Thermal Treatment of UK Irradiated Graphite Waste: the 14C Story

Robert Numa Worth, Greg Black, Lorraine McDermott (Collaborator), Abbie Jones (Collaborator), Paul Mummery (Collaborator), Anthony J Wickham (Collaborator)

    Research output: Contribution to journalArticlepeer-review

    Abstract

    Approximately 96,000 tonnes of the UK Higher Activity Waste (HAW) inventory consists of irradiated nuclear graphite. The current NDA baseline strategy for irradiated graphite in England and Wales is isolation in a future Geological Disposal Facility, with Scottish policy endorsing an alternative decision of near surface long-term storage. Irradiated graphite disposal routes in the UK remain under review, however, as there are concerns surrounding timing and whether deep geological disposal is the most appropriate course of action for graphite[1]. An alternative waste management solution is treatment prior to disposal to separate mobile radioactive isotopes such as 3H and 14C from the bulk material, allowing for HAW volume reduction and concentration, and potential considerable cost saving through reduced interim storage requirements. Readily available facilities could be utilised for disposal of any remaining bulk material if waste classification reduction can be achieved. This programme of research employs controlled thermal oxidation to investigate the factors affecting radioisotope removal from Oldbury Magnox reactor graphite, with particular emphasis on studying the mobility of prominent volatile isotopes such as 3H and 14C, and the inventory of non-volatile isotopes such as 60Co – these are the highest specific activity radioisotopes at the end of a graphite-moderated reactor’s working lifetime. The effects of treatment variables such as oxidation time, temperature and gas composition are under investigation in order to fully understand and optimise radioisotope removal with minimal weight loss, and to drive forward the evaluation and application of thermal treatment processes to UK irradiated graphite. Preliminary results demonstrate that up to 45% 3H and 50% 14C can be removed from irradiated graphite with a corresponding 40% sample weight loss caused by thermal oxidation at 700oC. The performance of this removal process can be significantly enhanced with lower temperature oxidation at 600oC; the specific activity of 3H and 14C in the off-gas, and thus the radioisotope removal efficiency, increases from order 104 Bq/g to 105 Bq/g following this temperature reduction, incurring a penalty to treatment duration. [1] NDA, SMS/TS/D1-HAW-10/001/B Higher Activity Waste – Operational Management Strategy (Gate A&B) – v2.0, August 2013
    Original languageEnglish
    JournalNuclear Future
    Issue numberJanuary/February 2015
    Publication statusPublished - 5 Jan 2015

    Keywords

    • Irradiated Graphite, Graphite Waste, Carbon-14, 14C, Thermal Treatment, Decontamination

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