2, Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
The phenomena describing the formation of brittle hydrides in nuclear fuel cladding induced by corrosion during reactor operation have been well-established in the literature for years. Further, the degradation of the mechanical properties has been a notorious consequence of these hydrides. As the first barrier against fuel release, it is paramount that nuclear fuel cladding maintains its structural integrity during reactor use and post-use storage. For this reason, developing a method to obtain material properties to evaluate the likelihood of failure has significant merits, especially a method that can test tubes in their customary state without significant modification.
An effective method to determine mechanical properties such as Young’s modulus, yield stress, strain-hardening, and fracture energy of as-received unirradiated and hydrided unirradiated Zircaloy-4 tubes in the hoop direction at room temperature was developed for ring compression tests through experimental and computational work. The hydrogen content was controlled up to 500wppm where failure is isolated at the 3 o’clock position. Since the crack location is isolated, cohesive elements were implemented to model fracture behavior. Length dependence of ring samples was determined experimentally where geometric features were tracked using projection digital image correlation (DIC) along with conventional load-displacement data. Computational work was performed in Abaqus using 2D and 3D models to complement this experimental work. Further, sensitivity analysis was performed on the methodology in this work to determine how strongly the input parameters affect simulation results. As expected, increasing hydrogen content caused embrittlement of the Zircaloy-4 samples tested. The integrated approach will be described and demonstrated to identify the most influential material properties on the damage tolerance of hydrided Zircaloy-4 tubes.