homi bhabha centenary year oct. 30, 2008- recycle · pdf filerecycle fuel fabrication in india...

RECYCLE FUEL FABRICATION RECYCLE FUEL FABRICATION IN INDIAIN INDIA

H.S. KAMATH,Nuclear Fuels Group,

Bhabha Atomic Research Centre,Trombay, Mumbai 400 085

IAEA TM on “Country Nuclear Fuel Cycle Profiles” at Fukui, Japan on 1-2 Dec, 2008

Homi BhabhaCentenary Year

Oct. 30, 2008-Oct. 30, 2009

Dec.1-2, 2008 IAEA TM, Fukui 2

OutlineOutline1. Introduction.2. MOX fuel fabrication for BWR & PHWR.3. Mixed Carbide fuel fabrication for FBTR.4. PFBR MOX fuel development &

fabrication.5. AHWR MOX fuel development6. Concluding Remarks.


29/5/06 - 01/06/06 TWGNFCO, Vienna 3

““There is no power as costly There is no power as costly as Noas No--PowerPower”” – Homi Bhabha

10 100 1000 1000030

40

50

60

70

80 Japan

U.S.A.

India (1951-60)

India (1961 -70)

India (1980-85)

India (1997)

Source of the Data:World Bank, 1999Li

fe E

xpec

tanc

y at

Bir

th (y

ears

)

Electricity Consumption per Capita (kWh/year)

Nuclear Energy wouldprovide Prosperity to developing Nations--FIRST UN Conf PUAE,Geneva,1955


Three Stage Indian Nuclear Three Stage Indian Nuclear Power Power ProgrammeProgramme1. To maximize the energy potential

of Uranium and empower Thorium.

2. Self-Reliance for Energy security andEnergyIndependence.

3. Minimize emissions to environment while meeting the huge demand for electricity.

Closed fuel cycle is the integral part of this strategy

H.J. Bhabha(1909-1966)


Growing Economy & Growing Energy NeedsGrowing Economy & Growing Energy Needs

India is one of the fastest growing economy.- Recorded 9% GDPgrowth IN 2007-2008.Population is over 1 billionbut India’s power consumption is about 600 kWh well below the global average of 2500 kWh.Demand for energy continues to rise because of the growth in economy & population.

India’s GDP growth for the last decade& Population growth rates

0

1

2

3

4

5

6

7

8

9

10

2003-04 2004-05 2005-06 2006-07 2007-08

GDP

GD

P gr

owth

rate

(%)

CSO, Govt of India

Population Growth Rate of India



IndiaIndia’’s GDP to Grow Even During s GDP to Grow Even During Current Global Financial CrisisCurrent Global Financial Crisis

2/3 of India’s one billion people live in energy deficiency.Out of 600,000 villages, about 25% have no grid electricity .Domestic savings rate is 35% of GDP.Worlds Largest young demographic profile.

-0.5

0.5

1.5

2.5

3.5

4.5

5.5

6.5

7.5

8.5

9.5

India US Japan

20082009

Projected GDP growth ratesfor India, US & Japan

IMF Projections(Nov. 2008)

7.8%

6.3%

1.4%0.5% 0%

(-?)0%(-?)


Indian Installed Electric CapacityIndian Installed Electric CapacityCoal 55% 68,300 MWGas 10% 12,300 MWOil 1% 1,200 MWHydro 26% 32,135 MWRenewable 4.5% 6,150 MWNuclear 3.5% 4,120 MWTotal ~130 GW

Projected Installed Power CapacityProjected per capita (kWh)


Dec.1-2, 2008

IndiaIndia’’s Green House Gas s Green House Gas EmissionsEmissions

• India’s green house gas emission (GHG) is the lowest in per-capita terms.

• 4% of the World GHG emission in spite of having 17% of the World population.

• Rapid economic growth is leading to increased demand for electricity. Nuclear energy has to play an important role.

0%

5%

10%

15%

20%

25%

30%

35%

1980 1985 1990 1995 2000 2005

Sha

re o

f Glo

bal C

O2

Emis

sion

s (%

)

US

Western Europe

China

India

Share of global CO2 emission

Environmental burden



Indian Nuclear Energy Indian Nuclear Energy Resource PositionResource Position

ResourceResource QuantityQuantity(Tonnes)

Energy PotentialEnergy Potential(GWe-Yr)

Uranium 61,000 328 in PHWR42,330 in FBR

Thorium 300,000 155,500 in Breeders

Uranium is 1% of the World resources.Thorium resource is one of the largest in the World.



