Overview of the Thorium Programme in India P.K. Vijayan, I.V. Dulera, P.D. Krishnani, K.K. Vaze, S. Basu and R.K. Sinha*, Bhabha Atomic Research Centre, Mumbai India Department of Atomic Energy, Mumbai, India* [email protected] iThEC13-International Thorium Energy Conference, Geneva, Switzerland, October 27-31, 2013 1 CONTENTS Introduction Relevance of thorium to long term Indian nuclear power programme Advantages and challenges of thorium fuel cycle Indian experience with thorium fuel cycle Mining, thorium fuel fabrication, Thorium bundle irradiation, reprocessing and Built research reactors based on U233 fuel Indian programme on thorium based reactors AHWR and Indian HTR Programme Concluding remarks 2 Introduction Thorium is three to four times more abundant than uranium in the earth’s crust Relatively easy to mine India has large reserves in the form of monazite sands along its southern and eastern coastal areas Relevance of thorium to long term Indian nuclear power programme 4 The goal of three stage Indian nuclear power programme is resource sustainability- Accordingly power generation in 3rd stage is predominantly dependent on thorium based fuel PHWR FBTR AHWR Th About 42000 GWe- 300 GWe-Year Year U fueled Nat. U Electricity Th PHWRs Dep. U Electricity >155000 Pu Fueled GWe-Year Fast Breeders Pu U233 Electricity U233 Fueled Reactors H2/ Transport Pu U233 fuel Power generation primarily by PHWR Expanding power programme Thorium utilisation for Building fissile inventory for stage 2 Building U233 inventory Sustainable power programme Stage 1 Stage 2 Stage 3 5 Current Status of Indian Three Stage Nuclear Power Programme 95 Globally Advanced 90 86 Globally Unique 84 84 90 91 90 89 85 79 Technology 80 75 72 75 69 70 World class Availability65 60 performance 55 50 1995- 1996- 1997- 1998- 1999- 2000- 2001- 2002- 2003- 2004- 2005- 96 97 98 99 00 01 02 03 04 05 06 Stage – I PHWRs Stage - III Stage - II • 18 – Operating Thorium Based Reactors Fast Breeder Reactors • 4x700 MWe - Under construction • 40 MWth FBTR - • 30 kWth KAMINI- Operating • Several others planned • 300 MWe AHWR: Pre- Operating since 1985 • Gestation period has licensing safety appraisal by Technology Objectives been reduced AERB completed, Site • POWER POTENTIAL realised selection in progress 10 GWe LWRs • 500 MWe PFBR- POWER POTENTIAL IS • 2 BWRs Operating Under Construction >95% VERY LARGE • 1 VVER– Achieved criticality complete • MSBRs – Being evaluated as (July, 2013) & commercial an option for large scale operation (Oct. 2013) • TOTAL POWER • 1 VVER- under deployment construction (~96%) POTENTIAL 42,000 GWe-year 6 Optimising the use of Domestic Nuclear Resources PFBR (500 MWe) 600 Reactor building Capital cost 69840 (Rs/kWe) 500 UEC (Rs/kWh) 3.22 Construction 7 years period 400 Project sanctioned in 2003 Further development 300 being pursued to reduce doubling time Installed capacity (GWe) capacity Installed 200 and UEC Power profile Growth with Pu-U FBRs 100 of PHWR Russia is the only programme other country with a larger FBR under 0 construction 1980 1995 2010 2025 2040 2055 2070/operation Year Results of a case study; assumptions 60000 te Uranium and short doubling time FBRs beyond 2021 Optimising the use of Domestic Nuclear Resources 600 Further growth 500 with thorium 400 300 Installed capacity (GWe) capacity Installed 200 Power profile Growth with Pu-U FBRs 100 of PHWR programme 0 1980 1995 2010 2025 2040 2055 2070 Year Results of a case study; assumptions 60000 te Uranium and short doubling time FBRs beyond 2021 Optimising the use of Domestic Nuclear Resources 75 Premature introduction of 600 thorium hampers the growth DetailedGrowth calculations with have shownPeaks at 50that thoriumPu-U FBRs can be deployed 36on GWe Further 500 a large scale about 3 decades only growth with afterBest effortintroduction use of of FBRs with thorium 400 shortPu with doubling Th in time 25 Installed Capacity (GWe) Capacity Installed PHWRs and MSRs 300 Installed capacity (GWe) capacity Installed PHWR(Nat U) 200 0 Growth with Power1980 profile1995 2010 2025 2040 2055 PuYear-U FBRs 100 of PHWR programme 0 1980 1995 2010 2025 2040 2055 2070 Year Results of a case study; assumptions 60000 te Uranium and short doubling time FBRs beyond 2021 A Road Map for Deployment of Thorium Based Reactors Premature deployment of thorium leads to sub-optimal use of indigenous energy resource. Necessary to build-up a significant level of fissile material before launching thorium cycle in a big way for the third stage Incorporation of thorium in the blankets of metallic fuelled fast breeder reactors – after significant FBR capacity built-up Full core and blanket thorium FBRs only after U-shortage felt Thorium based reactors expected to be deployed beyond 2070 AHWR – Thorium fuel cycle demonstrator by 2022 Surplus 233U formed