Abstract
Nuclear Power Programme in India is based on “closed fuel cycle”. Closed fuel cycle involves reprocessing and recycling of Spent Nuclear Fuel (SNF) coming out of nuclear reactors. During reprocessing, uranium &plutonium, constituting bulk of the SNF are separated and subsequently recycled. The remaining small portion constitutes high level radioactive waste containing most of the fission products and minor actinides. A three-step strategy involving immobilization, interim storage followed by ultimate disposal has been adopted in India for management of High Level Waste (HLW). Borosilicate glass matrix has been identified for immobilization of HLW owing to optimal waste loading, adequate leach resistance and long term stability of the product. Glass formulations are developed by suitable addition of modifiers to accommodate compositional variation in the waste. Detailed characterization studies are carried out to understand the structural modifications in the three dimensional network of incorporation of waste constituents followed by product characterization to ascertain the properties of the conditioned vitrified waste product (VWP). On the strength of these indigenous developments, Indian vitrification facilities are in operation at Trombay, Tarapur and Kalpakkam. Different types of melters like metallic, Joule Heated Ceramic Melter (JHCM) have been successfully deployed on industrial scale for vitrification of HLW. A Cold Crucible Induction Melter (CCIM) has also been developed. An interim storage facility is in operation for storage and surveillance of VWP. A site selection programme has been initiated to identify a few suitable geological domains for identifying a candidate disposal site for high level and long lived radioactive wastes. Presently granites have been studied extensively as a host rock & as a natural barrier. Long term evaluation of vitrified high level waste under geological conditions and its comparison with natural analogues (basaltic & obsidian) is being pursued in India.
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