P.Chellapandi, P.Puthiyavinayagam, T.Jeyakumar S.Chetal and Baldev Raj Indira Gandhi Centre for Atomic Research - 603102

IAEA-Technical Meeting on ‘Design, Manufacturing and Irradiation Behavior of Fast Reactor Fuels’ 30 May-3 June 2011, IPPE, Russia Scope of Presentation

 Nuclear Power & FBR Programme in  Economic advantages of high burnups  Int. experience on achieving high burnup  Roadmap of enhancing the burnup  Experience with carbide & oxide fuels  Highlights of R&D  Future Plans India’s Nuclear Roadmap

70000 • PHWRs from indigenous Uranium Nuclear Power Capacity • PHWRs from imported Uranium 60000 Projection (in MWe) • Imported LWR to the max. extent of 40 GW(e) 50000 • PHWRs from spent enriched U from LWRs 40000 (undersafeguard) 30000 • FBRs from reprocessed Pu and U from PHWR 20000 • FBRs from reprocessed Pu and U from LWR (undersafeguard) 10000 • U-233- Thermal / Fast Reactors 0 2010 2012 2017 2022 2032

• India has indigenous nuclear power program (4780 MW out of 20 reactors) and expects to have 20,000 MWe nuclear capacity on line by 2020 and 63,000 MWe by 2032. • Now, foreign technology and fuel are expected to boost India's nuclear power plans considerably. All plants will have high indigenous engineering content. • India has a vision of becoming a world leader in nuclear technology due to its expertise in fast reactors and thorium fuel cycle. FBR Programme in India

• Indigenous Design & Construction Future FBR • Comprehensiveness in development of • 1000 MWe • Pool Type Design, R&D and Construction • Metallic fuel • High Emphasis on Scientific Breakthroughs • Serial constr. • Indegenous • Synthesis of Operating Experiences • Beyond 2025 • Synthesis of Emerging Concepts (Ex.GENIV) • Focus on National & International Weight in t No. Component PFBR CFBR CFBR 07 08 01 Main vessel 134 116 Collaborations 09 10 02 Core support 44.8 36 structure •500 MWe 03 Grid Plate 76 34 • Emphasis on high quality human resources 04 Core -- -- 06 •Pool Type 05 Inner vessel 61 55 •UO2-PuO2 Ø11950 06 Transfer Arm -- -- 07 Large Rotatable -- -- • Creation of environment for enabling 13725 Plug •3 twin units 08 SRP/Control Plug -- -- 04 05 11 innovations 09 IHX -- -- 10 Primary Pump -- -- •Indigenous 03 01 11 Anchor safety 110 95 •From 2023… 02 12 vessel • Marching towards world leadership by 2025 12 Thermal -- -- Insulation

PFBR FBTR •1250 MWt • 40 MWt •500 MWe • 13.5 MWe •Pool Type • Loop type •UO2-PuO2 • PuC – UC •Indigenous • Design: CEA •From 2013.. • Since 1985 Economic Befefits of High Burnup

High Burnup is desired to FR Burn-up (MWd/Kg) • Reduce Fuel Cycle Cost - to lower unit energy cost • Minimise Waste Generation

- less minor actinides and fission products • Reduce Man Rem Exposure

Cost Projection (%) EFR EPR Capital 71 55

O & M 19 17 cost cycle fuel Relative Fuel Cycle 10 20

For PFBR with 100 GWd/t (peak) Unit Energy Cost Rs. 3.22 Fuel Cycle Cost Rs. 0.73 Target - 200 GWd/t gradually Fuel Cycle Cost Variation enhanced to 250 GWd/ti Worldwide Experience on Burnup

Standard MOX fuel Experimental fuel Country or Burn-up Maximum group of No. of pins Type of fuel reached burn-up Main reactors countries irradiated pellets MWd/t MWd/t Oxide Western Phenix, PFR, Solid & 265 000 135 000 200 000 Europe KNK-II annular Unites States 64 000 130 000 200 000 FFTF Leading pins Japan 50 000 100 000 120 000 Joyo Solid CIS 13 000 135 000 240 000 BOR-60 Vibro-pac 1 800 100 000 - BN-350 Solid &annular Solid & 1 500 100 000 - BN-600 annular Country Reactor Burnup (% h.A) BR-10 5 Russia BOR-60 10 Carbide Germany KNK-II - USA EBR-II 20.7 India FBTR 17

