Indian Nuclear Power Programme - Role of Thorium and the ... · A vital role for thorium in India India has a rather unique global position in nuclear resource. Has vast reserves

Indian Nuclear Power Programme - Role ofThorium and the Challenges Ahead

P.K. Vijayana, S. Basua and R K Sinhab

a Bhabha Atomic Research Centre, Mumbai, Indiab Department of Atomic Energy, OYC, Mumbai 400001, India

1

P.K. Vijayana, S. Basua and R K Sinhab

a Bhabha Atomic Research Centre, Mumbai, Indiab Department of Atomic Energy, OYC, Mumbai 400001, India

THEC15 - International Thorium Energy Conference,NUB, Mumbai, India, October 12-15, 2015

CONTENTS

Importance of thorium in Indian nuclear power programme

Indian experience with thorium fuel cycle Mining, Fuel Fabrication, Fuel Irradiation, Reprocessing, waste management and use of U233 fuel

Thorium based advanced reactors AHWR MSR

Other Thorium systems ADS HTR

Concluding remarks

Importance of thorium in Indian nuclear power programme

Indian experience with thorium fuel cycle Mining, Fuel Fabrication, Fuel Irradiation, Reprocessing, waste management and use of U233 fuel

Thorium based advanced reactors AHWR MSR

Other Thorium systems ADS HTR

Concluding remarks

2

A vital role for thorium in India

India has a rather unique global position in nuclear resource.Has vast reserves of thorium, but a modest reserve of uranium.

This not just provides a greater incentive for large-scale use ofthorium, but also calls for deployment of thorium based systemsmuch earlier than that contemplated by others.

The long-term programme is devised to suit this unique position.The key factors in it to ensure resource utilisation in asustainable manner are

Adoption of closed fuel cycle

Use of Breeder Reactors

Development of self-sustaining thorium based reactors

India has a rather unique global position in nuclear resource.Has vast reserves of thorium, but a modest reserve of uranium.

This not just provides a greater incentive for large-scale use ofthorium, but also calls for deployment of thorium based systemsmuch earlier than that contemplated by others.

The long-term programme is devised to suit this unique position.The key factors in it to ensure resource utilisation in asustainable manner are

Adoption of closed fuel cycle

Use of Breeder Reactors

Development of self-sustaining thorium based reactors

Three stage Indian nuclear power programme

Electricity

U fueledPHWRs

Pu FueledFast Breeders

Nat. U

Dep. U

Pu

Th

Th

U233 FueledReactors

Pu

U233

Electricity

Electricity

U233

44

Thorium utilisation forSustainable power programme

Pu

Stage 1Stage 1 Stage 2Stage 2 Stage 3Stage 3

Expanding power programmeBuilding U233 inventory

U233

Power generation primarily by PHWRBuilding fissile inventory for stage 2

(Th-233U) based systems constitute the third stage

TARAPUR-1&2 RAJASTHAN-1to 6 MADRAS-1&2 NARORA-1&2

PHWRsPHWRs

•• 1818 unitsunits underunder operationoperation

•• 44 xx 700700 MWeMWe –– underunder advancedadvanced stagesstagesofof constructionconstruction

•• MoreMore underunder planningplanning && constructionconstruction

LWRsLWRs

•• 22 BWRsBWRs underunder operationoperation

•• 11 VVERVVER underunder operationoperation andand 11 VVERVVERinin finalfinal stagestage ofof constructionconstruction


First Stage: Well established(Commercial Operation under NPCIL)

KAKRAPARA-1&2 KAIGA-1 to 4

TARAPUR 3&4

PHWRsPHWRs

•• 1818 unitsunits underunder operationoperation

•• 44 xx 700700 MWeMWe –– underunder advancedadvanced stagesstagesofof constructionconstruction


LWRsLWRs

•• 22 BWRsBWRs underunder operationoperation

•• 11 VVERVVER underunder operationoperation andand 11 VVERVVERinin finalfinal stagestage ofof constructionconstruction

•• MoreMore underunder planningplanning && constructionconstructionKK 1&2 (2x1000 MWe)

Second Stage : Well Underway

IGCAR, Kalpakkam established to focus on fast reactor programme

FBTR, Kalpakkam PFBR, Kalpakkam

Fast Breeder Test Reactor (FBTR) is under operation since 1985.

