Removal of carbon-14 from irradiated graphite

• Published in: Journal of nuclear materials 2014-08, Vol.451 (1-3), p.328-335
• Main Authors: Dunzik-Gougar, Mary Lou, Smith, Tara E.
• Format: Article
• Language: English
• ISSN: 0022-3115; 1873-4820
• DOI: 10.1016/j.jnucmat.2014.03.018

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 (14C), with a half-life of 5730years. Study of irradiated graphite from some nuclear reactors indicates 14C is concentrated on the outer 5mm of the graphite structure. The aim of the research presented here is to develop a practical method by which 14C can be removed. In parallel with these efforts, the same irradiated graphite material is being characterized to identify the chemical form of 14C in irradiated graphite. A nuclear-grade graphite, NBG-18, and a high-surface-area graphite foam, POCOFoam®, were exposed to liquid nitrogen (to increase the quantity of 14C precursor) and neutron-irradiated (1013neutrons/cm2/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°C and 1400°C to evaluate the selective removal of 14C. Thermal treatment also was performed with the addition of 3 and 5vol% oxygen at temperatures 700°C and 1400°C. Thermal treatment experiments were evaluated for the effective selective removal of 14C. Lower temperatures and oxygen levels correlated to more efficient 14C removal.