A process for recycling spent nuclear fuel cladding, a zirconium alloy (Zircaloy), into a metal powder that may be used for advanced nuclear fuel applications, was investigated to determine if it is a viable strategy. The process begins with hydriding the Zircaloy cladding hulls after the spent nuclear fuel has been dissolved from the cladding. The addition of hydrogen atoms to the zirconium matrix stresses the lattice and forms brittle zirconium hydride, which is easily pulverized into a powder. The dehydriding process removes hydrogen by heating the powder in a vacuum, resulting in a zirconium metal powder. The two main objectives of this research are to investigate the dehydriding process and to design, build and demonstrate a specialized piece of equipment to process the zirconium from cladding hulls to metal powder without intermediate handling. The hydriding process (known from literature) took place in a 95 percent argon - 5 percent hydrogen atmosphere at 500 degrees C while the dehydriding process conditions were researched with a Thermogavimetric Analyzer (TGA). Data from the TGA showed the dehydriding process requires vacuum conditions (~0.001 bar) and 800 degrees C environment to decompose the zirconium hydride. Zirconium metal powder was created in two separate experiments with different milling times, 45 minutes (coarse powder) and 12 hours (fine powder). Both powders were analyzed by three separate analytical methods, X-Ray Diffraction (XRD), size characterization and digital micrographs. XRD analysis proved that the process produced a zirconium metal. Additionally, visual observations of the samples silvery color confirmed the presence of zirconium metal. The presence on zirconium metal in the two samples confirmed the operation of the hydriding / milling / hydriding machine. Further refining of the hydride / milling / dehydride machine could make this process commercially favorable when compared to the high cost of storing nuclear waste and its components. An additional important point is that this process can easily be used on other metals that are subject to hydrogen embrittlement, knowing the relevant temperatures and pressures associated with the hydriding / dehydriding of that particular metal.

The sintering behavior of uranium and uranium-zirconium alloys in the alpha phase were characterized in this research. Metal uranium powder was produced from pieces of depleted uranium metal acquired from the Y-12 plant via hydriding/dehydriding process. The size distribution and morphology of the uranium powder produced by this method were determined by digital optical microscopy. Once the characteristics of the source uranium powder were known, uranium and uranium-zirconium pellets were pressed using a dual-action punch and die. The majority of these pellets were sintered isothermally, first in the alpha phase near 650°C, then in the gamma phase near 800°C. In addition, a few pellets were sintered using more exotic temperature profiles. Pellet shrinkage was continuously measured in situ during sintering. The isothermal shrinkage rates and sintering temperatures for each pellet were fit to a simple model for the initial phase of sintering of spherical powders. The material specific constants required by this model, including the activation energy of the process, were determined for both uranium and uranium-zirconium. Following sintering, pellets were sectioned, mounted, and polished for imaging by electron microscopy. Based on these results, the porosity and microstructure of the sintered pellets were analyzed. The porosity of the uranium-zirconium pellets was consistently lower than that of the pure uranium pellets. In addition, some formation of an alloyed phase of uranium and zirconium was observed. The research presented within this thesis is a continuation of a previous project; however, this research has produced many new results not previously seen. In addition, a number of issues left unresolved by the previous project have been addressed and solved. Most notably, the low original output of the hydride/dehydride powder production system has been increased by an order of magnitude, the actual characteristics of the powder have been measured and determined, shrinkage data was successfully converted into a sintering model, an alloyed phase of uranium and zirconium was produced, and pellet cracking due to delamination has been eliminated.

Nuclear fuel cycles (NFC) are the collection of interconnected processes which generate electricity through nuclear power. Due to the high degree of coupling between components even in the simplest cycles, the need for a dynamic fuel cycle simulator and analysis framework arises. The work presented herein develops essential physics models of nuclear power reactors and incorporate them into a NFC simulation framework. First, a one-energy group reactor model is demonstrated. This essential physics model is then to simulate a sampling fuel cycles which are perturbations of well known base-case cycles. Because the NFC may now be simulated quickly, stochastically modeling many fuel cycle realizations dramatically expands the parameter space which may be analyzed. Finally, a multigroup reactor model which incorporates spectral changes as a function of burnup is presented to increase the fidelity of the original one-group reactor. These methods form a suite of modeling technologies which reach from the lowest levels (individual components) to the highest (inter-cycle comparisons). Prior to the development of this model suite, such broad-ranging analysis had been unrealistic to perform. The work here thus presents a new, multi-scale approach to fuel cycle system design. / text

