P.Chellapandi, P.Puthiyavinayagam, T.Jeyakumar

S.Chetal and Baldev Raj

Indira Gandhi Centre for Atomic Research

Kalpakkam - 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 India

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

• 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.

0

10000

20000

30000

40000

50000

60000

70000

2010 2012 2017 2022 2032

Nuclear Power Capacity

Projection (in MWe)

• PHWRs from indigenous Uranium

• PHWRs from imported Uranium

• Imported LWR to the max. extent of 40 GW(e)

• PHWRs from spent enriched U from LWRs

(undersafeguard)

• FBRs from reprocessed Pu and U from PHWR

• FBRs from reprocessed Pu and U from LWR

(undersafeguard)

• U-233-Thorium Thermal / Fast Reactors

FBR Programme in India

• Indigenous Design & Construction

• Comprehensiveness in development of Design, R&D and Construction

• High Emphasis on Scientific Breakthroughs

• Synthesis of Operating Experiences

• Synthesis of Emerging Concepts (Ex.GENIV)

• Focus on National & International Collaborations

• Emphasis on high quality human resources

• Creation of environment for enabling innovations

• Marching towards world leadership by 2025

PFBR

•1250 MWt

•500 MWe

•Pool Type

•UO2-PuO2

•Indigenous

•From 2013..

----Thermal

Insulation

12

95110Anchor safety

vessel

11

----Core04

5561Inner vessel05

----Transfer Arm06

----Large Rotatable

Plug

07

----SRP/Control Plug08

----IHX09

----Primary Pump10

3644.8Core support

structure

02

3476Grid Plate03

01

No.

116134Main vessel

CFBRPFBR

Weight in tComponent

10 09

07 08

06

04 05

03

02 01

11

12

13

72

5

Ø11950

CFBR

•500 MWe

•Pool Type

•UO2-PuO2

•3 twin units

•Indigenous

•From 2023…

Future FBR

• 1000 MWe

• Pool Type

• Metallic fuel

• Serial constr.

• Indegenous

• Beyond 2025

FBTR • 40 MWt

• 13.5 MWe

• Loop type

• PuC – UC

• Design: CEA

• Since 1985

Economic Befefits of High Burnup

High Burnup is desired to

• Reduce Fuel Cycle Cost

- to lower unit energy cost

• Minimise Waste Generation

- less minor actinides and

fission products

• Reduce Man Rem Exposure

Fuel Cycle Cost Variation

Cost Projection (%) EFR EPR

Capital 71 55

O & M 19 17

Fuel Cycle 10 20

FR Burn-up (MWd/Kg)

Re

lati

ve

fu

el c

yc

le c

os

t

Target - 200 GWd/t gradually

enhanced to 250 GWd/ti

For PFBR with 100 GWd/t (peak)

Unit Energy Cost Rs. 3.22

Fuel Cycle Cost Rs. 0.73

Country or

group of

countries

Standard MOX fuel Experimental fuel

No. of pins

irradiated

Burn-up

reached

MWd/t

Maximum

burn-up

MWd/t

Main reactors Type of fuel

pellets

Western

Europe 265 000 135 000 200 000

Phenix, PFR,

KNK-II

Solid &

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

1 500 100 000 - BN-600 Solid &

annular

Country Reactor Burnup (% h.A)

Russia BR-10

BOR-60

5

10

Germany KNK-II -

USA EBR-II 20.7

India FBTR 17

Carbide

Oxide

200 GWd/t is realisable. Burnup on whole core basis is important

Worldwide Experience on Burnup

Issues Related to High Burnup

In-Pile Behaviour of Fuel

Element & Subassembly

- FCMI

- FCCI

Degradation in

Mechanical Properties

Fuel Cycle Aspects

Multi Disciplinary Expertise

• Reactor physics

• Fuel Properties

• Material Development

• Core engineering

PIE

Modelling

• Reprocessing and

Waste Management

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.

Integrated and Synergistic Approach is Essential

Governing Parameters

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 %.

FBTR would operate further around 10 EFPY for irradiation tests on

metallic fuel, sol-gel fuel and advanced clad materials

• 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: 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

Parameters Active core

bottom

Active core

Middle Active Core top

Coolant Temp (Av.)

º C 340 405 515

Coolant Pressure

MPa 0.34 0.25 0.17

Neutron Flux

n/cm2/sec 1.25X 1015 1.71X 1015 0.98X 1015

Burnup

(GWd/t) DPA

Peak Fluence

(n/cm2)

100 46.8 0.783E+23

150 70.2 1.17E+23

154 72.5 1.21E+23

160 74.9 1.25E+23

165 77.2 1.28E+23

Temp & DPA variation over SA length

0

100

200

300

400

500

600

0 50 100 150 200 250 300 350Axial length of SA (mm)

Tem

pera

ture

( C

) .

