​​​The DOE-ART Graphite R&D program is the primary nuclear graphite research program for the USA. This program focuses on research and development activities necessary to qualify and license graphite components for use within nuclear applications, specifically within advanced reactor designs such as High Temperature Reactor designs. The data generated within the ART Graphite program is intended to be used in conjunction with other publicly available nuclear graphite data such as is contained within the ​IAEA Nuclear Graphite Knowledge Base​. The ART Graphite program is divided into 5 primary research areas providing a combination of data, analysis reports, and pertinent references to describe and explain the trends within the data.

  • Unirradiated (Baseline): Establish as-manufactured (Baseline) values for unirradiated material properties that can be used to determine the quantitative changes during irradiation and degradation during nuclear applications. 
  • Irradiation (AGC Experiment): Establish evolution of material property changes due to irradiation dose and temperature. The AGC Experiment is an irradiation creep experiment which provides creep data for selected graphite grades.  
  • Degradation: Establish effects of irradiation, oxidation, and molten salt interaction on graphite behavior.
  • Behavior models: Predictive and degradation models for graphite behavior.
  • Licensing and code: Papers and data supporting ASME code development and NRC license assessment.​
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