The work within this thesis focuses on the study of alloy formation using an active liquid metal electrode for fundamental analysis and for the extraction and separation of the lanthanides and actinides in a pyroprocessing system. The electrochemical work herein is performed in a molten salt of lithium chloride and potassium chloride at its eutectic point (LKE). This salt is a likely candidate for pyroprocessing due to its relatively low melting point and resistance to degradation on exposure to high levels of radiation. The active electrode material under examination is bismuth due to its propensity to alloy with other elements, its relatively low melting point, high density and non-toxicity. The alloying processes studied are those of bismuth-lithium and bismuth-cerium. Lithium is the limiting reduction reaction defining the negative solvent limit in LKE. As a result, understanding the processes that would occur if the electrode were to be pushed to such negative potentials is of significant importance. Cerium is a commonly-used surrogate for plutonium, which is an element of relatively high concentration in waste nuclear fuel and is of significant interest to the nuclear international community in waste fuel recycling. This work examines the alloying processes in terms of which intermetallic compounds are formed and by what mechanisms. This is achieved through the use of co-deposition on a macro tungsten rod, employing a number of electrochemical techniques to extract pertinent information. Lithium electrodeposition and alloying with bismuth (at the negative solvent limit) was found to form BiLim alloy with increasing m at more reducing potentials, followed by the deposition of near pure lithium. Mixing of these two then gave rise to specific bismuth-lithium alloys and the apparent ejection of a lithium metal fog into the molten salt, which resulted in the chemical reduction of Bi3+ and the loss of the bismuth electrodeposition current. When electrodepositing cerium on, and alloying with, bismuth, the formation of intermetallic compounds is governed by potential with a maximum BiCem stoichiometry of m = 1 with equimolar Bi3+ and Ce3+. However, at concentrations of cerium greater than that of bismuth, alloys much richer in cerium were also deposited at more negative potentials. There is evidence that deposited cerium may also escape into solution and chemically react with Bi3+. In-house microelectrodes are also developed and used for this purpose, both through co-deposition and direct alloy formation on a liquid bismuth thin-film microelectrode. This work demonstrates that these devices provide a richness of information due to their highly beneficial microelectrode properties. A means of controllably depositing bismuth from an aqueous plating bath, without dendrite formation, on both platinum and tungsten microelectrodes was devised. This was followed by electrodeposition of bismuth films on these devices in LKE. Platinum was found to be an active electrode material, alloying with bismuth, while tungsten remained inert. Nonetheless, both electrode types produced characteristic microelectrode behaviour, which was successfully used to determine the diffusion coefficient of bismuth in LKE. A comparison of bismuth-cerium and cerium alloying on a thin film liquid bismuth microelectrode found that the latter indicated the formation of BiCe2 where only BiCe had been seen previously during co-deposition in an equivalent salt. This is thought to be due to the thin film liquid bismuth microelectrode configuration with enhanced Ce3+ mass transport. This response was also used to calculate the diffusion coefficient of cerium inside the bismuth film, which was found to be slightly slower than for Ce3+ in LKE.