in these FBRs could drive HTRs including MSBRs 10 Important attributes and advantages of thorium fuel cycle 11 ThO2 - Physical and Chemical properties Relatively inert. Does not oxidise unlike UO2, which oxidizes easily to U3O8 and UO3. Does not react with water. Higher thermal conductivity and lower co-efficient of thermal expansion compared to UO2. Melting point 3350 °C as against 2800 °C for UO2. Fission gas release rate one order of magnitude lower than that of UO2. Good radiation resistance and dimensional stability Failed UO ThO2-4%PuO2 Failed ThO2-4%PuO2 2 UO2-4%PuO2 Advantages of 233U-Thorium: Important Neutronic Characteristics U238 4 233 U235 U has Th232 an value (greater than 2.0) that remains Pu239 nearly constant over a wide 3 U233 239Pu 233 energy range, in thermal as well as U 235 239 epithermal regions, unlike U and Pu. 2 This facilitates achievement of high conversion ratios with thorium utilisation 1 238 in reactors operating in the 235U U 232Th thermal/epithermal spectrum. neutron neutron absorbed, Number Number neutrons of per : Number of neutrons released per thermal 0 0.1ev 1ev 10ev 100ev 1kev 10kev 100kev 1MeV neutron absorbed Incident neutron energy Fast Thermal Neutrons Neutrons 8 12 7 Cross section for capture of Capture cross section of 232 10 thermal neutrons in Th is 233U is much smaller than 6 typically 2.47 times that in 8 235U and 239Pu, the fission 5 238U. Thus thorium offers cross section being of the 4 greater competition to 6 same order implying lower capture of the neutrons and ) 3 c non-fissile absorption lower losses to structural / 4 f leading to higher isotopes. 2 and other parasitic 2 This favours the feasibility 1 materials leading to an 233 sections sections ( Ratio of fission Ratio of and fission capture cross of multiple recycling of U, improvement in conversion 0 0 232 238 233U 235U 239Pu as compared to plutonium. Th U of 232Th to 233U. The neutronic characteristics of 233U-Th lead to: • High potential for achieving near-breeding condition in thermal reactors • High potential for tolerating higher parasitic absorption (e.g. in light water and structural materials) Advantages of Thorium : Presence of 232U offers an Intrinsic Barrier to Proliferation 232U is formed via (n, 2n) reactions, from 232Th, 233Pa and 233U. The half-life of 232U is about 69 years. The daughter products (208Tl and 212Bi) of 232U are high energy gamma emitting isotopes. Due to presence of 232U in separated 233U, thorium offers good proliferation-resistant characteristics. Advantage of thorium over uranium becomes evident if thorium based fuels are used at high burnups 5.0 There is a growing global interest in thorium for various 4.0 LEU reasons such as Energy advantage 3.0 Relatively stable core Thorium-U235 fuel reactivity 2.0 Low minor actinide generation Proliferation 1.0 resistance U235 enrichment in initial fuel wt % wt fuel initial in enrichment U235 0.0 0 20 40 60 80 100 Discharge burnup GWd/te Performance potential of homogeneous mixture of fertile materials with 235U in a PHWR By virtue of being lower in the periodic table than uranium, the long-lived minor actinides resulting from burnup are in much lower quantity with the thorium cycle 15 Challenges of thorium fuel cycle 16 Challenges of thorium fuel cycle Melting point of ThO2 (3350 °C) is much higher than that of UO2 (2800°C). Much higher sintering temperature (>1700°C) is required to produce high density ThO2 and ThO2–based mixed oxide fuels. • Admixing of ‘sintering aid’ (CaO, MgO, Nb2O5, etc) is required for achieving the desired pellet density at lower temperature. ThO2 and ThO2 based mixed oxide fuels are relatively inert and do not dissolve easily in concentrated nitric acid. Addition of small quantities of HF in conc. HNO3 is essential which cause corrosion of stainless steel equipment and piping in reprocessing plants. The corrosion problem is mitigated with addition of aluminium nitrate. Requires long dissolution time in boiling THOREX solution [13M HNO3+0.05M HF+0.1M Al(NO3)3] for ThO2 based fuel The irradiated Th or Th–based fuels contain significant amount of 232U, having a half-life of 68.9 years and is associated with strong gamma 212 208 emitting daughter products, Bi and Tl with very short half-life. There is significant build-up of radiation dose with storage of spent Th – 233 based fuel or separated U, This necessitates remote and automated reprocessing and re-fabrication in heavily shielded hot cells, increasing the cost of fuel cycle activities. Challenges of thorium fuel cycle In the conversion chain of 232Th to 233U, 233Pa is formed as an intermediate, which has a relatively longer half-life (~27 days) as compared to 239Np (2.35 days) in the uranium fuel cycle thereby requiring longer cooling time of at least one year for completing the decay of 233Pa to 233U.