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Fuel cells are one of the most promising clean energy technologies under development. But a constraining factor in their further development is related to operating temperature ranges of current fuel cell systems, which is either low or high temperature. The intermediate temperature (200¡ÃƒÂ£C to 600¡ÃƒÂ£ C) would be the most desirable temperature range for a fuel cell for most applications, but there is no existing mature fuel cell technology in this range, mainly because of an absence of appropriate electrolytes. An effort to develop an intermediate-temperature molten-salt electrolyte fuel cell (IT-MSFC) was undertaken in this study. As a start, molten KOH was used as an electrolyte around 200¡ÃƒÂ£ C supported on a porous matrix. Tests used Pt loaded carbon cloth to be the electrode-catalyst layer, hydrogen and oxygen as fuel. The major challenge for this fuel cell was to hold electrolyte within a suitable porous support layer, without crossover of fuel gas during operation. Performance was short-lived, thus several ceramic materials were investigated in this research, including Zirconia felt, Zirconia disk, and porous NiO. To evaluate the properties of KOH molten salts working for IT-MSFCs, the performances were compared to fuel cell tests with KOH saturated solution and phosphoric acid with the same electrolyte support. KOH molten salt has large potential to work as electrolyte, with an open circuit voltage (OCV) of 1.0 V, and had linear performance curve between 1.0 V and 0.6 V, which is characteristic of fuel cells with low kinetic overpotentials. The highest performance was got by using porous NiO support in certain porosity range. Longevity of the fuel cell was a little better than the former, but still far from practical application. The result suggested that the capillarity, permeability and compatibility of support material are essential for performance of this type of fuel cell. Besides the problem of electrolyte II retention by the support matrix, unsuitable water management, degradation of the gas diffusion layer and catalyst may also reduce the fuel cell performance. Although this work is at a preliminary stage, it has demonstrated the immense potential of IT-MSFC, and a great deal of additional work will be required to produce a practical fuel cell.