200 GWd/t is realisable. Burnup on whole core basis is important Issues Related to High Burnup Limitation to achieve high burnup comes from the current generation core structural materials owing to its excessive deformation and degradation due to irradiation rather than from the fuel. Thus, the capability to achieve high burn up is limited by the capability to accommodate higher neutron doses. Governing Parameters Multi Disciplinary Expertise  In-Pile Behaviour of Fuel • Reactor physics Element & Subassembly • Fuel Properties - FCMI • Material Development - FCCI • Core engineering  Degradation in PIE Mechanical Properties Modelling  Fuel Cycle Aspects • Reprocessing and Waste Management

Integrated and Synergistic Approach is Essential Experience with Carbide Fuel FBTR and Its Fuel Cycle • FBTR is in operation since 1985: completed more than 25 years of operation • It uses a unique U,Pu mixed carbide fuel with high Pu content (Mk-I:70%, Mk-II: 55 %) • Over 1000 fuel pins have reached an international record in burn-up (165 GWd/t) without any failure in the core • PFBR test fuel SA attained allowable burnup of 112 GWd/t. The performance of FBTR has been excellent in recent years, with plant availability during each campaign exceeding 80 %. • Since no direct data was available, physical, chemical and irradiation behaviour of fuel were obtained through extensive out-of-pile tests and with the gradual increase of burnup in the reactor. FBTR would operate further around 10 EFPY for irradiation tests on metallic fuel, sol-gel fuel and advanced clad materials FBTR: Mk-I Carbide Fuel Subassembly

SALIENT DATA Fuel : (70%Pu-30%U)C Pellet dia : 4.18 mm Pin OD/ID : 5.1 / 4.36 mm Peak Linear Power : 400 W/cm Active Core Height : 320 mm Clad Material : 20%CW 316 Wrapper Material : 20%CW316L No. of Pins : 61 Width Across Flats : 49.8 mm Height : 1661.5 mm Mark-I SA Performance

Active core Active core Parameters Active Core top bottom Middle Coolant Temp (Av.) 340 405 515 º C Coolant Pressure 0.34 0.25 0.17 MPa Neutron Flux 1.25X 1015 1.71X 1015 0.98X 1015 n/cm2/sec

Temp & DPA variation over SA length Burnup Peak Fluence 600 80 DPA 2 70 (GWd/t) (n/cm ) 500 60 400 100 46.8 0.783E+23 50 150 70.2 1.17E+23 300 40 30

200 Fluence. (DPA)

154 72.5 1.21E+23 20 Temperature ( C ) . Temperature ( )C 100 160 74.9 1.25E+23 10 0 0 0 50 100 150 200 250 300 350 165 77.2 1.28E+23 Axial length of SA (mm) Micrographs of Fuel Pin Cross Section

25GWd/t 50GWd/t 100GWd/t 155GWd/t

Micrographs of fuel pin cross section at the centre of fuel column

 Radial cracking at low burn-ups in free swelling regime 155GWd/t

 Progressive reduction in fuel clad gap with burn-up

 Cracking pattern changes from radial to circumferential cracking with closure of fuel clad gap

 Complete closure of fuel-clad gap along the entire fuel column at 155 GWd/t burnup

 Porosity free dense zone at the outer rim of the fuel

 Swelling of fuel accommodated by porosities & clad swelling End of fuel column X –radiography & Neutron radiography

Plenum Fuel column Fuel column

Plenum X-radiographs N-radiographs

12

10 Max. FG release ~ 16 % FG Pressure ~ 20 bars 8

6 (155 GWd/t)

4

2 Average increase in increase stackAverage length(mm) Burn-up in GWd/t 0

0 50 100 150 200 Av. increase in stack length stack in increase Av. Burn-up(GWd/t) 5 Fuel column elongation (X-ray)

Higher axial swelling in the restrained swelling phase & low fission gas release and plenum pressure Performance of Fuel Clad and Wrapper 20 % CW SS316 Burnup Max Fluence Peakdpa 155 GWd/t 1.2 x 1023 n/cm2 83

. • 155 GWd/t Burnup Fuel assembly and Fuel Pins

11.5%

(485 C) Progressive increase in ∆d ∆d /d %) dimensions of

clad & / V % V 3.5% wrapper with ∆ (430 C)

dpa

Diametral strain ( strain Diametral

Dimensional Changes in Wrapper & Clad Void Swelling of FBTR Clad & Wrapper Structural Material Properties: PIE Data