Closed Fuel CycleClosed Fuel Cycle

1. Resource extension & Sustainability.

2. Waste Classification & Isolation.3. Reduction in Repository space.4. Proliferation Resistance – No Pu

mines- Threat reduction for the Future generation.


PRESENT STAGE OF NUCLEAR POWER PROGRAMPRESENT STAGE OF NUCLEAR POWER PROGRAM

Stage - IPHWRs

• 15 Operating• 3- Under construction• Several others planned• POWER POTENTIAL ≅

10,000 MWe

LWRs• 2 BWRs Operating• 2 VVERs under

construction

Stage - IIFast Breeder Reactors

• 40 MWth FBTR - OperatingTechnology Objectives realised

• 500 MWe PFBR- Under Construction

• POWER POTENTIAL ≅530,000 MWe

Stage - IIIThorium Based Reactors

• 30 kWth KAMINI- Operating

• 300 MWe AHWR- Under development

• POWER POTENTIAL ≅ Very Large. Availability of ADS can enable early introduction of Thorium on a large scale

9186848479

7569

72

90

50

55

60

65

70

75

80

85

90

95

1995-96 1996-97 1997-98 1998-99 1999-00 2000-01 2001-02 2002-03 2003-04

Ava

ilabi

lity

Fac

tor

(%)



MOX Fuel for TAPSMOX Fuel for TAPS

Development for MOX fuel for TAPS:Phase I: [Early 80’s]

Fabrication of MOX fuel experimental clusters for irradiation in Cirus

Phase II: [Mid 90’s]Industrial scale MOX fabrication plant set-up at Tarapur.



PuPu bearing Fuels for bearing Fuels for Thermal ReactorsThermal ReactorsMOX Fuel for BWRs

Several lead MOX Fuel Assemblies irradiated in TAPS-1 and TAPS-2 (BWRs) successfully to the design burn-up.

MOX fabrication plant at Tarapuruses ATTRITOR technology for highly homogeneous fuel.



MOX Fuels for MOX Fuels for PHWRsPHWRsDevelopment of high burnDevelopment of high burn--up fuels for use in up fuels for use in PHWRsPHWRs

Fuel subassembly inside glove box

End plate welding of PHWRMOX bundle at AFFF

50 MOX fuel bundles fabricated at AFFF,BARC,Tarapur.All the MOX bundles are irradiated in KAPS-1 and peak

burn up achieved is 20,000Mwd/t.Demonstration of capability to manufacture & utilize

MOX fuels for PHWRS



Fast Reactors Fast Reactors –– Powering Powering the Futurethe Future

Fast Breeder Test Reactor at Kalpakkam (40MW(th)) is the test bed for development of fuel, blanket and structural materials for FBR programme.

FBTR achieved criticality on October 18, 1985 with unique Pu rich mixed carbide fuel developed by BARC.

The peak burn-up achieved on MK I is 155 GWD/T.



Fuelling FBTR, Fuelling FBTR, KalpakkamKalpakkam

FBTR at Kalpakkam is the cradle for development of LMFR in India.

An indigenous Pu rich (U-Pu)MOX fuel was first considered and flow-sheet developed.

Difficulties with respect to(a) Poor thermal conductivity(b) Higher oxygen potential

raising concern on fuel-coolant and fuel clad compatibility issues.

Pu rich carbide appeared attractive option.



Fabrication of Mixed Carbide Fabrication of Mixed Carbide Fuel for FBTR at Fuel for FBTR at KalpakkamKalpakkam

a) (U0.3Pu0.7) Mixed Carbide (MK I) and (U0.45Pu0.55) Mixed Carbide fuel (MK II) was chosen as fuel for FBTR at Kalpakkam.

b) High thermal conductivity and metal densityin MC lead to compact core & high breedingratio.

c) Performance of this fuel has been proved to be excellent and is at present beyond 150,000 MWD/T.

d) Technology is highly complex but can be adopted for small cores.


• Lower solidus temperature,Lower solidus temperature,

•• Lower thermal conductivity,Lower thermal conductivity,

•• ‘‘OO’’ and and ‘‘NN’’ pick up from cover gas pick up from cover gas

•• ‘‘PuPu’’ loss during fabrication.loss during fabrication.