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

The molten salt reactor (MSR) is one of the advanced nuclear reactors expected to be alternatives to the conventional water-cooled nuclear reactor systems. Despite many advantages of MSRs, properties of molten salts have not been sufficiently measured in previous studies. In addition, the corrosion of structural alloys by molten salt is the biggest challenge for the operation of MSRs. This study focuses on measurements of thermophysical and thermodynamic properties of fluoride salt fuels, salt purification, and the degradation of structural materials in static and flowing molten-salt fuels. For the measurements of properties, phase transition, specific heat capacity, vapor pressure, contact angle on nuclear-grade graphite, and density were measured. The methodologies for the property measurements used in this study were validated by measuring the properties of metals or salts that have been well studied. For the flow-induced corrosion tests, the salt flow with different velocities was simulated by rotating the stainless steel 316H (SS316H) specimens in molten NaF-KF-UF4 (FUNaK) contained in glassy carbon crucibles at 1073 K. Salt samples were intermittently collected to monitor concentration changes of corrosion products in the salt, and surfaces and cross-sections of post-test SS316H specimens were analyzed to study their corrosion behaviors. Different batches of FUNaK were synthesized using different methods of purification, such as thermal purification, U-metal purification, and hydrofluorination with electrochemical purification (chemical purification) to study impacts of salt purification on the corrosion of SS316H. The corrosion test of SS316H by thermally purified FUNaK showed that the Fe concentration increased at the beginning and then decreased while the Cr concentration continued increasing while the rate decreased. In addition, (Cr, Fe)7C3 layers, Cr-metal particles, and dendritic structures concentrated with Cr and Fe were observed on the glassy carbon crucible after the 2 m/s test. The U-metal purification and hydrofluorination with electrochemical purification reduced concentrations of oxygen and hydrogen in FUNaK and mitigated the corrosion of SS316H significantly. The infiltration of the fluoride fuel salts into graphite and the fluorination of graphite by the salts at different pressures and temperatures were also studied. The salt infiltration into graphite at pressures above its threshold pressure was observed, and the formation of carbon fluorides on the surface of post-test graphite specimens was identified. / Doctor of Philosophy / As conventional water-cooled nuclear power systems showed safety issues, the Generation IV International Forum was established to expedite the development of next-generation nuclear reactor systems. Among the six advanced nuclear reactors, the molten salt reactor (MSR) stands out for its remarkable technical advantages, including low operating pressures and increased efficiency resulting from higher operating temperatures compared to water-cooled nuclear systems. Despite their advantages, further studies need to be conducted to develop and operate MSRs, as properties of molten salts have not been comprehensively measured in previous studies, and the corrosion of structural materials by molten salt is a significant challenge to their operation. The corrosion of alloys by molten salt can be attributed to many different factors, and the level of impurities in salt is an important factor directly linked to corrosion. Thus, the purification of salt is imperative to mitigate the corrosion of MSRs and needs to be well studied. In this study, methodologies for measuring thermophysical and thermodynamic properties of fluoride fuel salts were developed and validated using reference data. In addition, the corrosion of stainless steel 316H (SS316H) in a flowing fuel salt was also studied. Although various corrosion tests with static molten salts have been conducted, studies on corrosion of alloys in flowing molten salt fuels containing uranium fluorides are still limited. This study addresses this gap by developing a test apparatus equipped with a rotating disk to simulate the flow of molten salt on the surface of alloy specimens. Different batches of fuel salts with varying impurity levels, especially oxygen and hydrogen, were prepared using different purification methods. These salts were then used for corrosion tests under the same conditions, such as temperature and time duration, to explore the impacts of the non-metallic impurities on the corrosion of SS316H. The findings revealed that the salts with lower levels of oxygen and hydrogen caused less corrosion of SS316H, underscoring that the purification of salt is indispensable to the mitigation of corrosion in MSRs. This study also explored interactions of molten-salt fuels with graphite which is a promising candidate for a moderator or reflector of MSRs for enhancing neutron economy for thermal nuclear reactors. A high-pressure graphite-infiltration test apparatus was developed to investigate infiltration of fluoride fuel salts into graphite and the fluorination of graphite.

No description available.

Microelectrodes have been found to be a valuable tool in a variety of analytical studies. Their advantages over macro-sized electrodes are well known, including their enhanced mass transport properties (due to their ubiquitous hemispherical diffusion) which lead to steady state responses without external convection. They also exhibit high signal-to-noise ratios (greater sensitivities), furthering their analytical application. Microelectrode arrays are analytical devices with multiple electrodes. There are suitable for practical sensing with all the benefits of microelectrodes but with greater currents, leading to greater ease of measurement. To produce a reliable electroanalytical device the microelectrode response must be reproducible, a fundamental property based on the quality control of their production. Square microelectrode and array fabrication techniques have been developed for this purpose. This research discusses the fabrication and development of closely spaced arrays of square microelectrodes. Simulated and measured responses are compared and used to characterize electrode and array responses by cyclic voltammetry, electrical impedance spectroscopy and current-time transients. Measurements on variably spaced arrays allow insight into overlap of hemispherical diffusion from individual electrodes and the subsequent effect including peak current output on the array device. By studying these devices key insights into the mass transport properties of single square microelectrodes and microelectrode arrays were gained. This study also prepares and develops microelectrodes from materials appropriate for use in the extreme environments of molten salts and concentrated nitric acid solutions. These robust electrodes were developed for use in hydro- and pyro-chemical techniques for nuclear fuel reprocessing. These results demonstrate the practical uses for microelectrode systems across a wide range of chemical systems and in extreme conditions.

