Report (M. Eng.)--Virginia Polytechnic Institute and State University, 1991. / Includes bibliographical references (leaves 27-28). Also available via the Internet.

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The next generation of nuclear power sources, Gen. IV, will include an emphasis on small, modular reactor (SMR) designs, which will allow for standardized, factory-based manufacturing and flexibility in the design of power plants by utilizing one or several modular reactor units in parallel. One of the reactor concepts being investigated is the Molten Salt Reactor concept (MSR), which utilizes a molten salt flow loop to cool the reactor and transfer heat to the power conversion cycle (PCS).Here, the use of a supercritical carbon dioxide (S-CO2) Brayton cycle is assumed for that PCS. The purpose of this thesis is to investigate the heat exchange between these two systems and to determine the suitability of a common heat exchanger concept, the shell-and-tube heat exchanger (STHE). This was accomplished using a code written in Python programming language that optimized the geometry ofa baffled STHE for a range of conditions the reflect MSR power plants currently in the design or concept stages. Star-CCM+ computational fluid dynamics (CFD)software was used to visualize the flow patterns of molten salt and CO2 in these STHE designs, and it was also used to determine heat transfer coefficients and pressure drops. These values were compared to those calculated by the optimizer code in order to validate its results. Finally, modularity analysis was performed for these STHE designs. Trends were generalized from these results that will contribute to judgments about the suitability of STHE’s for use with MSR’s and S-CO2.

Molten salt reactors (MSRs) using chloride-based salt-matrixes as coolants or fuels are a promising option for advanced nuclear reactors, but the extreme temperatures and corrosivity of molten salts pose a challenge for implementation. Molten MgCl2-NaCl-KCl is a viable candidate for MSRs that is considered in this work. Thermochemical properties are derived from electrochemical tests that aid in characterizing the properties of salts. To study these properties, some work has proposed using a three-electrode system with a reference electrode housed in a ceramic membrane. This research aims to develop a stable high-temperature reference electrode using a ceramic membrane that is then applied to develop an on-line monitoring system of Ni-Cr alloy corrosion in chloride salt. A mullite tube used as the membrane of a Ni(II)/Ni reference electrode in molten MgCl2-NaCl-KCl is studied. The performance of two different membrane thicknesses (1.325mm and 0.255mm) was studied in temperature ranges from 635oC to 835oC and data collected on the calculated formal potential of the Ni(II)/Ni system. Tests indicated that the results were stable and repeatable, and the formal potential for both systems differed from the previous experimental data by 0.12V at most, indicating that the system can be applied as an effective reference electrode. Using the reference electrode, on-line monitoring the corrosion of Ni-15wt.%Cr, Ni-20wt.%Cr, and Ni-30wt.%Cr was studied for 120 hours in MgCl2-NaCl-KCl. The on-line measurements showed the concentration changes of dissolved Cr and Ni by corrosion in the bulk molten salt. This work confirms that Ni(II)/Ni reference electrodes with a mullite tube membrane are stable and effective in molten chloride salt systems, particularly MgCl2-NaCl-KCl. The mullite membrane prepared by the manufacturer may be used directly for electrochemical applications without polishing, simplifying the reference electrode manufacturing process, and making it easier to replicate. The use of a Ni(II)/Ni reference electrode provides an avenue to study a different range of salt systems than previous reference electrodes allowed, particularly alloys in chloride salts at high temperatures. This work also confirms that the mullite tube may be used to perform on-line analysis of alloy corrosion in high temperature molten chloride salts. The study of Ni-Cr alloys in chloride salts better prepares the nuclear industry to select coolant salts and alloy containers with the best set of thermochemical and corrosion resistant characteristics for MSRs. / Master of Science / The United States receives approximately 18% of its energy from nuclear technology. Many of the reactors supplying this energy are at the end of their lifecycle and the decommissioning of some of these plants has already begun. In order to replace this older generation of nuclear reactors, a safer and cheaper option has been suggested: Molten Salt Reactors. Molten salt reactors (MSRs) using high temperature salts as a fuel or coolant are a promising option, but the extreme conditions of molten salts pose a challenge for construction and use of MSRs. Molten MgCl2-NaCl-KCl is a salt being considered for MSR application, and is considered in this work. Properties of the salts considered for MSRs are being studied diligently before implementation of these reactors. Electrochemical tests are used to study and monitor these properties. These electrochemical tests use a three-electrode system with a reference electrode housed in a membrane. In this work, a mullite tube is used as a ceramic membrane for a reference electrode in molten MgCl2-NaCl-KCl. The performance of two different membrane thicknesses (1.325mm and 0.255mm) was studied in temperature ranges from 635oC to 835oC. Results indicate that the system is an effective reference electrode. Using this innovative reference electrode, a method of monitoring on-line corrosion of Ni-15wt.%Cr, Ni-20wt.%Cr, and Ni-30wt.%Cr alloys was studied for 120-hour time periods during exposure to MgCl2-NaCl-KCl. This work confirms that reference electrodes with a mullite membrane may be used for electrochemical applications when studying molten chloride salts. The use of a Ni(II)/Ni reference electrode with a mullite membrane provides an avenue to study a different range of salt systems than previous reference electrodes and ceramics allowed, particularly chloride salts. Additionally, this mullite membrane Ni(II)/Ni reference electrode system may be used for monitoring on-line corrosion of Ni-Cr alloys in chloride salt systems.

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.

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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The Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR) was modeled using the neutronics analysis code SCALE6.0 and the thermal-hydraulics and kinetics modeling code RELAP5-3D with objective to devise, analyze, and evaluate the feasibility and stability of a start-up procedure for this reactor using natural circulation of the coolant and under the Loss Of Offsite Power (LOOP) conditions. This Generation IV reactor design has been studied by research facilities worldwide for almost a decade. While neutronics and thermal-hydraulics analyses have been previously performed to show the performance of the reactor during normal operation and for shutdown scenarios, no study has heretofore been published to examine the active or passive start-up of the reactor. The fuel temperature (Doppler) and coolant density coefficient of reactivity of the LS-VHTR were examined using the CSAS6 module of the SCALE6.0 code. Negative Doppler and coolant density feedback coefficients were calculated. Two initial RELAP5 simulations were run to obtain the steady-state conditions of the model and to predict the changes of the thermal-hydraulic parameters during the shutdown of the reactor. Next, a series of step reactivity additions to the core were simulated to determine how much reactivity can be inserted without jeopardizing safety and the stability of the core. Finally, a start-up procedure was developed, and the restart of the reactor with natural convection of the coolant was simulated. The results of the simulations demonstrated the potential of a passive start-up of the LS-VHTR.

Nowadays there is research into molten salt reactors. The use of chlorine-based salts, which would be more available than known fluoride salts, is envisaged. The subject of research is not only the chemical and physical properties of chloride salts, but also their behavior in the neutron field and the influence of neutron balance inside the reactor. Many properties can also be determined using calculations that draw information from scientific nuclear libraries (endf). The purpose of this work is to compare important nuclear libraries with each other, and also to compare the reaction rates calculated from the library data with the reaction rates obtained by self-measurement. The preview will include a description of the necessary activities associated with the preparation of measurements, instructions for compiling the computer program NJOY and the process of the measurement itself. At the end of the work will be summarized the results and statements of which nuclear library is the closest in its values to the results of experiments.