The electrorefiner (ER) is the heart of pyroprocessing technology operating at a high-temperature (723 K – 773 K) to separate uranium from Experimental Breeder Reactor-II (EBR-II) used metallic fuel. One of the most common electroanalytical methods for determining the thermodynamic and electrochemical behavior of elemental species in the eutectic molten salt LiCl-KCl inside ER is cyclic voltammetry (CV). Information from CV can possibly be used to estimate diffusion coefficients, apparent standard potentials, transfer coefficients, and numbers of electron transferred. Therefore, predicting the trace of each species from the CV method in an absence of experimental data is important for safeguarding this technology. This work focused on the development an interactive computational design for the CV method by analyzing available uranium chloride data sets (1 to 10 wt%) in a LiCl-KCl molten salt at 773 K under different scan rates to help elucidating, improving, and providing robustness in detection analysis. A principle method and a computational code have been developed by using electrochemical fundamentals and coupling various variables such as: the diffusion coefficients, formal potentials, and process time duration. Although this developed computational model works moderately well with reported uranium data sets, it experiences difficulty in tracing zirconium data sets due to their complex CV structures. Therefore, an artificial neural intelligent (ANI) data analysis has been proposed to resolve this issue and to provide comparative study to the precursor computational modeling development. For this purpose, ANI has been applied on 0.5 to 5 wt% of zirconium chloride in LiCl-KCl eutectic molten salt at 773 K under different scan rates to mimic the system and provide current and potential simulated data sets for the unseen data. In addition, a Graphical User Interface (GUI) through the commercial software Matlab was created to provide a controllable environment for different users. The computational code shows a limitation in high concentration CV prediction, capturing the adsorption peaks, and provides a dissimilarity. However, the model is able to capture the important anodic and cathodic peaks of uranium chloride CV which is the main focus of this study. Furthermore, the developed code is able to calculate the concentration of each species as a function of time. Due to the complexity of the CV of zirconium chloride, the computational model is used to predict the probability reactions occurring at each peak. The resulting study reveals that the reaction at the highest anodic peak is related to the combination of 70% Zr/Zr+4 and 30% Zr/Zr+2 for the 1.07 wt% and 2.49 wt% zirconium chloride and 30% Zr/Zr+4 and 70% Zr/Zr+2 combination for 4.98 wt% ZrCl4. The proposed alternative ANI method has demonstrated its capability in predicting the trend of species in a new situation with a high accuracy on predictions without any dissimilarity. Two final structures from zirconium chloride study which high accuracy (that is, a low error) are related to [9, 15, 10]-18 and [10, 11, 25]-19. These two final structures have been applied on uranium chloride salt experimental data sets to further validate the ANI’s ability and concept. Three different fixed data combinations were considered. The result indicates that by increasing the number of training data sets it does not necessarily help improving the prediction process. ANI implementation outcome on uranium chloride data set illustrates a good prediction with a specific fixed data combination and [9, 15, 10]-18 structure. Thus, it can be concluded that ANI is a promising method for safeguarding pyroprocessing technology due to its robustness in predicting the CV plots with high accuracy.

Over the past several years as interest in cleaner energy sources has grown nuclear power has come to the forefront. However, as interest in nuclear power grows so does the concern over the amount of high-level radioactive waste produced. Currently, the most popular way to deal with spent nuclear fuel is interim storage until a viable treatment option becomes available. Simply waiting for spent fuel to become safe to handle will take thousands of years and is not a reasonable long-term solution. We will soon run out of space in our spent fuel pools and while more dry storage space can be found it is not an ideal solution. One answer to this problem is the reprocessing of spent nuclear fuel. This could be done with either the plutonium uranium reduction extraction (PUREX) method or the pyroprocessing method. Since PUREX does not have the same level of built-in proliferation resistance as pyroprocessing, pyroprocessing is starting to be seen as a good alternative method. Pyroprocessing would take the spent nuclear fuel from a light water reactor and make it into a metal-based fuel that could be used in certain advanced reactors. Molten salt reactors are of particular interest when it comes to reprocessing spent nuclear fuel because of their unique property of using a liquid fuel. Molten salt reactors and spent fuel reprocessors could be directly connected which would save both time and money as little storage and transportation would need to be considered. Regardless of how and where the used nuclear fuel is being recycled it is important to be able to keep track of the major actinides and fission products in the fuel as it moves through the process. Electrochemical concentration measurements are straightforward and well understood in static cases when there is only a single element to consider. When additional elements are added, or the system is flowing rather than static, things get slightly more complicated but are still decently well understood. However, in the case of spent fuel reprocessing the system is both be flowing and contains much more than a single element. This case is not well understood and is what this study attempts to understand. Two different rotating electrodes were designed to simulate flowing conditions in an electrochemical cell. The first was a