Issues Carbide fuelIssues Carbide fuel

•• Clad CarburizationClad Carburization

•• SwellingSwelling

•• Susceptible to Oxidation and Hydrolysis by Susceptible to Oxidation and Hydrolysis by Moisture and Oxygen Moisture and Oxygen

•• PyrophorocityPyrophorocity

High ‘Pu’ carbide

Carbide fuels in generalCarbide fuels in general

Compare to Compare to low low ““PuPu”” fuelfuel



•

Challenges ?Challenges ?

Unique composition; highly Unique composition; highly enriched enriched ‘‘PuPu’’ carbide carbide

Lack of OutLack of Out-- ofof-- pile and Inpile and In--pile Data in literaturepile Data in literature



Evaluation of Fuel propertiesEvaluation of Fuel properties

Thermal Expansion

Thermal Conductivity

Hot Hardness

Solidus Temperature

Out-of-pile fuel-clad-coolant compatibility

Post Irradiation Examination (PIE)


400 800 1200 1600 20000.0

0.5

1.0

1.5

2.0

2.5

3.0

Temperature (K)

Perc

enta

ge T

herm

al E

xpan

sion

(%)

Experimental data Polynomial fitted curve

Thermal expansion as a function of temperature for sintered hyperstoichiometric (Pu0.55U0.45)C

Thermal ExpansionThermal ExpansionHomi Bhabha

Centenary Year

Dec.1-2, 2008 IAEA TM, Fukui 2229/5/06 - 01/06/06 TWGNFCO, Vienna 22

400 800 1200 1600 20004

8

12

16

20

MK I MK II

Ther

mal

Con

duct

ivity

, W m

-1 K

-1

Temperature, K

Thermal Conductivity Hot hardness

ThermophysicalThermophysical PropertiesProperties

200 400 600 800 1000 1200 1400 1600

103

104

(U0.3Pu0.7)C - MKI (U0.31Pu0.69)C0.93 Ref (9) (U0.79Pu0.21)C Ref (9) (U0.45Pu0.55)C - MKII

Log

Har

dnes

s (M

Pa)

Temperature (K)

BARC

40

30

20

10

0

-10

-20

0.1 0.2 0.3 0.4 0.5 0.6 0.718

16

14

12

10

8

6

4

2

ΔGC (

KJ/

mol

) C potential

Log

P (M

Pa)

Pu content

CO pressure Pu pressure N2 pressure

Effect of Effect of ‘‘PuPu’’ content on content on ‘‘CC’’ potential, potential, ‘‘COCO’’pressure, pressure, ‘‘PuPu’’ pressure and pressure and ‘‘NN’’ pressurepressure



SS316 tube with MKI at 973K, 1000hrs.

SS316 / MKI (Class A) at 973K,1000hrs.)

P/C interface

P/C interface

The average hardness of reference samples : 179 VHN (at 973K for 1000hrs) and 152 VHN(1123K for 1000hrs) The hardness data of as received material was 230VHN.

Fuel Clad chemical compatibilityFuel Clad chemical compatibility

INDIA

UO2 Graphite PuO2

Milling & Grinding

Tableting

Carbothermic ReductionAt 1673K in vacuum

Crushing

Milling

Compaction

SinteringAt 1923K in Ar-8%H2

Pellet Inspection

Process control

Process control

Process Flow Sheet for Fabrication of (U,Pu)C PelletsProcess Flow Sheet for Fabrication of (U,Pu)C Pellets

XRD

Carbothermic synthesis

Sintering

MK II

MK II



Reject RecyclingReject Recycling

Dry Methoda) Mechanical

Pulverization and re-consolidation

b) Thermal Pulverization to oxide phase and re-carbothermicreduction

1200 1300 1400 1500 160010

20

30

40

50

60

70

1470oC

DTA

sig

nal (μV

)

Temperature (oC)

1200 1300 1400 1500 160025

30

35

40

45

50

1440oCD

TA s

igna

l (μV

)

Temperature (oC)

MO2+C

UO2+ PuO2 + C

BARC



Sintered Pellets

Mixed Carbide Fabrication lineMixed Carbide Fabrication line


(a)

(c) (d)Fuel- Clad gap: 20-30 μm

No increase in clad diameterSwelling rate: 1.5% Fission gas release: 5 to 6%No measurable increase in the flat-to-flat

Clad diameter increase by 1.2 – 1.6%No fuel-clad gap, No visible clad carburization�

