Accident tolerant fuel (ATF) cladding is one of the most active area for research to advance the global contribution of nuclear energy since ATF will ensure and enhance the reactor safety allowing long refueling cycles and higher burnup. Silicon Carbide (SiC) is one material that can potentially be a solution to this long standing challenge. This study offers both computational reactor physics modeling and experimental investigation to examine the behavior of SiC in nuclear reactor environment. With recent interest in small modular reactors (SMR), this study selected SMR compact core as a reference assembly for Computational Burnup Modeling (CBM). Results of CBM suggest that SiC cladding will provide the highest burnup and maximum uranium utilization. Therefore, subsequent experimental investigations were focused on SiC only. Rhode Island Nuclear Science Center (RINSC)'s nuclear research reactor was used to investigate effects of irradiation on sintered tubular SiC material samples under pure steam at 1 atm, initially at 120 °C, followed by high temperatures between 850 °C and 1350 °C to simulate harsh reactor environment.

Experimental analyzes of (a) weight loss and (b) burst testing are considered for SiC samples. For the weight loss experiment, we have considered both types of samples: (1) Non-irradiated and (2) Neutron-irradiated samples. Weight loss was found to be dependent on sample geometry. Irradiated samples show ∼2%-10% higher weight loss per area than that of the non-irradiated samples. Besides, considering the medium flow rate (less than 10 g/min), it has been observed that the irradiated samples exhibit ∼10%-40% higher weight loss than that of the non-irradiated samples over temperatures of 120 °C, 850 °C to 1350 °C. It is also seen that material loss rates are generally more sensitive at higher temperatures than at lower temperatures, and irradiated samples are more prone to weight loss. Furthermore, for the burst testing experimental investigations, experiments were conducted for (1) As-received samples (2) Non-irradiated samples and (3) Irradiated samples. Considering the three types of samples, it was observed that the maximum and minimum peak load values are ∼70% higher and ∼60% lower than the average peak load while fracture hoop stress is consistently ∼70%-75% higher than the internal pressure for all as-received samples. Experimental and computational investigations suggest that SiC is a viable candidate for ATF cladding material.