0

10

20

30

40

50

60

70

80

Flu

en

ce (

DP

A)

.

Radial cracking at low burn-ups in free swelling regime 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

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

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

End of fuel column

155GWd/t

Micrographs of Fuel Pin Cross Section

5

Max. FG release ~ 16 %

FG Pressure ~ 20 bars

(155 GWd/t)

Fuel column elongation (X-ray)

0

2

4

6

8

10

12

0 50 100 150 200

Burn-up(GWd/t)

Ave

rag

e in

crea

se i

n s

tack

len

gth

(mm

)

Burn-up in GWd/t

Av.

inc

rea

se in

sta

ck

le

ng

th

X –radiography & Neutron radiography

Plenum

Fuel column

Fuel column Plenum

Higher axial swelling in the restrained swelling phase & low fission gas release

and plenum pressure

X-radiographs N-radiographs

(485 C)

(430 C)

11.5%

3.5%

Performance of Fuel Clad and Wrapper

Burnup Max Fluence Peakdpa

155 GWd/t 1.2 x 1023 n/cm2 83

D

iam

etr

al s

train

(∆

d /

d %

)

•

155 GWd/t Burnup Fuel assembly and Fuel Pins

20 % CW SS316

Dimensional Changes in Wrapper & Clad

∆V

/ V

%

Void Swelling of FBTR Clad & Wrapper

Progressive

increase in

dimensions of

clad &

wrapper with

dpa

Structural Material Properties: PIE Data

Cladding

Wrapper

∆V

/ V

%

81 dpa

40 dpa

30 dpa

Virgin

TEM studies

100 n

m

100 n

m

100 n

m

500 n

m

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

Salient PIE Results of 100 GWd/t Fuel

Burnup

(GWd/t)

Total Clad Strain Clad Diametrical Increase

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 -

Swelling and Creep Strains in Clad

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

SA Active Core

Length (mm)

Theoretical Normalised At Max. B.U Limit

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

Dilation Width Across Flat

Burnup Limited to 165 GMd/t based Handling consideration (< 1 mm)

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

Cladding Failure in 17th Campaign

• 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)O2

Pellet OD/ID : 5.52 / 1.9 mm

Pin OD/ID : 6.6 / 5.7 mm

Peak Linear Power : 250 W/cm

Active Core Height : 430 mm

Clad Material : 20%CW D9

Wrapper Material : 20%CWD9 /

20% CW316L

No. of Pins : 37

Width Across Flats : 49.8 mm

Height : 1661.5 mm

CDF Vs Burn-up of MOX fuel test pin

0

0.1

0.2

0.3

0.4

0.5

0 20 40 60 80 100 120

Burn-up (GWd/t)

CD

F

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)

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

PFBR Core & Fuel Subassembly

Fuel SA - 181

Blanket SA - 120

Total SA - 1758

Salient Details

Advanced Clad and Wrapper Materials

Parameter Current Stage-1 Stage-2 Stage-3 Stage-4

Target Burnup

GWd/t

100 <150 150 200 200

Fuel Oxide Oxide Oxide Oxide Metallic

Clad material D9 IFAC-1 SS

(Indian fast

reactor

advanced

core material)

IFAC-1

SS

9-18 Cr

ODS

steels

T91

F-M steel

Wrapper

Material

D9 IFAC-1 SS T9 F-M

steel T9 F-M

steel

T9

F-M steel

Linear Power,

W/cm 450 450 450 500 >500

• 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

• 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

Design Approach for Future FBRs

Linear Power to Melt

Design Safety Limit

Allowable Linear Power

Over Power Margin

Margin for Uncertainties

Through hot spot factors

Rationalisation of hot spot factors for fixing the Design

Safety Limit & Allowable Linear Power

700 W/cm

518 W/cm

450 W/cm

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

Linear Power = 370 W/cm

0

300

600

900

1200

1500

1800

2100

2400

Tinlet Tna Tco Tci Ts Tcen

Te

mp

era

ture

(K

)

Old Hotspot temperature Rationalized Hotspot temperature

Tmelt = 2123 K

Benefits of Rationalisation of HSF

Thermal Hydraulics Analysis of 217 Fuel Pin Bundle

Validation through CFD Simulations

Fuel pin Spacer wire Hexcan

• Temperature

distribution has been

estimated considering

axial as well as pin-to-

pin variations in heat

generation rate.