500 MWe Prototype Fast Breeder Reactor (PFBR) in final stages of construction.

Fast Reactor Fuel Cycle Facility (FRFCF) is under construction at Kalpakkam.

BHAVINI has initiated work on CFBRs.

R&D on metallic fuels in an advanced stage to achieve shorter doubling period.

Third Stage : Ground being prepared

In Third stage, U-233 obtained from FBRs of second stage will beused in self-sustaining reactors and will require only thorium feed.

Technology established for all aspects of thorium fuel cycle, albeitat laboratory scale.

A few U-233 based research reactors established – KAMINI is theonly one operating now.

AHWR designed and developed for industrial scale technologydemonstration AHWR critical facility established Several large scale test facilities established for design validation

MSBR – seems to be the most suitable candidate for the 3rd stage Self sustaining Electricity and hydrogen generation

In Third stage, U-233 obtained from FBRs of second stage will beused in self-sustaining reactors and will require only thorium feed.

Technology established for all aspects of thorium fuel cycle, albeitat laboratory scale.

A few U-233 based research reactors established – KAMINI is theonly one operating now.

AHWR designed and developed for industrial scale technologydemonstration AHWR critical facility established Several large scale test facilities established for design validation

MSBR – seems to be the most suitable candidate for the 3rd stage Self sustaining Electricity and hydrogen generation

7

Experience with Thorium in IndiaExperience with Thorium in India

8

Thorium resource and experience with thorium

India has among the largest resource of thorium in theworld. The beach sands of India contain rich deposits ofmonazite (thorium ore). About 11.93 million tonnes ofmonazite has so far been established, which contains about1.07 Million tonnes of Thorium oxide.

• Work has been carried out on all aspects of thorium fuel cycle.• Studies include mining, ore conversion, fuel fabrication,

irradiation in reactors, reprocessing, re-fabrication and wastemanagement .

Summary of thorium fuel cycle experience

Fabrication

J-rods of CIRUSThO2 fuel for DhruvaThoria fuel bundles for PHWRThoria fuel assemblies for FBTR blanket

Irradiation

CIRUS J-rod positionDhruva regular fuel locationPHWR initial core flux flatteningFBTR blanketExperimental thoria based MOX fuel pins

Irradiation

CIRUS J-rod positionDhruva regular fuel locationPHWR initial core flux flatteningFBTR blanketExperimental thoria based MOX fuel pins

Reprocessing J-rods of CIRUS at BARC & IGCARPHWR Thoria bundles in the new facility PRTRF

Utilisation of U-233 KAMINI plate type fuelIrradiation of PFBR Test fuel pins in FBTR

10

Mining & Extraction of Thorium

Environment Friendly Dredging Operation

Process of producing thoria from monazite has been establishedand several tonnes of nuclear grade thoria powder produced.

Thorium Extraction LaboratoryMonazite Processing Plant at OSCOM

0

100

200

300

400

500

0 5 10 15Burnup (GWd/t)

Bun

dle

Pow

er (k

W)

High Flux

Reactor No. ofbundles

Madras- I 4

Thoria bundles in Initial Core of PHWRs forflux flattening

Thoria Pellets

0

100

200

300

400

500

0 5 10 15Burnup (GWd/t)

Bun

dle

Pow

er (k

W)

Low Flux

MediumFlux

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

Test Irradiations and PIE of ThoriaBased MOX fuels

(Th-4%Pu) MOX fuel pins of BWR design (Th-6.75%Pu) MOX fuel pins of PHWR design (Th-1%Pu),(Th-8%Pu) & (Th-LEU)MOX fuel pins of AHWR design

13

Fission products & Uranium Isotopic measurements were used to validatereactor physics computer codes.

Examinations also showed better performance characteristics of thoria basedfuels in comparison to that of UO2.

Engineering Scale ReprocessingFacilities

Uranium Thorium Separation Facility(UTSF)

Thorium assemblies irradiated inresearch reactors have beenreprocessed in UTSF.

