This thesis investigates the possibility of using magnetic flow meters to measure the flowrate of molten salts in Seaborg Technologies Compact Molten Salt Reactor (CMSR).There is a need to accurately measure the flow rate in salt circulation systems to ensureproper operation of the entire facility. The requirements and criteria for the operationof a magnetic flow meter are studied, from which a model is constructed in COMSOLMultiphysics. The flow meter characteristics are analysed in COMSOL by performingsteady-state magnetohydrodynamic (MHD) simulations and by doing a sensitivity anal-ysis of the velocity field and the magnetic field strength. The induced electric potentialdifference in the flow meter when the reactor is at a maximum designed thermal power isin the range of 65 mV when using a normal inlet flow profile. The effect of the velocityfield is studied for two velocity profiles, and it indicates that the velocity profile alters theinduced potential difference even though the mass flowrate is the same. The magneticfield strength increases the electric potential difference when it is increasing, which isaccording to theory. The results indicate that magnetic flow meters are a viable optionfor Seaborg’s CMSR. However, further analysis is needed regarding the materials usedto ensure proper operation of the flow meter.

This thesis documents research done for a transient hot wire system that will be used in future thermal conductivity measurements of molten salts. Research done with molten salts have been limited because of erroneous measurement capabilities, but the current research strives to introduce a new technique to accurately record thermal conductivity over a wide range of temperatures. This work follows up on past transient hot wire researchers whose designs and tests produced an instrument that can measure the thermal conductivity of molten metals up to 750 K. The transient hot wire (THW) technique has been selected to be used in molten salt to derive thermal conductivity values. While running a THW test in molten salts is outside the scope of this thesis, a modular system has been created for the use of running transient hot wire test that allows for a robust and repeatable testing. A PEGDA/galinstan sensor is used for the validation of the system. A robust GUI has been created to automate the experimental procedure in a glovebox environment. The inverse finite element method has been paired with a non linear fit script to optimize calculations and reduce run times. Test have been done to determine the thermal conductivity of PEGDA. The overall uncertainty of the thermal conductivity measured with the PEGDA sensor is estimated to be ±5% at a 95% confidence level. With a THW system implemented and validated a sensor has been designed to work in molten salts. A model has been created in two separate FEA programs to validate design changes and material properties. The sensor is made up of a chemical vapor deposition (CVD) diamond substrate and tungsten wires to overcome corrosion and heat challenges introduced when measuring molten salts. New manufacturing processes have been designed to allow the technique to use these materials in the THW sensor design. The selected material properties of the sensor and extensive finite element work have laid down the ground work for future experimentation and understanding of the thermal properties of molten salts. It is predicted that the CVD diamond (carbide) apparatus design will use the THW techniques to operate with an estimated accuracy of ±3% over a wide range of temperatures, from ambient up to 1200 K. Manufacturing of the diamond-tungsten sensor have proven the viability of depositing tungsten wire onto CVD diamond and growing a secondary layer of CVD diamond over the tungsten wire.

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.

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Corrosion is a major challenge in the use of molten fluoride salt as a coolant in molten salt reactors (MSRs). A promising way of satisfying the two requirements of high strength and corrosion resistance is to clad structural alloys with a corrosion resistant material. Four candidate cladding and structural alloy combinations—stainless steel 316L and Incoloy 800H structural alloys either diffusion bonded to Hastelloy N or electroplated with nickel—were thermally aged at 700 °C for two to eight months. Based on measured concentration profles, the diffusion resistance of the four material combinations was compared and diffusion results were extrapolated to an end of reactor lifetime. The most important conclusion from this work is that Hastelloy N is highly likely to be signifcantly more diffusion resistant than nickel. The difference in diffusion resistance between Incoloy 800H and stainless steel 316L is relatively small. Two methods were used for extrapolating experimental diffusion results: (1) a diffusion model and calculated diffusion coeffcients and (2) simulations with Thermo-Calc DICTRA. Some simulations were carried out with a corrosion boundary condition of near-zero chromium concentration, demonstrating the potential of simulations for predicting diffusionlimited corrosion in molten fluoride salts. A surprising result of these simulations is that decreasing the thickness of Ni plating did not increase the thickness of diffusion zones in underlying structural alloys. / Thesis / Master of Applied Science (MASc)

