Investigation of nitride fuels and their properties has, for decades, been propelled on the basis of their desirable high metal densities and high thermal conductivities, both of which oer intrinsic advantages to performance, economy, and safety in fast and light water reactor systems. In this time several key obstacles have been identied as impeding the implementation of these fuels for commercial applications; namely chemical interactions with air and steam, the noted diculty in sintering of the material, and the high costs associated with the enrichment of 15N. The combination of these limitations, historically, led to the well founded conclusion that the most appropriate use of nitride fuels was in the fast reactor fuel cycle, where the cost burdens associated with them is substantially less. Indeed, it is within this context that the vast majority of work on nitrides has been and continues to be done. Nevertheless, following the 2011 Fukushima-Daiichi nuclear accident, a concerted governmental-industrial eort was embarked upon to explore the alternatives of so-called \accident tolerant" and \high performance" fuels. These fuels would, at the same time, improve the response of the fuel-clad system to severe accidents and improve the economy of operation for light water reactor systems. Among the various candidates proposed are uranium nitride, uranium silicide, and a third \uranium nitride-silicide" composite featuring a mixture of the former. In this thesis a method has been established for the synthesis, fabrication, and characterization of high purity uranium nitride, and uranium nitride-silicide composites, prepared by the spark plasma sintering (SPS) technique. A specic result has been to isolate the impact of the processing parameters on the microstructure of representative fuel pellets, essentially permitting any conceivable microstructure of interest to be fabricated. This has enabled the development of a highly reproducible technique for the production of pellets with microstructures tailored towards any desired porosity between 88-99.9%TD, any grain size between 6-24 μm, and, in the case of  the uranium nitride-silicide composite, a silicide-coated UN matrix. This has permitted the evaluation of these microstructural characteristics on the performance of these materials, specically with respect to their role as accident tolerant fuels. This has generated results which have tightly coupled nitride performance with pellet microstructure, with important implications for the use of nitrides in water-cooled reactors. / Under artionden har forskning om nitridbranseln och dess egenskaper bedrivits pa grundval av nitridbransletsatravarda egenskaper avseende dess hoga metall tathet och hog varmeledningsformaga. Dessa egenskaper besitter vasentliga fordelar avseende prestanda, ekonomi och sakerhet for metallkylda som lattvatten reaktorer. Genom forskning har aven centrala begr ansningar identierats for implementering av nitridbranslen for kommersiellt bruk. Begransningar avser den kemiska interaktionen med luft och vattenanga, en uppmarksammad svarighet att sintring av materialet samt hoga kostnader forknippade med den nodvandiga anrikningen av 15-N. Kombinationen av dessa begransningar resulterade, tidigare, i en valgrundad slutsats att nitridbranslet mest andamalsenliga anvandningsomrade var i karnbranslecykeln for snabba reaktorer. Detta da kostnaderna forenade med implementeringen av branslet ar avsevart lagre. Inom detta sammanhang har majoriteten av forskning avseende nitrider bedrivits och fortskrider an idag. Dock, efter karnkraftsolyckan i Fukushima-Daiichi 2011, inleddes en samlad industriell och statlig anstrangning for att undersoka alternativ till sa kallade \olyckstoleranta" och \hogpresterande" branslen. Dessa branslen skulle samtidigt forbattra reaktionstiden for bransleinkapsling systemet mot allvarliga olyckor samt forbattra driftsekonomin av lattvattenreaktorer. Foreslagna kandidater ar urannitrid, uransilicid och en tredje \uran nitrid-silicid", komposit bestaende av en blandning av de foregaende. Genom denna avhandling har en metod faststallts for syntes, tillverkning och karaktarisering av uran nitrid av hog renhet samt uran nitrid-silicid kompositer, forberedda med tekniken SPS (Spark Plasma Sintering). Ett specikt resultat har varit att isolera eekten av processparametrar pa mikrostrukturen pa representativa branslekutsar. Detta mojliggor, i princip, framstallningen av alla tankbara mikrostrukturer utav intresse for tillverkning. Vidare har detta mojliggjort utvecklingen av en hogeligen reproducerbar  teknik for framstallningen av branslekutsar med mikrostrukturer skraddarsydda for onskad porositet mellan 88 och 99.9 % TD, och kornstorlek mellan 6 och 24 μm. Dartill har en metod for att belagga en matris av uran nitrid-silicid framarbetats. Detta har mojliggjort utvarderingen av dessa mikrostrukturella parametrars paverkan pa materialens prestanda, sarskilt avseende dess roll som olyckstoleranta branslen. Detta har genererat resultat som ar tatt sammanlankat nitridbranslets prestanda till kutsens mikrostruktur, med viktiga konsekvenser for den potentiella anvandningen av nitrider i lattvatten reaktorer. / <p>QC 20170210</p>

