Nuclear Fuel Cycle Indian Scenario
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1 Nuclear Recycle Group Nuclear Fuel Cycle Indian Scenario KAILASH AGARWAL Associate Director, NUCLEAR RECYCLE GROUP BHABHA ATOMIC RESEARCH CENTRE DEPARTMENT OF ATOMIC ENERGY MUMBAI, INDIA Advanced Fuel Cycle to Improve the Sustainability of Nuclear Power through the Minimisation of High Level Waste International Atomic Energy Agency, Vienna, Austria October 17-19, 2017
2 Outline Indian nuclear fuel cycle Reprocessing Block Diagram Low and Intermediate level liquid waste management High level liquid waste management Partitioning new path way Wealth from Waste Evolving Fuel Cycle 2
3 Fresh unraium fuel (as oxide) Reactor India s Option: Closed Fuel Cycle valuable materials 96% + U: 94-96% Pu: ~1% SPENT FUEL + Wastes 4% FP: 3-4% MA: ~0.1% Radiotoxicity of wastes after 1000 years Reprocessing Resource utilisation recovery of fissile material for energy source and optimal utilisation of resources Recycling Waste burden minimisation (long lived radio-nuclides) Waste Volume minimisation Conditioning Isolation of radioactivity from human environment 3
4 Indian Nuclear Fuel Cycle: Today Reprocessing Facility, Tarapur Reprocessing an important activity for success of Indian Nuclear Power Programme U (Natural) Reprocessing of Fast Reactor Spent Fuel UOX Fuel PHWR Non α Waste (L&IL) Reprocessing (PUREX) α Bearing HLW Heavy Metal Mixed Carbide Fuel for Fast breeder Test Reactor Mixed Oxide Fuel for Prototype Fast Breeder Reactor Volume Reduction Near Surface Disposal Vitrification Interim Storage Repository
5 Head End Aqueous Reprocessing Block Diagram Fuel Receiving Fuel Disassembly Shearing Receiving Fuel Dissolution Feed Preparation Accountability Separation Process Conversion 1 St Cycle Solvent Extraction 2 nd Cycle Solvent Extraction 3 rd Cycle Solvent Extraction 4 th Cycle Solvent Extraction 5 th Cycle Solvent Extraction Special product U/Pu/Np Conversion U Conversion Support System Cell ventilation Vessel Off Gas System Dissolver Off Gas Cold Chemical Make-up Nitric Acid Recovery Solvent Recovery System Process Control & Accountability Robotics and In Cell Maintenance HLW & ILW System Solid Waste LLLW System 5 Off-gas Treatment Recycle & Feed System Control Remote Maintenance Waste Treatment
6 Low level liquid wastes - Deployment of New Technology for minimisation of discharges Zeolite based adsorption- Ion Exchange Pilot Plant Reverse Osmosis Pilot Plant Nano-sorbent for Tc removal Pilot scale plant of 100 m 3 /day Capacity is under operation. High capacity plant is under construction Challenge of removal of traces long lived fission products i.e. Tc 99 is successfully overcome. Ion exchange and membrane based separation process for low level liquid waste for Near Zero Discharge 6
7 Intermediate level liquid wastes - Treatment to eliminate ILW ILLW Treatment HLLW Low Volume High concentration of activity LLLW High Volume Negligible amount of activity Treatment Options 7 Evaporation Ion Exchange
8 HLLW Management Three step process Canister Welding Shielded Transport Cask Glass Pouring HLW Deep Disposal Air cooled Storage HLLW from Reprocessing Geological Disposal Repository 8
9 Vitrification of High Level Liquid Waste Evolution of Melter Technologies Induction Heated Metallic Melter, WIP, Trombay Capable of handling variation in waste characteristics Ease of operation, handling & decommissioning Joule Heated Ceramic Melter, AVS, Tarapur Continuous operation High throughput Better life Adopted in WIP, K and INRP Cold Crucible Induction Melter, Trombay Higher temperatures feasible Suitable for future matrices Expected long melter life Matured experience of more than 2 decades IHMM and JHCM are under regular operations CCIM under demonstration trials India is among the six countries who have mastered the vitrification technology 9
