Microstructural characterization of the CGB graphite grade from the molten salt reactor experiment

J. David Arregui-Mena, Philip D. Edmondson, David Cullen, Samara Levine, Cristian Contescu, Yutai Katoh, Nidia Gallego

Research output: Contribution to journalArticlepeer-review

6 Scopus citations

Abstract

In the 1960s, the Molten Salt Reactor Experiment demonstrated the feasibility of molten salt reactors for civil applications. The Molten Salt Reactor Experiment used a nuclear graphite grade known as CGB in the core as the fast neutron moderator. For application in this purpose, CGB graphite underwent additional impregnation steps that are not typically used in conventional nuclear graphites. These additional impregnations significantly lowered the permeability and reduced the size of pore openings in CGB graphite, for the purpose of inhibiting molten salt ingression into the graphite. Although several publications report the mechanical properties and irradiation response of CGB graphite, little information has been published about the microstructure or sealant used for this graphite grade. Here the results of multiple advanced microscopy techniques are presented with the objective to investigate the unique microstructure and sealing technology of legacy material from the Molten Salt Reactor Experiment so it can be potentially readopted for modern reactors. The microstructural characterization shows that some pores were fully sealed by the impregnation whereas other regions contained foam-like materials that partially filled the void content. This manuscript also investigates and documents the nature and possible source of the sealant material. The pore sealing methods used for CGB are discussed and compared with other common technologies used to reduce the impregnation of molten salts into graphite components. Knowledge of CGB graphite's microstructure can enable development of new pore infiltration technologies to improve in operando behavior advancing the next generation of molten salt reactors.

Original languageEnglish
Article number154421
JournalJournal of Nuclear Materials
Volume582
DOIs
StatePublished - Aug 15 2023

Funding

This work was supported by the Office of Nuclear Energy under DOE Idaho Operations Office Contract DE-AC07- 051D14517 as part of a Nuclear Science User Facilities experiment. A portion of this research used the resources of the Low Activation Materials Development and Analysis Laboratory (LAMDA), a DOE Office of Science research facility operated by the Oak Ridge National Laboratory (ORNL). Support from the Advanced Reactor Technologies program of DOE Office of Nuclear Energy is also acknowledged. Oak Ridge National Laboratory is managed by UT-Battelle under contract DE-AC05–00OR22725. The authors would like to thank Hughie Spinoza for his valuable comments and discussion.

FundersFunder number
Nuclear Energy
U.S. Department of EnergyDE-AC07- 051D14517
Office of Nuclear Energy
Oak Ridge National Laboratory
UT-BattelleDE-AC05–00OR22725

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