Innovative Extension of the EDF Inventory Methodology for Irradiated Graphite Sleeves - 16084

About 17,000 t of EDF irradiated graphite wastes will be produced from the decommissioning of the six French gas-cooled nuclear reactors. Determining the radionuclide content of this graphite is of capital importance for safety purposes and in order to determine the best way to manage them. A method using an identification calculation-measurement process has been proposed in order to assess a radiological inventory as precise as possible and has been validated by ANDRA, the French governmental radioactive waste management agency and by IRSN, the technical support of the French nuclear safety authority. As presented at WM2013, this method is based on (a) radionuclide measurements from sufficient samples from a statistical point of view and (b) activation coefficients calculated from the neutron flux histories of the measured samples. For that purpose, it is important to know the precise location of each measured graphite sample, which is easy with pile sampling. About 2,000 t of EDF irradiated graphite is composed of sleeves which contained fuel cartridges and they have been stored in silos of St-Laurent site. The general method needs to know the precise history including neutron flux of any sampled sleeve. As there is no possibility to individually identify sleeves, such a requirement on sleeves in a silo is impossible. The paper presents both the general method for piles and an innovative way to replace that knowledge for long lived radionuclide of irradiated sleeves. (authors)