Investigation of γ lithium aluminate as tritium breeding material for a fusion reactor blanket

Abstract

The assessment of γ-LiAlO 2 , a potential ceramic breeder, is in progress at CEA, within the framework of the EEC fusion technology program. The performance of this material, with respect to relevant issues, is being investigated (i.e. fabrication, unirradiated material properties, neutron irradiation damage and tritium extraction). Using the selected fabrication route, pure and homogeneous materials are obtained covering a wide range of microstructures. Optimization of material microstructure with respect to tritium release is underway and upscaling of the fabrication capacity is being considered. Investigation of properties — physico-chemical, mechanical, thermal, compatibility with structural materials and beryllium s underway and the data needed for design purposes are being collected. Experiments were designed to investigate the neutron irradiation and the tritium release behaviours of the material: — effects of neutron irradiation are studied by measuring properties changes after irradiations in the core of the Osiris reactor corresponding to burn-ups of 1 to 2% of the total lithium. — tritium extraction is studied both in annealing experiments and in in-pile experiments in Siloe or Melusine reactors in conditions representative of blanket operation. For aluminate pellets investigated so far in the 450 °C–600 °C range, results show that neutron radiation damage is negligible and that tritium residence time may be less than one day.