Power Reactor Thorium Reprocessing Facility(PRTRF)

Commissioned in January 2015 for reprocessingof thoria fuel bundles irradiated in PHWRs

Cell view of PRTRF

Laser Assisted FuelBundle DismantlingShielded glove boxes

Experience with 233U fuel

PURNIMA II

Experiments with uranyl nitrate solution

containing 233U reflected by BeO blocks.

PURNIMA III

Experiments were performed with 233U-Al

Dispersion plate type Fuel

These measurements helped in finalising the

core of KAMINI reactor.

KAMINI

A 30 KW reactor based on 233U fuel in the

form of U-Al alloy

The only operating reactor in the world with233U as fuel.

Fuel Irradiation testing in FBTR

PFBR design pin with (233U+ DU+ Pu) MOX

fuel was tested

PURNIMA II

Experiments with uranyl nitrate solution

containing 233U reflected by BeO blocks.

PURNIMA III

Experiments were performed with 233U-Al

Dispersion plate type Fuel

These measurements helped in finalising the

core of KAMINI reactor.

KAMINI

A 30 KW reactor based on 233U fuel in the

form of U-Al alloy

The only operating reactor in the world with233U as fuel.

Fuel Irradiation testing in FBTR

PFBR design pin with (233U+ DU+ Pu) MOX

fuel was testedKAMINI

Thorium Based Advanced ReactorsThorium Based Advanced Reactors

16

A road map for deployment of thorium basedreactors

Premature deployment of thorium in the Indian programme leads tosub-optimal use of the available resource.

Necessary to build-up a significant level of nuclear generationcapacity and fissile material before launching thorium cycle in thethird stage

Incorporation of thorium in the blankets of metallic fuelled fastbreeder reactors – when significant FBR capacity is built-up

233U formed in these FBRs could drive self-sustaining MSRs

Thorium based self-sustaining reactors are expected to be deployedbeyond 2070

AHWR – A Thorium fuel cycle demonstration reactor

Premature deployment of thorium in the Indian programme leads tosub-optimal use of the available resource.

Necessary to build-up a significant level of nuclear generationcapacity and fissile material before launching thorium cycle in thethird stage

Incorporation of thorium in the blankets of metallic fuelled fastbreeder reactors – when significant FBR capacity is built-up

233U formed in these FBRs could drive self-sustaining MSRs

Thorium based self-sustaining reactors are expected to be deployedbeyond 2070

AHWR – A Thorium fuel cycle demonstration reactor17

1980 1995 2010 2025 2040 2055 20700

100

200

300

400

500

600

Insta

lled

capa

city

(GW

e)

Year

Optimising the use of Domestic NuclearResources

Further growthwith thorium

Premature introduction ofthorium hampers the growth

Detailed calculations have shown that thorium can bedeployed on a large scale about three decades afterintroduction of FBRs with short doubling time

1980 1995 2010 2025 2040 2055 20700

100

200

300

400

500

600

Insta

lled

capa

city

(GW

e)

Year

Power profileof PHWRprogramme

Growth withPu-U FBRs

Further growthwith thorium

Advanced Heavy Water Reactor (AHWR)

AHWR is being set up as a technology demonstration reactorkeeping in view the long term deployment of Thorium basedreactors in the third phase.

The reactor design also addresses many key issues required forsustainable development of nuclear energy like Enhanced safety Proliferation concern Minimise waste burden Maximise resource utilisation

AHWR is based on a relatively more mature technology of solid-fuelbased water-cooled-reactor-system. This will be an important step inachieving long-term sustainability with thorium in molten salt reactorsand accelerator driven systems.

AHWR is being set up as a technology demonstration reactorkeeping in view the long term deployment of Thorium basedreactors in the third phase.

The reactor design also addresses many key issues required forsustainable development of nuclear energy like Enhanced safety Proliferation concern Minimise waste burden Maximise resource utilisation

AHWR is based on a relatively more mature technology of solid-fuelbased water-cooled-reactor-system. This will be an important step inachieving long-term sustainability with thorium in molten salt reactorsand accelerator driven systems.

Thorium has been proposed to be used worldwide inmany reactor systems (1/2): LWRs and HWRs

Ref.No.

