Development of Silicon Carbide Fuel Rod Cladding Having Improved Impact Resistance

The zirconium-alloy fuel rod cladding typically found in pressurized water reactor (PWR) fuel assemblies can overheat and burst in a loss of coolant accident, exposing the nuclear fuel, and reacting with steam in the core to form an explosive hydrogen mixture that can challenge containment integrity and threaten public health and safety. Accident-tolerant fuel (ATF) aims to prevent this by replacing the zirconium-alloy fuel rod cladding with alternate, less reactive materials. Silicon carbide (SiC) is one of the leading materials being looked to with advancements that include employing multiple layers of solid and fibrous SiC that are combined to provide circumferential and axial tensile strength, prevention of crack propagation, and hermeticity. However, these processes do not fully overcome the inherent brittleness of SiC, and despite its demonstration of corrosion resistance and steady-state behavior under irradiation at PWR operating conditions, its ability to withstand mechanical shocks remains substantially less than the metallic cladding it would replace. How the Problem is Being Addressed: The purpose of this project is to develop metal-reinforced SiC cladding that will withstand mechanical shock without failure while minimizing the amount of metal added to the ATF. Various ductile metals will be embedded in the SiC composite layers during fabrication of fuel rod test segments by either overwrapping the existing SiC fibers with metal fibers or by coating them with a thin metal layer. Phase I will focus on subjecting these test segments to characterization and a series of thermal and mechanical tests, culminating in metallurgical examination of the segments and reporting of their performance results. In Phase II, the bench-scale fabrication and characterization work will be extended to full scale test hardware culminating in fuel assembly impact tests performed at reactor operating temperatures and a full-lateral-scale, partial-length, long-duration flow test. Reactor test specimens will also be fabricated for follow-on irradiation testing.