Cladding

Wrapper 81 dpa TEM studies 40 dpa 30 dpa

Virgin

100 100

nm

100 100

100 100

nm

V / / V % V

nm

500 500

∆

nm

Salient PIE Results of 100 GWd/t Fuel

 EXCELLENT PERFORMANCE OF FUEL / CLAD / HEXCAN  NO FUEL/CLAD GAP SEEN AT THE CENTRE OF FUEL COLUMN  AVERAGE INCREASE IN STACK LENGTH ~ 1.73%  FISSION GAS RELEASE ESTIMATED TO BE BETWEEN 3-14%  INTERNAL CLAD CARBURISATION NOT OBSERVED  MAXIMUM INCREASE IN CLAD DIAMETER- 1.6%  RESIDUAL DUCTILITY OF 3% OBSERVED ON THE CLAD TUBE  CLAD VOLUMETRIC SWELLING ESTIMATED TO BE 4.4% FUEL HAS REACHED 150 GWd/t BURN-UP WITHOUT FAILURE Swelling and Creep Strains in Clad

Burnup Total Clad Strain Clad Diametrical Increase (GWd/t) Predicted PIE Predicted PIE

50 0.2% 0.3% 0.01 mm 0.018 mm

100 1.9% 1.6% 0.09 mm 0.08 mm

150 5.7 % - 0.27 mm -

154 6.2% 5% 0.274 mm 0.27

160 6.7% - 0.32 mm -

165 7.2% - 0.34 mm -

Burnup Limited to 165 GMd/t based on residual ductility consideration (<10 %) Dilation Width Across Flat

SA Active Core Theoretical Normalised At Max. B.U Limit Length (mm) 154GWd/t 154 GWd/t 164.2 GWd/t 0 0.008 0.004 0.004 32 0.009 0.004 0.005 64 0.010 0.005 0.006 96 0.029 0.017 0.019 128 0.104 0.072 0.080 160 0.245 0.199 0.224 192 0.455 0.405 0.462 224 0.647 0.596 0.690 256 0.701 0.650 0.765 288 0.575 0.525 0.633 320 0.343 0.297 0.362

Burnup Limited to 165 GMd/t based Handling consideration (< 1 mm) Cladding Failure in 17th Campaign

Campaign started in Dec ‘10 with 48 fuel SA 27 MK-I, 13 MK-II & 8 MOX

In Feb 2011, at 18 MWt with TG feeding the grid, scram took place on DND signal provided for detection of fuel clad failure

• From DND signals and the ratio of Kr85 / Kr87 in cover gas samples, failed fuel identified to be one with a burn-up of >100 GWd/t. • From the DND signals during the flux tilting experiment with operation of control rod at 2 MWt power, the failed fuel was identified in the first instance itself, with only a single fuel handling • The failed fuel has been presently discharged for further investigation and PIE. Experience with Oxide Fuel FBTR: MOX SA for Future Hybrid Core

SALIENT DATA Fuel : (44%Pu-56%U)O 2 CDF Vs Burn-up of MOX fuel test pin Pellet OD/ID : 5.52 / 1.9 mm 0.5 Pin OD/ID : 6.6 / 5.7 mm 0.4 Peak Linear Power : 250 W/cm Active Core Height : 430 mm 0.3

Clad Material : 20%CW D9 CDF 0.2 Wrapper Material : 20%CWD9 / 20% CW316L 0.1 No. of Pins : 37 0 Width Across Flats : 49.8 mm 0 20 40 60 80 100 120 Height : 1661.5 mm Burn-up (GWd/t)

SA was discharged at 112 GWd/t (> design value of 100 GWd/t) PIE is in progress to get more information about fuel behaviour Road Map for Achieving High Burnup (Materials and Design) PFBR Core & Fuel Subassembly

Fuel SA - 181 Blanket SA - 120 Total SA - 1758

Salient Details

Fuel : (Pu-U)O2 Pellet OD/ID : 5.55/1.8 mm Pin OD/ID : 6.6/5.7 mm Peak Linear Power : 450 W/cm Active core height : 1000 mm Breeding Ratio : 1.05 Clad & Wrapper : 20 % CW D9 No.of Pins : 217 Width Across Flats : 131.3 Peak target Burnup : 100 GWd/t Peak neutron dose : 85 dpa Advanced Clad and Wrapper Materials Parameter Current Stage-1 Stage-2 Stage-3 Stage-4 Target Burnup 100 <150 150 200 200 GWd/t Fuel Oxide Oxide Oxide Oxide Metallic Clad material D9 IFAC-1 SS IFAC-1 9-18 Cr T91 (Indian fast SS ODS F-M steel reactor steels advanced core material) Wrapper D9 IFAC-1 SS T9 F-M T9 F-M T9 Material steel steel F-M steel Linear Power, 450 450 450 500 >500 W/cm

