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DOE OSTI2013

Inclusion property of Cs, Sr, and Ba impurities in LiCl crystal formed by layer-melt crystallization

Pyroprocessing is one of the promising technologies enabling the recycling of spent nuclear fuels from a commercial light water reactor (LWR). In general, pyroprocessing uses dry molten salts as electrolytes. In particular, LiCl waste salt after pyroprocessing contains highly radioactive I/II group fission products mainly composed of Cs, Sr, and Ba impurities. Therefore, it is beneficial to reuse LiCl salt in the pyroprocessing as an electrolyte for economic and environmental issues. Herein, to understand the inclusion property of impurities within LiCl crystal, the physical properties such as lattice parameter change, bulk modulus, and substitution enthalpy of a LiCl crystal having 0-6 at% Cs{sup +} or Ba{sup 2+} impurities under existence of 1 at% Sr{sup 2+} impurity were calculated via the first-principles density functional theory. The substitution enthalpy of LiCl crystals having 1 at% Sr{sup 2+} showed slightly decreased value than those without Sr{sup 2+} impurity. Therefore, through the substitution enthalpy calculation, it is expected that impurities will be incorporated within LiCl crystal as co-existed form rather than as a single component form. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Sampling and Analysis Plan for Investigation of Undeclared Nuclear Activities - 20104

The nuclear-centered energy policy in South Korea is moving towards a nuclear-free era. It also facilitates an action or progress on the denuclearization of the Korean peninsula. The Nuclear Safety and Security Commission (NSSC) launched a new R and D project which attempts to verify nuclear activities via analyzing environmental samples. Environmental samples possibly contain significant radiological information relevant to past undeclared activities. In order to precisely verify them, it is essential to provide 1) the concrete information of target facilities, 2) appropriate methodologies and procedures, and 3) well-suited instrumentation and equipment corresponding to sample signatures. It is also important to provide scientific reliability or validity in any process such as sample collection, handling, transportation and laboratory analysis. Therefore, the information of nuclear facilities (fuel fabrication plant, graphite moderated reactor, reprocessing plant) in North Korea are reviewed in order to determine parameters for appropriate sampling methods representing a facility's characteristics. Secondly, a sampling procedure for radioactive elements is being developed based on international and domestic standards such as ASTM (American Society for Testing and Materials), KS (Korean industrial Standards), and International Atomic Energy Agency (IAEA) manuals. The procedure contains some specific information including sampling process, sampling tools, amount of sample, sample identification, and other related information. Developing the qualified national guidelines is most important since sampling and analyzing results can be used as evidences of nuclear activities. The guideline developed through this work will clearly define various methodologies such as document verification, sampling procedures, collection methods, sample storage criteria, chain of custody, and transportation conditions. The final objective of this work is to establish the infrastructure and regulatory framework for assessment of nuclear activities, and provide a national guideline for sampling and analysis. This will bring confidence in the results and play a critical role in complete denuclearization progress in South Korea. (authors)

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2015

Zircaloy Cladding as Starting Precursor for Ceramic Waste-Forms - 15562

Zircaloy cladding of spent fuel is regular waste-stream related to the process of reprocessing. It remains in the apparatus after spent fuel dissolution. The most common technique of zircaloy cladding treatment is cementation and disposal in metallic containers. In view of the high cost of the cladding material (mainly zirconium), there have been a number of proposals for this material to be recovered and re-used. However, the residual contamination of the claddings following the dissolution of the fuel and the chemical stability of the zirconium militate against any proposal for a recycling process that might be economically reasonable. At the same time there is a lot of information available about synthesis of mineral-like materials based on zirconium for immobilization of long-lived radionuclides. The object of this work was to study experimentally electrochemical dissolution of fuel rod cladding for further obtaining starting precursor for the synthesis of ceramic-waste forms. The results of experiments on electrochemical dissolution of zirconium and zirconium alloy (fuel rod cladding) in nitric acid solutions are presented. The peak dissolution rate of specimens is shown to be observed in dilute nitric acid at elevated temperatures. The rate of dissolution rises with temperature, reaching ∼40-70 mg/cm{sup 2}.h at 90 deg. C. The current yield at this point is 0.6-0.7 g/A.h. The results of analysis of the surface oxide film on specimens and residues formed during dissolution are also presented. An electrochemical dissolution of zircaloy claddings is the common technique for preparation of zirconium solutions to be used for ceramic waste form synthesis. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