The first Molten Salt Reactor (MSR) was designed and tested at Oak Ridge National Laboratory (ORNL) in the 1960's, but recent technological advancements now allow for new components, such as heat exchangers, to be created for the next generation of MSR's and molten salt-cooled reactors. The primary (fuel salt-to-secondary salt) heat exchanger (PHX) design is shown here to make dramatic improvements over traditional shell-and-tube heat exchangers when changed to a compact heat exchanger design. While this paper focuses on the application of compact heat exchangers on a Molten Salt Reactor, many of the analyses and results are similarly applicable to other fluid-to-fluid heat xchangers. The heat exchanger design in this study seeks to find a middle-ground between shell- and-tube designs and new ultra-efficient, ultra-compact designs. Complex channel geometries and microscale dimensions in modern compact heat exchangers do not allow routine maintenance to be performed by standard procedures, so extended surfaces will be omitted and hydraulic diameters will be kept in the minichannel regime (minimum channel dimension between 200 μm and 3 mm) to allow for high-frequency eddy current inspection methods to be developed. High aspect ratio rectangular channel cross-sections are used. Various plant layouts of smaller heat exchanger banks in a "modular" design are introduced. FLUENT was used within ANSYS Workbench to find optimized heat transfer and hydrodynamic performance. With similar boundary conditions to ORNL's Molten Salt Breeder Reactor's shell-and-tube design, the compact heat exchanger interest in this thesis will lessen volume requirements, lower fuel salt volume, and decrease material usage. / Master of Science

Nuclear energy has gained crucial importance since it has a minor impact on climate change and greenhouse gas releases; additionally, the other energy sources are insufficient to reach the world's energy needs without nuclear energy. Another sign that the Generation IV International Forum (Kelly, Gen IV International Forum: A decade of progress through international cooperation, 2014) has pointed out is to utilize uranium resources to the maximum and recycle spent nuclear fuel through burn-up in the Generation IV reactor designs, one of which is the molten salt reactor (MSR). Therefore, the MSR can use the spent nuclear fuel as a fresh fuel when the actinides recycle. That reprocessing of spent fuel could be one of the opportunities to contribute to future nuclear energy goals. This study aims to develop a modified pyroprocessing method to prepare molten salt fuels for MSR from spent oxide nuclear fuel that was burned in light water reactors (LWRs). The process diagram illustrated as (1) spent fuel treatment, (2) chopping and voloxidation of spent oxide fuel, (3) oxide reduction of spent fuel, and then depending on the fuel structure and composition for the MSR, it continues by one or two of the following; – electrorefining, – chlorination, and – fluorination. The subject of this study focused on oxide reduction in two categories: chemical reduction and electrochemical reduction. The system designs have been optimized in calcium salts since they have high calcium metal and calcium oxide solubility. The significant results indicated that both methods would substantially reduce the solid uranium dioxide pellet. The chemical reduction will reduce the total solid pellet at 850oC in the composition of 55.73mol%CaCl2-12.37mol%CaF2-26.58mol%Ca-5.32mol%UO2 over 12 hours. The total reduction in the electrochemical test is seen at 850oC during 12 hours with a salt composition of 79mol%CaCl2-17mol%CaF2-4mol%CaO. These oxide reduction mechanisms are convenient ways to reprocess spent oxide fuel from LWRs to utilize in the MSR. Additionally, the reduced fuel is also applicable to using other next-generation reactors. The prospect of this research is the explicit comparison between chemical and electrochemical methods in calcium salts. / M.S. / Nuclear energy is a crucial energy production to meet the world’s future energy needs. The 6 (six) next-generation reactor design has been determined based on their sustainability, economic, and peaceful application for the world. One of those designs is molten salt reactors (MSRs) which have more attention due to their fuel choice. Most MSRs use the reprocessed fuel from current reactors or the fuel with the breeder blanket that creates more fuel while the reactor operates. This study aims to provide a diagram showing the various steps involved in the preparation of molten salt fuel from spent oxide fuel, which is a mainly utilized form of fuel in current and previous operations. The flowsheet’s first step is the treatment of spent fuel that releases most of the decay heat. The second step is that spent fuel chopping and voloxidation, which meets the requirements of removing gas products and cladding material from used fuel. Afterward, the spent oxide fuel reduces into its metal form chemically or electrochemically in oxide reduction. Then, the molten salt fuel could be fabricated in n one or two more steps from reduced metals: electrorefining, chlorination, or fluorination. Chlorination and fluorination pass through the specific gas components to convert the metal forms into salt. Electrorefining could be applied to arrange the composition of the reduced metal, and this stage is strongly dependent on the MSR designs; it may get eliminated due to its unnecessity. The oxide mechanisms mentioned above were examined under different design conditions to acquire a total reduction of the fuel pellet in calcium salts. The chemical reduction and electroreduction experiments have shown the reduced whole pellet at 850oC with two different salt mixtures. The design impacts of the reduction mechanism were discussed extensively between chemical and electrochemical reductions to identify the benefits and limitations.

