PARTITIONING TECHNOLOGY A Developmental Strategy for Sustainable Radioactive Waste Management

Transcription

1 Nuclear Recycle Group PARTITIONING TECHNOLOGY A Developmental Strategy for Sustainable Radioactive Waste Management SMITHA MANOHAR, NRG, BARC, INDIA Technical Meeting Towards Nuclear Energy System Sustainability Waste Management & Environmental Stressors IAEA, Vienna, May 2016

2 Indian Nuclear Fuel Cycle : Today U (natural) Reprocessing of carbide fuel UOX fuel PHWR Non α waste (L&IL) PUREX Heavy metal α Bearing HLW 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 Reprocessing Facility, Tarapur Vitrification Facility, Trombay Vitrified Waste Storage Facility, Tarapur

3 Present Management of High Level Waste: 3-Stage Programme Canister Welding Shielded Transport Cask HLW Glass Pouring Deep Disposal Air cooled Storage HLW from Reprocessing Deep Geological Repository

4 Joule Heated Ceramic Melter, WIP, Tarapur, India

5 Cold Cruicible - Advanced Melter Glass under melting Cold Crucible Demonstration Facility (WIP, Trombay, India) Glass under pouring

6 Vitrified Waste Product Storage Facility,Tarapur Objectives: a.to allow dissipation of decay heat b.continuous surveillance of VWP Salient Features: 1760 locations for storage of VWP overpacks Peak radioactive inventory when vault is full 1.2 billion Curies Natural draft air cooling induced by stack MOC SS304L Heat removal capacity 3 MW Elaborate temperature measurements for air, overpack surface, concrete, air borne activity 24 Ø x 650 HT

7 Partitioning of minor actinides from HLLW Partitioning of minor actinides from HLLW leads to reduction of radio toxicity of the waste addressing: Long term concerns for repositories, simplified repository design increase in public acceptance Such a strategy also favours recovery of useful fission products Minor actinides so separated can be transmuted in accelerator driven sub critical systems (ADSS) or fast breeder reactors (FBRs) Special tailor made matrices for separated radionuclide Alternate development Demonstration of partitioning process: consequence of industrial scale availability of novel solvents

8 Engineering Scale Test Facility Partitioning of high level liquid waste Setting up of an engineering scale test facility is perceived as a first technological milestone towards adoption of the Partitioning process on an industrial scale. Multi tier R&D efforts on laboratory scale followed by bench scale and inactive engineering scale have culminated into a flow sheet for adoption with actual waste. Sustained developmental efforts have resulted in the setting up and commissioning of Actinide Separation Demonstration Facility, Tarapur Solvent Extraction Facility at Trombay

9 Operational Engineering Scale Partitioning Facilities in India Actinide Separation Demonstration Facility : Tarapur Operational since Nov & being operated on demonstration basis Three cycles operated simultaneously at Processing rate ~ 30L/hr Demonstrated actinide partitioning from HLW along with An/Ln separation All secondary waste streams suitably addressed Solvent Extraction Facility, Trombay Operational since January 2015 & is under continuous operation Continuously pre-treats high salt loaded legacy waste at Trombay leading to separation of active constituents from bulk of the inactive constituents Recovery of Cs-137 and its conversion to caesium pencil Spin Off Strontium-90 recovery for societal benefits Spin Off

10 Solvent Systems Deployed ASDF Tarapur Cycle 1 U Separation 30 % TBP in n-dodecane Cycle 2 An + RE separation 0.2 M TEHDGA in 30% IDA in dodecane Cycle 3 An/Ln Separation 0.2 M D2EHPA in n-dodecane (TALSPEAK) Solvent Extraction Facility, Trombay Cycle 1 U Separation 30 % TBP in n-dodecane Cycle 2 Cs Separation 0.03 M Calyx Crown (CC6) in 50% IDA in dodecane Cycle 3 Sr, Ce and An Separation 0.4 M TEHDGA in 10% IDA in dodecane 10