•CFD calculations have

been performed for

typical fuel and blanket

SA in all the flow zones

and hotspot factors

have been estimated.

Structured mesh for 217 fuel

pins with helical wires

Sodium temperature variation

at various cross sections along

fuel SA height (every 200 mm in

active zone)

Parameter Present Future

Cycle length

(full power days)

180 270

Fuel enrichment (%) 21/28 23/31

Fraction of core

discharge per cycle

1/3 1/4

Peak fuel burnup

(MWd/t)

100,000 200,000

Improved Oxide Core For High Burnup

R&D Highlights

Swelling studies in IFAC-1 ---- Effect of P and Si

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

100 dpa

700 750 800 850 900 9500.0

2.5

5.0

Sw

ellin

g (

%)

Temperature (K)

Si 0.9 Si 0.75

Optimisation of F-M Steel for Wrapper Application

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

9% Cr steels shows

minimum shift in DBTT

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

Key Issues in the Development of ODS Steel

1 2

3

300

nm

Substrate

Rp =

5hf

39 40 41 42 43 44 45 46

0

500

1000

1500

2000

2500

3000

3500

4000

4500

5000

5500

6000

RT

A100

A200

A300

A400

A500

A600

A700

A720

A760

2 ( Degrees)

Co

un

ts

Studies on ODS Alloys : Fe-0.3%Y2O3

0 10000 20000 30000 40000 50000

0.0

90.0µ

180.0µ

270.0µ

360.0µ

450.0µ0 10000 20000 30000 40000 50000

0.0

0.2

0.4

0.6

0.8

1.0

No

rmalised

Dam

ag

e p

rofi

le / A

Depth (A)

Nu

mb

er

of

Imp

lan

ted

He/A

Depth (A)

30000 Al

38000 Al

5 MeV

Fe Damage

~ 1.8 µm

thick damage

layer

No dissolution at 25 dpa;

Dissolution seen at 100 dpa

High Temp. XRD across bcc to fcc transition in Fe

TEM Studies on stability of nano- precipitates under irradiation and interactions with dilocations

Dual beam irradiation : 5 MeV Fe + and Helium at 600°C

Dual indentor nanomechanical studies to estimate yield strength & strain hardening exponent in irradiated sample

Alloy powder, Plate-like,

varying sizes <300 m

{100} {110} {111}

NO TEXTURE 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

0

100

200

300

400

fre

qu

en

cy

of

the d

isp

ers

oid

s

Size (nm)2.5 4.5 6.5 8.5 10.5 12.5 15

Development of 9Cr-2W-0.1C-0.2Ti-0.35Y2O3 ODS Clad Tube

Characterisation

200 400 600 800 1000 1200-200

0

200

400

600

800

1000

1200

YS (NFC)

UTS (NFC)

YS (IGCAR)

UTS (IGCAR)

YS (Japanese data)

UTS (Japanese data)

Str

en

gth

, M

Pa

Temperature, K

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

Clad tube 0.12 8.85 2.01 0.21 0.36 0.01 0.01 0.12

100 1000 10000

30

40

50

60708090

200

300

Grade 91

India ODS clad tube

9Cr-ODS Japan

Alloy D9

Ap

pli

ed

str

ess,

MP

a

Rupture life, hour

700 oCTest in progress

Clad-tubes with 6.6 mm O.D., 0.45 mm thick and 4.2 m length have been successfully produced

Development of 9Cr-2W-0.1C-0.2Ti-0.35Y2O3 ODS Clad Tube

Mechanical properties

PFBR Core

Radial Shields:

9 Rows (SS & B4C)

609 SS + 417 B4C

Advantage of using ferroboron

• Reduction of 1 row of SA

• Reduction in no. of SA - 145

• Reduction in MV diameter by ~350

mm

• High temperature metallurgical

interaction tests with 304L SS have

shown good compatibility

• Low cost

• Estimated cost redn - ~ 40 Crores

CFBR Core

Radial Shields:

8 Rows (Ferroboron)

• Ferroboron is used as a master

alloy in steel industry as an additive

for boron.

• Commercial ferroboron has 15-18

wt% boron

• Available in form of lumps, granules

and powder

• Bulk density: ~4 g/cm3

Bulk Shield Reduction through Advanced Shielding Material

881 FERRO BORON SA

Thermal and epithermal neutron flux attenuation in Ferro

Boron slabs

1.00E-04

1.00E-03

1.00E-02

1.00E-01

1.00E+00

0 20 40 60 80 100 120

Thickness(cm) x Density (g/cm3)

Neu

tron

atten

uatio

n

Mn Equivalent flux

Na Equivalent flux

Gold Equivalent flux

Fast neutron flux attenuation in ferro boron slabs.