Reactor type CoolantCoreinlettemp.

Coreoutlettemp.

PowerApplication Country

Fraction ofpower fromTh

MWth MWe

1. CANDU-SCWR

Superciticalwater(H2O)

367 C 597 C 2540>44%

eff.Electricity Canada -

2. RBWR-AC (BWR) Water283 C(7.25MPa)

Voidfraction:75%

3926 1356 Electricity USA -

3. APR-1400 (PWR) Water 290.6 C 323.9 C 3983 1455 ElectricityRep. ofKorea

-

20

3. APR-1400 (PWR) Water 290.6 C 323.9 C 3983 1455 ElectricityRep. ofKorea

-

4. PWR Water 293 C 15.5 MPa 3587 - Pu burning USA -

5.PHWR(SSET)

D2O 249 C 297 C 756 220 Electricity India -

6.AHWR-Pu

Water 259 C 285 C 920 300 Electricity India60%

AHWR-LEU 37.5%

7.BWR (LEU-Th)

Water - - 2027 658 Electricity India28.25%

BWR (Pu(RG)-Th) 27.5%

8.PWR (LEU-Th)

Water - - 3000 1000 Electricity India28.5%

PWR (Pu(RG)-Th) 18%

Advanced Heavy Water Reactor (AHWR)

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

CALANDRIA

STEAM DRUM

INCLINED FUELTRANSFER MACHINE

FUELLINGMACHINE

GRAVITY DRIVENWATER POOL (GDWP)

•Power level- 300 Mwe•Coolant- Boiling light water•Heat removal by natural circulation•Moderator: Heavy water• Lattice pitch: 225 mm square•Number of coolant channels: 452•Two independent functionally diverse

shut down systems•Passive Poison Injection directly into

moderator as an additional shutdownsystem

REACTOR BUILDING FUEL BUILDING

Design Objectives1. Thorium utilisation & Energy

Security2. Incorporation of Passive Safety

Systems3. Plant location in a populated area4. Electric Power output – 300 MWe5. Design life of 100 years

•Power level- 300 Mwe•Coolant- Boiling light water•Heat removal by natural circulation•Moderator: Heavy water• Lattice pitch: 225 mm square•Number of coolant channels: 452•Two independent functionally diverse

shut down systems•Passive Poison Injection directly into

moderator as an additional shutdownsystem

Test Facilities for AHWR Design Validation

HPNCLModerator & liquidpoison distribution

PCCTF

U-duct withtransparent sectionU-duct withtransparent sectionPCITF

PCLFPTIL/CHIL

IC

SD

JC

22

SDTF

FCS

BFST

HEADER

PBC

SFP

JC SB

AA

AHWR Critical FacilityPARTH, TarapurBuilding housing ITL-2

ITL

Several test facilities have been setup for AHWR design validation. Some ofthese are devoted to the study of specific phenomena. Major test facilitiesinclude 3 MW BWL, ITL and AHWR critical facility. The PARTH facility at R&DCentre Tarapur is the latest of these which also includes ATTF & FMTF.

FMTF

Fuel cycle for AHWR

Closed fuel cycle to maximise energy generation from thoria

Recycling of self-generated 233U and thoria

External fissile feed of plutonium

Closed fuel cycle to maximise energy generation from thoria

Recycling of self-generated 233U and thoria

External fissile feed of plutonium

Central Rod

(Th-233U) MOXFuel Pins

(Th- Pu) MOXFuel Pins

AHWR

FuelBuilding

Fresh Fuel forInitial Core /

Transition Core

Reactor Facilities

Fuel Fabrication Facilities

Reprocessing FacilitiesIn-CoreComponents

A fuel cycle facility co-located with the reactor is planned to facilitate recycle of 233U

Fuel fabrication, reprocessing & waste management

Shielded and glove-box types of facilities

AHWR Fuel Cycle Facility

(Th-Pu) MOXFuel Fab. Plant

FuelBuilding

AssemblyPlant

Reprocessing & Waste Management Plant

In-CoreComponents

(Th -233U) MOXFuel Fab. Plant

Design Goals of AHWR-EM

Power uprate to reduce the unit energy cost

Diverse cooling system to eliminate hydrogen generation and fuel

melt and emergency planning in public domain

Air cooled condenser as ultimate heat sink

Feed water heating and DM water production using solar power

Power generation from decay heat steam

Use of thermoelectric generators/solar PV to charge battery

Passive shutdown in case of wired shutdown system failure

Wireless transmission of safety significant parameters to BCR










25

Strategies for generation of U233

AHWR is loaded initially with Pu based MOX fuel only. The spent fuel from the AHWR is reprocessed to recycle 233U. There is a

gradual introduction of 233U based fuel in the reactor.