11 Development of a Facility for partitioning of HLW Solvent system Characterization Extraction properties Physical properties Bulk availability of solvent Cell worthy contactor Degradation behaviour Performance Evaluation (Inactive Engineering Studies) Conv. Engg inputs (radiochemical) Solvent effectiveness Contactor effectiveness Metering effectiveness Phase monitoring Terminal Stream management On-site constraints Engineering Flow sheet Equipment sizing/layout Active Facility Design

12 Partitioning Processes: Engineering considerations Process based on Conventional SX process for Minimizing losses Maximizing loading leading to higher concentration of product Equipment selection based on Relatively small concentration of radionuclides requirement of larger no of stages Handling HLW No maintainable/moving parts Redundancy Retrofitting in already available cells Limited head room availability Management of secondary streams Uranium Product : reprocessing plant for further processing Other secondary Streams (Cs-137- Societal Benefits) Spent Solvent : Storage, treatment & Management

13 The Actinide Separation Demonstration Facility, Tarapur & Solvent Extraction Facility, Trombay Throughput to match vitrification process Multi step process based on solvent extraction All secondary streams suitably addressed Facilities has been designed as per the conventional basis of design of radiochemical plant Spent Solvent Treatment Facility Actinide Partitioning Demonstration Facility Overall Layout of the integrated Facility Actinide Partitioning Demonstration Facility Solvent Extraction Facility, Trombay

14 Steps for production of Cs-137 Glass pencil Blood Irradiator Separation of Cs Solution from HLW Na- Borosilicate Glass Vitrification of Cs Solution Shielded cask for transportion of Cs pencil Welding of lid on Cs glass pencil Production of Cs glass pencil

15 Overall Block Diagram for ASDF/SSMF, SSSF, Tarapur 30% TBP TEHDGA D2EHPA HLW U Removal Cycle U Lean HLW MA Separation Cycle Ac/Ln Product Ac/Ln Group Separation Cycle Ac. Product Storage Rare earth solution for vitrification/cementation U Product to ROP MA Lean HLW U Lean HLW for Vitrification Recovered dodecane casking Spent Solvent from PREFRE Storage Spent Solvent Management Facility Aq. After Alkaline Hydrolysis & Carbonate solution for immobilization Novel solvent casking

16 Partitioning flowsheet : Addressing High Salt Bearing HLW (Legacy Waste) Basic process : Decontamination of HLW by Solvent Extraction HLW Gross : 6 mci/l Gross : 8.77 Ci/L Solvent Extraction 1 st Cycle : U 2 nd Cycle : Cs 3 rd cycle : Ce, Sr, An Raffinate (ILW, Short Lived) To Tk4 Gross : 5-7 x 10-4 mci/l after neutralization Gross : mci/l Spent Solvents Cs Sr, Ce, An Solvent Washings Washed Solvents Evaporation & vitrification U Evaporation & vitrification Cementation Recycle/ Treatment Conversion to Cs Pencil U Solution Recycle VWP (HL) CWP (L&IL)

17 Commissioning of the Partitioning Facilities Performance of the cycles Metering of feed & solvent Stable operation of mixer-settlers with Detection & maintenance of interphase Mass transfer performance check (extraction & stripping of Markers) instrumentation parameters mixer-settlers & density pots Tanks aq./org. detection and separation provisions in the various equipments Density pots Rework tank mixer-settlers installed at site along with allied piping Service area for the facilities

18 Process Performance

19 Actinide Separation Demonstration Facility (ASDF), Tarapur ASDF which is based on indigenous efforts, inaugurated by Hon. President of India on November 15, Has treated 13 m 3 of High Level Liquid Waste. (HLLW ) till date Performance Cycle % removal of residual U& Pu Cycle % removal of alpha activity Cycle % An/Ln separation Incell view of ASDF With the successful commissioning of this facility, India has taken a lead role towards reducing the long term concerns of HLLW, an endeavor pursued globally.

