A Perspective on Fuels for A Perspective on Fuels for Indian Fast Breeder ProgrammeIndian Fast Breeder Programme

Baldev RajBaldev RajDistinguished Scientist and Director, IGCARDistinguished Scientist and Director, IGCAR

International Conference on

“Characterization and Quality Control of Nuclear Fuels (CQCNF-2009)”

February 18- 20, 2009 Hyderabad, India

Fast Reactors Fast Reactors -- Sustainability and CapabilitiesSustainability and Capabilities

• Effective utilisation of U resource

• Effective utilisation of thorium to convert into U233

• Minor actinide burning

• Can provide critical liquid metal technology and high temperature design

inputs for ADS, fusion and HTR

• Can provide an efficient means of reducing the quantity and toxicity of

radioactive waste requiring ultimate disposal

• FBR can be designed to incinerate high level wastes arising from the

reprocessing of spent fuel

Neutrons produced per neutron absorbed (Neutrons produced per neutron absorbed (ηη) for different isotopes in Thermal and Fast Reactors) for different isotopes in Thermal and Fast Reactors

1 0- 3

1 0- 1

1 01

1 03

1 05

1 07

0

1

2

3

4

5

P u - 2 3 9

U - 2 3 3

U - 2 3 5

ηη ηη

E n e r g y ( e V )

1 0- 2

1 00

1 02

1 04

1 06

1 08

0

1

2

3

4

5

Most neutrons in FBR

have energies near here

2.20

2.04

235 U

2.35

2.26

233 U

2.75Fast Reactor spectrum

2.06Thermal Reactor spectrum

239 PuNeutron Spectrum

UtilisationUtilisation & Growth Potential with Fast Reactors& Growth Potential with Fast Reactors

FBR Fuel Performance ParametersFBR Fuel Performance Parameters

� High Burnup

- lower fuel cycle cost

� Higher Linear Power

- higher power output

� Higher Breeding Ratio

- faster growth

� Cost Effectiveness

- economy

Issues to be addressed for high performance

� In-pile Fuel Behaviour

� Structural material behaviour

� Design measures and design optimization

Breeding

Indian Fast Breeder Reactor ProgramIndian Fast Breeder Reactor Program

� India started FBR programme with the construction of FBTR

� FBTR is a 40 MWt (13.5 MWe) loop type reactor. The design is

same as that of Rapsodie-Fortissimo except for incorporation of

SG and TG (agreement signed with CEA, France in 1969).

� FBTR is in operation since 1985.

� 500 MWe Fast Breeder Reactor Project (PFBR) through

Indigenous design and construction

� Govt. granted financial sanction for construction in Sep 2003.

� Construction of PFBR has been undertaken by BHAVINI.

� PFBR will be commissioned by 2010.

� Beyond PFBR: 4 units of 500 MWe FBR (twin unit concept) similar

to PFBR with improved economy and enhanced safety by 2020.

� Subsequent reactors would be 1000 MWe units with metallic fuel

Basis for Big Leap in FBR ProgramBasis for Big Leap in FBR Program

FBTR

PFBR

1200 MWt

500 MWe

Pool Type

Fuel: UO2-PuO2

40 MWt

13.5 MWe

Loop type

Fuel: PuC - UC

FBTR PFBR• 380 r-y worldwide FBR operational experience

• Rich experience with MOX fuel

• 30 y of focused R&D programme involving

extensive testing and validation

• Material and Manufacturing Technology

Development and Demonstration

• Science based technology

• Peer Reviews

• Synergism among DAE, R&D Institutions and

Industries

Confidence on PFBR Project Confidence on PFBR Project

� Technology with strong R&D backup

� Manufacturing technology development completed prior to start of

project

� Capability of Indian industries to manufacture high technology

nuclear components demonstrated (main vessel, safety vessel, steam

generator, grid plate) and cost close to project estimates

PFBR will be commissioned

by Sept 2010

Safety vessel

successfully erected on

June 24th,2008 – A

major mile stone

Safety Vessel Erection Grid Plate delivery

Main Vessel

Fuel Cycle for PFBRFuel Cycle for PFBR

• Initial fuel requirement of

PFBR will be met from

Plutonium obtained from

the PHWRs (planned to be

used for the first series of

FBRs)

• The fuel cycle of PFBR

would be closed by

constructing a Fast Reactor

Fuel Cycle Facility (FRFCF)

at Kalpakkam.