Fission gas release: 14 %Fuel swelling : <1% per atom percent burn-up

25

25 G

Wd/

tG

Wd/

t11

0 11

0 G

Wd/

tG

Wd/

t50

50

GW

d/t

GW

d/t

155

155

GW

d/t

GW

d/t

No increase in clad diameter,Fuel swelling: 1.2 % per at% burn-upFission gas release: NegligibleLHR raised to 400 W/cm after 35 GWd/t

Diametral strain of fuel pin : 5%

Maximum internal pin pressure due to fission gas : 2.09 MPa. ;A distinct zone with no porosity near periphery due to creep of fuel

PIEPIEof of MCMCFuelFuel



(U(U--44% 44% PuPu) MOX fuel ) MOX fuel

For Hybrid core of FBTRFor Hybrid core of FBTR



Phase StabilityPhase Stability

ThermoThermo--physical Propertiesphysical Properties

Thermal conductivity / diffusivity, thermal expansion, high temperature hardness

FuelFuel-- Coolant Chemical compatibilityCoolant Chemical compatibility

Issues of (UIssues of (U-- 44% PU)MOX fuel44% PU)MOX fuelHomi Bhabha

Centenary Year

1000 1200 1400 1600 18000.5

1.0

1.5

2.0

2.5

3.0

3.5

4.0

4.5

5.0

5.5

T.Jarvis

MOX-Proposed by IGCAR (calculated) UO2 J.K.Fink 1

The

rmal

Con

duct

ivity

(W/m

K)

Temperature (K)

MOX (U0.55 Pu0.45)O2 (RMD)

UO2 Control Sample (RMD)

Thermal conductivity ofThermal conductivity of(U(U0.550.55PuPu0.450.45)O)O2 2 pelletpellet

XX--ray radiograph of MOX (ray radiograph of MOX (UO2--45%PuO45%PuO22) fuel ) fuel ––Sodium compatibility capsulesSodium compatibility capsules

(U(U0.550.55PuPu0.450.45)O)O2 2 pellets pellets After compatibility testsAfter compatibility tests

XRD Pattern of (UOXRD Pattern of (UO22--45% PuO45% PuO22) fuel) fuel

Photomicrograph of MOX pellets after Photomicrograph of MOX pellets after compatibility tests at 700compatibility tests at 700ooC for 400 hC for 400 h


X-ray radiograph of fuel-coolant compatibility capsules loaded with MOX (PFBR type) pellets (700C after 400 hrs.)

XX--ray radiographyray radiographyHomi BhabhaCentenary Year


Na + MOX

Na

breach

MOX

clad

D9 clad

MOX800C, 260hrs. 700C,450hrs

x200

700C,350hrs

x200

Clad breach

MOX

Na

700C,350hrs

x200

MICROSTRUCTURESMICROSTRUCTURES


FBTR OperationFBTR Operation

1. FBTR now operates with expanded hybrid core of mixed carbide and high PuMOX.

2. 20% of the core is High Pu (45%Pu) MOX.

3. Experimental PFBRMOX fuel assembly at the centre of FBTR.

FBTR Hybrid Core configuration

FBTRHomi Bhabha

Centenary Year


Prototype PFBR FA Prototype PFBR FA Irradiation in FBTRIrradiation in FBTR

37 pins of UO2-29%PuO2fabricated at A3F(T), BARC are under irradiation in FBTR.

U233 is added to simulate high linear rating of PFBR with same (U-Pu) chemical composition

FA loaded in the centre of FBTR core

Burn-up reached - 80 GWd/t at 450 W/cm & continuing

37 pin Experimental PFBR MOXFuel

PFBR type MOX pins for experimental irradiation



MOX Fuel for PFBR: MOX Fuel for PFBR: Why MOX?Why MOX?1) Industrial scale operational

experience in fuel cycle facilities2) Safe fuel cycle operations and proven

safety response of the MOX fuel in the reactors

3) Economic competitiveness4) High burn-up potential5) Closed fuel cycle with minimum in

and out of core inventory of Pu


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/cmActive core height : 1000 mmBreeding Ratio : 1.05Clad & Wrapper : 20 % CW D9No.of Pins : 217Width Across Flats : 131.3 mmPeak target Burnup : 100 GWd/tPeak neutron dose : 85 dpa

PFBR FUEL SUBASSEMBLYPFBR FUEL SUBASSEMBLYSalient Details

No of Fuel SA : 181

Total SA : 1758

ERECTION OF SAFETY VESSEL IN PFBR ERECTION OF SAFETY VESSEL IN PFBR

Safety vessel successfully erected on June 24th,2008 – A major mile stone


Artistic View of PFBRArtistic View of PFBRHomi Bhabha

Centenary Year

CRO Dep. UO2 Powder PuO2 Powder SS Tubes & Hardwares

Dep.UO2 Pellets

Weighing

Mixing & Milling

Pre-compaction&

Granulation

Final Compaction

Sintering

MOX Pellets Degassing

Stack Making and Loading

Top End PlugWelding (He)

Decontamination

Wire Wrapping &Spot Welding

Fuel Pin Inspection & Storage

Powder Charac. & Chem. Analysis

PuO2 Enrich.