• 316 SS, limited to 50,000 MWd/t • D9 - 15Cr-15Ni-Ti limited to 100,000 MWd/t • IFAC1 - 14Cr-15Ni-Ti-Si-P limited to 150,000 MWd/t • 9-18 Cr-1Mo ODS steels upto 200,000 MWd/t Design Approach for Future FBRs

• Rationalisation of hot spot factors • Increase in residence time with excessive reactivity • Increasing the fuel handling cycle length • Increased fissile enrichment zones, • Optimum core restraint design for reduced bowing & dilation of wrapper • Optimization of inter-wrapper gap • Increasing the fission gas plenum • Increasing the pellet density and decreasing smeared density, • Optimum O/M ratio, • Optimum fuel pin diameter, • Optimum subassembly size, • Differential axial enrichment • Rationalization of over power margin Rationalisation of hot spot factors for fixing the Design Safety Limit & Allowable Linear Power

700 W/cm Linear Power to Melt

Margin for Uncertainties Through hot spot factors 518 W/cm Design Safety Limit

Over Power Margin 450 W/cm Allowable Linear Power

Rationalization has resulted in lowering of the fuel centre line hot spot temperature by about 12-15% and the clad midwall hot spot temperature by about 3-5%, in general. Safety margin is improved Benefits of Rationalisation of HSF

Linear Power = 370 W/cm

Old Hotspot temperature Rationalized Hotspot temperature 2400 Tmelt = 2123 K 2100 1800 1500 1200 900 600

Temperature(K) 300 0 Tinlet Tna Tco Tci Ts Tcen Validation through CFD Simulations

Thermal Hydraulics Analysis of 217 Fuel Pin Bundle

• Temperature distribution has been estimated considering axial as well as pin-to- pin variations in heat generation rate. Fuel pin Spacer wire Hexcan • CFD calculations have Structured mesh for 217 fuel been performed for pins with helical wires typical fuel and blanket SA in all the flow zones Sodium temperature variation at various cross sections along and hotspot factors fuel SA height (every 200 mm in have been estimated. active zone) Improved Oxide Core For High Burnup

Parameter Present Future Cycle length 180 270 (full power days) Fuel enrichment (%) 21/28 23/31 Fraction of core 1/3 1/4 discharge per cycle Peak fuel burnup 100,000 200,000 (MWd/t) R&D Highlights Swelling studies in IFAC-1 ---- Effect of P and Si

5.0 Si 0.9 100 dpa Si 0.75

2.5

Swelling (%) Swelling

0.0 700 750 800 850 900 950 Temperature (K)

Peak swelling 3.9% and 2.5% for 0.026%P and 0.04%P respectively. P effective in suppressing swelling at temperature above 850 K due to needle like phosphide precipitates. P in solution is effective for swelling at lower temperatures

30 appm He pre-implanted + 5 MeV Ni++ ion irradiation; Damage rate: 7x10-3 dpa/s

Optimised composition: 0.26Ti, 0.75Si, 0.04P (IFAC-1) Weldability studies and irradiation in FBTR in progress. IFAC1 - 14Cr-15Ni-Ti-Si-P targeted to 150,000 MWd/t

Optimisation of F-M Steel for Wrapper Application

9% Cr steels shows minimum shift in DBTT

Optimised composition (T9) Si 0.4 – 0.6, P <0.005, S< 0.005 Normalising at 970 – 1000 C Tempering at 740 -760 C IFAC 1 clad + P9 wrapper Key Issues in the Development of ODS Steel

Clad tubes with improved creep strength ODS alloy composition (Ferritic vs F-M) Processing route (upsetting/forging/extrusion) Particle size & distribution and vol. fraction of dispersoid Microstructural stability – thermal and irradiation Anisotropy and mechanical properties (strength/ductility, toughness) Weldability Fuel side compatibility Sodium compatibility Reprocessing compatibility Studies on ODS Alloys : Fe-0.3%Y2O3

6000 Depth (A) 5500 0 10000 20000 30000 40000 50000 5000 RT 450.0µ 1.0 A100 4500 A200 4000 30000 Al A300 3500 38000 Al A400 360.0µ 0.8 A500 3000 A600 5 MeV A700 2500

