Abstract
The motivation for this thesis is to develop a repeatable, and accurate testing method to evaluate the thermal properties of the Silicon Carbide (SiC) composite and tin system. The primary focus is on using SiC composite as a cladding material for nuclear fuel, which properties can vary depending its manufacturing process. Experimental testing is necessary to account for the voids in the SiC composite matrix and its heat transfer interactions with tin at elevated temperatures. The information gathered will advance fuel development and reactor design in the nuclear industry.As the U.S. and global energy demands continue to rise, nuclear energy remains a crucial source, providing a significant portion of electricity. To meet these demands, improvements in nuclear materials are essential. Historically, zircaloy has been the preferred cladding material due to its mechanical and thermal properties. However, it has several drawbacks, including embrittlement and high neutron absorption. In response, the Accident Tolerant Fuel (ATF) program, initiated in 2012, has been exploring new cladding materials like SiC, which offers excellent thermal, mechanical, and chemical properties.
This research specifically investigates SiC composites, reinforced with SiC fibers, for their improved fracture toughness and pseudo-ductile behavior. A novel experimental setup, designed to create a radial heat transfer scenario, involves concentric cylinders of molybdenum, tin, and SiC composite. Utilizing transient state methodology, the initial data collection focuses on understanding the heat transfer experienced through a SiC composite cladding specimen and tin system, with plans for future work to include steady-state systems.
To further understand the system's heat transfer, finite element models were created and validated against analytical solutions. Sensitivity studies identified the significant impact of SiC cladding properties on heat transfer.