Characterisation and treatment of irradiated graphite waste

Decommissioning of the UK’s Magnox and Advanced Gas Cooled reactors, research reactors and plutonium production reactors will produce approximately 90,000 tonnes of graphite waste, this by far the greatest volume produced worldwide from one country. The management of this graphite waste will require complex planning and consideration due to the volume and nature of the material, therefore radioactive graphite core dismantling and the management of radioactive graphite waste is an important issue in the UK. The current UK plan for graphite disposal is that this material will be packaged as Intermediate Level Waste (ILW) for deep geological repository disposal. It has not yet been shown that the current baseline represents the optimum solution in terms of safety, cost and protection of the environment. In order to evaluate the possibility of reducing costs, it is important to consider alternative decommissioning methods. To make informed decisions of how best to dispose of large volumes of irradiated graphite waste, it is necessary to understand fully its microstructural and radioisotopic character and the consequent effectiveness of the various proposed preparative treatment options. The aim of this research project is to develop and demonstrate advanced immobilisation a spectrum of graphite wastes, in order to reduce this volume of waste ILW to that of Low Level Waste. Introduction The main issue associated with the disposal of irradiated graphite is the large volume of relatively low active waste to be dealt with [1, 2]. Current data released by the Nuclear Decommissioning Authority (NDA) [3] reports that 56,000 tonnes Magnox, 25,000 tonnes AGR plus graphite from other sources including Windscale piles and test reactors will lead to 90,000 tonnes of nuclear waste graphite in the UK. To date the favoured option for the 90,000 tonnes of irradiated nuclear graphite waste is for deep geological disposal. In order to evaluate the possibility of reducing costs, it is important to consider alternative methods. There are potential concerns associated with several radioactive isotopes in this material; these are namely H, C, Cl, Co, Eu and Cs. One of the most important isotopes of concern is C [4], mainly produced during irradiation through the transmutation of the N. C has a long half life of 5730 years and there are environmental concerns related to possible release of this isotope during decommissioning. In recent review, it was concluded that it may be possible to reduce the activity of the radioactive graphite waste [5]. If applied, this would lead to the reduction in volume of intermediate level waste. In order to make informed decisions of how best to dispose of large volumes of irradiated graphite waste from the nuclear programme, it is necessary to understand fully its character and the consequent effectiveness of the various proposed decontamination and immobilisation treatments these include various methods of encapsulation, and alternative preparative treatments for radiological reduction to LLW including disposal [6], reprocessing [7] and decontamination [8]. Graphite is used in reactors mostly as a moderator and reflector to slow down fast-neutrons released from fission events. In early low temperature reactors, radiation-damage effects caused problems including swelling in the Hanford and Brookhaven reactors and accumulation of stored energy in the Windscale, BEPO, and X-10 reactors [9]. Irradiation-induced dimensional change [10] is also an underlying cause of the lifetime limits for higher temperature reactors, it is characterised by the creation and aggregation of large numbers of point defects which change and distort the surrounding lattice. This can lead to significant bulk dimensional changes and significantly alters the physical properties such as thermal and electrical conductivity.