Scaled up Process Report – Apparatus and Model

Advanced voloxidation with NO 2 is a proposed process for used nuclear fuel head-end reprocessing scheme that converts UO 2 to higher oxides, and it also converts partitioning volatile fission products into the gas phase, thus facilitating fuel dissolution and actinide recovery. NO 2 voloxidation is being studied on small batches of UO 2 Simfuel, but the real test of process feasibility will be when it is scaled up to work with >100 g of irradiated material. This report discusses the aspects of scale-up that must be considered for NO 2 voloxidation, including development of a stirred reactor to promote agitation of the mixture during processing, online process monitoring, and automation controls. Brief details on parallel efforts are also provided in this report, including (a) demonstration of iodine release from Simfuel made by Spark Plasma Sintering, and (b) development of an order-of-magnitude scale-up to react 100 g of UO 2 Simfuel in a metal reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Irradiated FA Structural Materials Conditioning by Induction-Slag Remelting in a Cold Crucible

A method of managing one type of radwaste forming during BREST-300 reactor operation – the structural materials of FA – is described. In the module developed for reprocessing, the fissile materials losses together with all forms of waste must not exceed 0.1% of the initial amount, in connection with which the maximum admissible content of facade materials and structural materials sent into long-term storage must not exceed 0.001% by weight. Only the plutonium content may reach 0.015% by weight, which makes it necessary to perform additional purification and return fissile materials into the nuclear fuel cycle. A method of reprocessing the structural materials of irradiated fuel elements is proposed on the basis of induction-slag remelting in a cold crucible, which makes it possible to conduct purification to the desired index value. The operating regime of the melting unit and the qualitative composition of the employed flux that make it possible to reach the required level of purification by removal of fissile materials are described.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Testing of an Iodine and Tritium Removal System for Advanced Tritium Pretreatment Off-Gas

The reprocessing of used nuclear fuel would release volatile radionuclides into the off-gas of a processing plant, including 3 H, 14 C, 85 Kr, and 129 I. For a reprocessing plant sited within the United States and governed by US regulations, the abatement of some or all of these radionuclides from the plant off-gas would be required before discharge of that gas to the environment. A potential simplification of the off-gas management system can be achieved through an efficient advanced tritium pretreatment (ATPT) step in which the UO 2 fuel is oxidized by NO 2 prior to dissolution. Oxidation of the UO 2 fuel matrix results in the release of tritium contained within. Up-front removal of tritium from the fuel in a pretreatment step, followed by the abatement of tritium from the advanced tritium pretreatment off-gas (ATPTOG) stream, can minimize or eliminate the distribution of tritium throughout the plant, also decreasing or eliminating the need for tritium capture on multiple off-gas streams. The use of NO 2 as the oxidant allows tritium pretreatment to be performed at lower temperatures and may result in the quantitative release of iodine from the fuel. A total of five tests examined AgA as a potential iodine sorbent at an operating temperature of 150°C and both 3A MS and silica gel as potential tritium sorbents with an operating temperature of 40°C. 3A MS and silica gel are intended for use as regenerative sorbents, in which the adsorbed tritium (as tritiated water or tritiated nitric acid) is loaded onto the sorbent at operating temperature and then subsequently desorbed at high temperature into a small-volume gas stream for recovery and conversion to a waste form. This testing leaves many fundamental questions regarding the ATPTOG iodine and tritium abatement systems unanswered. These include questions regarding sorbent capacity and sorbate desorption. Overall, this testing indicates that tritium and iodine abatement from an ATPTOG stream by solid sorbents is currently at a low stage of technological development and that deployment of such an abatement system would require substantial additional research and development.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