0.01

0.1

1

0 20 40 60 80 100

Thickness (cm) x Density (g/cm3)

Neu

tron

atten

uatio

n

Rh Equivalent flux

In Equivalent flux

Potential of ferroboron as shield material tested in neutron attenuation experiments in KAMINI

Ferro boron as the shielding material

Metallographic characterization

of ferro-boron powder has shown

boron as iron borides

Hot-sodium experiment has

shown that ferro boron does not

react with liquid Na upto 650 o C.

Metallurgical and chemical

compatibility studies in progress

High temperature clad-ferro boron

(304lss/feb) interaction tests at

high temperatures with varying

time durations; estimation of clad

penetration depth at 700oC in

progress

EXPERIMENTS WITH VARYING B CONTENT IN PROGRESS

Numerical Simulations

Analyses leading to cost benefits

0

20

40

60

80

100

120

120 80 40 10

Axial Distance from top - mm

Gap

- m

icro

ns

Cold gap - Pre irradiation Cold gap - post irradiation

Time to reach full power

Irradiation expt in FBTR

Numerical Analysis & expt both

confirm that LHR can be taken to 450

W/ cm in few days

-25

-20

-15

-10

-5

0

5

10

15

20

25

Clad_ID % He Pellet_ID Pellet_OD Pellet

Density

O/M ratio

Parameters affecting LPM

Ch

an

ge

in L

PM

(%

)

% LPM change for -2.5% variation in all parameters

% LPM change for +2.5% variation in all parameters

Key parameters to reach high

linear power

Clad ID & Pellet OD

Influence of fab.parameters on LPM

Allowable Manufacturing Deviations Pellet Defects Acceptability

Analysis defines pellet depth for

defect area. Improves in pellet

recovery without compromising

safety

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

0

100

200

300

400

500

600

700

800

900

1000

1100

1200

1300

1400

1500

1600

0 1 2 3 4

No of w eld repaires

Ruptu

re L

ife(H

ours)

No of weld repairs

Ru

ptu

re

life

-h

No of weld repairs

Reductio

n f

acto

rWeld strength reduction factor

The rupture lives measured for the

various test cases, based on the test

data and analysis, a design approach

is proposed for the inclusion in RCC-

MR:2007 edition.

Acceptability of refused end plug weld

Status on Metallic Fuel Studies

Metallic Fuel Pin Design Concepts

LINER

Sodium Bonded

U-Pu-Zr(6/10%)

No liner

75 % smeared

density

Top Plenum

Mechanical bonded

U-15Pu (4 grooves)

Zr- 4 Liner

75 % smeared

density

Bottom plenum

Mechanical bonded

U-15Pu ( 2 grooves)

Zr- 4 Liner

85 % smeared

density

Bottom plenum

Sodium Bonded

U-15Pu (No Zr in

fuel)

Zr- 4 Liner

75 % smeared

density

Top Plenum

Mechanical Bonded

Fuel Pin Cross

Section

CLAD

FUE

L

Advantages:

High BR, Low T, High linear rating, Inherent safety

Doubling time : 30 y for oxide , 12 y for metal and

8 ys for improved metal (without Zr)

Linear Power

- 450 W/ cm

Clad

- T91

Irradiation

Capsule irradiation – 3 pins

SA irradiation – 37 pins

Prototype scale -217 pins

Target Burnup

-150 GWd/ t

Pin Irradiation in

FBTR

Subassembly

Irradiation in FBTR

Full Core Metallic

Fuel in FBTR

Metallic Fuel

320 MWe Design

Metallic Fuel

1000 MWe Design

Metallic Fuel Development

Road Map

Expt Pin - Schematic Salient Highlights

Metallic Fuel Sodium Bonding Facilities

Argon Glove Box

for Sodium Handling

Sodium wire extruder

Sodium wire extrusion

into PVC tube

Pin welding fixture

Sodium Bonding

Furnace with Vibrator

Dummy Fuel Pin

Developmental Facilities

established in BARC and

IGCAR to demonstrate the

technology

Purification tower arrangement Glove Box Train arrangement Fuel Fabrication facility

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

Metallic Fuel Mechanical Bonding Facilities

Facilities for R&D on Fuel Pin Behavior

Induction heating

• 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

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