With the increasing research on advanced reactors, molten salt reactors have been recognized for their potential. As with any advanced reactor concept, each component and material must be thoroughly investigated before any reactors of that type are created. One of the most pressing issues in MSR research is that of the salts themselves. Though there are a multitude of salts to choose from when designing such a reactor, many of these salts lack the extensive research required to fully understand them. Across the decades there have been many studies that have investigated select molten salts, but there are a few problems with many of those studies. Those problems are the following: prior papers use obsolete and less reliable methods for their measurements, the papers don't investigate the salts across a wide enough range of temperatures nor at varying compositions, and finally many of the salts that are seen as candidates today were not given as much attention when molten salt reactors were first conceptualized which has resulted in a lack of research on them. Indeed, the research into these salts is lacking in many ways. This study seeks to investigate a collection of promising coolant salts in depth with acknowledgment to those past studies. LiF-NaF-KF (46.5-11.5-42.0 mol%) will be used as a calibration standard and for the purpose of verifying our methodology. Specifically, FLiNaK was used in the development of volume-height curves as calibration for density measurements. NaOH-KOH of four different compositions ( 0.5-0.5mol%, 0.55-0.45mol%, 0.6-0.4mol%, and 0.65-0.35 mol%) will be evaluated for their densities and heat capacities. And finally, BeF2-NaF(43-57mol%) will be evaluated within the question of if the properties are desirable enough that the dangers posed by beryllium are an acceptable risk. BeF2-NaF will have melting point, heat capacity, density, and vapor pressure measurements performed. Additionally, extensive impurity analysis and removal (via an HF gas system) was done to our BeF2-NaF samples. The melting point and heat capacity were evaluated using dynamic scanning calorimetry (DSC), the vapor pressure was evaluated using thermogravimetric analysis (TGA), and the density was measured using a system similar to the Arrhenius method that measures height. / Master of Science / Decades have passed since the discussion of nuclear energy began. Although great progress has been made in the field, the nuclear reactors in use today consist mainly of boiling water reactors (BWRs) or pressurized water reactors (PWRs). As reliable as these reactors have become, one can no longer ignore the fact that there is a multitude of other options for how a reactor can be built and operated. Options that provide greater safety and more energy output. Many reactor concepts of the past were discounted for the extensive research that would be required to make use of them. However, as time has passed and technology has improved, that research has become more and more possible. Many advanced reactors are the result of that attention to the reactor concepts and materials of the past that couldn't be given the attention that they deserve until now. Molten salt reactors (MSRs) are one of those promising concepts. However, before they can be built every part of the reactor, from the structure to the materials, must be entirely understood. One of the most pressing issues in MSR research is the properties of the salts in consideration for use. Though there are a multitude of salts to choose from when designing such a reactor, many of these salts lack the extensive research required to fully understand them. Across the decades there have been many studies that have investigated select molten salts, but there are a few problems with many of those studies. Those problems are the following: the papers are so old that the methods that were used are now obsolete, the papers don't investigate the salts across a wide enough range of temperatures nor at varying compositions, and finally many of the salts that are seen as candidates today were not given as much attention when molten salt reactors were first conceptualized which has resulted in a lack of research on them. Indeed, the research into these salts is lacking in many ways. This study seeks to investigate a collection of promising coolant salts in depth with acknowledgment to those past studies. LiF-NaF-KF will be used as a calibration standard and for the purpose of verifying our methodology. A multitude of different compositions of NaOH-KOH will be evaluated for their densities and heat capacities. And finally, BeF2-NaF will be evaluated within the question of if the properties are desirable enough that the dangers posed by beryllium are an acceptable risk. BeF2-NaF will have melting point, heat capacity, density, and vapor pressure measurements performed. Additionally, extensive impurity analysis and removal was done to our BeF2-NaF samples.

This work demonstrates how ICP-MS can be employed to evaluate cesium volatilization from LiF-BeF2 (Flibe) salt with ab initio molecular dynamics studies used as corroborating data to better understand cesium behavior. Using mixtures of Flibe with 2 mol% CsF, it was found that cesium was stable within the salt melt at temperatures between 500-650°C over a time span of 8 hours. At 800°C, cesium vaporized from the salt at a rate of 0.83% / hr with a mass flux of 0.0023 g Cs / cm2hr. The atomistic modeling results show poor solvation of Cs at 500°C and 800°C, with stability preferred at 650°C. Specialized equipment and procedures were needed to enable this work, especially those required for safe handling of beryllium containing salts. The methods, custom equipment, and important considerations for working with high-temperature fluoride salts are detailed in this thesis.