DEVELOPMENT OF AN APPROACH FOR DOWNBLENDING AND DISPOSITION OF IRRADIATED GRAPHITE FUEL

The IGR reactor, a pulse-type test experimental reactor, reached full power in 1961. The reactor contains a center axial experimental ampoule that is subjected to a large pulse of neutrons (neutron flux maximum density of 7×1016 n/cm2·s, peak power 10GW) when the reactor is operated. The fuel form for the IGR reactor is graphite cubes and rods impregnated with highly enriched (90%) uranyl-dinitrate (UO2(NO3)2·6H2O). This fuel is eligible for return to Russia for reprocessing under the Russian Research Reactor Fuel Return program. However due to some technical and licensing issues it was decided to immobilize this fuel in Kazakhstan. The disposition approach thus selected for the spent HEU fuel involves down-blending of the fuel followed by cementation. Development of the process for down-blending and subsequent cementation requires a process that satisfies many criteria, including fissile content, compressive strength, permeability, and containment of radioactivity, among others. One key goal of the final product is to meet the criteria for termination of safeguards. This paper will describe the overall challenge associated with disposition of irradiated HEU graphite spent fuel, the identification of the objectives of the disposition process, and the constraints associated with development of the process and the waste form.

Gaines, Kris↗

DOE OSTI2016

A Scintillation-Fiber Based Cosmic-ray Muon Tomography System for Imaging Dry Storage Cask

Because of the growth of the nuclear power industry in the United States and the policy decision to ban reprocessing of commercial spent nuclear fuel, the spent fuel inventory at commercial reactor sites has been increasing. Spent nuclear fuel presents extra complexity of security and proliferation issues. These fuel assemblies are practically inaccessible for inspection purposes, as reopening a DSC would require special facilities and would be tremendously expensive. Cosmic-ray muon Tomography has been studied in the past for various applications and was demonstrated to be a good non-destructive assay technique for high-Z materials, such as spent nuclear fuel assemblies in dry storage casks (DSC). Compared to conventional imaging techniques, muon imaging has some advantages due to its negligible health risk and high penetrating power. In this paper, we report our simulation study on imaging a DSC using muon Tomography technique. The Point of Closest Approach (POCA) and the Maximum-Likelihood Expectation- Maximization (ML-EM) algorithms are implemented and evaluated. A scaled-down prototype muon tomography system is proposed and studied. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

High-Assay Low Enriched Uranium Pyroprocessing and Electrometallurgical Treatment at the Fuel Conditioning Facility

Fuel Conditioning Facility (FCF) fulfills part of INL’s mission by reprocessing irradiated sodium-bonded fuel from the EBR-ll reactor and participates in research to further explore the fuel cycle for recovery of fuel from a variety of current reactors in use today. The fuel cycle of a nuclear reactor under the current non-proliferation act, leaves unused fuel in the spent fuel rods. The rods are safely stored until their fuel can be recovered. ? Fuel recovery from the EBR-II fuel pins has several steps to separate the uranium from the salts. The pyroprocessing technology involves high - temperatures, chemical and electrochemical methods for separating the unused uranium from the salts, fission products, and used fuel. The recovered uranium is then remixed and formed into ingots or Regulus with an enrichment of < 20% U-235. This final product is a source to fuel current nuclear reactors as well as research and development for future nuclear reactors

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2024

Iodine Mass Tracking Research and Development Needs for Pyrochemical Fuel Cycles

This report was generated jointly by Argonne National Laboratory (ANL) and Idaho National Laboratory to provide a high-level summary of the current knowledge on the behavior of fission product iodine during reprocessing of used nuclear fuel, as well as provide recommended path forward for research and development activities to fill particular knowledge gaps. The focus of this work is on pyrochemical processing as is applied to light water reactor (LWR) oxide-based used nuclear fuels (UNF), however, some discussion of electrorefiner behavior from metal fuel processing equipment is also included.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Preferred Concepts for Disposal of the UK Inventory of Depleted, Natural and Low Enriched Uranium - 16382