This study focuses on techniques that can be used to fuel next generation reactors. The first two studies are new techniques for recycling used nuclear fuel (UNF) and the third is a method of separating uranium (U) from lithium fluoride (LiF) and thorium fluoride (ThF4) salt also known as FLiTh for a thorium (Th) fuel cycle. The first technique proposed for UNF recycling was to use the cladding as an anode to oxidize the zircaloy and dissolve it into a LiF, sodium fluoride (NaF), zirconium fluoride (ZrF4) salt. Zirconium (Zr) was also reduced and deposited on a tungsten (W) cathode at the same time transporting the Zr through the salt. As commercial zircaloy would be contaminated with UNF oxides, and the oxides will not oxidize as part of the electrochemical process, they would be left at the anode as the Zr is dissolved away. This means the deposited Zr, on the cathode, can be disposed of as low-level waste (LLW) or recycled back into the nuclear industry instead of being stored as high-level waste (HLW). The next technique was fluorination of UNF oxides using ZrF4. Using the same LiF-NaF-ZrF4 salt, uranium oxide (UO2), lanthanum oxide (La2O3), and yttrium oxide (Y2O3) were fluorinated into uranium fluoride (UF4), lanthanum fluoride (LaF3), and yttrium fluoride (YF3). By sampling and recording the change in concentration over time, the reaction rate of all three oxides was determined and a temperature dependent reaction rate was reported from 500°C to 650°C. A zirconium oxide (ZrO2) product layer developed on UO2, but it only slowed down the fluorination process but did not stop it. UO2 and Y2O3 fluorinated entirely but La2O3 did not. The solubility limit of LaF3 in the salt was determined to be the reason the reaction did not go to completion. The last technique was the electrochemical separation of U from FLiTh, to simulate irradiated Th that decays to protactinium (Pa). A constant, albeit small current, was used to deposit U on a W electrode without Th depositing with it. A liquid metal bismuth (Bi) electrode was used as well, and a constant current resulted in Th depositing with the U. To get just U to deposit, the current needed to be applied for a time and then no current applied for a time so the system could reach equilibrium. By cycling these two steps it was possible to get U to deposit in Bi without Th. / Doctor of Philosophy / This study focused on techniques useful to the fabrication of next generation reactor fuels. The first focus was on new techniques for recycling used nuclear fuel (UNF). Nuclear waste currently needs to be stored for hundreds of thousands of years to reach background radiotoxicity levels. If plutonium (Pu) is removed from the waste this time is limited to ten thousand years and if the other transuranics (TRU) are removed the waste only needs to be stored for 300 years to reach background radiotoxicity levels. As recycling UNF can make such a drastic difference, developing techniques for this are of utmost importance. The first technique studied was to show that the zirconium (Zr) in zircaloy cladding could be oxidized and transported through salt. This was done by applying a current between a zircaloy anode and tungsten (W) cathode, dissolving the cladding into the salt. The salt used was lithium fluoride (LiF), sodium fluoride (NaF), and zirconium fluoride (ZrF4) salt called FLiNaZr. This transported Zr through the salt and then deposited it on W. If this process was done with zircaloy contaminated with used nuclear fuel (UNF) oxides, the oxides would not dissolve into the salt as part of the process and would be left behind at the anode as Zr is transported through the salt, effectively separating the two. This alone leads to a 25% reduction in the weight of the UNF that needs to be stored. The next technique studied was converting the UNF oxides into fluorides. This was done by having it react with ZrF4 to make zirconium oxide (ZrO2) and UNF fluorides. The oxides studied here were uranium oxide (UO2), yttrium oxide (Y2O3), and lanthanum oxide (La2O3). UO2 and Y2O3 reacted until no material was left but La2O3 did not. This was due to lanthanum fluoride (LaF3) having a solubility limit in the salt that made it impossible for more to be made and stopping the reacting. The reaction rate for each oxide was found and the order of the reaction rates was Y2O3>UO2>La2O3. This process was a success and should be studied more to ensure it will work with all oxides found in UNF. The last technique studied was electrochemically separating uranium (U) from lithium fluoride and thorium fluoride (ThF4) salt. Thorium (Th) is another nuclear material, and while it cannot fission in a reactor it can be turned into an isotope of U, U-233, that can. To do this Th must be irradiated so it turns into protactinium (Pa) which can then be separated from the salt. In this study U was a surrogate for Pa as it is too radioactive to handle in this lab. First, an inert W electrode was used to deposit U metal, and once it was successful a liquid metal bismuth (Bi) electrode was used. A small constant current was able to deposit U on W without co-deposition of Th. For a Bi electrode, an alternating time of applying current and then letting the system rest was needed to deposit U without co-deposition of Th.

