ICONE19-43691 Thermalhydraulic Analysis of Uranium Carbide (UC) Fuel in 54 and 64-Element Fuel Bundles for SCWRs

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  • Qureshi Arif
    Faculty of Energy Systems and Nuclear Science University of Ontario Institute of Technology
  • Draper Shona
    Faculty of Energy Systems and Nuclear Science University of Ontario Institute of Technology
  • Abdalla Ayman
    Faculty of Energy Systems and Nuclear Science University of Ontario Institute of Technology
  • King Krysten
    Faculty of Energy Systems and Nuclear Science University of Ontario Institute of Technology
  • Peiman Wargha
    Faculty of Energy Systems and Nuclear Science University of Ontario Institute of Technology
  • Pioro Igor
    Faculty of Energy Systems and Nuclear Science University of Ontario Institute of Technology
  • Joel Jon
    Faculty of Engineering University of Waterloo
  • Gabriel Kamiel
    Faculty of Engineering and Applied Science University of Ontario Institute of Technology

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The objective of this paper is to investigate a possibility of using Uranium Carbide (UC) as a nuclear fuel in the newly developed 54 and 64-element fuel bundles with smaller outer diameter pins (8.5 mm and 9.1 mm, respectively), at supercritical conditions of a generic SuperCritical Watercooled Reactor (SCWR). Uranium Carbide has been used in numerous applications such as fuel in pebble bed reactors. The focus of this paper is to calculate the fuel centerline and cladding temperature profiles in order to ensure that the new fuel-bundle design complies with the industry accepted limits of 1850℃ for the fuel centerline temperature and the design limit of 850℃ for the cladding temperature. As a candidate fuel for SCWRs, Uranium Carbide has a high melting point, does not undergo a phase change at high temperatures, and possesses high dimensional stability under irradiation. Furthermore, Uranium Carbide has a significantly higher thermal conductivity and greater fissile element density than Uranium Dioxide (UO_2), making it a potential candidate for use in SCWRs. To examine the Uranium Carbide fuel within the operating conditions of SCWRs, the 54 and 64-element fuel bundles were utilized. Further, the fuel centerline and cladding temperature profiles were calculated under several axial heat flux profiles based on an average thermal power per channel of 8.5 MWth. To show that the 64-element bundle is the best available option, profiles of the Variant-20 and 54-element bundle were analyzed. Results indicated that the 64-element bundle attains major improvements in fuel centerline temperature and sheath temperature compared with previously designed fuel bundles as well as an improved safety margin over the Variant-20 and 54-element fuel bundle.

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