Depleted, natural and low enriched uranium (DNLEU) forms a significant quantity of the UK inventory of higher activity radioactive materials. The great majority comprises uranic materials produced from uranium enrichment operations and from reprocessing of spent nuclear fuels. The most recent estimate of the total quantity (stocks plus arisings) of UK-owned DNLEU from civil nuclear operations is 170,000 tU. Current UK Government strategy is for DNLEU to be stored indefinitely in existing or new facilities, as required. However, Government strategy for long-term management of uranic materials is under review and a change could potentially cause this material to be reclassified as a waste. An Integrated Project Team (IPT) was formed in 2012 with the objective of identifying and addressing the issues associated with the potential need to dispose of DNLEU, in order to inform wider decision-making on the long-term management of UK DNLEU. Key technical issues considered by the IPT relate to the following: 1. The need for an improved understanding of the UK DNLEU inventory. 2. The identification of preferred chemical and physical form(s), and preferred packaging and disposal concepts for DNLEU - the main subject of this paper. 3. The approach to safety assessment of DNLEU disposal. This paper summarises the disposal concepts under consideration by the IPT, and the rationale for identification of preferred concepts for disposal of DNLEU inventory components. The conclusions will be of interest to other programmes and countries with a requirement to manage large quantities of similar materials. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Natural Transmutation of Actinides via the Fission Reaction in the Closed Thorium–Uranium–Plutonium Fuel Cycle

It is shown for a closed thorium–uranium–plutonium fuel cycle that, upon processing of one metric ton of irradiated fuel after each four-year campaign, the radioactive wastes contain ~54 kg of fission products, ~0.8 kg of thorium, ~0.10 kg of uranium isotopes, ~0.005 kg of plutonium isotopes, ~0.002 kg of neptunium, and “trace” amounts of americium and curium isotopes. This qualitatively simplifies the handling of high-level wastes in nuclear power engineering.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Performance of a Steel/Oxide Composite Waste Form for Combined Waste Steams from Advanced Electrochemical Processes

The results of electrochemical corrosion tests and modeling activities performed collaboratively by researchers at the University of Illinois at Chicago and Argonne National Laboratory as part of workpackage NU-13-IL-UIC-0203-02 are summarized herein. The overall objective of the project was to develop and demonstrate testing and modeling approaches that could be used to evaluate the use of composite alloy/ceramic materials as high-level durable waste forms. Several prototypical composite waste form materials were made from stainless steels representing fuel cladding, reagent metals representing metallic fuel waste streams, and reagent oxides representing oxide fuel waste streams to study the microstructures and corrosion behaviors of the oxide and alloy phases. Microelectrodes fabricated from small specimens of the composite materials were used in a series of electrochemical tests to assess the corrosion behaviors of the constituent phases and phase boundaries in an aggressive acid brine solution at various imposed surface potentials. The microstructures were characterized in detail before and after the electrochemical tests to relate the electrochemical responses to changes in both the electrode surface and the solution composition. The results of microscopic, electrochemical, and solution analyses were used to develop equivalent circuit and physical models representing the measured corrosion behaviors of the different materials pertinent to long-term corrosion behavior. This report provides details regarding (1) the production of the composite materials, (2) the protocol for the electrochemical measurements and interpretations of the responses of multi-phase alloy and oxide composites, (3) relating corrosion behaviors to microstructures of multi-phase alloys based on 316L stainless steel and HT9 (410 stainless steel was used as a substitute) with added Mo, Ni, and/or Mn, and (4) modeling the corrosion behaviors and rates of several alloy/oxide composite materials made with added lanthanide and uranium oxides. These analyses show the corrosion behaviors of the alloy/ceramic composite materials are very similar to the corrosion behaviors of multi-phase alloy waste forms, and that the presence of oxide inclusions does not impact the corrosion behaviors of the alloy phases. Mixing with metallic waste streams is beneficial to lanthanide and uranium oxides in that they react with Zr in the fuel waste to form highly durable zirconates. The measured corrosion behaviors suggest properly formulated composite materials would be suitable waste forms for combined metallic and oxide waste streams generated during electrometallurgical reprocessing of spent nuclear fuel. Electrochemical methods are suitable for evaluating the durability and modeling long-term behavior of composite waste forms: the degradation model developed for metallic waste forms can be applied to the alloy phases formed in the composite and an affinity-based mineral dissolution model can be applied to the ceramic phases.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

