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Lattice Physics Calculations for Alternative Fuels for the Canadian SCWR

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Lattice Physics Calculations for Alternative Fuels for the Canadian SCWR Lattice Physics Calculations for Alternative Fuels for the Canadian SCWR LATTICE PHYSICS CALCULATIONS FOR ALTERNATIVE FUELS FOR THE CANADIAN SCWR By EDWARD MATTHEW GLANFIELD, B.Sc. A Thesis Submitted to the School of Graduate Studies in Partial Fulfilment of the Requirements for the Degree Master of Applied Science McMaster University © Copyright by E.M. Glanfield, July 2017 Master of Applied Science, 2017 Department of Engineering Physics McMaster University Hamilton, Ontario, Canada TITLE: Comparison of Fuel Types for the Pressure-Tube Super-Critical Water Cooled Reactor AUTHOR: Edward Matthew Glanfield, B.Sc. SUPERVISOR: Dr. David Novog NUMBER OF PAGES: ix, 122 Abstract The Canadian-designed Super-Critical Water-cooled Reactor (SCWR) has been in development since the early 2000s and has undergone many design changes since its initial conception. Initially the reactor was designed to use a fuel bundle modeled after the Canada Deuterium Uranium (CANDU) pressurized heavy water reactor bundles. Over time, this fuel bundle evolved into its current design of a re-entrant coolant channel with two fuel rings in the outer portion of the channel. Throughout the development of this design, the fuel composition has remained relatively constant, with only minor changes to enrichment. The initial SCWR design was intended to utilize a fuel mix of reactor grade plutonium and thorium. The thorium filler is an easily attainable material, being about three times more common than uranium. More importantly, the majority of thorium is made of 232Th, which breeds more fissile material than 238U in the SCWR reactor. This leads to production of 233U that can significantly extend the lifetime of fuel in the reactor, compared to fuels with either 238U or no fertile source. Even though the plutonium-thorium fuel cycle seems generally advantageous compared to traditional uranium fuels, or even other fuel types, it may not ultimately be preferable to other fuels in a specific reactor design such as the SCWR. The geometry, operation conditions, and cycle length can all have a large influence on the performance of a specific fuel in a reactor. Therefore it is important to compare the performance of different fuels when utilized in the reactor of interest. The SCALE suite of programs is used in this work to evaluate the performance of plutonium-thorium (PuTh), mixed oxide (MOX), and enriched uranium (UO2) fuels in the SCWR reactor. TRITON was used to deplete the fuel and obtain nuclide concentrations in 100 day intervals over the lifetime of the fuel. NEWT was used to perform lattice calculations for the determination of reactivity coefficients. In both cases, the ENDF/B-VII nuclear data library was used. In order to achieve similar burnups for all three fuel types, the enrichment of the UO2 and MOX fuels was adjusted. The standard PuTh fuel mix used in this work has an average enrichment of 10% plutonium, of which around 50% is 239Pu. An ii average enrichment of 7.6% 235U was used for the uranium fuel, while the MOX fuel retained the same enrichment as the PuTh. All three fuel types share the same ratio of enrichment between the inner and outer fuel rings. Besides adjusting the composition of the fuel, all other reactor parameters were kept identical. Varying the fuel type could change the operating temperatures and other thermalhydraulic properties of the reactor, but this was not considered for ease of computation as well as allowing for more direct comparison of the neutronic differences between fuels. As an alternative, the feedback coefficients for each fuel type were also investigated. The reactivity changes with burnup of both the PuTh and MOX fuels are extremely similar. The UO2 fuel, in contrast, decreases in reactivity at a faster rate. This results in the uranium fuel requiring a higher k-eff at zero burn-up to maintain the same cycle length. This also indicates that the PuTh and MOX fuels are being used more efficiently in the reactor. Additionally, a reactor with a more stable reactivity over time is generally easier to refuel. In general, the UO2 fuel has feedback coefficients of reactivity closer to zero than either the PuTh or MOX fuels. This is most pronounced in the Coolant Void Reactivity coefficients (CVR), where the uranium fuel has a significantly smaller coefficient than the others. The most notable exception to this is the Coolant Temperature Coefficient of reactivity (CTC), where the uranium fuel has a higher reactivity coefficient at large burnups. It is important to note that in nearly all cases the sign of the reactivity coefficients does not change with fuel selection, meaning that reactor behaviour could be expected to be quite similar between the three fuels. The originally proposed PuTh fuel seems adequate for use in the SCWR compared to the two alternative fuels. It shows no significant advantages or disadvantages in regards to the reactivity coefficients analyzed in this work. The MOX fuel performs very similarly to the PuTh option, with many reactivity coefficients being nearly the same. Its main benefit would be the availability of existing operational experience in comparison to PuTh fuels. An enriched uranium fuel benefits the most from historic operational experience, and in general has the most optimal reactivity coefficients. However, it burns less efficiently, requiring the fuel to be enriched to around 8.5%, well above the level currently used in power reactors. iii From this work there does not seem to be a single, best fuel choice. The multitude of factors considered are a small portion of those that influence the choice of fuel in an operating reactor. Furthermore, some studies have demonstrated the need for advanced fuel materials that would lower the centreline melting temperature and hence some form of nitride or accident tolerant fuel may be needed in the future. Additionally, the fuel that stands out based on a single criterion does not necessarily perform well in other criteria, and further analysis would be beneficial. iv Acknowledgements Without the guidance of David Novog, this thesis would not have been possible. He is responsible for giving me the opportunity to pursue a Master's degree, providing assistance and patience throughout the process, and ensuring that I have completed it successfully. His experience and counsel were invaluable. Although their knowledge and advice was frequently questionable, I have to thank the familiar faces that I would see each week at WEB. They welcomed and introduced me to Canada, and provided a variety of wonderful extracurricular distractions during my time at McMaster. I also thank the people I met through PNB, Makeshift, and FIRST Canada for their friendship and support. I would also like to thank Haley Kragness for bringing me to Canada, providing advice and support whenever needed, and for being there from the beginning to the end. And most of all, I appreciate the constant support of my family over the two decades of my education. v Table of Contents Abstract ............................................................................................................................... ii Acknowledgements............................................................................................................. v Table of Contents ............................................................................................................... vi List of Figures ................................................................................................................... viii List of Tables ...................................................................................................................... ix Chapter 1: Introduction ................................................................................................... 1 1.1 Development of Canadian Nuclear Industry ....................................................... 1 1.2 Development of the SCWR ................................................................................. 2 1.3 Current Design .................................................................................................... 7 1.4 SCWR Fuel ......................................................................................................... 12 Chapter 2: Theory .......................................................................................................... 14 2.1 Nuclides of Interest ........................................................................................... 14 2.2 Fuels .................................................................................................................. 15 2.3 Multiplication Factor ......................................................................................... 16 2.4 Coefficients of Reactivity .................................................................................. 17 2.5 Path of a Neutron.............................................................................................. 20 2.6 Dancoff Factors ................................................................................................. 21 Chapter 3: Literature Review ......................................................................................... 22 3.1 Work on Early Designs ...................................................................................... 22 3.2 Work on the 64-Element Design ....................................................................... 23 3.3 Variations on the 64-Element Design ..............................................................
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