt.%. From tensile creep tests on Zircaloy-2 and Zircaloy-4 enriched in oxygen, similar oxygen-induced exponential strengthening of the α phase was reported between 700 and 1400°C for contents up to 1.5 wt.% [10,11]. The following empirical relationship has been proposed to describe such effect:
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Electrochemical behaviour of thorium(IV) in molten LiF–CaF2 medium on inert and reactive electrodes

Electrochemical behaviour of thorium(IV) in molten LiF–CaF2 medium on inert and reactive electrodes

In this work, the behaviour of liquid Bi, which has a low melting point (271 ◦ C) and a high density (10.05) at melting temperature, was investigated. A specific technical change of the experimental set-up, detailed in Fig. 12 , was needed in order to operate with a liquid pool as work- ing electrode: the bismuth was placed in a boron nitride (MCSE, quality AX05 oxide free) compartment with an open window in the BN wall immersed in the melt. The electrical contact was obtained using a molybdenum wire, insulated with boron nitride, immersed in the liquid.
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Neodymium electrowinning into copper-neodymium alloys by mixed oxide reduction in molten fluoride media

Neodymium electrowinning into copper-neodymium alloys by mixed oxide reduction in molten fluoride media

- low electrical conductivity ( s 650 ! C,0,2barO2 = 1,45 % 10 "6 V "1 cm "1 ) [33] ; - unfavourable Pilling-Bedworth coefficient (V Nd /V Nd2O3 = 0.89). Pilling–Bedworth rule considerations [40] , as illustrated by Li et al. [41] and Gibilaro et al. [42] , indicate that during oxide reduction, if the molar volume of the formed metal Vm is smaller than the molar volume of the oxide Vo, the metal obtained is porous enough to allow the molten salt electrolyte accessing the underlying oxide. For neodymium, the metal to oxide molar volume ratio is V Nd /V Nd 2 O 3 = 0.89, meaning that volume
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