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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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