• Co-location of the facility

with reactor would reduce

cost due to transport and

also avoid security issues

• Basic technology required

for the facility is available

Layout has been planned in such a way

that expansion is possible to meet the

requirements of two more 500 MWe

FBRs to be built at Kalpakkam at later

date.

FRFCF facility – Bird’s eye view

Initial Design Limits

Burnup - 25 - 50 GWd/t

Linear Heat Rating - 250- 320 W/cm

Mixed carbide fuel with high Pu content (U0.3 Pu 0.7)C chosen

based on

Theoretical studies

Experimental studies

Availability of basic expertise in fabrication

Literature data on irradiation performance

Pre-irradiation data base

20 % cold worked ASS 316 as structural material

Design of Fuel and Structural Materials for FBTRDesign of Fuel and Structural Materials for FBTR

BurnBurn--up Evolution of FBTR carbide fuel up Evolution of FBTR carbide fuel

CARBIDE FUEL PERFORMANCE

Out Of Pile

Experiments

In Reactor

Experiments

Post-Irradiation

Examinations Modeling &

Analysis

Fuel column elongation with burnup from X-radiography

0

2

4

6

8

10

12

0 50 100 150 200

Burn-up(GWd/t)

Average increase in stack length(mm)

Burn-up in GWd/t

Av. increase in stack length

Trend in fission gas release

at different burnups

Performance of FBTR Carbide FuelPerformance of FBTR Carbide FuelX –radiography & Neutron radiography of high burnup fuel

Plenum

Fuel columnFuel columnPlenum

Higher axial swelling in the restrained swelling phase

Low fission gas release and plenum pressure

X-radiographs N-radiographs

5

Max. FG release ~ 16 %

FG Pressure ~ 20 bars

(155 GWd/t)

17

5

Performance of FBTR Carbide FuelPerformance of FBTR Carbide FuelMicrostructure of fuel pin cross section after

different burn-ups

� 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 circumferentialcracking 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 accomodated by porosities & cladswelling

Micrographs of fuel pin cross section at the centre of fuel columncentre of fuel columnafter 25 & 50 & 100 GWd/t burn-up

155 GWd/t – CENTRE of the fuel column

155 GWd/t – END of the fuel column

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

(485 C)

(430 C)

11.5%

3.5%

Performance of FBTR fuel Clad and WrapperPerformance of FBTR fuel Clad and Wrapper

Burnup

Reached

Max

Fluence

Peak

dpa

155 GWd/t 1.2 x 1023 n/cm2 83

Diametralstrain (∆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

Mechanical Properties & Microstructure Evolution of SS316: PMechanical Properties & Microstructure Evolution of SS316: PIE Data IE Data

UTS vs dpa

Uniform Elongation vs dpaCladding

Variation in Room Temperature (RT) tensile properties of hexagonal wrapper with dpa

Wrapper

Exploring the possibility of higher burnup (>155 GWd/t) for FBTR with carbide fuel

∆V / V %

81 dpa

40 dpa

30 dpa

Virgin

TEM studies

100 nm

100 nm

100 nm

500 nm

FBTR fuel burnup will be enhanced up to 165 GWd/t

based on PIE and Thermo-Mechanical analysis

Life Extension of FBTR FuelLife Extension of FBTR Fuel

Performance Evaluation

through PIE

Thermo-Mechanical Modeling

and Analysis

Phase

Stability

Strength &

ductilityDuct dilation

& interaction

Fission gas

release

Metal phase

formation

Porosity

exhaustion

Clad carburisation

Swelling

FUEL STRUCTURAL

MATERIALS

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 mm

Peak target Burnup : 100 GWd/t

Peak neutron dose : 85 dpa

PFBR Fuel SubassemblyPFBR Fuel SubassemblySalient Details

No of Fuel SA : 181

Total SA : 1758

PFBR Fuel Design ParametersPFBR Fuel Design Parameters

PFBR core design incorporates the following

measures to assure the target performance

parameters

� Number of fuel enrichment zones

� Annular pellet

� Adequate and optimum inter subassembly gap

� Optimum pin bundle porosity within a subassembly

� Judicious choice of fuel chemical parameters

� Judicious choice of pellet smeared density

� Optimum core restraint system design

� Rational choice of materials for clad and wrapper

PFBR Salient Results for Target BurnPFBR Salient Results for Target Burn--upup

� IAEA CRP programme on Static Core Mechanics behaviour

� Inter-comparison of codes on benchmark problems – Single SA to Core Sector

� Indian code MABOW code matches well with UK, German, Russian codes

Salient Results

Max bowing : 24.6 / 16.1 mm

Control rod bowing : 20.1 / 16.3mm

(Operation/shutdown)