Degassing

Bottom End PlugWelding (Ar)

Loading in Tubes

Degassing

Visual He Leak

Metrology X-Radiography

U,Pu - Analysis,Dissol.

Test,Visual,Dimensional, Linear Mass,Metallography, α - Autorad.

O/M, Total Gas, Metallic & Non-metallic Impurities

Stack Length Check

Visual, He Leak, XGAR, Metrology, X-Radiography, γ - Scan, Cover Gas in Pins

Contamination check

To CSP

Oversize pellets

C.Grinding

Degassed depleted UO2 pellets

PFBR Fuel PinPFBR Fuel PinFabrication Fabrication FlowsheetFlowsheet withwithQC StepsQC Steps


Conveying of powder containers at AFFFConveying of powder containers at AFFF

COMPACTION PRESS AND COMPACTED PELLETSCOMPACTION PRESS AND COMPACTED PELLETS

Inspection of EndInspection of End--plug Welding plug Welding of Dof D--9 Fuel Clad Tubes9 Fuel Clad Tubes

Solidification cracking Defect free weld zone

Probe Water Tank

Clad

Plug

Beam Path

Slit

X-Ray Tube

Fluoroscopy

Pin carriage

REAL TIME MOTION RADIOGRAPHYREAL TIME MOTION RADIOGRAPHY ULTRASONIC END CAP WELD ULTRASONIC END CAP WELD INSPECTIONINSPECTION

Monitor

Metallographic Evaluation

TIG welding Machineinside a glove-box


PFBR FUEL PINSPFBR FUEL PINS


Metallic fuels for future Metallic fuels for future FBRsFBRs

Faster Growth Rate (BR 1.3 – 1.4 )

Enhanced Reactor Safety

Integrated Fuel Cycle

BARC



Development of Metallic FuelDevelopment of Metallic Fuel

1) Thermophysical, chemical, compatibility and thermodynamicsstudies of metallic fuels.

2) Development of technology for fabrication of fuel/blanket columns for sodium bonded and mechanical bonded fuel and demonstration with U/U-Zr alloy.


Metallic Fuels for FBRSMetallic Fuels for FBRS

Ref metallic fuel for FBR - U-15Pu-10Zr ( Sodium Bonded)

• zirconium increases the liquidus/solidus temperature of (U,Pu) alloy

• retards interdiffusion of fuel and clad constituents

Design for high burn-up metallic fuels• reduction in smear density to 75 % to accommodate swelling

and to reduce FCMI

• provision of adequate gas plenum for fission gas release

• measures to reduce FCCI



New Designs of Metallic fuelsNew Designs of Metallic fuels

1. Barrier clad to minimise fuel clad chemical interactions.

2. Mechanical bonding instead of sodium bonding.

3. New U-Pu alloys ( Ternary or Binary ).


300 400 500 600 700 800

750800850900950

1000105011001150120012501300

Tmelt=1243 K Clad limit = 923 K Centreline clad,T91

Tem

pera

ture

(K)

Linear Heat Rate (W/cm)

Maximum LHR for Power to Melt and Eutectic Maximum LHR for Power to Melt and Eutectic formation with T91 Claddingformation with T91 Cladding



ASAS--PILGERED CLAD TUBE WITH LINERPILGERED CLAD TUBE WITH LINER

OD = 6.645 – 6.650 mm

ID = 5.39 – 5.44 mm

LINER THICKNESS = 137 – 149 μ

NO GAP WAS FOUND BETWEEN SS TUBE & ZIRCALOY LINER



Why Mechanical Bond?Why Mechanical Bond?1. In case of metallic fuels better heat transfer

is ensured by better contact of fuel meat with clad by Swaging operation.