Transition from first criticality to equilibrium core configuration takes about15-20 years. ( i.e. when entire core has the standard composite cluster with (Th-Pu)MOX & (Th-233U)MOX fuels and core average burnup is ~20GWd/t)

The annual requirement of 233U during equilibrium core is about 85 kgs.

Studies have been carried out for generation of 233U in other reactors, whichcan be used in AHWR and the transition can be reduced to less than 10 years. Loading of thoria bundles in PHWRs Loading of thoria assemblies in FBTR/PFBR

233U generated can also be used to setup experimental/prototype MSRs to getexperience prior to large scale deployment.

26

AHWR is loaded initially with Pu based MOX fuel only. The spent fuel from the AHWR is reprocessed to recycle 233U. There is a

gradual introduction of 233U based fuel in the reactor.

Transition from first criticality to equilibrium core configuration takes about15-20 years. ( i.e. when entire core has the standard composite cluster with (Th-Pu)MOX & (Th-233U)MOX fuels and core average burnup is ~20GWd/t)

The annual requirement of 233U during equilibrium core is about 85 kgs.

Studies have been carried out for generation of 233U in other reactors, whichcan be used in AHWR and the transition can be reduced to less than 10 years. Loading of thoria bundles in PHWRs Loading of thoria assemblies in FBTR/PFBR

233U generated can also be used to setup experimental/prototype MSRs to getexperience prior to large scale deployment.

OptionsGeneration of U-233 in each

PHWR unit per year

220 MWe 540 MWe

2 Thoria + 10 Nat. U bundles inperipheral channels of the core

4 kg 9 kg

6 Thoria bundles + 6 SEU bundles in eachchannel of the entire core

9 kg 28 kg

Studies on Generation of U-233 in PHWRs

6 Thoria bundles + 6 SEU bundles in eachchannel of the entire core

9 kg 28 kg

(Th + Pu) MOX fuel bundles in the entirecore

150 kg 340 kg

Entire core loaded with fuel bundles having1 thoria pin and remaining SEU pins

8 kg 12 kg

Entire core loaded with fuel bundles having7 thoria pin and remaining SEU pins

55 kg 80 kg

27

Studies on generation of U-233 in FastReactors

Core Core

RadialBlanket

FBTR

In reflector(ring 9 to ring 12)

23 kg in 5 FPY

In reflector(ring 6 to ring 12)

46 kg in 5 FPY

PFBRIn Radial Blanket

(117 assemblies)105 kg per year

In Axial Blanket 38 kg per year

PFBR Core ArrangementFBTR Core Arrangement

Reduction in Transition PeriodCase :Initial 233U inventory of 200 kg & 40 kg/Year generation

Years No. of fuel clusters required annually for refueling 233U Inventory in kg(Initial Inventory= 200 kg &annually 40 kg/Year)(Th,Pu,233U) MOX (Th,Pu) MOX Total

1 0 0 0 200 + 40 – 00 = 240 kg

2 0 116 116 240 + 40 – 00 = 280 kg

3 0 92 92 280 + 40 – 00 = 320 kg

4 0 64 64 320 + 40 – 00 = 360 kg

5 0 48 48 360 + 40 – 00 = 400 kg

6 38 10 48 400 + 40 – 68 = 372 kg

7 48 372 + 40 – 86 = 326 kg7 48 0 48 372 + 40 – 86 = 326 kg

8 48 0 48 326 + 40 – 86 = 280 kg

9 48 0 48 280 + 40 – 86 = 234 kg

10 48 0 48 234 + 40 – 86 = 188 kg

11 48 0 48 188 + 40 – 86 = 142 kg

12 48 0 48 142 + 40 – 86 = 96 kg

13 48 0 48 96 + 40 – 86 = 50 kg

14 48 0 48 50 + 40 – 86 = 4 kg

15 48 0 48 4 + 40 + 96* – 86 = 36 kg

* 96 kg 233U is obtained from reprocessing of 48 fuel clusters discharged in 8th year

