Development of vitrification technology in India for immobilization of high-level radioactive liquid waste

In Indian Nuclear Energy Program spent nuclear fuel (SNF) is considered as a vital resource material. The strategy of reprocessing of SNF has been adopted to recover and recycle uranium and plutonium for optimum resource utilization. In this process, high-level radioactive liquid waste (HLW) is generated as the first cycle raffinate and contains about 99% of the radioactivity generated in the entire nuclear fuel cycle. This waste is characterized by the presence of nitric acid (1-4 M), corrosion products (Ni, Cr, Fe), nitrate salts of Na, Al, etc., dissolved organics and their degradation products. The radioactivity content varies from 8-100 Ci/1 depending upon the type of fuel, fuel burn-up and off-reactor cooling. A three-step strategy has been adopted in India for management of HLW. This involves: (i) immobilization of waste in inert stable matrix, (ii) interim storage of the immobilized waste product under cooling and surveillance and (iii) disposal in deep geological formations. Borosilicate glass matrix has been selected for immobilization of HLW owing to its stability over long period of time and the technologies based on glass matrix are able to meet the other demands of the vitrification process such as amenability for remote control processing in shielded hot-cells, ability to retain problematic species present in the waste which are likely to create phase separation (sulphates, phosphates) or release due to volatilization (radionuclides of Ru, Cs etc) and minimum secondary waste generation