Thermodynamic modeling of the KCl-LiCl-NaCl-UCl3 system for molten salt electrolysis and reprocessing of spent nuclear fuel
Not Available
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Engineering topics
Fredrickson, Guy L.
Publications and source records attributed to Fredrickson, Guy L..
Chlorination of HALEU regulus casting dross
A scoping study was performed for chlorinating dross formed during uranium casting operations. Here, the purpose is to minimize the losses of uranium to dross wastes. The dross is primarily a mixture of uranium metal and uranium oxide with a minor fraction of crucible and crucible coating materials. The reaction chemistries were performed in a carrier salt of LiCl-KCl eutectic at 500 °C. The addition of FeCl 2 chlorinated the uranium metal to UCl 3 , by reducing the FeCl 2 to iron metal. After the uranium metal was chlorinated, zirconium metal was added to the salt. The residual FeCl 2 chlorinated the zirconium metal to ZrCl 4 , by reducing the FeCl 2 to iron metal. In turn, the ZrCl 4 chlorinated the uranium oxide to UCl 3 , by converting the ZrCl 4 to zirconium oxide. The effectiveness of the chlorination reactions was qualitatively verified by cyclic voltammograms that indicated the presence or absence of FeCl 2 and UCl 3 in the salt.
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Synthesis and characterization of super occluded LiCl-KCl in zeolite-4A as a chloride salt waste form intermediate
Here, this paper reports the hygroscopic properties of eutectic LiCl-KCl after absorption into zeolite-4A, up to salt loadings of 75 wt%. Samples of the salt occluded zeolite were hydrated in a humidity chamber at constant temperature and relative humidity for up to 100 h. At up to 45 wt% salt loading, the un-occluded phase of salt consisted primarily of NaCl, which forms when the Na + ions present in the zeolite framework exchange with Li + and K + ions from the eutectic LiCl-KCl. This results in minimal water absorption and corrosion of contacted stainless steel. At greater than 45 wt% salt loading, water absorption and corrosion progressively worsened. The mixture has a significant amount of excess LiCl-KCl, making it highly hygroscopic. This study reveals an option for the intermediate treatment of waste salt from spent nuclear fuel electrorefiners that could facilitate it to be stored in a non-inert atmosphere for extended periods of time before final conversion into a permanent waste form.
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History and status of spent fuel treatment at the INL Fuel Conditioning Facility
Experimental Breeder Reactor II (EBR-II) had an illustrious operational history from 1964 to 1994 as one of the first sodium-cooled fast reactors to use sodium-bonded metallic driver fuels and blankets. EBR-II was a testbed for fuel development and reactor design principles. Following shutdown, it was decided to use electrometallurgical methods to treat the spent fuel and blanket materials for final disposition. Here, this paper provides a brief history of U.S. liquid metal cooled reactor development with emphasis on EBR-II, and a technical description of the electrometallurgical methods.
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Methods and systems for aluminum electroplating
Systems and methods for coating a metallic component are provided. In one embodiment, a metallic coating may be disposed in a plating bath comprising AlBr3. The metallic coating may be coupled with, or configured as, a working electrode. A counter electrode formed of aluminum may be disposed within the plating bath. An electric current may be applied between the two electrodes resulting in the electrodeposition of aluminum on the metallic component. In one particular embodiment, the plating bath may include LiBr, KBr and CsBr, with AlBr3 being present in an amount of approximately 80 percent or greater by weight. Various types of metals may be coated with aluminum using embodiments of the present disclosure. Additionally, the methods and systems described herein are amenable to coating of complex geometries.
Tripathy, Prabhat K.↗
Electrochemical measurement and analysis of YCl 3 , ScCl 3 , GdCl 3 and MgCl 2 in molten eutectic LiCl-KCl
For this work, electrochemical measurements of YCl 3 , ScCl 3 , GdCl 3 and MgCl 2 were obtained in molten LiCl-KCl eutectic at 773 K using a three-electrode configuration. Tungsten rods were used as the working and counter electrodes. Two quasi- and one reference electrode(s) (glassy carbon, nickel wire, and Ni/NiO) were used to collect electrochemical measurements of the analyte salts. Cyclic voltammetry data were analyzed to calculate values of diffusion coefficient, exchange current density, and charge transfer resistance and were determined to be on the order of 10 –5 cm 2 s –1 , 10 –2 A cm –2 , and 10 0 Ω; respectively. In general, the values of diffusion coefficients were found to be consistent with those reported of high-temperature molten salts. Relatively large values of exchange current density corresponded with smaller values of charge transfer resistance. These values were found to be reasonable in comparison to results available in the literature. The measured cyclic voltammograms were normalized with respect to both lithium reduction and chloride ion oxidation potentials. Such a normalization technique is effective for comparing experimentally obtained cyclic voltammetry data to those that have been published. An analysis of experimentally obtained results indicates the relation of electrode reactions (e.g., differences of reduction potentials) are independent of the choice of reference electrode. Additionally, the choice of reference electrode did not affect the electrochemical window, exchange current density, or diffusion coefficient values.
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Deliquescence of Eutectic LiCl-KCl Diluted with NaCl for Interim Waste Salt Storage
Molten eutectic LiCl-KCl salt is a widely used electrolyte for electrorefining uranium from spent nuclear fuel. Due to the hygroscopic nature of this salt, such operations must be performed under controlled atmospheric conditions, and waste salts require careful storage to avoid deliquescence and corrosion of container materials. This study investigated a potential processing path for reducing the degree of deliquescence through dilution to varying extents with NaCl. Here, the hydration behavior of LiCl-KCl salts diluted with NaCl was evaluated in terms of mass gain due to water absorption, degree of deliquescence (including first appearances of standing water), and evidence of corrosion to stainless steel containers in a humid air environment (40°C, 20% relative humidity). In this humid air environment, pure eutectic LiCl-KCl exhibited a 50 mass % increase due to water absorption and showed evidence of standing water after 24 h. Waste salt diluted with NaCl required loadings of 89 mass % NaCl in order to prevent deliquescence and exhibited a 3 mass % increase due to water absorption. After periodic observation for 48 h, standing water was observed near all ingots with the exception of the 89 mass % NaCl samples. Dilution with 89% NaCl was also found to reduce evidence of corrosion when stored in stainless steel crucibles. While dilution with NaCl greatly decreases steady-state hydration, the storage volume is increased ~10× through this procedure.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Nuclear Fuels and Reprocessing Technologies: A U.S. Perspective
Reprocessing and/or waste management issues are of concern to the “back end” of the nuclear fuel cycle. Of course, there are a great many “nuclear fuel cycle” scenarios to consider; if not in practice, then at least in theory. The simplest conceptually is the “once through” fuel cycle in which the spent fuel is discarded. The more complex fuel cycle scenarios involve reprocessing spent nuclear fuels and a family of nuclear reactor technologies to accommodate burning and breeding for various military and commercial needs. Therefore, the selection of a specific “fuel cycle” is what ultimately imposes the engineering requirements of the reprocessing and waste management technologies. No one part is independent of the other parts in a fuel cycle flowsheet; all parts are fully integrated. This paper presents a summary of radiochemical processes, nuclear reactor technologies, nuclear fuel types, and the reprocessing technologies that serve the different nuclear fuel types. Comprehending how this series of topics are related to each other is a prerequisite to understanding the requirements of any reprocessing strategy. The summary materials presented here are selective, as opposed to comprehensive. More detailed information on any one subject can be found in the reference materials.
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