McMaster University MASTER OF APPLIED SCIENCES (2017) Hamilton, Ontario (Engineering Physics) TITLE: Comparative study of accident-tolerant fuel for a CANDU lattice AUTHOR: Simon Younan, B.Eng. (McMaster University) SUPERVISOR: Dr. David Novog NUMBER OF PAGES: xiii, 120 / Since the Fukushima accident in 2011, there have been an increasing number of studies on the use of accident-tolerant fuel (ATF) in light water reactors to mitigate the consequences of a future severe accident, by better retaining fission products and/or providing operators more time to implement emergency measures. However, few studies exist for CANDU reactors in this regard. The goal of this study is to determine how different types of ATF are expected to behave in a CANDU lattice when compared to the current UO2 fuels. In particular, this study focuses on neutronic parameters calculated using the Serpent 2 code, but also models heat transfer and stylized accident scenarios. The ATF concepts tested include UO2-SiC composites, UN and UN-based composites, U-9Mo, and fully ceramic microencapsulated (FCM) fuel, along with SiC and SS-coated cladding. Four general conclusions can be drawn: 1. Fuel temperature are lower for ATF as compared to traditional fuels. UO2-SiC composite fuel exhibits a moderate temperature reduction compared to UO2, particularly for fresh fuel. Other ATF fuel materials exhibit a substantial decrease in fuel temperature compared to UO2. The lower fuel temperatures are also accompanied by lower melting temperatures for some fuels, hence each design requires specific assessments on safety. 2. As most ATF have a poorer neutron economy compared to standard fuel designs, enrichment is required to use ATF in a CANDU, particularly for UN and FCM fuel compositions. Coolant void reactivity (CVR) is lowest with FCM fuel and highest with U-9Mo fuel. Fuel temperature coefficient (FTC) is most negative for fuel containing UN or U-9Mo. 3. Changing the cladding material from zircaloy to SiC slightly improves neutron economy, while a FeCrAl surface layer impairs neutron economy. The impact of many ATF sheath materials is to greatly reduce or eliminate hydrogen production in some severe accidents. A specific assessment on hydrogen production was not performed in this study. 4. In stylized accident scenarios, all fuels exhibit only a small temperature spike due to the reactivity insertion of the LOCA as the reactor shutdown limits the power excursion. For cases where Emergency Core Cooling functions as designed, fuel and channel failures are precluded for both traditional fuels and ATF. For cases with impairment of ECC, most ATF fuels show lower fuel temperatures than UO2 fuels and adequate heat removal to the pressure-calandria tube fuel channel. The exception would be Mo-based fuels that reach the melting point prior to establishing an adequately high sheath temperature to sustain radiative heat removal to the PT-CT assembly. / Thesis / Master of Applied Science (MASc) / Since the Fukushima accident in 2011, there have been an increasing number of studies on the use of accident-tolerant fuel in nuclear reactors to mitigate the consequences of a future severe accident, reducing the likelihood and severity of a radiation release. Canadian reactors are of the CANDU design, which differs greatly from the reactors most recent studies have focused on. The goal of this study is to determine the feasibility of using accident-tolerant fuel in CANDU reactors, studying different types. In general, the goal of accident-tolerant fuels in CANDU reactors would be to reduce fuel temperatures and improve fission product retention, reducing the likelihood/magnitude of radioactive releases in a severe accident. However, nearly all types of accident-tolerant fuel would also require the uranium to be slightly enriched as opposed to the current fuel which is based on naturally-occurring uranium. This study outlines the results obtained by computer modelling of accident-tolerant fuel in a CANDU reactor, including the enrichment requirements, changes to important reactivity feedbacks, and impacts on accident performance.