2. Chemical reactivity of sodium during fabrication needs stringent safety measures.

3. Additional head end problems during reprocessing.

4. Non-bond problems – challenges for fabrication, quality control & performance.

5. F.G. column only at the top and also F.G. mobility in sodium.



Preparation of U-Zr alloys by Arc Melting

Induction melting with addition of Pu & Injection Casting

Demoulding

Length cutting

ECT for defect detection

Length, diameter & weight measurement

Sodium/Mechanical bonding with clad

Metallic Fuel for Fast Reactors

Swaging Drawing

ECT for Bond Inspection Injection Castingof Uranium

R&D on Fuel Fabrication for R&D on Fuel Fabrication for FBRsFBRs

8mm diameter (Zr clad)6 mm diameter



INJECTION CASTING INJECTION CASTING SETSET--UPUP

Mould Pallet up-down mechanism

Top Clamp

Top Chamber

View Port

Bottom Clamp

Bottom Chamber

Control Panel


INJECTIONINJECTION--CAST & SWAGED URANIUM RODSCAST & SWAGED URANIUM RODS

LENGTH = 160 mm, DIAMETER = 4.67±0.04



COCO--SWAGED SSSWAGED SS--ROD WITH CLAD/LINERROD WITH CLAD/LINER

Zircaloy SS Clad

No Gap

Fuel



Thermophysical & compatibility studies of Thermophysical & compatibility studies of Metallic Fuels Metallic Fuels

Thermo physical Property

Evaluation Laboratory

has the following facilities:

1. Thermal conductivity

2. Specific heat

3. Coefficient of linear expansion

4. Chemical compatibility


Hardness profile along mechanically Hardness profile along mechanically bonded U/bonded U/ZrZr 4/D4/D--9 composite9 composite

0 100 200 300 400 500 600 700 800 900 10000

50

100

150

200

250

300

350

400

450

U metal Zr-4 D-9 cladHAR

DN

ESS(

Kg/

mm

2 )

DISTANCE FROM D-9 OUTER EDGE(μm)

UZr-4D9



SEM of U-Zr-D9 Metallic fuel pin after out-of-pile chemical compatibility test showing no reaction layer at the interfaces.

[thicknesses of ‘Zr’ barrier layer:150μm and D9 cladding material: 450μm ].

a b c(a) As received (b) 650C, 500 hrs; (c) 650C, 1500hrs

UZr

D9 U

U

ZrZr

D9D9

Chemical Compatibility StudiesChemical Compatibility StudiesHomi Bhabha

Centenary Year


880C, 15hrs.

onset of reaction

880C, 15hrs.

Chemical Compatibility StudiesChemical Compatibility StudiesHomi Bhabha

Centenary Year


ZrZr

T91T91

U U

U U ZrZr U U

ZrZr

700oC

500hrs 1000hrs 1500hrs

T91T91

ZrZr

U U T91T91ZrZrU U T91T91 ZrZr

U U

650oC

UU

OutOut--ofof--pile Chemical compatibility studies [ U/Zr/T91]pile Chemical compatibility studies [ U/Zr/T91]



0 200 400 600 800 1000

0.0

0.5

1.0

1.5

2.0

2.5

3.0

Uranium (Radial) D9 (Radial) Zr (Radial) T 91

Ther

mal

Exp

ansi

on (%

)

Temperature ( 0C )

Thermal expansion values of U, Thermal expansion values of U, ZrZr, D, D--9 and T9 and T--91 91



Thorium based MOX Fuels Thorium based MOX Fuels in Nuclear Power in Nuclear Power

a) Prominent Role in Future Nuclear Power Programme.

b) Non-Proliferation Aspects.c) Spent Fuel – Almost a Stable Waste

Form.d) Lower Minor Actinide Generation. e) Lower Reactivity Change with Burn-up.f) Abundance of Resource.



Three times more abundant than uranium

Better Performance Characteristics• Higher melting point

• Better thermal conductivity

• Lower fission gas release

• Good radiation resistance and dimensional stability

Better Chemical Stability• Reduced fuel deterioration in the event of failure• No oxidation during permanent disposal in repository• Poses problem in dissolution during reprocessing

Attributes of ThoriumAttributes of ThoriumHomi Bhabha

Centenary Year


0

100

200

300

400

500

0 5 10 15Burnup (GWd/t)

Bun

dle

Pow

er (k

W)

Low Flux

Medium Flux

High Flux

PHWR PHWR ThoriaThoria BundlesBundles

Reactor No. of bundles

Madras- I 4Kakrapar-I 35Kakrapar-II 35Rajasthan- II 18Rajasthan -III 35Kaiga-II 35Rajasthan-IV 35Kaiga-I 35



Irradiation Experience Of (Irradiation Experience Of (ThTh--PuPu) MOX) MOXin PWL, CIRUSin PWL, CIRUS

Cluster Fuel Clad type GWd/t kW/m

AC-6 (Th-4%Pu)O2

Free standing(TAPS-BWR) 18.5 40

BC-8(Th-6.75%Pu)O2

ThO2

Collapsible(PHWR) 10.2 42


Major Design Objectives• Power output – 300 MWe with 500

m3/d of desalinated water.• A large fraction (65%) of power

from thorium. • Extensive deployment of passive

safety features – 3 days grace period, and no need for planning off-site emergency measures.

