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1 This article was downloaded by: [ ] On: 22 March 214, At: 5:25 Publisher: Taylor & Francis Informa Ltd Registered in England and Wales Registered Number: Registered office: Mortimer House, Mortimer Street, London W1T 3JH, UK Journal of Nuclear Science and Technology Publication details, including instructions for authors and subscription information: Development of Electrochemical Separation Methods in Molten LiF-NaF-KF for the Molten Salt Reactor Fuel Cycle Pavel SOUČEK a, František LISÝ a, Radka TULÁČKOVÁ a, Jan UHLÍŘ a & Rudolf MRÁZ b a Nuclear Research Institute Řež plc, Husinec-Řež 13, Czech Republic b Institute of Chemical Technology Prague, Technická 195, , Praha 6, Czech Republic Published online: 5 Jan 212. To cite this article: Pavel SOUČEK, František LISÝ, Radka TULÁČKOVÁ, Jan UHLÍŘ & Rudolf MRÁZ (25) Development of Electrochemical Separation Methods in Molten LiF-NaF-KF for the Molten Salt Reactor Fuel Cycle, Journal of Nuclear Science and Technology, 42:12, , DOI: 1.18/ To link to this article: PLEASE SCROLL DOWN FOR ARTICLE Taylor & Francis makes every effort to ensure the accuracy of all the information (the Content ) contained in the publications on our platform. However, Taylor & Francis, our agents, and our licensors make no representations or warranties whatsoever as to the accuracy, completeness, or suitability for any purpose of the Content. Any opinions and views expressed in this publication are the opinions and views of the authors, and are not the views of or endorsed by Taylor & Francis. The accuracy of the Content should not be relied upon and should be independently verified with primary sources of information. Taylor and Francis shall not be liable for any losses, actions, claims, proceedings, demands, costs, expenses, damages, and other liabilities whatsoever or howsoever caused arising directly or indirectly in connection with, in relation to or arising out of the use of the Content. This article may be used for research, teaching, and private study purposes. Any substantial or systematic reproduction, redistribution, reselling, loan, sub-licensing, systematic supply, or distribution in any form to anyone is expressly forbidden. Terms & Conditions of access and use can be found at

2 Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 42, No. 12, p (December 25) ORIGINAL PAPER Development of Electrochemical Separation Methods in Molten LiF-NaF-KF for the Molten Salt Reactor Fuel Cycle Pavel SOUČEK 1;, František LISÝ 1, Radka TULÁČKOVÁ 1, Jan UHLÍŘ 1 and Rudolf MRÁZ 2 1 Nuclear Research Institute Řež plc, Husinec-Řež 13, Czech Republic 2 Institute of Chemical Technology Prague, Technická 195, Praha 6, Czech Republic (Received April 5, 25 and accepted in revised form October 4, 25) Electrochemical methods for the separation of fission products from fission material in molten fluoride salt media have been studied in the context of their application within the framework of the developed Molten Salt Reactor fuel cycle. The separation possibilities of selected actinides (U, Th) and lanthanides (Nd, Eu, Gd) in molten eutectic LiF- NaF-KF at 53 C were evaluated by means of cyclic voltammetry. The applicability of different electrochemical techniques is discussed with reference to the new results from this study, and a basic flow sheet for the Molten Salt Reactor fuel cycle is outlined. KEYWORDS: spent fuel reprocessing, pyrochemical partitioning, molten salt reactor, electrochemical separation, molten fluorides I. Introduction Corresponding author, Tel , Fax , liz@ujv.cz Fig. 1 The general scheme of the MSR fuel reprocessing system The need to minimize the impact of nuclear power on the environment, and the benefits from the saving of raw materials for nuclear fuel production, has engendered a raised interest in the development of advanced new generation reactor types. The Molten Salt Reactor (MSR) represents a very important technology that addresses both intended aims significant decrease of radiotoxic waste from Light Water Reactors (LWRs) originally appointed for final disposal, and effective utilization of raw materials for electricity and heat production with minimized waste generation. The present R&D deals with two main applications of the MSRs:. An energy producing breeding system that works under the 232 Th- 233 U fuel cycle, studied e.g. within the framework of Generation IV reactor types. 