UC Berkeley

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 Reactors (HTGRs). Graphite versatility in nuclear systems stems from its unique combination of mechanical, thermal, chemical, and neutronic properties. These properties are influenced by operational parameters like temperature, radiation, and chemical environment. In FHRs and MSRs, graphite can interact with the salt through multiple mechanisms, including salt-infiltration in graphite pores, chemical reactions with salt constituents, and tribo-chemical wear. The goal of this Ph.D. dissertation is to investigate mechanisms of interaction of fluoride salts with graphite in FHRs and assess their impact on salt reactor engineering. Chemical interactions between the salt and graphite are studied by exposing a graphite sample to 2LiF-BeF2 (FLiBe) salt and to the cover gas above the salt at 700°C for 240 hours. Chemical and microstructural characterization of the samples highlights formation of two types of C-F bonds upon exposure, with different degrees and mechanisms of fluorination upon salt and gas exposure. Infiltration of salt in graphite pores is examined by reviewing literature on infiltration and its effect and by studying salt wetting on graphite. Contact angles for salt on graphite are measured under variable conditions of graphite surface finish and salt chemistry, and used to predict salt infiltration. Wear and friction of graphite-graphite contacts at conditions relevant to pebble-bed FHR operation is studied through tribology experiments in argon and in FLiBe. Characterization via SEM/EDS, polarized light microscopy, and Raman spectroscopy is employed to seek a mechanistic understanding. Different mechanisms of lubrication are observed in the tests: in argon, graphite is observed to self-lubricate by forming a tribo-film that remains stable at high temperature in argon; in FLiBe, boundary lubrication is observed and postulated to be associated with C-F bond formation at graphite crystallite edges.

The interactions between graphite and tritium are studied. Tritium production rates in FHRs are quantified to be three orders of magnitude larger compared to light water reactors. A literature review is performed to investigate the thermodynamics and kinetics of the hydrogen-graphite interaction; the findings are employed to develop an improved model for hydrogen uptake and transport in graphite, which is used to extract tritium transport parameters from experimental studies.The experiments conducted in this dissertation indicate that the presence of the salt impacts graphite engineering performance in the reactor and after discharge in multiple ways, from providing increased lubrication to impacting graphite surface chemistry. As a further development, exploration of other areas where the salt could have an effect, including evolution of oxidation and graphite reactive sites upon neutron irradiation, in the presence of salt-exposure, is recommended.