• Design life of 100 years.• Easily replaceable coolant

channels.

AHWR is a vertical pressure tube type, boiling light water cooled and heavy water moderated reactor using 233U-Th MOX and Pu-Th MOX fuel.

Advanced Heavy Water ReactorAdvanced Heavy Water Reactor

Salient Features of Pressure tube

• 120mm ID x 6300 mm length• Replaceable through top end-fitting• Unique shape by Pilgering route.• - Thicker at one end, tapering at

the other • Controlled cold work to achieve

required tensile properties.

Gravity driven water pool

Steam drumInclined fueltransfer machine

Fuel storage bay

Fuelling machine

Core


Thorium UtilizationThorium Utilization--Initiated by Initiated by PuPu

Advanced Heavy Water Reactor: 300 MWe

Equilibrium Core:Fuel cluster containing 54 pins arranged in 3 rings of 12, 18 and 24 pinsThO2-3.25%PuO2 (outer 24 pins)ThO2-3.75%U233O2 (18 pins)ThO2-3.00%U233O2 (12 pins)Initial core will have fuel assemblies containing all (Th-Pu)MOX pins.

Displacer Rod

(Th-Pu)O2(24 pins)

(Th-U233)O2(30 pins)

Fuel Cluster Cross section

AHWRFuelCluster

Bottom tie plate

Top tie plate

Fuel pin Displacerrod

Water rod



Proliferation Resistance Proliferation Resistance Characteristics of AHWRCharacteristics of AHWR

1. High energy gamma from U233 daughters makes U232 handling difficult.

2. Reprocessing & re-fabrication needs high technology.

3. Denaturing of U233 is possible with U238.

Isotope Feed isotopic %

Discharge Isotopic %

U232U233U234U235U236

-92.466.930.600.04

>1000 ppm

8512.871.970.14

Isotope Feed isotopic %

Discharge Isotopic %

Pu239Pu240Pu241Pu242

68.724.65.261.30

6.550.2921.5021.70

U232U233U234U235

----

>0.192.466.930.606

(Th-Pu)MOX Pins (Th-U233)MOX Pins



AHWR Critical FacilityAHWR Critical Facility

Validate reactor physics design on various Fuel types.Reference core consists of 19 element U metallic fuel clusters - 65 Nos.Thoria, (Th-Pu) & (Th-U233)MOX fuel clusters to be introduced in phases for Physics study.Criticality achieved in April 2008.

Fuel for AHWR Critical Facility Built at Trombay to conduct lattice physics experiments for validation of AHWR



232U (68 years)α - decay

α - decay

β- - decay

α - decay

α - decay

α - decay

228Th (1.9116 years )

212Pb (10.64 hours )

224Ra (3.66days)

220Rn (55.6 seconds )

216Po( 0.145 second)

208Pb(stable)

212Bi (60.55 minutes )

208Tl ( 3.053 minutes) 212 Po (0.299 μ seconds)

35..94 %α - decay

64.06 %β - decay

β- - decay α - decay

UU232232 Decay ChainDecay Chain

Fabrication of (Th,233U)O2 fuel is difficultbecause it usually contains daughters of232U (half-life 73.6 years) namely, 212Bi and 208Tl, which emit strong gamma radiations.