1). A transmuter that incinerates plutonium and minor actinides from reprocessed LWR spent fuel, simultaneously producing energy, and studied within the Czech national P&T programme in the framework of the project MSTR-SPHINX (Molten Salt Transmutation Reactor Spent Hot Fuel Incineration by Neutron Flux). 2) In both these cases of MSR technology, one of the most important prerequisites for its long-term operation ability is an integration of their fuel reprocessing, i.e. closing of the fuel cycle. The fuel dissolved in the carrier melt circulates rather quickly through the active zone of the reactor and according to the gradual burn-up, the content of fission products in the carrier melt increases. In addition to the feeding with the fresh fuel, the corresponding part of the main fuel stream should be reprocessed by removal of the required amount of the fission products to maintain suitable ratio between fission material and fission products in the reactor core. Due to the minimal cooling period of the fuel, the pyrochemical separation processes are perhaps the only suitable for the on-line reprocessing technology. A general scheme illustrating the main components of the MSR fuel reprocessing system is shown in Fig. 1. Although the feasibility of the reactor technology has been successfully demonstrated during the Molten Salt Reactor Experiment 3) in Oak Ridge National Laboratory, USA, the reprocessing unit for the real fuel has never been operated and it is necessary to continue with an investigation of the applicability of the pyrochemical separation methods. Electrochemical techniques represent one of the most promising methods for mastering the separation process, together with molten salt/liquid metal extraction techniques. In the field of pyrochemical electroseparation methods, recent attention has been mainly focussed on the investigation of systems utilizing a molten chloride medium. The carrier medium of the MSR consists of a molten fluoride salt mixture. For the specific application of the electroseparation method within the context of MSR fuel reprocessing, it would be very convenient to keep the medium based on the same chemical form during the whole MSR fuel cycle. Apart from the above-mentioned usage of electrochemical separation methods within the back-end of the MSR 117

3 118 P. SOUČEK et al. Fig. 2 The fundamental scheme of the MSTR fuel cycle fuel cycle, it is also possible to utilize these methods in the process of MSR fuel preparation. Within the MSTR- SPHINX project, the pyrochemical methods have been planned to be used for the LWR fuel reprocessing. With the exception of electrochemical methods from molten fluoride salt media, the Fluoride Volatility Method (FVM) has been developed in the laboratories of Nuclear Research Institute Řež plc (NRI). The use of the FVM is particularly effective for the removal of components that form volatile fluorides, mainly uranium, and for the conversion of fuel components into their fluoride form. 4) The formed non-volatile fluorides are dissolved in a suitable fluoride melt, and then subsequently processed by pyrochemical methods the combination of electrochemical separation and molten salt/liquid metal extraction has been considered. The fundamental scheme of the proposed MSTR fuel cycle is shown in Fig. 2. Spent fuel is fluorinated in the Fluoride Volatility process and non-volatile fluorides of Pu, MA and FP are sent to electroseparation unit. In this process residual uranium and FP are removed. MA are led into Molten Salt Transmutation Reactor in the form of liquid fuel. Circulating fuel is continually reprocessed to maintain a suitable ratio of MA vs. FP in molten fuel. A very similar flow sheet can be drawn for the case of the Th-U fuel cycle, where application of pyrochemical separation techniques is also advantageous for preparation of the fuel. Uranium-233 as a good fissile material arises from 233 Pa free decay but in the reactor core non-wanted reaction to 234 Pa occures, which decay does not result in 233 U. Moreover, 233 Pa is also a significant neutron poison. The specific problem in this fuel cycle is the separation of protactinium in addition to the removal of fission products from the circulating fuel. A scheme of the fuel cycle for MSR Th-U breeder is shown in Fig. 3. Before the designing of any particular flow sheet