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, 99 TcO 4 – , 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 TcO 4 – or its well-known surrogate ReO 4 – and transuranic/lanthanide species. In the current study, we investigated the coordination of ReO 4 – /TcO 4 – 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 ReO 4 – /TcO 4 – directly connects to Pu IV , Pu VI O 2 2 + , Ce III , and Ce IV in the terminal and bridging coordination modes, leading to 1-, 2-, and 3-dimensional frameworks. For example, ReO 4 – coordination to Pu(IV) formed a 1D chain or 2D framework, isostructural with previously isolated Th(IV) compounds. However, Pu VI O 2 2 + alternating with ReO 4 – led to a unique 1D chain, different from the prior-reported U(VI)/Np(VI)-ReO 4 – /TcO 4 – structures. Coordination of ReO 4 – /TcO 4 – with Ce(III) promotes the assembly of 3D frameworks. Finally, attempted synthesis of a Ce(IV)-ReO 4 – compound resulted in a 2D framework with a mixed-valence Ce III/IV . The highly acidic reaction conditions supported the reduction of both Ce IV and Tc VII , challenging isolation of compounds featuring these species. Only one TcO 4 -containing structure was obtained in this study (Ce III –TcO 4 3D framework), vs the six total Ce/Pu-ReO 4 compounds. Furthermore, our three Pu-ReO 4 crystal structures are the first reported and translated to atomic-level information about Pu-TcO 4 coordination in nuclear fuel reprocessing scenarios, in addition to broadening our knowledge of bonding trends in the early, high-valence actinides.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2019

Adsorbents and adsorption models for capture of Kr and Xe gas mixtures in fixed-bed columns

Off-gases produced during the reprocessing of used nuclear fuel (UNF) include 129-I 2 , 3-HHO, 14-CO 2 , 85-Kr, and 135-Xe, which are volatilized out into the off-gas. In order to meet regulatory requirements for reprocessing plant emissions, these gases must be captured and removed from the o?-gas stream prior to o?-gas emission. Of particular interest are the noble gases, Kr and Xe, which can be fairly difficult to remove from the off-gas due to their low chemical reactivity. Thus, this work is focused on utilizing engineered adsorbents, AgZ-PAN and HZ-PAN, to capture Kr and Xe from a mixed-gas stream at relatively low temperatures (191–295 K) and various flow rates (50–2000 mL/min). Isothermal data for Kr and Xe on each adsorbent are analyzed to produce the Langmuir parameters needed to model the mixture adsorption capacities at relevant temperatures using the Extended Langmuir model. Those parameters are then incorporated into a fixed-bed adsorption model developed in this work using the Mulitphysics Object-Oriented Simulation Environment (MOOSE). That model is used to simulate breakthrough times for Kr and Xe in packed columns of AgZ-PAN and HZ-PAN, ranging in length from 6 to 20 in., at relevant temperatures and flow rates. Breakthrough times varied from nearly instantaneous for Kr in AgZ-PAN to 30 h for Xe in HZ-PAN. After the developed model was validated by comparisons with experimental break-through data, the model framework was used to simulate the performance of multiple fixed-bed columns connected in series.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Survey of Uranium Nitride and Mixed Oxide Fuels for Microreactor Applications

Different fuel systems are being proposed for use in microreactors including TRISO fuel and metallic U-Zr based fuel. TRISO fuel is considered the main fuel option for most of the industry teams. Given the need for compact size core and potential for long fuel life of a microreactor, high density fuels such as metallic and nitride fuels are potential options. Of interest here is the uranium nitride (UN) option, which can allow for higher fissile material content and higher thermal conductivity compared to conventional uranium oxide fuel, and also has higher melting temperature compared to metallic fuel. Meanwhile, the current availability is limited for high assay low enriched uranium (HALEU) that is needed for high density fuels, which motivates the consideration of using Pu as a potential replacement for HALEU until adequate production capacity is in place. Current Pu availability is mainly attributed to the inventory of excess weapons Pu rather than through reprocessing of spent nuclear fuel. This inventory of excess Pu can be used in both metallic and oxide fuel systems to replace HALEU. Of interest here, the oxide form, that is the mixed oxide form of PuO 2 and UO 2 (MOX). In this report, the options of using UN or MOX, in both pellet and TRISO fuel forms, in microreactors are evaluated in relation to their properties, fuel performance and irradiation data, as well as fabrication. Gaps related to those areas are identified for both fuel systems, to guide future activities by DOE programs such as the advanced fuels campaign (AFC), to enable their use in microreactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Cosmic Ray Muon Radiography of Spent Fuel in Dry Storage Casks