Therefore the fabrication of the thorium fuel requires operations in the shielded glove-boxes to protect the operators from radiation

Difficulties in making (ThDifficulties in making (Th--U)OU)O22 FuelFuel

212Bi 0.7-1.8 MeV208Tl 2.6 MeV

Gamma Energies of 232U daughters



Fabrication Technology for Fabrication Technology for (Th(Th--UU233233) MOX) MOX

Sr.No Technique Flow sheet and

type of facility1 Powder-Pellet Route Fully Shielded Facility2 Sol-gel microsphere

pelletization Fully shielded facility

3 Impregnation (vacuum/microwave of gel/pellet)

Partially shielded facility

4 CAP process(Coated agglomerate

pelletization)Partially shielded facility


Impregnation SetImpregnation Set--up Inside the Glove Boxup Inside the Glove Box

Microstructure of ThO2-4%UO2Fuel made by Impregnation



ThO2 UO2

Co- Milling (Planetary ball mill)

Precompaction(140 MPa)

Granulation

Final Compaction(350 MPa)

Centreless grinding

Encapsulation(in zircaloy cladding tube)

SinteringAt 1923K in Ar-8%H2

ThO2 granules

Coating with UO2 powder

Final Compaction(350 MPa)

Centreless grinding

Encapsulation(in zircaloy cladding tube)

SinteringAt 1673K in Air

P/M route

CAP

FlowFlow--sheet for fabricationsheet for fabricationof (Th,U)Oof (Th,U)O22 pelletspellets


Fresh ThO2Granules

(precompaction)

Fresh ThO2Agglomerates

(Extrusion)

Fresh ThO2Microspheres

(So-gel Technique)

Coating/blendingwith U233 oxide/

Co-ppt/Master mix

Compaction in rotary press

Sintering in air

Encapsulation

Co-ppted ThO2-50% U3O8 or

pure U233 oxide

UnshieldedFacility

Shielded Facility

Flow sheet for (ThFlow sheet for (Th--UU233233)O)O22 pellet using CAP Processpellet using CAP Process

Duplex Duplex GrainstructureGrainstructure

The interface between two colonies of fine grains indicating that diffusion is complete between two agglomerates

Initial U3O8 coating area

Initial ThO2 particle

Initial ThO2 particle



Uranium distributionUranium distributionElemental scan for U Mα, ThMα, and O Kα for ThO2-3.75% UO2 pellet

O Kα

Th Mα

U Mα

Fine grains

Coarse grains

Wt % Wt %Th 84.770 84.236

U 3.120 3.656

O 12.110 12.108

Element

Atom distribution in coarse grains and fine grains for ThO2-4%UO2 pellet determined by EPMA



Rock in sand MicrostructureRock in sand Microstructure

The strength of the pellet at room temperature is related to grain size by the Hall- Petchrelation. Accordingly, the smaller grain sized pellets will have higher strength. But at high temperature, the grain boundaries become weaker than grain matrix. Since the pellets of smaller grain size have larger grain boundary areas, these pellets become softer than pellets with larger grain size. Also as the grain size decreases, the creep rate of the fuel increases. Therefore, pellets with smaller grain size have higher creep rate and better plasticity at high temperatures. These pellets will reduce the PCMI.

The basic requirementsfor the high perform-ance of a fuel are:

a) ”Soft pellets”– To reducePellet clad mechanicalinteraction (PCMI).

b) Large grain size –To reduce fission gas release(FGR)


Dec.1-2, 2008 IAEA TM, Fukui 781000 1200 1400 1600 1800

0

2

4

6

8

T.Jarvis

The

rmal

Con

duct

ivity

(W/m

K)

Temperature (K)

95% T.D. ThO2 RMD ThO2+2% UO2 RMD ThO2+4% UO2 RMD UO2 RMD

Thermal Conductivity Data of (Th,U)OThermal Conductivity Data of (Th,U)O22Generated at RMDGenerated at RMD


Recycle Fuels in INDIARecycle Fuels in INDIA

PHWR

FBTR

AHWR

Pu Bearing MOX Fuel Fabrication for Thermal & Fast ReactorsScope of work

(U~4%Pu) MOX for BWRs (TAPS)(U~0.4%Pu) MOX for PHWRs (KAPS)

First stage

(U-45%Pu) MOX for FBTR(U-29%Pu) MOX for PFBR(U-70%Pu) MC for FBTR(U-Pu-Zr ) Metal for FBR

Second stage

(Th-3%Pu) MOX - AHWR(Th-3%U233)MOX -AHWRThird stage



CLOSED FUEL CYCLE OPTIONCLOSED FUEL CYCLE OPTIONIS ESSENTIAL FOR INDIAIS ESSENTIAL FOR INDIA

“ Development of Nuclear Energy based on a closed fuel cycle enabling fuller use of Uranium & Thorium is the only way to meet the development aspirations of over a billion people while meeting the low emission norms.”



Thank YouThank You

homi bhabha centenary year oct. 30, 2008- recycle · pdf filerecycle fuel fabrication in india...

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