Development and adoption of low sodium glass frit for vitrification of high level radioactive liquid waste at Tarapur

High level Liquid Waste (HLW) is generated during the reprocessing of spent nuclear fuel which is used to recover uranium and plutonium. More than 99% of the fission product activity generated during the burning of nuclear fuel in the reactor is present in HLW. For the efficient management of HLW by vitrification, sodium borosilicate glass has been adopted worldwide. Sodium oxide acts as modifier in glass matrix and variation in its concentration may vary the properties of the glass and hence the melter parameters. The HLW presently used for vitrification has higher concentration of sodium. As the composition of the base glass is fixed the concentration of Na in the HLW is one of the limiting factors for the waste loading for the vitrification process. Present article gives a brief account of the formulation of a base glass frit with lower sodium content and the feedback after implementing in the vitrification plant. (author)