Session: MF-29-01 Mechanical Properties of Nuclear Graphite and their Implementation in Codes and Standards (Joint with CS-15)

Paper Number: 106613

106613 - Graphite - Molten Salt Considerations for Components in Nuclear Applications 

The new high-temperature reactor (HTR) designs being considered for future Generation IV nuclear reactor deployment include designs using molten salt as the primary coolant. These molten salt–cooled graphite core designs pose new material compatibility challenges that are not considered within the gas-cooled HTR designs that have been previously built and operated. Although early indications from the Molten Salt Reactor Experiment (MSRE) in the 1960s were that molten salts could be considered chemically inert to graphite, recent studies revealed additional physical and thermal interactions that the molten salt imposes that may be just as significant as the chemical reactivity. Specifically, molten salt intrusion into the open pore structure of nuclear graphite grades can cause additional internal stresses within the microstructure, exacerbating the stress accumulation from irradiation-induced dimensional change. Additionally, designs using a molten salt–containing liquid fuel could produce hot spots within graphite structural components, causing local thermal stresses. Abrasion and erosion concerns are magnified with molten salt because of their extremely high density (some salts have higher densities than the structural graphite components). Finally, the graphite–graphite and fuel pebble–graphite tribological behavior are distinctly different within the molten salt from the inert gas environments and must be investigated. These topics and others are currently under investigation within the US Department of Energy Advanced Reactor Technologies graphite program and will be discussed in depth.

Presenting Author: William Windes Idaho National Laboratory

Presenting Author Biography: Dr. William Windes is a Division Fellow at the Idaho National Laboratory and is the current Chair of the ASME Nonmetallic Materials and Design Working Group. He has over 30 years of extreme material research and development and has been the Technical Lead for the DOE Advanced Reactor Technologies (ART) Graphite R&D Program

Authors:

Nidia C. Gallego Oak Ridge National Laboratory
Josina W. Geringer Oak Ridge National Laboratory
William Windes Idaho National Laboratory