The swelling mechanisms of U3Si2 under neutron irradiation in reactor conditions are not unequivocally known. The limited experimental evidence that is available suggests that the main driver of the swelling in this material would be fission gases accumulation at crystalline grain boundaries. The steps that lead to the accumulation of fission gases at these locations are multiple and complex. However, gradually, the gaseous fission products migrate by diffusion. Upon reaching a grain boundary, which acts as a trap, the gaseous fission products start to accumulate, thus leading to formation of bubbles and hence to swelling. Therefore, a quantitative model of swelling requires the incorporation of phenomena that increase the presence of grain boundaries and decrease grain sizes, thus creating sites for bubble formation and growth. It is assumed that grain boundary formation results from the conversion of stored energy from accumulated dislocations into energy for the formation of new grain boundaries.This thesis attempts to develop a quantitative model for grain subdivision in U3Si2 based on the above mentioned phenomena to verify the presence of this mechanism and to use in conjunction with swelling codes to evaluate the total swelling of the pellet in the reactor during its lifetime.

Fuel for nuclear reactors with an increased thermal conductivity offers the potential for lower fuel operating temperatures and reduced fission gas release rates. Uranium dioxide (UO2) based composites offer a method of achieving a higher thermal conductivity. Silicon carbide (SiC) and molybdenum (Mo) have been identified as potential candidates for use in a composite fuel material. Uranium dioxide composites were manufactured with the inclusion of whiskers and granules of SiC up to a 30 vol% loading. The manufacturing route used was based on the current process employed to commercially manufacture UO2 fuel, by reductive sintering. Composites containing Mo were manufactured via spark plasma sintering and included loadings of up to 10 vol% Mo. The composites were characterised on their microstructural properties and where appropriate the thermal conductivity was determined by laser flash analysis. The composites containing SiC achieved low densities, 95%TD. The microstructure contained channel like structures of Mo, due to the use of an agglomerated UO2 precursor powder. An increased thermal conductivity was determined for the molybdenum composites. At the maximum measurement temperature of 800°C the increase was found to be 68% in the 10 vol% composites compared to UO2.

In a worldwide pursuit for more Accident Tolerant nuclear Fuel (ATF), the quest to obtain and certify alternative nuclear fuel cladding tubes for light-water nuclear power reactors is still a key challenge. One of the facets in this program to develop more ATF is the heat transfer evaluation between the various proposed clad tubes manufactured from suitable replacement materials and the current problematic zirconium-alloy based clad tubes used in nuclear power reactors. For the heat transfer analysis, the accurate measurement of the temperature on the heat transfer surface of heated tubes to be tested was one of the important objectives for the effective analysis of the heat transfer characteristics to the water coolant. After extensive investigations, a suitable technique was developed and validated against recognised forced-convection heat transfer correlations. The results showed that this technique was well suited for external forced convection heat transfer studies from heated surfaces exposed to forced convection water coolants. / Dissertation (MSc)--University of Pretoria, 2019. / Mechanical and Aeronautical Engineering / MSc (Applied Science - Mechanics) / Unrestricted

The search for Accident tolerant fuels (ATF) which is the first part of this thesis is currently one of the most actual topics in the field of nuclear fuels. These fuels must be first successfully tested in operational and also accident conditions for their possible inclusion in commercial use. Following part of the thesis specifically focuses on the boiling crisis in nuclear reactors which can damage the nuclear fuel cladding. Therefore, it is necessary to know the critical heat flux value and the departure from nuclear boiling ratio. Calculations which determine critical heal flux value are placed in the practical part of the thesis. Calculations are compared with the data obtained during experiments. The ALTHAMC12 and the other correlations which are based on the previous measurements are used for the computational analysis.