This study evaluates the evolution of the composition and thermal behavior of the Mark-IV electrorefiner salt at Idaho National Laboratory during continued processing of sodium-bonded Experimental Breeder Reactor II driver fuel. MASTERS flowsheet simulations were used to project salt composition changes during the remaining campaign, and thermodynamic equilibrium calculations were coupled to those compositions to estimate phase behavior, liquidus temperature, and operating margin. Thermal measurements were performed on actual Mark-IV electrorefiner salt: differential scanning calorimetry used a December 2022 dip sample, and bulk-cooling experiments used an approximately 100 g salt sample recovered directly from the Mark-IV salt cell in March 2026. The measured liquidus range of 408 to 437 ∘C brackets the model-projected initial liquidus of approximately 422 ∘C, providing quantitative experimental support for the thermodynamic calculations. Under current operating assumptions, the liquidus temperature is projected to remain essentially constant at approximately 425 ∘C throughout the present campaign, rising by less than 3 ∘C from the initial value of 422 ∘C to 425 ∘C at end of campaign, while remaining well below the nominal 500 ∘C electrorefiner operating temperature, corresponding to a thermal margin of approximately 75 ∘C. Alternative processing scenarios further show that operating choices can materially affect salt inventory, composition, and liquidus behavior. These results demonstrate the value of combining process simulation, thermal measurements, and thermodynamic interpretation to support electrorefiner campaign planning and salt management.

Corrosion mitigation has long been a challenge for the long-term service of both pyrochemical reprocessing and molten salt reactor applications. This work analyzed corrosion coupons made of 2.25Cr-1Mo, American Society of Mechanical Engineers Steel Alloy 387 (ASME SA387), Grade 22, Class 2 steel, that were installed in a uranium electrorefiner containing molten chloride salt at 500 degrees C. Seven of these coupons were removed in 2019 (i.e., after having been submerged in molten salt for 25 years) for analysis and characterization using neutron imaging, optical microscopy, scanning electron microscopy, and wave/energy dispersive x-ray spectroscopy. The results showed that the surface of all seven coupons degraded between 30 and 175 mu m without evidence of brittle fracture. There was no evidence of accelerated corrosion or fractures, and the average steel surface loss was <5 m per year. These results indicate that the environmental conditions inside the electrorefiner were favorable for the steel alloy, making these environmental conditions the primary driver for low corrosion.

This paper investigates an electrochemical process designed to control mixing kinetics in liquid electrolytes. The process utilizes an electrochemical cell with anode and cathode chambers connected by a channel without a membrane. This configuration depletes multiple electroactive species in the cathode chamber while selectively replenishing them in the anode chamber. A key application is the removal of impurity species from the cathode chamber electrolyte and the selective replenishment of primary electroactive species in the anode chamber. Aqueous cell experiments were conducted to validate the theoretical model, and simulations were performed for molten salt electrolytes. This approach minimizes waste by reducing the need for additional electrolyte supplies and extends the utilization of anion species, contributing to environmentally sustainable electrochemical materials processing.

Sodium-cooled fast reactors utilize metallic fuels that include bond-sodium within the fuel element. Molten salt electrolysis at 773 K with eutectic KCl-LiCl mixed with UCl3 as an electrolyte can recover the actinides from spent fuel. The critical factor affecting the useful life of the electrolyte is the increase in liquidus temperature from the accumulation of lanthanides, actinides, and sodium. Therefore, thermodynamic modeling of the KCl-LiCl-NaClUCl3 system was carried out by considering experimental data from the present work and literature as input. The liquidus and solidus temperatures for the two ternary systems, KCl-NaCl-UCl3 and LiCl-NaCl-UCl3, were determined using differential scanning calorimetry. The thermodynamic parameters for pure UCl3 were optimized for liquid and solid states over a wide temperature range. Several constituent binary (AkCl-UCl3; Ak: K, Li, Na) and ternary (KCl-LiCl-UCl3, KCl-NaCl-UCl3 and LiCl-NaCl-UCl3) systems were assessed or reassessed in this work. A new intermediate phase (K3UCl6) was included in the reassessment for the KCl-UCl3 system. There is good agreement between the experimental and calculated thermochemical and phase diagram data for all the systems optimized in the present work. This work is beneficial to determine the effect of NaCl on the liquidus temperature and other thermodynamic properties of KCl-LiCl electrolyte mixed with UCl3 for improving the efficiency of molten salt electrolysis.

The Mk-IV electrorefiner has been in service since 1996 recovering uranium from spent fuels from the Experimental Breeder Rector II and Fast Flux Test Facility sodium-cooled fast reactors. The electrorefiner vessel includes a cadmium pool beneath the salt. The voltage signal between the cadmium pool and reference electrode (Ag/AgCl type) provides information about the chemical condition of the cadmium pool with respect to uranium and zirconium saturation. This information is used to guide process control and sampling decisions. This paper describes how the voltage signal is related to chemistry by the analyses of data from Mk-IV operations and confirmatory experimentation.

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.

A scoping study was performed for chlorinating dross formed during uranium casting operations. 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 degrees C. The addition of FeCl2 chlorinated the uranium metal to UCl3, by reducing the FeCl2 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 ZrCl4, by reducing the FeCl2 to iron metal. In turn, the ZrCl4 chlorinated the uranium oxide to UCl3, by converting the ZrCl4 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 UCl3 in the salt.