tungsten rotating disk electrode (RDE) and the second was a graphite RDE. We were not able to fully insulate the tungsten RDE and were therefore unable to achieve reliable results. Because of this the tungsten design was put aside in favor of the graphite design, which did prove to be sufficiently insulated. The graphite RDE was tested in two different salt systems: LiCl-KCl-NiCl2-CrCl2 and LiCl-KCl-EuCl3-SmCl3. In the nickel-chromium system the graphite RDE produced the expected results. The calculated nickel concentration was found to be within 10% of the measured concentration. Calculations of the chromium concentration, however, were not possible due to the deposition of nickel on the graphite surface, which increased the surface area of the working electrode. When the graphite RDE was tested in the second system it was first tested in the ternary salt LiCl-KCl-EuCl3 and was able to produce decent results. The concentration of europium calculated from the scan was within 10% of the measured value. When the RDE was tested in the LiCl-KCl-EuCl3-SmCl3 salt the results did not come out as expected. Several rather noisy CV curves were obtained and no alterations to the cell seemed to affect them. At this point it was determined that the reason for the confused scans was a connection problem that could not be remedied within the time frame of this study. While this study does not accomplish the task it set out to do, it is a good step in the direction toward understanding flowing systems containing more than a single element of interest and has successfully designed a reliable graphite RDE. / Master of Science / As interest in nuclear power continues to grow, so does the concern over the amount of high-level nuclear waste produced. More nuclear power means more nuclear reactors and thus more spent nuclear fuel to be dealt with. Currently most used nuclear fuel ends up in interim storage facilities where it is meant to wait until it is safe to handle, which could take several thousand years, or until a reliable disposal method is determined. On this path the amount of spent fuel that requires storage will quickly overrun the amount of storage space safely available. One way to reduce the amount of nuclear waste is to reprocess it to be used as fuel for different types of reactors. The pyroprocessing method takes the spent nuclear fuel from a typical light water reactor and recycles it into fuel that can be used in certain types of advanced reactors, such as molten salt reactors (MSR) and sodium-cooled fast reactors (SFR). The reprocessing system works to separate the usable actinide elements, such as uranium and plutonium, from any fission products or other contaminants. During these processes it is important to be able to keep track of the concentrations of each of these different elements to ensure proper separation. This study examines the use of two rotating disk electrode (RDE) designs that are meant to simulate the flowing conditions found in many reprocessing systems. These RDEs were to be used to measure the concentrations of different elements in molten salt systems. The first design, a tungsten RDE, could not be properly insulated and thus was unable to produce reliable results when tested in the electrochemical cell. The second design was a graphite RDE. This design did prove to be properly insulated and was able to produce good results when tested in the cell. The graphite RDE was tested in both LiCl-KCl-NiCl2-CrCl2 and LiCl-KCl-EuCl3-SmCl3. In the first system the concentration of nickel was correctly calculated using the data collected with the graphite RDE, while the chromium concentration could not be due to the nickel deposition on the graphite. In the second system, good results were obtained before the SmCl3 was added to the salt. At this point a connection error became apparent and reliable results were no longer possible. Further study is needed to understand the LiCl-KCl-EuCl3-SmCl3 system using the graphite RDE.

Nuclear energy presents environmental benefits, yet the challenge of radioactive waste management persists. Advanced solutions, such as Molten Salt Reactors (MSRs), require a more profound understanding of molten salt chemistry. This research aims to develop tools, including a depletion simulator, molten salt electrochemical simulator, and a fluoride-based thermodynamic reference electrode for electrochemical purification. The computationally inexpensive depletion simulator allows for exploration into extraction and processing strategies for molten salt reactors. An illustrative case study on Mo-99 production from MSRs demonstrates the practical application of the theoretical framework, emphasizing the need for optimization in extraction effectiveness and separation difficulty. The electrochemical simulator, employing first-principles models, contributes to both nuclear technology and the broader field of electrochemistry. Detailed analyses of linear sweep voltammetry (LSV) for uranium deposition, coupled with numerical simulations for diffusion coefficient measurements, enhance precision in experimental methodologies. The study into fluoride-based thermodynamic reference electrodes provides validation of boron nitride as a viable ion-exchange membrane permeation of oxide impurities as a contributing factor to reference electrode failure, and an investigation of an alternative reference electrode chemistry based on the equilibrium between U3+ and U4+. This novel reference electrode chemistry enabled electrochemical purification of fluoride-based salts which were characterized with square wave voltammetry and have less than 30 ppm O2-. In summary, this work not only advances theoretical understanding but also provides practical tools for nuclear energy and electrochemical processes. Its interdisciplinary approach of integrating theory, computation, and experimentation represents a significant stride toward the responsible and balanced utilization of nuclear power to address global energy needs and challenges.

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