Preliminary study on oxidation of nuclear grade graphite

A test methodology has been defined in order to assess the risk of bulk-oxidation of nuclear graphite with a focus on particular operations, such as cutting of graphite components during decommissioning of nuclear reactors. Graphite is used for its properties in nuclear reactors, as a fuel element matrix, moderator, structural and reflector material. The experimental tests have been performed using a horizontal TGA-DTA furnace and both isothermal and nonisothermal tests were performed on non-irradiated specimens. In isothermal tests, testing temperature range was 550 – 750 °C, the selected temperature has been reached heating the furnace under N2 atmosphere at 5 °C/min, then oxidation has been evaluated at constant temperature in air (100 mL/min) for at least 180 min. Non-isothermal tests have been performed either under N2 atmosphere or air and samples have been heated at 5 °C/min from 30 to 950 °C. As several properties may affect the oxidation resistance of nuclear graphite, such as impurities, grain size, pore structure, graphitization degree, antioxidation treatments, the testing protocol has been preliminary applied to non-irradiated graphite samples from two different graphite reactors. Preliminary experimental results revealed the excellent ability of the tested graphite samples for oxidation resistance with lower oxidation rates and longer oxidation times compared with some literature results.