Removal of carbon-14 from irradiated graphite

Approximately 250,000 tonnes of irradiated graphite waste exists worldwide and that quantity is expected to increase with decommissioning of Generation II reactors and deployment of Generation IV gas-cooled, graphite moderated reactors. This situation indicates the need for a graphite waste management strategy. On of the isotopes of great concern for long-term disposal of irradiated graphite is carbon-14 (C-14), with a half-life of 5730 years. Study of irradiated graphite from some nuclear reactors indicates C-14 is concentrated on the outer 5 mm of the graphite structure. The aim of the research presented here is to develop a practical method by which C-14 can be removed. In parallel with these efforts, the same irradiated graphite material is being characterized to identify the chemical form of C-14 in irradiated graphite. A nuclear-grade graphite, NBG-18, and a high-surface-area graphite foam, POCOFoam (R), were exposed to liquid nitrogen (to increase the quantity of C-14 precursor) and neutron-irradiated (10(13) neutrons/cm(2)/s). During post-irradiation thermal treatment, graphite samples were heated in the presence of an inert carrier gas (with or without the addition of an oxidant gas), which carries off gaseous products released during treatment. Graphite gasification occurs via interaction with adsorbed oxygen complexes. Experiments in argon only were performed at 900 degrees C and 1400 degrees C to evaluate the selective removal of C-14. Thermal treatment also was performed with the addition of 3 and 5 vol% oxygen at temperatures 700 degrees C and 1400 degrees C. Thermal treatment experiments were evaluated for the effective selective removal of C-14. Lower temperatures and oxygen levels correlated to more efficient C-14 removal. (C) 2014 Elsevier B.V. All rights reserved.