Transition period reduces by about 10 years in this case

Thorium Fuel Cycle Studies for AHWRThorium Fuel Cycle Studies for AHWR

30

Reprocessing Studies for Thoria basedFuels

Shielded glove boxes

Development of THOREX Process Flowsheet

• Lab-scale simulation studiescarried out to evolveprocess.

• PRTRF will provide vitaloperational experience andinputs for optimisation ofprocess parameters andequipmentsCell view of PRTRF

Fuel Fabrication Development

Buildup of daughter products in 1kg ofseparated U-233 having 500 ppm of U-232

• Methods more amenable forautomation and remote fabrication

212Bi : 0.7-1.8 MeV208Tl : 2.6 MeV

Fuel Type Technique Type of facility

(Th-LEU) MOX Powder-Pellet Conventional UO2

(Th-Pu) MOX Powder-Pellet Glove Box

(Th-233U)MOXPowder-Pellet Fully shielded facility

Sol-Gel MicrospherePelletization Fully shielded facility

Pellet Impregnation Partially shielded facility

Coated AgglomeratePelletization Partially shielded facility

Studies on Waste Management forThoria Fuel Cycle

• Management of thorium bearing liquid wastes.

• Concerns of thoria & fluoride during vitrification

• Partitioning of high level wastes as in uranium cycle

Shielded cubicle for processing ofthorium lean waste Actinide Partitioning Demonstration Facility

Thorium Based Sustainable Reactor Systems -MSR

Thorium Based Sustainable Reactor Systems -MSR

34

Indian Molten Salt Breeder Reactor(IMSBR)

IMSBR is an attractive option being studied for large scale thorium utilisation in third stage.

Reactor Vessel

BlanketIntermediate saltline

• Breeding possible over wide neutron spectrum.• Self- sustaining thorium reactor with minimal fissile inventory .• On-line FP removal and 233Pa separation helps in neutron economy and breeding.• Low pressure systems due to high boiling point• Hydrogen formation as well as core melt is eliminated• Less complications in fuel cycle activities from 232U formation

Pool Type Option

Secondary coolant saltinlet/ outlet

Blanket saltInlet/outlet

Reflectors

Primary HX

Core

To fuel saltreprocessing

Motor of primarycirculation pump

Sacrificial salt layer

Reactor Vessel

Core SafetyVessel

Blanket SaltCirculation Pump

Blanket Salt HeatExchanger

Blanket Salt Dump Line

Freeze Valves

Fuel SaltDump Line

Intermediate Salt DumpLine

Fuel SaltIntermediate Heat

Exchanger

Fuel SaltCirculation Pump

Intermediate saltline

Fuel Salt

Blanket Salt

Intermediate Salt

InnerVessel

Loop-in-Tank Option

• Breeding possible over wide neutron spectrum.• Self- sustaining thorium reactor with minimal fissile inventory .• On-line FP removal and 233Pa separation helps in neutron economy and breeding.• Low pressure systems due to high boiling point• Hydrogen formation as well as core melt is eliminated• Less complications in fuel cycle activities from 232U formation

Studies for MSRsAreas being addresseda) Development of closely coupled neutron transport and CFD codes with capability to

account for online reprocessing systemb) Large scale salt preparation and purificationc) Physical property characterisation for molten saltsd) Materials for use with molten salts and their codal qualificatione) Batch mode offline reprocessing method, without requiring cooling of fuel saltf) Instrumentation for operation in high temperature, high radiation, molten salt environmentg) Online chemistry control techniques

Molten Active Fluoride SaltLoop, for thermal hydraulic

studies on active fluoride saltsFacility for preparation andpurification of flouride salts

