Vitrification of UK intermediate level radioactive wastes arising from site decommissioning. Initial laboratory trials

Abstract

Vitrification was considered as one potential treatment option for four types of wet intermediate level radioactive waste (wet ILW) arising from decommissioning of a UK Magnox nuclear power station. Here we discuss the results of initial laboratory scale trial vitrification studies using suitable glass compositions which were previously short listed from a matrix of 80 potential candidates. The results of the initial trials have (a) demonstrated the feasibility of vitrification of these wet ILWs at 35 wt% (dry) waste loading; (b) confirmed that the candidate glasses exhibit acceptable chemical durabilities; and (c) enabled further down-selection to three final candidate glasses which have undergone detailed analysis and testing, which will be discussed in a forthcoming publication. Waste loading of 35 wt% (dry) waste has been demonstrated for all four waste permutations under consideration, and results indicate that achievable final waste loading limits may be considerably higher. For most glasses studied Cs retention was 60–75%; however, this result was partly attributable to high volatilisation rates resulting from the especially high surface area / volume ratio of the small laboratory melts studied. Reductions in melting temperature from 1200°C are possible for the majority of the studied glasses, which should also increase Cs retention and reduce melt corrosivity