Molten chloride salts have been proposed to be used as the primary coolant in molten salt reactors, and as the heat transfer fluid in concentrated solar power plants in next generation energy system design. The corrosive properties of molten chloride salts make it challenging to find appropriate structural materials for plant/system realization. In this work, two corrosion mitigation strategies are investigated to determine the relative corrosion performance of high temperature alloys in molten chloride salt mixtures: (1) chemical purification of salt mixture using a Mg sacrificial anode and (2) developing a protective oxide layer on the surface of high temperature alloys after pre-oxidation. These corrosion inhibitors are studied in combination with each other to determine the relative corrosion performance of three high temperature alloys: Incoloy 800H (chromia former), Haynes 214 (alumina former), and Noram SX (silica former). The unprotected and pre-oxidized alloys were exposed to molten chloride salt (62.5 wt % KCl + 37.5 wt % MgCl2·6H2O) with and without 1.7 wt % Mg as a corrosion inhibitor for 100 h at 700 °C under inert Ar atmosphere. SEM-EDS characterization was used to compare cross-sections and surfaces of each alloy exposed to molten salt with and without Mg additions. SEM-EDS cross-sectional characterization revealed significant Cr depletion in each unprotected alloy, and reduced Cr depletion in alloys immersed in molten chloride salt mixtures with chemical purification included. The addition of Mg metal to the salt mixture resulted in the precipitation of MgO on the alloy surfaces. The oxide deposition of MgO on components may impact the thermal and mechanical performance of the system. Therefore, the addition of Mg should be optimized for use in an operational system. Cross-sectional analysis identified the dissolution of Cr2O3 and SiO2 oxide scales and a stable Al2O3 oxide scale post-exposure. / Thesis / Master of Applied Science (MASc)

This study encompasses various aspects of corrosion in chloride molten salt environments, employing electrochemical techniques and a forced convection loop. It explores corrosion thermodynamic properties, electrochemical corrosion kinetics, and flow-induced dynamic corrosion. The study developed a novel electrochemical method for measuring thermodynamic properties of corrosion products and develops a new analysis theory for potentiodynamic polarization data obtained from cathodic diffusion-controlled reactions. Additionally, the design and operation experience of a forced convection chloride molten salt loop is shared. Particularly, the study presents novel findings on the turbulent flow-induced corrosion phenomenon and mechanism of Fe-based alloys in Mg-based chloride molten salt. These outcomes provide valuable insights into the corrosion mechanisms and flow-induced corrosion of Fe-based alloys in chloride molten salt. The results and experiences shared in this paper have implications for the successful implementation of molten salt as an advanced heat transfer fluid and thermal energy storage material in high-temperature applications, benefiting the nuclear and concentrating solar communities. / Doctor of Philosophy / This study explores the corrosion behavior of materials chloride molten salt, which is used in advanced energy systems. By using advanced techniques, the researchers investigated how these materials react and corrode in different conditions. They developed new methods to measure the properties of the corrosion products and analyzed how different factors affect the corrosion process. Additionally, they shared their experiences in building and operating a flow loop to simulate these conditions. The study discovered interesting phenomena, such as how the flow of molten salt can cause corrosion in certain types of metals. These findings provide important insights for improving the use of molten salt as a heat transfer fluid and energy storage material in advanced energy technologies.