Abstract
Graphite is a ubiquitous material in nuclear engineering. Within Generation IV designs, graphite serves as a reflector or fuel element material in Fluoride-Salt-Cooled-High-Temperature Reactors (FHRs), Molten Salt Reactors (MSRs), and High-Temperature-Gas-Reactor (HTGRs). In fusion research, graphite was originally proposed as a divertor material and may be employed in the breeding blanket. Graphite versatility in nuclear systems stems from its unique combination of mechanical, thermal, chemical, and neutronic properties. These properties vary across graphite grades and are influenced by operational parameters like temperature, radiation, and chemical environment. In reactors that employ fluoride salts, graphite can interact with the salt through multiple mechanisms, including salt-infiltration in graphite pores, chemical reactions with salt constituents, and tribo-chemical wear. These mechanisms can have an impact on graphite’s integrity and functional performance in the reactor, including its ability to immobilize tritium, its irradiation-resistance, and its sensitivity to degradation under air ingress. This seminar will describe mechanisms of interactions of fluoride salts with graphite and assess their impact on reactor engineering. This talk will also discuss active research projects on molten salts and graphite at the University of Illinois.
Bio
Dr. L Vergari is an Assistant Professor in Nuclear, Plasma, and Radiological Engineering and the director of the ABC Lab at the University of Illinois Urbana-Champaign. Dr. Vergari holds a Ph.D. in Nuclear Engineering from the University of California Berkeley, a M.S. in Energy and Nuclear Engineering from Politecnico di Torino, a M.S. in Nuclear Engineering from Politecnico di Milano and a B.S. in Energy Engineering from Politecnico di Milano.
Useful links: Google Scholar, ResearchGate, NPRE Department
Contact information: vergari@illinois.edu.
