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No description available.

No / Production of titanium is challenging and expensive due to the energy energy-intensive and time-consuming processes used at present. Current commercial production method reduces titanium tetrachloride with magnesium or sodium to produce titanium metal. Several researchers have attempted electro-deposition of titanium from ionic solutions but have faced difficulties in eliminating multivalent titanium ions and highly reactive dendrite products. In this paper, we report, for the first time, the solid-solid phase transformation of titanium dioxide with calcium metal, under suitable conditions, to form solid titanium metal powder (>98% pure) without any oxygen impurity. On phase characterisation, it was found that homogeneous alpha-titanium was produced. The paper also includes the results and interpretations obtained using quantitative analysis, X-ray diffraction (XRD), scanning electron microscopy with energy dispersive X-ray spectroscopy (SEM-EDX) and phase diagram. The process is simple, green, rapid and cheap compared to the existing methods.

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)

No description available.

A molten carbonate fuel cell (MCFC) is an electro-chemical energy conversion technology that runs on natural gas and employs a molten salt electrolyte. In order to keep the electrolyte in this state, the cell must be kept at a temperature above 500 C, eliminating the need for precious metals as the catalyst. There has been only a limited amount of research on modelling the transport processes inside this device, mainly due to its limited ability for mobile applications. In this thesis, three one-dimensional models of a MCFC are presented based on different types of diffusion and convection. Comparisons between models are performed so as to assess their validity. Regarding ion transport, it is shown that there exists a limiting case for ion migration across the cathode that depends on the conductivity for the liquid potential. Finally, an optimization of the diffusivity across the cathode is carried out in an attempt to increase the cell performance and its longevity. / UOIT

No description available.

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.