Use this URL to cite or link to this record in EThOS: http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.713579
Title: Thermal treatment of Oldbury Magnox reactor irradiated graphite
Author: Worth, Robert
Awarding Body: University of Manchester
Current Institution: University of Manchester
Date of Award: 2016
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Abstract:
Approximately 96,000 tonnes of the UK Higher Activity Waste (HAW) inventory consists of irradiated nuclear graphite. The current Nuclear Decommissioning Authority (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. 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. Optimisation of an existing thermal treatment process at the Nuclear Graphite Research Group (NGRG) of the University of Manchester has been effected and a detailed review of the uncertainties associated with quantitative determination of radioisotope releases during thermal treatment of irradiated graphite samples has been conducted. Thermal treatment experiments in both an inert atmosphere and 1% oxygen in argon atmosphere have been conducted for temperatures ranging from 600°C to 800°C, and durations from 4 to 120 hours, to determine the effects of oxidation time and temperature, and the consequent oxidation characteristics on the release rate of prominent radioisotopes, with a focus on the release of 14C. Lower temperature treatments in an oxidising atmosphere have shown that a preferential release of 14C-enriched graphite can be achieved from the bulk material of Oldbury Magnox reactor irradiated graphite, with evidence demonstrating that this liberated 14C-enriched region is located at the graphite surfaces throughout the porous structure. A large proportion of radiocarbon found in this irradiated graphite, however, is uniformly distributed throughout the bulk material and cannot be selectively oxidised. It is found that prominent metallic radioisotopes such as 60Co are not mobile at these temperatures and remain in the bulk graphite material, inclusive of radioactive caesium which the literature suggests will volatilise. The preliminary results were undertaken as part of the EU FP7 EURATOM Project: CARBOWASTE.
Supervisor: Not available Sponsor: Engineering & Physical Sciences Research Council (EPSRC)
Qualification Name: Thesis (Ph.D.) Qualification Level: Doctoral
EThOS ID: uk.bl.ethos.713579  DOI: Not available
Keywords: radioactive waste ; irradiated graphite ; thermal treatment ; decontamination
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