for the MSR fuel reprocessing, the basic thermodynamic data corresponding to the separated compounds in suitable molten fluoride media have to be determined, and the separation possibilities have to be appraised with regard to all the available techniques. The main objectives of the presented work are (i) the assessment of the separation possibilities for the selected actinides (U, Th) and lanthanides (Nd, Eu and Gd) in an eutectic mixture of LiF-NaF-KF (FLINAK), and (ii) to outline the separation process with respect to the conclusions from this study. Due to the lack of published relevant thermodynamic data, the properties needed for this assessment were determined experimentally by means of the Cyclic Voltammetry method. All measurements were taken with the same experimental set-up to enable the direct comparison of the collected data, while the conditions were kept as close as possible to the conditions that are supposed to occur within the real reprocessing system. II. Experimental 1. Experimental Set-up The eutectic mixture LiF-NaF-KF (FLINAK), melting point 454 C, was selected as the carrier electrolyte for the electrochemical studies. The chemical composition and the main properties of the melt FLINAK at the working temperature 53 C are given in Table 1. 5) The raw materials powders of LiF, NaF and KF were conditioned before the melt preparation by desiccation under vacuum. All electrochemical experiments were carried out in a laboratory nickel electrolyser under an inert atmosphere in a three-electrode set-up without any separation of electrode areas. A sheet with an active area of 2. cm 2 was used as the working electrode, and a glassy carbon crucible served as a container for the melt, which was connected as the counter electrode. A specially designed reference electrode based on the Ni/Ni 2þ red-ox couple 6) was developed in our workplace in the NRI and successfully used during the measurements. The instrumental equipment consisted of a high power potentiostat Wenking HP 96-2 coupled by an analogue JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY

4 Development of Electrochemical Separation Methods in Molten LiF-NaF-KF for the Molten Salt Reactor Fuel Cycle 119 Table 1 Chemical composition and selected properties of the carrier melt FLINAK at the working temperature 53 C Compound LiF NaF KF Content Density (gcm 3 ) Fig. 3 mol% wt% Selected properties of the melt at the temperature 53 C 5) Dynamic viscosity (cp) scan generator Wenking MVS 98, both made by Bank Elektronik Intelligent Controls GmbH, Germany. The apparati were connected to a PC by a control interface and Interface Card KPCI 312 made by Keithley Instruments, Inc., USA. The system enabled the computer control of the electrochemical processes and simultaneously provided data acquisition and processing by software CPC-DA developed by the potentiostat producer. 2. Realized Measurements The following measurements were realized with the goal to assess separation possibilities of actinides from lanthanides for the future application within the MSR reprocessing system. Electrochemical behaviour of pure carrier melt FLI- NAK and of uranium species, thorium and selected lanthanides in the FLINAK were studied. Uranium and thorium are the key components of the MSR fuel and after its reprocessing they can be re-used. The lanthanides represent a very important group of fission products, which has to be continuously removed from the MSR fuel during its long-term operation. Neodymium, The fundamental scheme of the MSR Th-U breeder fuel cycle Electrical conductance ( 1 cm 1 ) gadolinium and europium were selected to represent this group of elements in the assessment of separation possibilities within the MSR spent fuel reprocessing. Because of considered future application, the experimental condition were kept similar to the presumed real parameters. The Linear Potential Sweep Cyclic Voltammetry Method was used as the measurement technique with scan rates from.25 mv/s up to 2 mv/s. A scan rate of 5 mv/s was used in the majority of experiments. The potentials stated in the following text are related to the potential corresponding to the used reference electrode (Ni/Ni 2þ, 1 molar NiF 2 solution in the melt FLINAK). An atmosphere of highly pure argon gas (99.998%) and a constant working temperature of 53 C were maintained in the electrochemical cell during the recording of all measurements. III. Results Surface tension (Nm 1 ) Pure Melt FLINAK The electrochemical behaviour of the melt FLINAK was observed on molybdenum, platinum and graphite working VOL. 42, NO. 12, DECEMBER 25