20 Performance of Solvent Extraction Facility, Trombay Salient Feature of Operation: HHLW Treated: 34,000 lit in 1 year 481 Kg of U recovered and transferred to PP Ci of Cs-137 recovered, concentrated and vitrified to produce Cs glass. Sr, Ln and An separated, concentrated and vitrified Raffinate activity (β-γ) ~ 2 mci/lit and (α) ~ 1 x 10-4 mci/lit. Six nos of VWP canisters produced resulting in loading of 5500 lit of HLW per canister. In cell view of solvent extraction system DF (β-γ) : 3500 DF (α) : > nos of Cs pencil produced and deployed for Blood irradiation

21 Handing over of Cs glass pencils to BRIT for their use as source in blood irradiator India is the first country to recover large quantities of Cs-137 (80000 Ci) from HLW and deploy it in vitrified glass form for irradiation purpose (50 nos pencils)

22 ASDF Process Performances Ist cycle performance Actual HLW (11.7 Ci/L β) 99.65%wrt U % wrt α (α :2 X 10-1 mci/l) Actual HLW(18.3 Ci/L β)* 98.35%wrt U 99.81% wrt α (α :3.60 X 10-1 mci/l) Actual HLW (54.1Ci/L β)* 98.4%wrt U 99.81% wrt α (α :1.28 mci/l) 2 nd cycle performance 3 rd cycle performance 98.5% wrt α (91.67% wrt RE) 86.4 % wrt α (α :4.9 X 10-2 mci/l) 97.66% wrt α (3.01 X 10-2 mci/l) * Run to test extent of alpha removal 22

23 Partitioning of HLW : A Changing Scenario 23

24 Partitioning of HLW : A changing Scenario Process flow sheets have been demonstrated for separation of the following Minor Actinides Fission products (Cs & Sr) Partitioning primarily aimed at : Transmutation of Minor actinides and some of the long live FPs Resource utilisation of FP s like Cs & Sr As a transition step can we incur benefits in the short time frame by vitrification of separated radionuclides (A preliminary study carried out ) Vitrification of heat generating Cs & Sr in borosilicate glass followed by vitrification of Minor actinides+ Rare Earths (Scenario 1) Vitrification of heat generating Cs & Sr in borosilicate glass followed by vitrification of separated Mas and cementation of RE stream (Scenario 2)

25 Brief comparison of management techniques (Trombay legacy HLW) HLW Loading Expected No of canisters for present inventory Direct Vitrification 250 Lit/canister (due to sulfate & presence of inactive salts) Solvent extraction based Pre-treatment 5500 Lit/canister Processing rate m 3 per year m 3 per year Fission product partitioning Not feasible Separation of Cs & Sr feasible, usable for societal applications This pre-treatment step which partitions the waste thereby separating active components from bulk inactive salts is unique in nature being the first kind in World.

26 HLW Management : Partitioning Scenarios (a Study) 1. Direct Vitrification 2. HLW partitioning Scenario 1 3. HLW partitioning Scenario 2 Alpha Glass for GDF 50 Kg/T 26 Kg/T (An & Ln - 8% Ln loading limit) 7 Kg/T (5 year cooled fuel) Cs-Sr Glass for SSDF - 20 Kg/T 20 Kg/T Lean FP Glass for NSDF - 22 Kg/T 22 Kg/T Cement Product for NSDF - - Short lived REs (?) Basis of Calculation Alpha loading based on α/gm: He accumulation limit. In case of Transmutation with multiple recycling this will reduce to substantial low value. Heat bearing βγ loading based on the centreline temperature limitation of glass (500 o C), the calculated maximum heat load for canister size of 100 kg is 2000 W. This will corresponds to Cs-Sr loading of ~4.5%. Focus Areas Separation & Management of long lived FP Tc 99 & I 129 needs to be addressed Disposal scenario for Cs to include long lived Cs 135 (based on its radio toxicity contribution)

27 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

28 Acknowledgements This work detailed out is a consolidation of efforts undertaken primarily by Nuclear Recycle Group and Nuclear Recycle Board. Acknowledgements are due to various NRG/BARC laboratories where the lab. studies & analysis works have been carried out. The support of the Nuclear Recycle Board in erection, installation & commissioning of the ASDF facility is deeply acknowledged. Acknowledgements are also due to the various safety committees including DSRC and BSC for working along with us to ensure safety of ASDF facility 28

29 Thank You For Your Kind Attention! "Not all of us can do great things. But we can do small things with great love."...mother Teresa