Abstract
The thermophysical performance and solid mechanics behavior of UO2-36.4vol % BeO fuel pellets cladded with Zircaloy, SiC, and FeCrAl, and Zircaloy cladding materials coated with SiC and FeCrAl, are investigated based on simulation results obtained by the CAMPUS code. In addition, the effect of coating thickness (0.5, 1 and 1.5 mm) on fuel performance and mechanical interaction is discussed. The modeling results show that Zircaloy claddings are more effective in decreasing fuel centerline temperature and fission gas release than other kinds of cladding material because of the smaller gap between cladding and fuel at the same burnup. SiC claddings and SiC-coated Zircaloy claddings possess smaller plenum pressure than other kinds of cladding. SiC claddings contribute more to fuel radial displacement but less to fuel axial displacement. FeCrAl claddings exhibit very different radial and axial displacements in different axial positions. FeCrAl-coated Zircaloy claddings have a lower fuel centerline temperature than Zircaloy claddings at burnup below 850 MWh/kg U, but a higher fuel centerline temperature at higher burnup. The gap between FeCrAl-coated Zircaloy claddings and fuel pellets closes earlier than that of Zircaloy claddings. SiC-coated claddings increase fuel radial and axial displacements, and cladding axial displacements of inner and outer cladding surfaces.
Highlights
Accident tolerant fuel (ATF) systems have attracted significant attention regarding the safety margins of commercial light water reactors (LWRs) since the Fukushima Daiichi nuclear accident in2011
Based on previous research by Fink [8] and Lucuta et al [9], the influences of dissolved fission products, precipitated fission products, porosity, deviation from stoichiometry, and radiation damage all must be considered in relation to the thermal conductivity of UO2 -BeO fuel during irradiation
The fuel performance and solid mechanics of UO2 -36.4vol % BeO fuel with Zircaloy, SiC, FeCrAl, and surface-coated Zircaloy cladding materials are investigated based on simulation results obtained using the CAMPUS code
Summary
Accident tolerant fuel (ATF) systems have attracted significant attention regarding the safety margins of commercial light water reactors (LWRs) since the Fukushima Daiichi nuclear accident in2011. The low thermal conductivity of UO2 fuel has become one of the major concerns limiting nuclear reactor performance and safety. Incorporation of UO2 with high thermal conductivity particles such as BeO or SiC could improve fuel thermal conductivity, one of the key criteria in balancing thermal energy and reactor safety demands [1,2]. Another way to mitigate against severe accidents is to develop enhanced strength and ductility ATF cladding, as this would alleviate the severity of reactivity-initiated accidents, and oxidation-resistant
Sign-in/Register to view highlights & summary Sign-in/Register to access full text options 
Talk to us
Join us for a 30 min session where you can share your feedback and ask us any queries you have
Schedule a callDisclaimer: All third-party content on this website/platform is and will remain the property of their respective owners and is provided on "as is" basis without any warranties, express or implied. Use of third-party content does not indicate any affiliation, sponsorship with or endorsement by them. Any references to third-party content is to identify the corresponding services and shall be considered fair use under The CopyrightLaw.