5 12 P. SOUČEK et al Fig. 4 Typical voltammogram of pure melt FLINAK measured on a Mo working electrode, scan rate 5 mv/s Fig. 5 Voltammogram of the system FLINAK-UF 4 (1. mol%) on a Mo working electrode, scan rate 5 mv/s electrodes. Two main conclusions resulted from the study 1) the usable potential range of the melt ( potential window ) was determined, and 2) the characterization of any impurities, if present. The latter measurement was performed before each experiment to enable the correct interpretation of the measured signal by comparison with that from a previously measured blank experiment. The cathodic limit is characterized by reduction of the least stable compound constituting the melt, which is accompanied by deposition of the respective metal on the working electrode surface. This process is characterized by the sharp current density decrease and the peak in the positive area of current density corresponds to the inverse reaction-dissolution of deposited metal. The shape of the voltammogram depends on the working electrode material, but the evaluated potential limit is approximately the same for all studied materials. The particular value for the most frequently used electrode material, molybdenum, was found to be 2:5 V. The gradual current decrease observed from a potential of 1:6 V is probably caused by the formation of a surface alloy of metal reduced from the component forming the melt and the material of the working electrode. Even though the red-ox potential E(M þ /M ), corresponding to the deposition of pure metal, is more negative, its partial reduction can be shifted to a more positive potential area when the mentioned surface alloy is formed. The typical voltammogram of the pure melt FLINAK measured on a molybdenum working electrode is shown in Fig Electrochemical Characteristics of Uranium Species The electrochemical behaviour of uranium was studied separately in positive and negative potential areas by measurement of the system FLINAK-UF 4 (1. mol%). The measured voltammograms of UF 4 correspond well to theoretical assumptions and they are also in very good agreement with the data published by Clayton et al. 7) A two-step reaction mechanism was detected when the deposition of uranium metal started from a potential of 1:75 V. The first pair of peaks, at a potential of 1:2 V, indicates the first reduction step from U 4þ to U 3þ. The typical voltammogram of the system FLINAK-UF 4 measured on a molybdenum working electrode is shown in Fig. 5. The electrodeposition of uranium from the melt FLINAK was experimentally studied at the controlled potential of 1:9 V. No uranium metal was found on the cathode surface after 1 min electrolysis when the current density varied in the range 25 to 4 macm 2. The problem with uranium deposition resides probably in the following facts:. Uranium metal adheres poorly to the electrode surface.. There is a formation of non-wanted anodic products (U 5þ and U 6þ ions) during the deposition process. With regard to uranium metal deposition we recommend that the working electrode has a structured surface and/or is capable of forming an alloy with uranium. We additionally recommend separation of the electrode areas by a diaphragm. The uranium species UF 4 and UO 2 were studied on a pyrolitic graphite working electrode in the positive potential area up to a potential of þ2:5 V in order to examine the formation of the above-mentioned uranium species at higher oxidation states. The assumptions were confirmed, and a two-step oxidation of U 4þ ions successively to U 5þ and U 6þ ions were recorded. Moreover, the formed uranium species (probably coordination compounds M x U y F z, M=Na, K, Li) were found to be electrochemically stable in the melt FLINAK. This fact is very important for the designing of the uranium electroseparation process because stable anodic products can migrate to the cathode and thus cause a decrease in the current efficiency. 3. Electrochemical Characteristics of ThF 4 The system FLINAK-ThF 4 (1. mol%) was investigated on a molybdenum working electrode. The typical voltammogram of this system is shown in Fig. 6. A similar two-step reaction mechanism was observed for the reduction of uranium, but the deposition of thorium metal started very close to the melt decomposition and is at least partially hindered by this action. A value of thorium metal deposition potential of 2: V was evaluated from the change of the current slope JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY

