The Next Generation Nuclear Plant (NGNP) Graphite R & D program is currently measuring irradiated material property changes in several grades of nuclear graphite for predicting their behavior and operating performance within the core of new Very High Temperature Reactor (VHTR) designs. The Advanced Graphite Creep (AGC) experiment consisting of six irradiation capsules will generate this irradiated graphite performance data for NGNP reactor operating conditions. All six AGC capsules in the experiment will be irradiated in the Advanced Test Reactor (ATR), disassembled in the Hot Fuel Examination Facility (HFEF), and examined at the INL Research Center (IRC) or Oak Ridge National Laboratory (ORNL). This is the first in a series of status reports on the progress of the AGC experiment. As the first capsule, AGC1 was irradiated from September 2009 to January 2011 to a maximum dose level of 6-7 dpa. The capsule was removed from ATR and transferred to the HFEF in April 2011 where the capsule was disassembled and test specimens extracted from the capsules. The first irradiated samples from AGC1 were shipped to the IRC in July 2011and initial post irradiation examination (PIE) activities were begun on the first 37 samples received. PIE activities continue for the remainder of the AGC1 specimen as they are received at the IRC.

The Next Generation Nuclear Plant (NGNP) Graphite Research and Development (R & D) Program is currently measuring irradiated material properties for predicting the behavior and operating performance of new nuclear graphite grades available for use within the cores of new very high temperature reactor designs. The Advanced Graphite Creep (AGC) experiment, consisting of six irradiation capsules, will generate irradiated graphite performance data for NGNP reactor operating conditions. The AGC experiment is designed to determine the changes to specific material properties such as thermal diffusivity, thermal expansion, elastic modulus, mechanical strength, irradiation induced dimensional change rate, and irradiation creep for a wide variety of nuclear grade graphite types over a range of high temperature, and moderate doses. A series of six capsules containing graphite test specimens will be used to expose graphite test samples to a dose range from 1 to 7 dpa at three different temperatures (600, 900, and 1200°C) as described in the Graphite Technology Development Plan. Since irradiation induced creep within graphite components is considered critical to determining the operational life of the graphite core, some of the samples will also be exposed to an applied load to determine the creep rate for each graphite type under both temperature and neutron flux. All six AGC capsules in the experiment will be irradiated in the Advanced Test Reactor (ATR). AGC-1 and AGC-2 will be irradiated in the south flux trap and AGC-3-AGC-6 will be irradiated in the east flux trap. The change in flux traps is due to NGNP irradiation priorities requiring the AGC experiment to be moved to accommodate Fuel irradiation experiments. After irradiation, all six AGC capsules will be cooled in the ATR Canal, sized for shipment, and shipped to the Materials and Fuels Complex (MFC) where the capsule will be disassembled in the Hot Fuel Examination Facility (HFEF). During disassembly, the metallic capsule will be machined open and the individual samples removed from the interior graphite body containing the samples. Samples removed from the capsule will be loaded in a shipping drum and shipped to the Idaho National Laboratory (INL) Research Center (IRC) for initial post-irradiation examination (PIE) and storage for any future testing at the newly completed Carbon Characterization Laboratory (CCL). All work was performed under an ASME NQA-1-2008;1a-2009 compliant quality assurance program.