Experimental Breeder Reactor II (EBR-II) operated 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. 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.

Molten salt consisting primarily of eutectic LiCl-KCl is currently being used in electrorefiners in the Fuel Conditioning Facility at Idaho National Laboratory. Options are currently being evaluated for storing this salt outside of the argon atmosphere hot cell. The hygroscopic nature of eutectic LiCl-KCl makes is susceptible to deliquescence in air followed by extreme corrosion of metallic cannisters. In this study, the effect of occluding the salt into a zeolite on water sorption/desorption was tested. Two zeolites were investigated: Na-Y and zeolite 4A. Na-Y was ineffective at occluding a high percentage of the salt at either 10 or 20wt% loading. Zeolite-4A was effective at occluding the salt with high efficiency at both loading levels. Weight gain in salt occluded zeolite-4A (SOZ) from water sorption at 20% relative humidity and 40 degrees C was 17wt% for 10% SOZ and 10wt% for 20% SOZ. In both cases, neither deliquescence nor corrosion occurred over a period of 31 days. After hydration, most of the water could be driven off by heating the hydrated salt occluded zeolite to 530 degrees C. However, some HCl forms during dehydration due to salt hydrolysis. Over a wide range of temperatures (320-700 degrees C) and ramp rates (5, 10, and 20 degrees C min(-1)), HCl formation was no more than 0.6% of the Cl- in the original salt.

In pyroprocessing spent nuclear fuels by electrorefining in molten LiCl-KCl salt, it is desired to monitor in real time the UCl 3 concentration in the salt for safeguards purposes. Current chemical analysis of the highly radioactive salt for electrorefining by an inductively coupled plasma technique is inconvenient and usually time-consuming in generating the salt composition results. In this paper, we evaluated whether a simple potentiometry approach can be used for real-time monitoring the concentration of GdCl 3 , which was used as a surrogate for UCl 3 , in LiCl-KCl-GdCl 3 salt by measuring the open circuit potential of a Gd metal electrode with respect to a Ag/AgCl reference electrode (RE) when GdCl 3 salt was incrementally added to the LiCl-KCl salt. Additions of LaCl 3 , CeCl 3 and NdCl 3 salts were used for evaluating the effects of other chloride salts on the selectivity of the Gd metal electrode vs Ag/AgCl RE. While using potentiometry to determine GdCl 3 concentrations, Gd metal was unexpectedly observed to be unstable and dissolved in LiCl-KCl salt when GdCl 3 is present.

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. 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 degrees 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 similar to 10x through this procedure.

Electrochemical measurements of YCl3, ScCl3, GdCl3 and MgCl2 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) Omega; 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 voltammo-grams 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.

The defining characteristic of molten salt reactors (MSRs) is the use of molten salts as the primary heat transfer medium, operating at or near atmospheric pressure, and at temperatures above 500 °C. The fuel may be dissolved into the circulating salt, allowing a possibility for online salt processing, or the fuel may be in a solid form immersed in the coolant salt. Fluoride salt systems have received the most attention and have been used in two demonstration reactors. Nevertheless, chloride salt systems are also proposed for development and deployment. Molten salt properties, their purification, and corrosion control in molten salt systems are important aspects of molten salts for reactor application.

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.

During the electrometallurgical pyroprocessing of used metallic nuclear fuel, the chopper operation cuts the fuel element into segments and distributes these segments between the anode basket and plenum can. This paper discusses a method of systematically assessing the mass balance of partitioned materials between the anode basket and plenum can. The proposed method fully accommodates the mass closeout measurements from the chopper operation and automates the assessment of chopper-induced material partitioning. With a set of reasonable assumptions, the proposed approach allows a formal uncertainty analysis on the partitioned masses as well. An illustrative calculation example adopted from the ongoing fissium fuel processing campaign is given.

Anodic stripping voltammetry (ASV) and cyclic voltammetry (CV) measurements at 773, 823, and 873 K were made of uranium trifluoride (UF3) in lithium fluoride-sodium fluoride-potassium fluoride eutectic (FLiNaK) and uranium trichloride (UCl3) in lithium chloride-potassium chloride eutectic (ClLiK). ASV data were used to estimate the charge transfer coefficients, exchange current densities, and activation energies of the uranium reactions. Charge transfer coefficients of both salt systems were within the range of 0.17 to 0.38. Exchange current densities in the fluoride and chloride salts were estimated within the range of 0.060 to 0.12 A cm(-2). Activation energy of uranium exchange current was 16.0 kJ mol(-1)in the fluoride salt and 28.9 kJ mol(-1)in the chloride salt. Kinetics of charge transfer were found to be faster in FLiNaK. Analyses of the CV data suggest the electrochemical system was diffusion controlled and irreversible in the chloride and fluoride salt mixtures. Diffusion coefficients of uranium in the range of temperatures were on the order of 10(-5)cm(2)s(-1)in both systems. Greater values of diffusivity in ClLiK are attributed to its lower density compared with FLiNaK. Activation energy of uranium diffusion in the fluoride and chloride salt mixtures were 53.4 and 89.4 kJ mol(-1), respectively.