6 Development of Electrochemical Separation Methods in Molten LiF-NaF-KF for the Molten Salt Reactor Fuel Cycle Fig. 6 Voltammogram of the system FLINAK-ThF 4 (1. mol%) on a Mo working electrode, scan rate 5 mv/s detected prior to the peak corresponding to the melt decomposition. The shape of the voltammogram indicates the codeposition of thorium and the metal that is reduced from the carrier melt component, together forming an alloy. As the reduction to thorium metal takes place at almost the same potential as the decomposition of the melt FLI- NAK, it was impossible to perform the electrolysis and examine the quantitative deposition of thorium. The first reduction step, detected at a potential of :6 V, was not rigorously investigated, however, the shape of this peak, as well as the significant irreversibility of the reaction, suggest the partial reduction of thorium and the formation of coordination compounds. It must be stressed that only a partial reduction occurs and not a full reduction to thorium metal. The measured voltammograms are in good qualitative agreement with the previously published data. 7) 4. Electrochemical Characteristics of Selected Lanthanides The measured electrochemical properties of all the studied compounds NdF 3, EuF 3 and GdF 3 are similar and can thus be characterized generally by the following description from the point of view that the deposition potentials of species are the most important properties: The electrochemical stability of the studied compounds is too high in order to accomplish the deposition of metals from the melt FLINAK. In the case of neodymium and gadolinium, the deposition potentials cannot even be determined in this melt. The beginning of europium metal deposition was observed from a potential of 1:95 V, close to the carrier melt decomposition and almost at the same potential as for the deposition of thorium. A twostep reaction mechanism was observed for all studied lanthanides on a molybdenum working electrode. Because the tops of peak pairs corresponding to the first reduction steps are of a big potential difference, the reactions seem to be quite irreversible. Due to the very positive value of their centre at potential about :9 V cannot be interpreted as deposition of metal. The presented evaluation is also supported by the literature data, where a two-step reduction mechanism for NdF Pt working electrode Mo working electrode Fig. 7 Comparison of voltammograms corresponding to the system FLINAK-NdF 3 measured on Mo and Pt working electrodes, scan rate 5 mv/s EuF3 GdF Fig. 8 Voltammograms of systems FLINAK-GdF 3 and FLINAK- EuF 3 (both 1. mol%) on a Mo working electrode, scan rates 5 mv/s on a molybdenum electrode was reported by Shiguan et al. in molten LiF-CaF 2. 8) This fact was indirectly confirmed by measurement of the system FLINAK-NdF 3 (1. mol%) on a platinum working electrode, where no electrochemical reaction was detected in the whole potential window of the melt FLINAK and therefore the reaction seems to be of one-step. Because similar electrochemical behaviour was reported to take place on a tungsten electrode in molten fluorides, 9) a working electrode material seems to be very important for the electrochemical behaviour of the lanthanides in molten fluoride media. The comparison of voltammograms corresponding to the system FLINAK-NdF 3 (1. mol%) measured on molybdenum and platinum working electrodes are shown in Fig. 7. The voltammograms corresponding to the systems FLINAK-GdF 3 and FLINAK-EuF 3 measured separately on a molybdenum working electrode are shown in Fig. 8. VOL. 42, NO. 12, DECEMBER 25