10 Vitrified Waste Product Storage Facility, Tarapur Objectives: a. To allow dissipation of decay heat b. Continuous surveillance of VWP Salient Features: Air Draft Cooling provided Elaborate temperature measurements for air, over pack surface, concrete, air borne activity 10
11 Partitioning New path for HLW management Separation of fission products Reduction in heat load Application for societal benefits SX, WIP, Trombay for recovery of Cs from HLLW Separation of minor actinides Reduction in long tem radio-toxicity Amenable for transmutation 11
12 Relative Radiotoxicity Reduction of Radiotoxicity Waste after reprocessing with actinide partitioning Waste after reprocessing without partitioning Radiotoxicity of Uranium Ore Time (Years)
13 Wealth from Nuclear Waste Recovery of 137 Cs, 90 Sr, 90 Y, 106 Ru for health care from Radioactive Waste 137 Cesium Blood Irradiator Cs recovery from acidic and alkaline waste Deployment of 137 Cs for blood irradiators Cs in borosilicate glass matrix - nondispersive form Hot Cell Facility for Cesium Recovery Cs-CALIX Crown 6 (Acidic stream) Resorcinol formaldehyde resin (Alkaline stream) Cesium pencil cage 90 Strontium Milking 90 Y for radiopharmaceutical applications Removal of Sr from High Level Liquid Waste based on multi step process deploying novel solvents on supported liquid membrane 106 Ruthenium Brachytherapy for the treatment of eye cancer choroidal melanoma and retinoblastom [o] SLM Generator Sr-DTBDCH18C6 13
14 Cesium 137 as Source of Irradition Characteristics Co-60 Cs-137 Half Life 5.27 Years Specific Activity Ci/g 30 Years 5-15 Ci/g for glass matrix Recovery of Cs-137 from HLLW large quantity of recovery of Cs-137 from HLLW using indigenously developed extractant Vitrified form of Cs Serves as a better radiation source with enhanced chemical durability and longer service life 102 nos of Cs glass pencils produced and deployed for Blood irradiation 14
15 Stronium 90 Radiopharmaceutical application Milking of Y-90 from Sr-90 Two step process: Recovery of pharmaceutical grade Sr-90 from HLW of desire concentration at Waste management facility using novel extractants Milking of Y-90 using supported liquid generator (SLM) at Medical facility / Laboratory Sr-DTBDCH18C6 Novel extractant for Sr recovery Supported Liquid Membrane for milking of Y-90 from Sr-90 Y-90 potential use for therapeutic applications 15
16 ~1000 meters Geological Disposal Multi-barrier Concept VWP Canister cap Canister Overpack Backfill Engineered Barrier Geological media ( granitic host rock) Natural Barrier 16
17 Evolving Fuel Cycle U (Natural) PHWR U ( ~ 1.2% Enriched) Disposal U (enriched) MOX Fuel LWR U (depleted) Recovery of FPs Pu MOX Fuel FR PUREX MA partitioning FPs Pu Heterogeneous Recycling Irradiated Pins (high specific activity, short life) Am, Np Pu Cm Storage ~ 100 years Pu
18 Thank You Volume of High Level Vitrified waste generated for power consumption of an average family for entire life Volume of waste if actinide is also separated from HLW Weight of vitrified waste glass: 125 gm Source: * (2013) # (2013) Per capita consumption of electricity: India 957 KW-hr * World Average 3104 KW-hr # America more than KW-hr #
19 Waste Volume Impact of Partitioning on Waste Volume Vitrified Waste for Disposal Nearly obviates the requirement of geological disposal facility Without Partitioning With Partitioning With Partitioning & Transmutation Bhabha Atomic Research Centre 11 th March 2013
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