This work investigated material corrosion by nuclear reactor coolants, including pressurized water reactor (PWR) coolant, boiling water reactor (BWR) coolant, high-temperature steam, lead-bismuth eutectic (LBE), and molten salt. Novel cladding materials for accident tolerant fuel (ATF) and additive manufacture (AM) Ni-based alloy were studied in water coolants. Similarly, the ATF material and Ni-based alloys were also examined under high-temperature steam to understand the corrosion behavior in beyond design basis accident (BDBA) scenarios. In addition to isothermal corrosion, stress corrosion cracking (SCC) and oxide layer in situ measurements were also conducted. Unlike conventional studies in liquid LBE that focused on Fe-based alloys, the present studies also investigated Ni-based alloys to explore the Ni content effects on the corrosion by LBE at high temperatures under saturated oxygen conditions. In molten salt environments, the corrosion behaviors of both Ni-based and Fe-based alloys were investigated. This study developed a redox potential range for mitigating corrosion by using a redox couple of UF4 /UF3 and a novel approach of potential measurements against F2/ F- potential experimentally. / Doctor of Philosophy / This work focuses on material degradation in harsh and extreme nuclear environments, including light water reactors and advanced reactors such as molten salt and liquid metal coolant reactors. Given the renewed interest in advanced nuclear reactors as a resource of clean energy, advanced material development, including structural, fuel, and coolant materials, has become a significant and trending research area. Based on our past experiences, we have seen the detrimental effects of material failure due to corrosion. Systems are inherently safe in the absence of material degradation. Nevertheless, this is an idealistic thought, as corrosion is inevitable. Therefore, this research focuses on corrosion mitigation, as absolute material preservation is impossible. This work includes corrosion studies in aqueous environments in light water reactors and advanced nuclear reactors under normal operation and extreme conditions such as accident environments. Much of this work provides insights into material corrosion behavior and mitigation that helps nuclear reactor operators ensure safe operations. Commercially available alloys such as SS316, Hast. X and Hast. N were examined in primary water, molten salt, high-temperature steam, and liquid bismuth environment and model Fe-Cr-Si alloys were investigated in water and steam to compare the corrosion mechanisms.
| Identifer | oai:union.ndltd.org:VTETD/oai:vtechworks.lib.vt.edu:10919/111923 |
| Date | 19 September 2022 |
| Creators | Leong, Amanda |
| Contributors | Mechanical Engineering, Zhang, Jinsuo, West, Robert L., Pierson, Mark Alan, Bai, Xianming |
| Publisher | Virginia Tech |
| Source Sets | Virginia Tech Theses and Dissertation |
| Language | English |
| Detected Language | English |
| Type | Dissertation |
| Format | ETD, application/pdf, application/pdf |
| Rights | In Copyright, http://rightsstatements.org/vocab/InC/1.0/ |
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