Lorraine McDermott

Development of Surface Marker System for the Observation of Microstructural Changes in Nuclear Graphite using Micro X-ray Tomography

Current energy demands and future energy needs are a growing industry which at present attracts a large amount of research and investment of which nuclear energy is an integral part. Eight new nuclear stations are proposed to be developed in the UK over the next ten years to meet this demand. In order for nuclear energy to sustain growth and development, nuclear decommissioning of first and second generation power stations needs to be addressed in the U.K. and worldwide. Presently the UK has 36 graphite moderated reactors as a result of the UK military and civil programs, which over the next twenty years will close. This will result in ?99?000 tonnes of irradiated graphite waste for which no current national decommissioning strategy exists. The main issues associated with this waste are the large volume and activation products associated. By far the greatest inventory is from 3H and 14C. An EU Euroatom FP7 Program; CARBOWASTE was established in 2008 with the aim of developing treatment and disposal options for graphite.This research is based within CARBOWASTE, the main objectives are to understand the mechanisms involved in the production, location and removal of radioisotopes from nuclear graphite. Computed X-ray Tomography (CT) will be used in order to quantify the initial porosity in conjunction with thermal treatment (ex situ) in order to eventually identify the location of 14C within the matrix of irradiated graphite, through the preferential chemically controlled oxidation of graphite. Unirradiated Pile Grade A graphite samples have been laser and manually marked in order align the samples prior to and post thermal treatment to determine the degree of porosity changes and weight loss under a range of thermal oxidation parameters

3H and 14C Release in UK Nuclear Graphite by Chemical Treatment

Irradiated nuclear graphite waste may account for up to 30% of any future UK geological ILW disposal facility (NDA, 2002). In order to make informed decisions of how best to dispose of large volumes of irradiated graphite waste within the UK nuclear programme, it is necessary to understand the nature and migration of isotopes present within the graphite and ultimately how these short and long term isotopes may behave prior to and during disposal. Various proposed decontamination and immobilisation treatments, within the EU Euroatom FP7 CARBOWASTE programme have been explored (von Lensa, 2008; von Lensa et al., 2011). Experiments have been carried out on UK irradiated graphite waste to investigate the removal of isotopic content prior to long term storage and to assess the long term leachability of isotopes. Several leaching conditions have been developed in order to remove 3H and 14C from the irradiated graphite. Leach results obtained from this analysis and differences observed under varying leaching conditions are discussed. Thermal analysis of the samples pre- and post-leaching has been performed to quantify the 14C and 3H inventory. Finally, the suitability of the developed chemical treatments before disposal is discussed.