A720 2000Counts Fe Damage A760 1500 270.0µ 0.6 1000

500

0 180.0µ 0.4

90.0µ ~ 1.8 µm 0.2

thick damage Number of Implanted He/A Implanted of Number 39 40 41 42 43 44 45 46 0.0 0.0 layer A / profile Damage Normalised 2 ( Degrees) 0 10000 20000 30000 40000 50000 High Temp. XRD across bcc Depth (A) to fcc transition in Fe Dual beam irradiation : 5 MeV Fe + and Helium at 600°C

1 300 2 nm Rp = TEM Studies on stability of nano- 5h precipitates under irradiation and Substrate f 3 interactions with dilocations No dissolution at 25 dpa; Dual indentor nanomechanical studies to estimate yield Dissolution seen at 100 dpa strength & strain hardening exponent in irradiated sample Development of 9Cr-2W-0.1C-0.2Ti-0.35Y2O3 ODS Clad Tube Alloy powder, Plate-like, Characterisation varying sizes <300 m

{100} {110} {111}

400

300

200

100

0 frequency of the dispersoids the frequencyof 2.5 4.5 6.5 8.5 10.5 12.5 15 Size (nm) EBSD/ pole figures 2.5-4.5 4.5-6.5 6.5-8.5 8.5-10.5 10.5-12.5 12.5-15 NO TEXTURE Development of 9Cr-2W-0.1C-0.2Ti-0.35Y2O3 ODS Clad Tube Mechanical properties

Element C Cr W Ti Y2O3 Mn N O (Total) Spec. 0.11 - 0.13 8.8 - 9.2 1.9 - 2.1 0.19 - 0.22 0.32 - 0.35 0.04 max < 0.01 4.2 m 0.12 8.85 2.01 0.21 0.36 0.01 0.01 0.12 Clad tube

300 Test in progress o 1200 700 C 200 1000

800

600 90

80 400 70 60 YS (NFC) 200 UTS (NFC) 50

YS (IGCAR) Grade 91 Strength, MPa Strength, Applied stress, MPa stress, Applied 40 UTS (IGCAR) India ODS clad tube 0 YS (Japanese data) 9Cr-ODS Japan UTS (Japanese data) 30 Alloy D9 -200 200 400 600 800 1000 1200 100 1000 10000 Temperature, K Rupture life, hour

Clad-tubes with 6.6 mm O.D., 0.45 mm thick and 4.2 m length have been successfully produced Bulk Shield Reduction through Advanced Shielding Material

PFBR Core Radial Shields:

9 Rows (SS & B4C)

CFBR Core Radial Shields: 8 Rows (Ferroboron)

609 SS + 417 B4C 881 FERRO BORON SA • Ferroboron is used as a master Advantage of using ferroboron alloy in steel industry as an additive • Reduction of 1 row of SA for boron. • Reduction in no. of SA - 145 • Reduction in MV diameter by ~350 • Commercial ferroboron has 15-18 mm wt% boron • High temperature metallurgical interaction tests with 304L SS have • Available in form of lumps, granules shown good compatibility and powder • Low cost • Estimated cost redn - ~ 40 Crores • Bulk density: ~4 g/cm3 Ferro boron as the shielding material

Potential of ferroboron as shield material tested in neutron attenuation experiments in KAMINI  Metallographic characterization 1 of ferro-boron powder has shown boron as iron borides 0.1 Rh Equivalent flux

In Equivalent flux  Hot-sodium experiment has Neutron attenuation attenuation Neutron 0.01 shown that ferro boron does not 0 20 40 60 80 100 o EXPERIMENTS WITH Thickness (cm) x Density (g/cm3) react with liquid Na upto 650 C. VARYING B CONTENT IN PROGRESS Fast neutron flux attenuation in ferro boron slabs.  Metallurgical and chemical 1.00E+00 Mn Equivalent flux compatibility studies in progress 1.00E-01 Na Equivalent flux Gold Equivalent flux  High temperature clad-ferro boron 1.00E-02 (304lss/feb) interaction tests at 1.00E-03 high temperatures with varying Neutron attenuation attenuation Neutron 1.00E-04 0 20 40 60 80 100 120 time durations; estimation of clad

3 o Thickness(cm) x Density (g/cm ) penetration depth at 700 C in Thermal and epithermal neutron flux attenuation in Ferro progress Boron slabs Numerical Simulations

Time to reach full power Influence of fab.parameters on LPM

Cold gap - Pre irradiation Cold gap - post irradiation % LPM change for -2.5% variation in all parameters % LPM change for +2.5% variation in all parameters 25 120 20