Thermal conductivity is a crucial parameter for nuclear fuel, as it sets an upper limit on reactor operating temperature to have safety margins. Uranium mononitride (UN) is a prospective fuel for fast reactors, for which limited experimental studies have been conducted, compared to the currently dominating light-water reactor fuel, uranium dioxide. The aim of this thesis is to determine the thermal conductivity in UN and to determine its porosity dependence. This was done by manufacturing dense and porous high-purity samples of UN and examining them with laser flash analysis, which with data on specific heat and thermal expansion gives the thermal conductivity. To analyse the result, a theoretical study of the phenomenology of thermal conductivity as well as a review and comparison with previous investigations were carried out. The porosity range was 0.1–31% of theoretical density. Thermal diffusivity data from laser flash analysis, thermal expansion data and specific heat data was collected for 25–1400 C. The laser flash data had high discrepancy at higher temperatures due to thermal instability in the device and deviations due to graphite deposition on the samples, but the low temperature data should be reliable. As the specific heat data was also of poor quality, literature data was used instead. As for the thermal diffusivity data, the calculated thermal conductivity for lower temperatures are more accurate. A modified version of the porosity model by Ondracek and Schulz was used to analyse the porosity dependence of the thermal conductivity, taking into account the different impacts of open and closed porosity. / Värmeledningsförmåga är en avgörande egenskap för kärnbränslen, eftersom det begränsar den maximala drifttemperaturen i reaktorn för att ha säkerhetsmarginaler. Uranmononitrid (UN) är ett framtida bränsle för snabba reaktorer. Jämfört med det dominerande bränslet i lättvattenreaktorer, urandioxid, har endast begränsade experimentella studier gjorts av UN. Målet med detta arbete är att bestämma värmeledningsförmågan i UN och bestämma dess porositetsberoende. Detta gjordes genom att tillverka kompakta och porösa prover av UN och undersöka dem med laserblixtmetoden, vilket tillsammans med värmekapacitet och värmeutvidgning ger värmeledningsförmågan. För att analysera resultatet gjordes en teoretisk studie av värmeledning såväl som en genomgång av och jämförelse med tidigare undersökningar. Provernas porositet sträckte sig från 0.1% till 31% av teoretisk densitet. Värmediffusivitetsdata från laserblixtmetoden, värmeutvidgningsdata och värmekapacitetsdata samlades in för 25–1400 C. Värdena från laserblixtmätningen hade hög diskrepans vid höga temperaturer p.g.a. termisk instabilitet i anordningen och avvikelser p.g.a. grafitavlagring på proverna, men data för låga temperaturer borde vara tillförlitliga. Eftersom resultaten från värmekapacitetsmätningen var av dålig kvalité, användes litteraturdata istället. Som en konsekvens av bristerna i mätningen av värmediffusivitet är presenterade data för värmeledningsförmåga mest exakta för låga temperaturer. En modifierad version av Ondracek-Schulz porositetsmodell användes för att analysera värmeledningsförmågans porositetsberoende genom att ta hänsyn till olika inverkan av öppen och sluten porositet.

No description available.

No description available.

No description available.

Thesis (M.S.)--University of Nevada, Reno, 2007. / "August, 2007." Includes bibliographical references. Online version available on the World Wide Web.

Thesis (Ph. D.)--University of Missouri-Columbia, 2007. / The entire dissertation/thesis text is included in the research.pdf file; the official abstract appears in the short.pdf file (which also appears in the research.pdf); a non-technical general description, or public abstract, appears in the public.pdf file. Title from title screen of research.pdf file (viewed on November 26, 2007 Vita. Includes bibliographical references.

Thesis (M.S.)--University of Nevada, Reno, 2007. / "December, 2007." Includes bibliographical references (leaves 25-28). Online version available on the World Wide Web.