In the absence of a permanent repository or reprocessing facilities, highly radioactive spent nuclear fuel bundles are typically placed in long-term storage in above ground casks. These dry storage casks are heavily shielded to prevent the escape of radiation from the contents to the environment. Casks under international safeguards are monitored to detect diversion of spent fuel. The heavy shielding of the cask walls also precludes inspection of cask contents with typical radiographic probes. When international safeguards inspectors need to verify cask contents (which may be necessary to recover from a loss of continuity of knowledge), the casks must be moved back into a cooling pool to be safely opened and visually inspected. This time-consuming procedure is invasive and potentially dangerous. A non-destructive radiographic method to verify a cask's contents in situ would therefore be a useful safeguards tool. Highly energetic cosmic ray muons are an alternative tomographic probe. Since muons are colorless leptons, they do not participate in strong interactions with nucleons, and their large mass suppresses radiative energy loss while passing through materials. These properties allow muons to penetrate dense shielding that defeats more typical probes like photons and neutrons. Measurements of the scattering of muons that pass through objects provide information on the object's internal structure. This summary describes the first measurements using cosmic ray muons to inspect the contents of a partially loaded dry storage cask. We show that partial defects in the cask loading can be identified with this technique. This data shows that cosmic ray muons are an effective radiographic probe of dry storage casks. This technique may be useful for international safeguards inspectors wishing to verify a cask's contents without opening it. The analysis techniques described here allow the column of missing fuel bundles to be determined, but not the row. Recording data from several positions around the cask would allow ambiguities in x-y position to be resolved. Alternatively, a detector could be designed such that it surrounds the cask and provides a full image of the casks interior. Improved measurements and the design of optimized cask imaging instruments are being explored. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Separation of Actinides from Lanthanides in Molten Fluorides by the Modulated-Current Electrolysis

Recently, the Molten Salt Reactor (MSR) is taking a renewed attraction as one of the reactor systems highlighted by the Generation IV International Forum. It is a non-classical (liquid fuel) system which can be operated as a breeder within the {sup 232}Th-{sup 233}U fuel cycle with the very interesting breeding factor. The necessary precondition for its proper functionality is the online reprocessing within the MSR's fuel cycle. A lot of effort is given to the development of the on-line pyrochemical reprocessing technology. Among the processes considered to be suitable for this process, the electrochemical separations are highlighted. Within this context, it is necessary to study electrochemical behavior of both fissile material and fission products in the molten salt systems and to develop the processes of their extraction from the system at various stages of the whole cycle. In this contribution, the study of the electrochemical separation of U from Gd in the LiF-CaF{sub 2} molten system by the modulated current electrolysis is presented. The study of MSR-related electrochemistry will continue as a part of the new Czech MSR R and D program. The knowledge of the electrochemical behavior of actinides and lanthanides in molten fluorides is necessary to the development of any electrochemical separation process. In our previous work, electrochemical characteristics of uranium, thorium, as well as that of several lanthanides, were measured in LiF-BeF{sub 2} (the basic melt considered for the MSR concept) and Be-free melts (LiF-NaFKF, LiF-NaF, LiF-CaF{sub 2}), and added to available literature data. If the fluoride carrier melt is electrochemically stable enough, it is possible to deposit the element of interest in its basic, metallic state (e.g. uranium in all types of fluoride melts tested) by the electrolysis. In the case of insufficient electrochemical stability of the fluoride melt in which the actinides and lanthanides are dissolved for the experiments, it is possible to use a reactive working electrode made of a material which creates alloys with the element to be deposited. An alloying reaction can shift the deposition potential in a favorable direction and therefore achieve the separation of an element from the system. The use of reactive electrode can also improve the adhesion of the deposit. When two or more electroactive species are present in the system at the same time, the efficiency of their separation by an electrolysis depends on how different the deposition potentials of these two species are. If we input certain modulation to the electrolytic current, we can improve overall result of an electrolytic experiment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