Areas being addresseda) Development of closely coupled neutron transport and CFD codes with capability to

account for online reprocessing systemb) Large scale salt preparation and purificationc) Physical property characterisation for molten saltsd) Materials for use with molten salts and their codal qualificatione) Batch mode offline reprocessing method, without requiring cooling of fuel saltf) Instrumentation for operation in high temperature, high radiation, molten salt environmentg) Online chemistry control techniques

Molten Salt NaturalCirculation Loop (MSNCL)

Molten Salt CorrosionTest Facility(MOSCOT)

Accelerator Driven sub-critical reactorSystems for Thorium Utilisation

Accelerator Driven Sub-critical reactor System(ADSS) is being developedfor Thorium utilisation as wellas for transmutation ofnuclear waste in dedicatedminor actinides burner reactorwith inherent safety againstpower excursions.

Development of high current high energy proton acceleratorStage-1: 30 mA 20 MeV Linac injector (LEHIPA)Stage-2: 1 GeV and 30 mA superconducting linac

ECR Ion Source Low Energy BeamTransmission line

RFQ

Target development studies related to heat removal andwindow damage (irradiation creep and void swelling)

Drift Tube Linac

37

Accelerator Driven Sub-critical reactor System(ADSS) is being developedfor Thorium utilisation as wellas for transmutation ofnuclear waste in dedicatedminor actinides burner reactorwith inherent safety againstpower excursions.

Target development studies related to heat removal andwindow damage (irradiation creep and void swelling)

Reactor Physics StudiesSubcritical test facility

LBE thermal-hydraulicexperimental test facility

High Temperature Reactor Programme

600 MWth, 1000 °C, TRISOcoated particle fuel

Combination of active andpassive systems for control &cooling

Pebble bed reactor concept withmolten Pb/ molten salt coolant

Medium life core Natural circulation of coolant for

reactor heat removal undernormal operation

Status: Reactor physics and thermal

hydraulic designs beingoptimised

Preliminary thermal andstress analysis carried out

Code under development forsimulating pebble motion

CentralReflector

De-FuellingChute

Side Reflector

BottomReflector

Core BarrelSupport

Fuellingpipe

CoolantOutlet

PebbleRetainingMeshPebbles andCoolant

CoolantInlet

ReactorVessel

Coolant

CentralReflector

De-FuellingChute

Side Reflector

BottomReflector

Core BarrelSupport

Fuellingpipe

CoolantOutlet


CoolantInlet

ReactorVessel

Coolant

ReactorVessel

Coolant

Compact High Temperature Reactor (CHTR)

Expansion Tank

Heat Exchanger

TRISO coating facility for fueldevelopment studies

Innovative High Temperature Reactor (IHTR)

Expansion Tank

38

600 MWth, 1000 °C, TRISOcoated particle fuel

Combination of active andpassive systems for control &cooling

Pebble bed reactor concept withmolten Pb/ molten salt coolant

Medium life core Natural circulation of coolant for

reactor heat removal undernormal operation

Status: Reactor physics and thermal

hydraulic designs beingoptimised

Preliminary thermal andstress analysis carried out

Code under development forsimulating pebble motion

CentralReflector

De-FuellingChute

Side Reflector

BottomReflector

Core BarrelSupport

Fuellingpipe

CoolantOutlet


CoolantInlet

ReactorVessel

Coolant

CentralReflector

De-FuellingChute

Side Reflector

BottomReflector

Core BarrelSupport

Fuellingpipe

CoolantOutlet


CoolantInlet

ReactorVessel

Coolant

ReactorVessel

Coolant

For commercial hydrogen generation 600 MWth, 1000 °C, TRISO coated particle fuel Pebble bed reactor concept with molten salt coolant Natural circulation of coolant

Heat Exchanger

Melt Tank

Sump

Heater

TRISO coating facility for fueldevelopment studies

Graphite compact developmentstudies with high packing fraction

Expansion Tank

Heater

Control Valve

Dump Tank

Cooler

(a)Lead-Bismuth eutectic loopfor thermal hydraulic studiesupto 550 C

(b)

Lead-Bismuth eutectic loopfor studies upto 1000 C,constructed of Nb-1%Zr-0.1%C – highesttemperature loop of its kindin the World

Kilo TemperatureLoop (KTL)

Liquid Metal Loop(Upto 550 C)

Concluding Remarks

The 3-stage INPP is a long term programme devised to utilise thelarge thorium resources in the country.