Max interaction load : 3500 N

Subassembly mass : 250 kg

Total extraction force : 6000 N

Dilation defines burn-up limit

Core

Sector

Resultant Bow ( top) Resultant Bow ( top) Extraction force Extraction force

Contact force (top) Contact force (top) Contact force (button) Contact force (button)

Ferritic / Martensitic Steels

9Cr-1Mo; Mod. 9Cr-1Mo-V-Nb9Cr-2Mo-V-Nb; 12Cr-1Mo-V-W;

Current generation

Immediate Future

Future

Oxide dispersion strengthened (ODS) steels

13Cr-1.5Mo-2.9Ti-1.8Ti2O3, 13Cr-1.5Mo-2.2Ti -0.9Ti2O3-0.5Y2O3, 12Cr-0.03C-2W- 0.3Ti-

0.24Y2O3,9Cr - 0.13C- 2W + Ti + Y2O3

Austenitic stainless steelAustenitic stainless steel

Type 316 & modificationsType 316 & modifications

15Cr15Cr--15Ni15Ni--TiTi--C (Alloy D9) & C (Alloy D9) &

its improved versionsits improved versions

Evolution of Structural Materials for FBRs

Rate Theory

Experimental Data

from ion irradiation

expts.

MD SimulationInter atomic

Potentials

KMC

Microstructure and

precipitate

evolution

Primary defect production

Diffusion of defects and

their reactions

Experiments to

Validate Models

PREDICTION OF IRRADIATION

BEHAVIOR OF MATERIALS

Approach For Advanced Material Development

Void swelling and Positron Annihilation Studies on

20% CW D9 Alloy with 0.15 and 0.25% Ti

TiC precipitates: The increase in average lifetime of positrons is due to the increase in the

number density of TiC precipitates (beyond 750 K in Sample A and beyond 850 K in Sample

B), which are effective in reducing the swelling. Thus, the swelling at Peak swelling

temperature is less in sample A (Ti/C=6). Also, the shift in peak swelling temperature is also

correlated with the onset of TiC precipitation.

Swelling studies at 100 dpaPositron Annihilation studies

Model alloy without Ti

Ti : 0.25%Ti/C = 6

Ti : 0.15% Ti/C = 4

(823K)(923K)

TiC

precipitates

Grain boundary engineering (GBE) in Alloy D9

As received specimen

CSL boundary = 55%

ΣΣΣΣ3 boundary = 44%

Random boundary

connectivity is significant

(a) b)

As received specimen (a) CSL+ Random (b) Random

(a) IQ (b) random boundary map of GBE sample

(a) b)

CSL boundary = 79%

ΣΣΣΣ3 boundary = 62%

ΣΣΣΣ9 boundary = 9%

Random boundary

connectivity is broken

Towards improving resistance to Swelling

0.16 0.20 0.24 0.28 0.32 0.36 0.40100

1000

4 5 6 7 8 9 10Ti/C

973 K

Rupture life,h

Titanium, wt%

250 MPa

200 MPa

175 MPa

Optimization of Ti/C ratio in D9I w.r.t. thermal creep properties

0.20 0.25 0.30 0.35 0.40

10-10

10-9

10-8

973 K

Minimum creep rate

, s

-1Titanium, wt%

250 MPa

200 MPa

175 MPa

Phosphorous = 0.025%; silicon = 0.75 %

Optimum level of titanium ~ 0.24 wt% (Ti/C ~ 6)

D9I-Effect of Phosphorus Addition on Swelling

Enhances sink strength and increases

the point defect recombination

Needle like Phosphide Precipitates (Fe2P)

700 750 800 850 900 950

1

2

3

4

Swelling (%)

Temp (K)

M58

G3088T

0.026%

0.048%

Neutron dataSimulation using heavy ions from accelerator30 appm Helium pre-implanted 5MeV Ni++ ion irradiation

:Damage rate : 7x10-3 dpa/s

Swelling at peak 3.9% and 2.5% for P=0.026 and P=0.046

wt% respectively.

P effective in suppressing swelling at temp > 800 K due to

needle like phosphide precipitates.

