This paper introduces a novel approach, developed through an academic and industrial collaboration, to the thermal treatment of nuclear graphite waste arising as a result of reactor decommissioning and oxide fuel assembly dismantling. A crucial part of the process is the thermal oxidation of the graphite via a plasma furnace. Laboratory scale treatment of the graphite found the oxidation rate to increase with temperature, with a significant increase in the CO/CO 2 production ratio at T > 1000 °C. There was also a linear increase of oxidation rate with air flow rate, up to 100 ml/min, after which, the process became less efficient due to oxygen wastage. Effects of graphite particle size, over the range 0.5–10.0 mm, was found to be relatively small and importantly the effect of the graphite provenance on the oxidation rate was also found to be minimal. Treatment of virgin and irradiated graphite analogues, under the same conditions, showed little difference in oxidation behaviour, providing confidence that this process could be scaled up and used effectively in the disposal of reactor cores. Scale-up of this work was carried out on a “nuclear-ready” full-scale industrial facility, demonstrating graphite feeding, furnace operation and graphite destruction successfully. Experiments showed that comparable conditions between lab scale and pilot-scale showed similar oxidation behaviour, with 71.6 kg (equating to 90%) of graphite gasified in 6 h, giving an oxidation rate of ∼12 kg/h. Engagement with UK regulators has indicated that it is likely to be desirable to further investigate the possibility that isolation and confinement of graphite-derived 14 C (t 1/2 = 5730 yrs) in a carbon sequestration scheme may be an improvement on the baseline strategy, which is to store it in its solid form as untreated graphite in a geological repository.
Theodosiou, A., Jones, A. N., Burton, D., Powell, M., Rogers, M., & Livesey, V. B. (2018). The complete oxidation of nuclear graphite waste via thermal treatment: An alternative to geological disposal. Journal of Nuclear Materials, 507, 208–217. https://doi.org/10.1016/j.jnucmat.2018.05.002