Brief overview of the 3-stage programme is given along with thecurrent status of the three stages.

India has developed technology for all aspects of thoriumutilisation, albeit on a lab scale.

Industrial-scale technology development and demonstration isexpected with the launch of AHWR.

The challenges of thorium fuel cycle has been described. The MSBR seems to be the most suitable candidate for the self-

sustainable thorium reactor. The conceptual design of IMSBR is briefly described along with its

advantages and challenges Work on HTRs is also briefly described.

The 3-stage INPP is a long term programme devised to utilise thelarge thorium resources in the country.

Brief overview of the 3-stage programme is given along with thecurrent status of the three stages.

India has developed technology for all aspects of thoriumutilisation, albeit on a lab scale.

Industrial-scale technology development and demonstration isexpected with the launch of AHWR.

The challenges of thorium fuel cycle has been described. The MSBR seems to be the most suitable candidate for the self-

sustainable thorium reactor. The conceptual design of IMSBR is briefly described along with its

advantages and challenges Work on HTRs is also briefly described.

39

Thank youThank you

40

41

Fuel Type Technique Type of facility

(Th-LEU) MOX Powder-Pellet Conventional UO2

(Th-Pu) MOX Powder-Pellet Glove Box

(Th-233U)MOXPowder-Pellet Fully shielded facility

Sol-Gel MicrospherePelletization Fully shielded facility

Pellet Impregnation Partially shielded facility

Coated AgglomeratePelletization Partially shielded facility

Advanced Fuel Fabrication Methods more amenablefor automation being studied

Coated AgglomeratePelletization (CAP)

Pellet Impregnation

Reprocessing Studies for thoria based fuels

Areas under study

DissolutionTHOREX for industrial scaleThird phase problemsThree component separation for irradiated (Th,Pu)MOXRemote handling technologies to tackle 232U associated gamma activity


• PRTRF will provide vitaloperational experience andinputs for optimisation ofprocess parameters andequipments

Development of THOREX Process Flowsheet


• PRTRF will provide vitaloperational experience andinputs for optimisation ofprocess parameters andequipments

Thorium has been proposed to be used worldwide inmany reactor systems (2/2): MSRs and LMRs

S.No.

Reactor type Coolant Coreinlettemp.

Core outlettemp.

Power ApplicationCountry

Fraction ofpower fromThorium

MWth MWe

9.MSFR/TMSR

Molten saltas fuel

N/A (750 C as meansalt temperature)

3000 1000 Electricity France

10. AMSTER 550 C 800 C 2250 1000Waste

management/Electricity

France

11. FUJI 567 C 707 C 450 200 Electricity Japan

12.ThorCon(DoAbleMSR)

564 C 704 C 550 250 Electricity USA ~25%

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ThorCon(DoAbleMSR)

13. GEM-STARMolten saltADS

650 C 750 C 500 220Biomass to

dieselUSA

14.MSTR -SPHINX

- - - -Waste

management/Electricity

Czech Rep.

15. AHTRMolten salt

650°C 700 C 3400 1500 Electricity USA

16. IHTR 600 C 1000 C 600 -Hydrogenproduction

India 45%

17. CHTR LBE 900 C 1000 C100kW

-Technology

demonstrationIndia 10%

18. IMSBRMolten saltas fuel

650 C 800C 1889 850 Electricity IndiaBeing

calculated

MSFR: Molten Salt Fast Reactor; TMSR: Thorium Molten Salt Reactor; AMSTER: Actinides Molten Salt transmuTER;GEM-STAR: Green Energy Multiplier - Subcritical Technology for Alternative Reactors; MSTR-SPHINX: Molten SaltTransmutation Reactor—. Spent Hot Fuel Incineration by neutron fluX)

Design Goals of AHWR-EM



















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Documents

Indian Nuclear Power Programme - Role of Thorium and the ... · A vital role for thorium in India India has a rather unique global position in nuclear resource. Has vast reserves