7 122 P. SOUČEK et al. Table 2 Summary of experimentally determined red-ox potentials vs. used Ni/Ni 2þ reference electrode in the melt FLI- NAK at a temperature of 53 C Red-ox couple U 3þ /U U 4þ /U 3þ U 5þ /U 4þ U 6þ /U 5þ Th xþ /Th Th 4þ /Th xþ E [V] vs. Ni/Ni 2þ 1:75 1:2 þ:4 þ1:4 2: :7 Red-ox couple Nd 2þ /Nd Nd 3þ /Nd 2þ Eu 2þ /Eu Eu 3þ /Eu 2þ Gd 2þ /Gd Gd 3þ /Gd 2þ E [V] vs. Ni/Ni 2þ < 2:5 1: 1:95 :75 < 2:5 1: Table 3 Summary of evaluated separation possibilities in the melt FLINAK Separated elements U-Nd, Gd U-Th U-Eu Th-Nd-Eu-Gd Nd-Eu-Gd Potential difference >3 mv 25 mv 2 mv n/a n/a Separation possibilities Yes Yes Limit value n/a n/a Probably not Probably not IV. Evaluation and Assessment of Separation Possibilities This study is aimed at the appraisal of separation possibilities under conditions that are as close as possible to the real parameters of the future MSR reprocessing system. The separation possibilities of the studied elements were assessed by comparison of their deposition potentials with regard to the corresponding reaction mechanism. All presented red-ox potentials were evaluated from experimentally measured voltammograms using the graphical method. Because the experimental working conditions are not equivalent to the standard states, presented potentials cannot be considered as the standard red-ox potentials. Moreover, the electrochemical reactions are not exactly defined in all cases, because the studied elements readily form coordination compounds with the solvent. For the above-mentioned purposes, the following assessment is correct, because all measurements were performed using the same experimental set-up and under the same working conditions. It was impossible to exactly determine the deposition potentials of neodymium and gadolinium, because their fluorides are more electrochemically stable than the carrier melt constituents. For this reason, their unknown deposition potentials had to be substituted by the decomposition potential of the melt for purposes of comparison. The summary of all found red-ox potentials in the melt FLINAK is given in Table 2. The following conclusions have resulted from the assessment (the minimum potential difference for successful quantitative separation was considered to be 2 mv):. It is thermodynamically feasible to remove uranium from the melt FLINAK by the electrodeposition process. The other studied elements, however, start, either, to be reduced too close to the carrier melt decomposition potential for removal (thorium and europium), or, alternatively, their reduction starts in a more negative potential area compared to the carrier melt decomposition potential (neodymium, gadolinium).. The selective separation of uranium from all studied lanthanides is feasible in the melt FLINAK.. The separation of thorium from europium is impossible, whilst the separation possibilities among thorium, neodymium and gadolinium cannot be determined in the melt FLINAK. However, according to the similarity in their electrochemical behaviour and with respect to available thermodynamic data, their separation will probably be impossible.. Similarly, it is impossible to determine the separation possibilities among the lanthanides in the used melt FLINAK, but, due to the same reason as mentioned above, we predict that the mutual separation among the group of lanthanides would not be thermodynamically feasible. The separation possibilities of the studied elements are summarized in Table 3 together with evaluated potential differences (if available). V. Conclusion According to the results obtained from this study, electrochemical methods seem to be able to effect the separation of actinides from lanthanides in molten salt FLINAK, but the following fact should be taken into account when designing a MSR fuel cycle flow sheet: Uranium is the only one from the studied elements that could be removed by deposition on the cathode. Separation of electrode areas by diaphragm or membrane is recommended in order to prevent migration of anodic products that are formed during the process. Also recommended is the use of an electrode with a structured surface or an electrode that forms an alloy with uranium metal. Thorium and all studied lanthanides are too electrochemically stable to be quantitatively removed by cathodic deposition from the melt FLINAK. For the particular application of the method within the on-line reprocessing of the MSR fuel, some important facts arose from the presented study. As mentioned above, the lanthanides should be removed from the main fuel circuit of the MSR and the actinides should remain there to be further processed in the reactor core. This aim is impossible to accomplish by using the cathodic deposition technique, because deposition potentials of the lanthanides are more neg- JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY

