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Meeting Materials in Nuclear Energy Systems (MiNES) 2025
Symposium Materials in Nuclear Energy Systems (MiNES) 2025
Presentation Title Preliminary Insights From Post Irradiation Examination of AGR 5/6/7 TRISO Fuel
Author(s) William Cureton, Tyler Gerczak, Grant Helmreich
On-Site Speaker (Planned) William Cureton
Abstract Scope The U.S. Advanced Gas Reactor Fuel Development and Qualification (AGR) Program has addressed TRISO fuel qualification through coordinated fuel fabrication, irradiation campaigns, and follow on examinations post irradiation. The AGR team at Idaho National Laboratory and Oak Ridge National Laboratory (ORNL) is conducting post irradiation examination (PIE) and high temperature safety testing on compacts from the final AGR irradiation experiment (AGR 5/6/7). This work highlights initial findings from destructive PIE of several as irradiated cylindrical compacts, as well as safety testing at temperatures at or exceeding 1600 °C. Particular focus is given to fission product release behavior at both the compact and individual particle scale, utilizing various experimental techniques contributing to a multiscale understanding of performance. Ongoing AGR 5/6/7 investigations are expanding knowledge of TRISO fuel behavior over a broader irradiation temperature and burn up range compared to earlier AGR experiments, under both normal operating conditions and simulated loss of coolant accident scenarios.
Proceedings Inclusion? Undecided

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