Liam Payne

A Study of the Oxidation Behaviour of Pile Grade A (PGA) Nuclear Graphite Using Thermogravimetric Analysis (TGA), Scanning Electron Microscopy (SEM) and X-Ray Tomography (XRT)

Pile grade A (PGA) graphite was used as a material for moderating and reflecting neutrons in the UK’s first generation Magnox nuclear power reactors. As all but one of these reactors are now shut down there is a need to understand the residual state of the material prior to decommissioning of the cores, in particular the location and concentration of key radio-contaminants such as 14C. The oxidation behaviour of unirradiated PGA graphite was studied, in the temperature range 600–1050°C, in air and nitrogen using thermogravimetric analysis, scanning electron microscopy and X-ray tomography to investigate the possibility of using thermal degradation techniques to examine 14C distribution within irradiated material. The thermal decomposition of PGA graphite was observed to follow the three oxidation regimes historically identified by previous workers with limited, uniform oxidation at temperatures below 600°C and substantial, external oxidation at higher temperatures. This work demonstrates that the different oxidation regimes of PGA graphite could be developed into a methodology to characterise the distribution and concentration of 14C in irradiated graphite by thermal treatment.

Examination of Surface Deposits on Oldbury Reactor Core Graphite to Determine the Concentration and Distribution of 14C

Pile Grade A graphite was used as a moderator and reflector material in the first generation of UK Magnox nuclear power reactors. As all of these reactors are now shut down there is a need to examine the concentration and distribution of long lived radioisotopes, such as 14C, to aid in understanding their behaviour in a geological disposal facility. A selection of irradiated graphite samples from Oldbury reactor one were examined where it was observed that Raman spectroscopy can distinguish between underlying graphite and a surface deposit found on exposed channel wall surfaces. The concentration of 14C in this deposit was examined by sequentially oxidising the graphite samples in air at low temperatures (450°C and 600°C) to remove the deposit and then the underlying graphite. The gases produced were captured in a series of bubbler solutions that were analysed using liquid scintillation counting. It was observed that the surface deposit was relatively enriched with 14C, with samples originating lower in the reactor exhibiting a higher concentration of 14C. Oxidation at 600°C showed that the remaining graphite material consisted of two fractions of 14C, a surface associated fraction and a graphite lattice associated fraction. The results presented correlate well with previous studies on irradiated graphite that suggest there are up to three fractions of 14C; a readily releasable fraction (corresponding to that removed by oxidation at 450°C in this study), a slowly releasable fraction (removed early at 600°C in this study), and an unreleasable fraction (removed later at 600°C in this study).

Evaluation of the use of magnetic sector secondary ion mass spectrometry to investigate 14C distribution in Magnox reactor core graphite

Abstract Large quantities of irradiated graphite will arise from the decommissioning of the UK’s Magnox power stations. Irradiated graphite contains 14C as well as other longer lived radionuclides (e.g. 36Cl). The potential use of magnetic sector secondary ion mass spectrometry (MS-SIMS) to examine the distribution of the 14C within trepanned graphite samples from a Magnox nuclear power station has been investigated. This work indicates that the methodology proposed has the potential to be used to analyse irradiated graphite samples with preliminary results highlighting a possible 14C enrichment in the carbonaceous deposit found on a channel wall sample. 14C concentrations in samples without this deposit were below the limits of detection of the instrument. The methodology used for these determinations ensured that possible mass interferences between 14C species and oxygen-bearing or nitrogen-bearing species were eliminated from the analysis. Future work will utilize the methodology proposed in this work on a larger number of samples.

The Impact of Alkaliphilic Biofilm Formation on the Release and Retention of Carbon Isotopes from Nuclear Reactor Graphite

Abstract14C is an important consideration within safety assessments for proposed geological disposal facilities for radioactive wastes, since it is capable of re-entering the biosphere through the generation of 14C bearing gases. The irradiation of graphite moderators in the UK gas-cooled nuclear power stations has led to the generation of a significant volume of 14C-containing intermediate level wastes. Some of this 14C is present as a carbonaceous deposit on channel wall surfaces. Within this study, the potential of biofilm growth upon irradiated and 13C doped graphite at alkaline pH was investigated. Complex biofilms were established on both active and simulant samples. High throughput sequencing showed the biofilms to be dominated by Alcaligenes sp at pH 9.5 and Dietzia sp at pH 11.0. Surface characterisation revealed that the biofilms were limited to growth upon the graphite surface with no penetration of the deeper porosity. Biofilm formation resulted in the generation of a low porosity surface layer without the removal or modification of the surface deposits or the release of the associated 14C/13C. Our results indicated that biofilm formation upon irradiated graphite is likely to occur at the pH values studied, without any additional release of the associated 14C.

Validation of a novel radiation mapping platform for the reduction of operator-induced shielding effects

With extensive remediation currently ongoing because of the Fukushima Daiichi Nuclear Power Plant accident, there exists the even greater need to provide a system with which the distribution of radiation (specifically radiocesium) can be rapidly determined across extensive areas, yet at high (metre or sub-metre) spatial resolutions. Although a range of potential survey methods have been utilised (e.g. fixed-wing aircraft, helicopter, vehicular and more-recently unmanned aerial vehicle) to characterise the distribution of radiological contamination, ground-based (on-foot) methods that employ human operatives to traverse sites of interest remains one of the primary methods through which to perform routine radiological site surveys. Through the application of a newly-developed platform carried as a backpack-contained unit, it was possible to map sites at twice the rate previously possible-reducing not only the exposure time of the operator to ionising radiation, but also dramatically reducing the level of radiation attenuation (introduced by the operator) onto the detector. Like magnetometry platforms used during geological ore prospecting, this system was similarly boom-based, extending sideways away from the central operator. While conventional radiological survey platforms require a correction be performed on the data to account for the carrier (aircraft, vehicle or human) interception and attenuation incident radiation-this system is shown to not require such a retrospective correction.