Development of an ASTM Graphite Oxidation Test Method
- ORNL
Oxidation behavior of graphite is of practical interest because of extended use of graphite materials in nuclear reactors. High temperature gas-cooled reactors are expected to become the nuclear reactors of the next generation. The most critical factor in their safe operation is an air-ingress accident, in which case the graphite materials in the moderator and reflector would come in contact with oxygen at a high temperature. Many results on graphite oxidation have been obtained from TGA measurements using commercial instruments, with sample sizes of a few hundred milligrams. They have demonstrated that graphite oxidation is in kinetic control regime at low temperatures, but becomes diffusion-limited at high temperatures. These effects are better understood from measurement results with large size samples, on which the shape and structural factors that control diffusion can be more clearly evidenced. An ASTM test for characterization of oxidation resistance of machined carbon and graphite materials is being developed with ORNL participation. The test recommends the use of large machined samples (~ 20 grams) in a dry air flow system. We will report on recent results and progress in this direction.
- Research Organization:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- Sponsoring Organization:
- USDOE Office of Nuclear Energy (NE)
- DOE Contract Number:
- DE-AC05-00OR22725
- OSTI ID:
- 965818
- Resource Relation:
- Conference: Carbon 2006 International Carbon Conference, Aberdeen, United Kingdom, 20060716, 20060721
- Country of Publication:
- United States
- Language:
- English
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