• 9-12% Cr-Mo ferritic martensitic steels

• Mod. 9Cr-1MoVNb (T91), 9Cr-1Mo (EM10), 9Cr-2MoVNb

(EM12),12Cr-1MoVW (HT9)

• Good choice for damage up to 200 dpa

- Very good choice for wrapper

- Creep resistance not adequate for clad

- Increase in DBTT (Lowest for. 9Cr-1Mo steels)

- Saturation in ∆∆∆∆DBTT with dose

- Manufacturing experience

Low Swelling Materials

Tem

perature (K)

Advantage of ODS alloy for Fuel Pin Clad

Dispersion Strengthened Alloys

0.33 nm

(222) YO

0.14 nm

(200)Fe

0.33 nm

(222) YO

0.14 nm

(200)Fe

Yttria-Titania-Oxide

Fe-9Cr-2W-0.2Ti-0.35Y2O3-0.1C (Mechanical Alloying / Extruded)

Collaborative research project: ARCI, NFC, IGCAR

1050 1100 1150400

450

500

550

600

650

Av. Vickers Macro-Hardness (HV)

Extrusion Temperature (oC)

ODS Alloy

Substructure

effect

Increase of Extrusion Temperature � Finer Martensite

laths (~100nm to ~20nm� Higher Hardness

Developmental ODS FM Alloy - CharacterisationFe-9Cr-2W-0.2Ti-0.35Y2O3-0.1C – Simoloyer 4h, Extruded at 1050oC at ARCI

DF

FIN

E M

AR

TE

NS

ITE

LA

TH

S

0

100

200

300

400

500

0 5 10 15 20 More

Size of Yttria (nm)

Frequency

DF

YTTRIA DISTRIBUTION

MOIRE FRINGE FROM YTTRIA

9Cr-ODS alloy under development through MA in collaboration with ARCI, NFC & IGCAR

Homogenous distribution of Yttria in ferrite matrix at higher Extrusion Temperature

Size distribution of Yttria peaks at ~ 5nm measured in extruded material

ODS Alloys - Indigenous Development

Collaborative research project: ARCI, NFC, IGCAR

Visual Inspection of ODS Clad Tubes at NFC

Feasibility of Production of ODS Alloy Clad Tube of

6.6 mm OD x 0.45 mm thk x 1.5 m has been demonstrated

Road Map for Development of Advanced

Clad and Wrapper Materials

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

Target Burnup

GWd/t

<150 >150 200 200

Fuel Oxide Oxide Oxide Metallic

Clad material D9I SS D9I SS ODS alloy T91

ferritic

steel

Wrapper

Material

D9I SS T9 ferritic

steel

T9 ferritic

steel

T91

ferritic

steel

Linear Power,

W/cm

450 450 500 450-500

• Oxide fuelled PBFR and subsequent four FBRs would adopt the

developments in the course of operation

• Metallic core of 1000 MWe (~ 2020)

Sol-gel based Fuel Fabrication

� Exploiting amenability for remote

operations to fabricate minor actinide

containing fuel for burning in fast

reactors

� Segregation of microspheres: to be

investigated by test fuel irradiation in

FBTR

� Laboratory Scale Facility set up for

test

fuel pin fabrication ( Collaborative effort

of BARC and IGCAR scientists)

Lab. Scale Facility

and Gelation set up

Pin loading and welding equipment

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

500 MWe Design

Metallic Fuel

1000 MWe Design

Metallic Fuel Development

Doubling time : 30 y for oxide , 12 y for metal and 8 ys for improved metal (without Zr)

Road Map

Expt Pin - Schematic Salient Highlights

Metallic Fuel Pin Design Concepts

CHALLENGES

� Sodium bonding

� Achieving good contact between fuel & liner

and liner & clad - Swaging

� Zr addition increases Tmelt but reduces

breeding & hence Zr to be minimized

� Clad eutectic formation – Clad temp

maintained at 650O C

� Low reactor outlet T ~ 510 O C

� Fuel Clad Chemical Interaction

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

FUEL

Metallic Test Fuel Pin Fabrication Facility

Purification tower arrangementGlove Box Train arrangement Fuel Fabrication facility

Co-swaged fuel rod with clad / liner

INJECTION-CAST, SWAGED & MACHINED

URANIUM RODS (demonstrated at BARC)

Length = 160 mm, Diameter= 4.67±0.04

Argon Glove Box

for Sodium Handling

Metallic Fuel Sodium Bonding Facilities

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

Challenges in Reactor Physics for Metallic Fuelled FBRs

� Metallic Fuel Behaviour under

Transient Conditions (experiments

similar to oxide fuel in CABRI &

TREAT to establish safety

margins)