100 15

80 10 5 60 0 Clad_ID % He Pellet_ID Pellet_OD Pellet O/M ratio 40 -5 Gap- microns Density -10

20 in Change (%) LPM -15 0 -20 120 80 40 10 -25 Axial Distance from top - mm Parameters affecting LPM

Irradiation expt in FBTR Key parameters to reach high Numerical Analysis & expt both linear power confirm that LHR can be taken to 450 W/ cm in few days Clad ID & Pellet OD

Analyses leading to cost benefits Allowable Manufacturing Deviations Pellet Defects Acceptability Acceptability of refused end plug weld

Rupture Life of Fuel pin end plug w elding,36 tubes simultanious testing

1600 1500

h 1400

- 1300 1200 1100 1000 900 800

700 Rupture Life(Hours) Rupture

Rupture Rupture life 600 500 400 300 200 100 0

Reductionfactor 0 1 2 3 4 No of weld repairs No of w eld repaires

Weld strength reduction factor

No of weld repairs The rupture lives measured for the various test cases, based on the test Analysis defines pellet depth for data and analysis, a design approach defect area. Improves in pellet is proposed for the inclusion in RCC- recovery without compromising MR:2007 edition. safety Status on Metallic Fuel Studies Metallic Fuel Pin Design Concepts Sodium Bonded Sodium Bonded Mechanical bonded Mechanical bonded U-15Pu (No Zr in U-Pu-Zr(6/10%) fuel) U-15Pu ( 2 grooves) U-15Pu (4 grooves) No liner Zr- 4 Liner Zr- 4 Liner Zr- 4 Liner 75 % smeared 75 % smeared 85 % smeared 75 % smeared density density density density Top Plenum Top Plenum Bottom plenum Bottom plenum CLAD Advantages: High BR, Low T, High linear rating, Inherent safety LINER Doubling time : 30 y for oxide , 12 y for metal and 8 ys for improved metal (without Zr)

FUE L Mechanical Bonded Fuel Pin Cross Section Metallic Fuel Development

Road Map Linear Power Pin Irradiation in - 450 W/ cm FBTR Clad

Subassembly - T91 Irradiation in FBTR Irradiation Capsule irradiation – 3 pins Full Core Metallic Fuel in FBTR SA irradiation – 37 pins Prototype scale -217 pins Target Burnup Metallic Fuel 320 MWe Design -150 GWd/ t

Metallic Fuel Expt Pin - Schematic Salient Highlights 1000 MWe Design Metallic Fuel Sodium Bonding Facilities

Sodium wire extrusion into PVC tube Sodium wire extruder Argon Glove Box for Sodium Handling

Dummy Fuel Pin

Developmental Facilities established in BARC and IGCAR to demonstrate the technology Pin welding fixture

Sodium Bonding Furnace with Vibrator Metallic Fuel Mechanical Bonding Facilities

Fuel Fabrication facility Glove Box Train arrangement Purification tower arrangement

Co-swaged fuel rod with clad / liner INJECTION-CAST, SWAGED & MACHINED URANIUM RODS (demonstrated at BARC Mumbai) Length = 160 mm, Diameter= 4.67±0.04 Facilities for R&D on Fuel Pin Behavior • FBTR for further about 10 EFPY • Radio Metallurgy Laboratory • Material Development Laboratory • PFBR from 2013… • JHR at Cadarache under CEA-DAE agreement and IGCAR is developing an innovative sodium loops for the irradiation tests of multiple specimens at high temperature in JHR. • 320 MWt Metallic Fast Reactor (MFR: to test the SA on 1:1 scale basis) from 2017…

A dedicated test facility for studying the oxide as well metallic fuel pin behaviour under rapid heating that would simulate various accident conditions leading to fuel melting Induction heating Epilogue: Success Mantra

• Well conceived roadmap for the development of fuels & structural materials and test facilities for enhancing burnup gradually to 200 GWd/t, subsequently 250 GMd/t • Experience from 400 r-y of FBRs worldwide including FBTR • Expertise developed on the essential domains such as material development including fuel, design, structural mechanics, core thermal hydraulics, numerical simulations of fuel behavior, PIE, manufacturing, testing and Evaluations • Availability of FBTR, PFBR and 320 MWt MFR over long period • Excellent coordination among various units in the Department of Atomic Energy, involved in design, manufacturing and R&D • National and International Collaborations INDIAN NUCLEAR PROGRAMME Towards sustainable energy