Method to assess the radionuclide inventory of irradiated graphite waste from gas-cooled reactors

Abstract

About 17,000 tons of irradiated graphite waste will be produced from the decommissioning of the six gas-cooled nuclear reactors operated by Electricite De France Limited company (EDF), an energetic utility ( http://www.edf.com/the-edf-group-42667.html ) in France. Determining the radionuclide content of this waste is an important legal commitment for both safety reasons and in order to determine the best suited management strategy. As evidenced by numerous studies nuclear graphite is a very pure material, however, it cannot be considered from an analytical viewpoint as a usual homogeneous material. Radionuclide measurements in irradiated graphite exhibit very high discrepancies especially when corresponding to precursors at trace level. This huge discrepancy cannot be avoided and can be easily explained by Pierre Gy’s theory of sampling of finely divided materials. The assessment of a radionuclide inventory only based on few number of radiochemical measurements leads in most of cases to a gross over or under-estimation that can be detrimental to graphite waste management. A method using an identification calculation–measurement process is proposed in order to assess the radionuclide inventory as precisely as possible whilst guaranteeing an upper margin corresponding to a 2.5 % risk of under-assessment. This method closely reflects the reality of the main phenomenon at radionuclide origin in a reactor, while also incorporating the secondary effects that can alter this result such as radionuclide or precursor release during reactor operation.