- Need for Experiments in a Test

Reactor

� Minimising the positive Na void

reactivity effect

- Use of Na plenum instead of top

axial blanket

- Reduction in Na volume fraction

in the core

� Improved Minor Actinide Cross-

sections

� Fresh 20% UO2

� 465W/cm,Pmax/P0=318%

� Axial fuel expansion is 3.25 cm in transient

� No clad failure

� Large fuel melting with slight clad deformation & central hole formation

� CABRI Test on Oxide

Pyrochemical Reprocessing for Spent Metallic Fuels

Electrotransport behaviour of Pu from

Liq.Cd Anode to Liq. Cd Cathode

Pu deposit on solid cathode in

lab. scale studies

# Ideally suited for metallic fuels

# Laboratory Scale Studies on Pu

based alloys

# Modelling the Process

PRAGAMAN- Code developed

Studies on Pyroprocess Flow Sheet Aspects

# Lab. scale studies on Ceramic Waste form- Glass bonded Sodalite

synthesised – Properties being studied

# Studies on metal Waste Form – Alloys of Zr with Mod.9Cr-1Mo cast

Leaching studies to be taken up

# Studies on Direct Electrochemical Reduction Process for

Actinide Oxides – Partial Reduction achieved for Uranium Oxide

DSC measurements on Glass Transition

Temperature of glass bonded Sodalite

UO2

pellet LiCl – Li2O melt

Pt

O2

O2-

RE

Direct Oxide Reduction Process

600 700 800 900 1000

He

at

flo

w (

arb

. u

nit

s)

Temperature(K)

BoroaluminoSilicate Glass

SLZ + glass physical mixture

Glass bonded Sodalite

Engineering Scale Facility for Electrorefining Studies

# Engineering Scale Facility for studies using 3 kg of U alloys set up

# Remote operation to be demonstrated

# Equipment to be housed inside the facility:

Pin Chopper, Electrorefiner, Distillation cum melting Chamber

# A crane and a power manipulator installed inside the containment box

for remotisation

Containment Box

Electrorefiner

inside

containment

box

Development of waste formsHigh-level waste from FBRs⇒⇒⇒⇒ high concentration of Actinides &

Noble Metals

* Difficult to fix in Borosilicate Glass

Ceramic & new Glass waste forms

under development at IGCAR

1. Synroc (poly phase titanates)

2. Monazite (single phase phosphate)

3. Iron Phosphate Glass

Up to 20 wt% simulated waste

(FBTR, 150 GWD/T) successfully

immobilized in these matrices.

Synroc canister with 1kg material, after HIP

(IGCAR – BARC – NCL – DMRL collab.)

SUMMARY

� Rich experience from operation of carbide fuel from

FBTR.

� Fuel performance issues are well understood and

robust roadmap has been drawn to address them

and engineer advanced materials to meet the targets.

� Confidence in going for moderately high burn-up in

PFBR with austenitics.

� Future FBR development focusses on high breeding

coupled with high burn-up.

� India has adopted a mission oriented approach in

meeting the challenges towards realisation of its

target and pursuing a path suiting its energy security

needs.

Global leadership in mega technology of high relevance to

India and World

Enhanced Synergy with

Academia,

Research and

Industry

Mega collaboration with

Indian Academic Institutes

and Research labs

Basic sc

ience,

scientifi

c breakt

hroughs

for challe

nging te

chnolog

y

Human resources (attracting, nurturing,

mentoring and motivating)

� FBTR life extension for next 20 years

� Robust PFBR

� Realising Fast Reactor Fuel Cycle Facility

� Design and development for 500 MWe FBRs with

improved economy and enhanced safety

� High performance fuel cycle technologies

� Significant Progress towards realisation of

Metal fueled reactor & associated fuel cycle

Challenges, Approaches and Targets

INDIAN NUCLEAR PROGRAMMEINDIAN NUCLEAR PROGRAMMEINDIAN NUCLEAR PROGRAMMEINDIAN NUCLEAR PROGRAMMEINDIAN NUCLEAR PROGRAMMEINDIAN NUCLEAR PROGRAMMEINDIAN NUCLEAR PROGRAMMEINDIAN NUCLEAR PROGRAMME

Towards sustainable energy Towards sustainable energy Towards sustainable energy Towards sustainable energy Towards sustainable energy Towards sustainable energy Towards sustainable energy Towards sustainable energy

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