8 Development of Electrochemical Separation Methods in Molten LiF-NaF-KF for the Molten Salt Reactor Fuel Cycle 123 Fig. 9 Proposed flow sheet for MSR fuel reprocessing back-end of MSR fuel cycle ative in comparison with those of the actinides. The combination of a) pre-reduction of all presented elements, and b) the subsequent group-selective separation by the anodic dissolution method, is proposed as a solution to this problem. The outline of the flow sheet for the MSR fuel reprocessing is shown in Fig. 9. It was proposed on the basis of the conclusions from this study and several presumptions that will have to be investigated. The following points belong among the most important areas for further study: The electrochemical behaviour of the actinides and lanthanides in a more stable melt (e.g. proposed LiF-CaF 2 ) and the behaviour of beryllium during molten salt/liquid metal extraction, which is supposed not to be extracted because of its very low solubility in liquid bismuth. 1) The melt FLINAK seems to be generally inconvenient for removal of the lanthanides by any electrochemical method due to its insufficient electrochemical stability. Fluorides with more negative deposition potentials should be used for the medium directly applicable to the MSR fuel online separation, e.g. the mixture LiF-CaF 2 is recommended for further studies. However, the melt FLINAK can be recommended for use during the following technological operations:. Pyrochemical preparation of the MSTR fuel, e.g. removal of residual uranium from the mixture resulting after application of the FVM (see Chap. I, scheme at Fig. 2). Purification of the fission products stream resulting from reprocessing of the MSR fuel by cathodic deposition of the non-separated actinides. The amount of presented actinides depends on the separation efficiency and on the number of separation steps or cycles. It is almost certainly economically more advantageous to process the mixture by another purification step, which should remove rather a small amount of actinides, than to return it back to the original separation process. On the other hand, the melt from the anodic dissolution separation step will be used for this operation according to the scheme shown in Fig. 9. Acknowledgments The project was realized thanks to the running projects under financial support of the Grant Agency of the Czech Republic within the Grant Project No. 14/3/155, the Radioactive Waste Repository Authority, the Czech Ministry of Industry and Trade and the PYROREP project of the EC/EURATOM 5th Framework Programme. References 1) A Technology Roadmap for Generation IV Nuclear Energy Systems, U.S. DOE Nuclear Energy Research Advisory Committee at the Generation IV International Forum, Dec. 22, GIF-2-, 2) J. Uhlir, V. Priman, Z. Frejtich, R&D of Molten-Salt Reactor Fuel Cycle in the Czech Republic, Proc. GLOBAL 23, New Orleans, U.S.A., Nov. 16 2, 23, (23). 3) M. W. Rosenthal, P. H. Haubenreich, H. E. McCoy, L. E. McNeese, Recent progress in molten-salt reactor development, At. Energy Rev., 9, 61 (1971). 4) M. Marecek, P. Novy, J. Uhlir, Technological verification of fluoride volatility method for front-end of molten salt transmutation reactor fuel cycle, Proc. GLOBAL 21, Paris, France, Sept. 9 13, 21, (21). 5) G. J. Janz, R. P. T. Carolyn, B. Allen, et al., Physical properties data compilations relevant to energy storage. I. Molten salts: Eutectic data, Nat. Stand. Ref. Data Ser., Nat. Bur. Stand. (U.S.), 61, Part I (1978). 6) P. Souček, F. Lisý, R. Zvejšková, et al., Electrochemical studies of selected actinides and lanthanides in molten LiF- NaF-KF, Proc. Euchem 24 Molten Salts Conf., Piechovice, VOL. 42, NO. 12, DECEMBER 25

9 124 P. SOUČEK et al. Poland, June 2 25, 24, (24). 7) F. R. Clayton, G. Mamantov, D. L. Manning, Electrochemical studies of uranium and thorium in molten LiF-NaF-KF at 5 C, J. Electrochem. Soc., 121[1], 86 9 (1974). 8) C. Shiguan, Y. Xiaoyong, Y. Zhogxing, et al., Rare Metals, 13, 46 (1994). 9) E. Stefanidaki, C. Hasiotis, C. Kontoyannis, Electrodeposition of neodymium from LiF-NdF 3 -Nd 2 O 3 melts, Electrochimica Acta, 46, (21). 1) L. M. Ferris, J. C. Mailen, J. J. Lawrance, et al., Equilibrium distribution of actinide and lanthanide elements between molten fluoride salts and liquid bismuth solutions, J. Inorg. Nucl. Chem., 32, (197). JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY