Abstract
Processes are being developed to immobilize high-level radioactive wastes in glass for increased safety during handling, storage and final disposal. A concurrent program is the characterization of the waste-containing glasses produced by these processes, particularly in relation to their expected long-term behavior as affected by thermal and radiation-induced alteration of the glass. A compilation of characterization data is presented for low-melting (less than 1100/sup 0/C) borosilicate waste glasses, which may contain 33 wt percent or more waste and which can be melted in situ in their stainless steel storage canisters. The data demonstrate that these low-melting glasses are relatively unaffected by the radiation doses to which they are exposed. The data also show that thermally induced devitrification and phase separation can be minimized by proper selection of glass composition.
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