MICROSTRUCTURAL CHARACTERIZATION AND MECHANICAL PROPERTY ASSESSMENT OF A NEUTRON IRRADIATED URANIUM-ZIRCONIUM NUCLEAR FUEL AND HT9 CLADDING
thesisposted on 30.07.2020, 15:00 authored by Jonova ThomasJonova Thomas
Metallic uranium-10 weight percent zirconium (U-10wt.%Zr) nuclear fuels are classified as potential fuels for fast breeder reactors as they possess a high fissile density and have increased compatibility with sodium, a frequently used reactor coolant. Despite their advantages when exposed to neutron irradiation in reactors, the fuels are subject to damage cascades and microstructural alterations. Fuel constituent re-distribution, phase transformation, fuel swelling, and fuel cladding chemical interactions (FCCI) are a few of the major interdependent microstructural alterations that occur in these fuels at the onset of neutron irradiation. The primary objective of this research is to understand the above-mentioned microstructural alterations in different regions of a neutron irradiated U-10wt.%Zr fuel and HT9 cladding that has achieved a cross-sectional burnup of 5.7 atomic percent (at%.). Additionally, this study also aims to provide a relationship between the microstructural alterations and local mechanical property changes exhibited at different regions of the HT9 cladding as a consequence of neutron irradiation, FCCI, and fission product migration.
To achieve this goal, a coordinated group of experiments was performed on the neutron irradiated U-10wt.%Zr/HT9 (fuel/cladding) at the nanoscale, microscale, and mesoscale, respectively. The experimental techniques used for microstructural analysis included the following: (1) transmission electron microscopy of focused ion beam (FIB) lamellas for nanoscale assessments, (2) serial sectioning of FIB cuboids for microscale assessments, and (3) synchrotron micro-computed tomography of FIB obelisks for mesoscale assessment. Following the microstructural assessments, nano-indentation experiments were performed on the neutron irradiated HT9 cladding to determine the changes in mechanical properties as a function of distance from cladding edge to FCCI locality, and the changes in mechanical properties as a consequence of several microstructural alterations. Furthermore, the results produced from the various experiments in this study were compared and correlated to existing literature (both in-reactor and out-of-reactor experiments), and new theories to explain the reason for the observed changes were established. This research also revealed several novel observations such as probable radiation induced segregation in fuels, localized fuel swelling and porosity distribution at different regions in the fuel, crystal structure of phases present at different regions in the fuel and their influence on pore morphologies, and nano mechanical properties of a neutron irradiated HT9 cladding.