Srikumar Banerjee Homi Bhabha Chair Professor

Bhabha Atomic Research Centre Mumbai

W.B. Lewis Lecture

Pacific Basic Nuclear Conference , Vancouver, Canada

August 25, 2014

Pressurised Heavy Water Reactor (PHWR) Technology

-It’s relevance today

Bhabha-Lewis Friendship

Shared Views

• Importance of Neutron economy

• Natural Uranium – Heavy Water Reactor

• Use of Thorium in the long run

• Sustainable nuclear energy future

Concept of Valubreeder (U-Th cycle)

Thorium thermal neutron near-breeder

Net fissile fuel credit can exceed basic inventory charges

Natural U + Th + slightly enriched U or Pu

3

W.B. Lewis

(1908-1987)

Fissile u

ranium

/ T

otal ur

anium

(%)

U233

U235

Total Fissile Uranium

Burn up GWd/t

Indian three stage programme

U fueled

PHWRs

Pu FueledFast Breeders

Nat. U

Dep. U

Pu

Th

Th

U233 Fueled

ReactorsPu

U233

Electricity

Electricity

Electricity

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

PHWR FBTR AHWR

Thorium in the centre stage

Power generation primarily by PHWR

Building fissile inventory for stage 2

Expanding power programme

Building U233 inventoryThorium utilisation for

Sustainable power programme

U233

300 GWe-Year

42000

GWe-Year

155000

GWe-Year

Homi Jehangir Bhabha

(1909-1966)

Neutron Inventory : The Deciding Factor For Choosing Appropriate Fuel Cycle

Using external fissile material U235, Pu or an external accelerator driven neutron source,Th-U233 cycle can be made self sustaining

U233 has excellent nuclear characteristics both in thermal and fast neutron spectrum.

Th is excellent host

for Pu and enables deeper burning of Pu.

Neutron economy is directly related to fissile

economy since it takes just one neutron to

make one new fissile atom – W.B. Lewis (1966)

Pressurised Heavy Water Reactor (PHWR) Technology

Conceptual Development

Making a power reactor with natural uranium-best fissile utilisation per ton of mined uranium

Heavy water moderator

Channel type reactor

Continuous bi-directional fuelling

Versatility for use of different fuel cycle

6

Natural Uranium

Slightly Enriched Uranium

UO2 + PuO2 MOX

Th-U233

“If I had to pick one person who contributed most to the success of the Canadian nuclear power program it would be W.B. Lewis” – J.L. Gray (Nuclear Journal of Canada )

169 Tons N.U. 5 Kg, MA 16 7 Tons Dep.U 650 Kg Pu

PHWR , 1000 MWe, 7000 MWd/T Fast Reactor

1.25 Tons F.P.

Storage

165 Tons Dep U

25 Tons 4% En U 45000 MWd/T

6000 MWd/T of natural U

190 Tons N.U.

Enrichment Plant

23 Kg, MA 250 Kg Pu

Annual Fuel Requirement for PHWR and PWR for 1000 Mwe

1.15 Tons F.P.

23.6 Tons R.U.

Best Fissile Utilisation Per Ton of Mined Uranium

PWR

1000 MWe

Pressurised Heavy Water Reactor Deployment Worldwide

8

Country No. of PHWRs

Canada 19

India 18

South Korea 4

China 2

Argentina 1+2

Romania 2

Pakistan 1

Total 49

Tarapur 540 MWe

Douglas Point to Rajasthan

9

Douglas Point Reactor, First Criticality : 15 November, 1966

RAPS-1, First Criticality : August 11, 1972

Evolution of PHWR Technology in India for 220 MWe Units

10

Core Physics

• Continuous discharge model to Multiple bundle shift model. • Optimum positioning of reactivity devices to obtain their desired worths. • Extensive experimentation in Narora Reactor

Experience gained utilised in power optimisation , fuel performance and loading Thoria, MOX & SEU fuels

Control and Shut Down Systems

• Moderator level control replaced by Shim Rods • Moderator dump replaced by 2 fast & independent shutdown systems • Automatic Liquid Poison Addition / Passive Injection system to augment worth of shutdown

systems

Materials

• New Fabrication Route for Zr-2.5Nb Pressure Tubes • Introduction of Seamless Calandria Tubes • Continuation of Zr-Nb-Cu Garter Spring • Spot Welded Spacers & Pads

Containment Design

• Single Containment Double Containment Steam Generator Entry Ports Stainless Steel Lining

for 220 Mwe

(1 of 2) RAPS (Same as DPGS)

Moderator dumping

NAPS 220MWe onwards

Two independent Shut Down Systems

Mechanical Shut off Rods

Liquid Shut off Rods

11

Dump tank

CALANDRIA

Poison Tank

Liquid Poison Tube Guide Tube

Rod Drive Mechanism

Shut off rod

Control & Shut Down Systems

(2 of 2)

12

GRAB

CALANDRIA

LPIS

ALPAS

ALPAS: Automatic Liquid Poison Addition System GRAB : Gravity Addition of Boron LPIS : Liquid Poison Injection System

Shut Down & Control Systems for 540 & 700 MWe PHWRs

Independent Control of 14 Zones

2 Independent Shut Down Systems

14

Evolution of PHWR technology in India –

Seismic design

UN-WELDED CALANDRIA-ENDSHIELDS ASSEMBLY

WELDED CALANDRIA-ENDSHIELDS ASSEMBLY

FUELLING MACHINE WITH COLUMNS

FUELLING MACHINE WITH BRIDGE & COLUMNS

Evolution of PHWR technology in India – Containment

NAPS- Double Containment RAPS- Single Containment

Suppression pool

Double Containment standardised for all 220MWe and 540 MWe

Evolution of PHWR technology in India – Containment

Penetration for Steam Generator replacement

Suppression pool

Containment pressure suppression by Spray

Passive decay heat removal

Lined containment

Evolution of PHWR technology in India – Containment

Large thermal inertia in

Beyond Design Basis Scenario

Calandria Vault Water 680 Te

Moderator 260 Te Time for operator action

Moderator 20 hrs.

Calandria vault water

57 hrs.

Data for 540 MWe

24

Pressure tube ballooning

Moderator as heat sink

Why PHWRs are relevant today ?

• Inherent Safety

• Economy in Small Size

• Uninterrupted Operation

• Choice of Fuel Cycle Options

Performance & Economics of Indian Pressurised Heavy Water Reactors (PHWR)

Indian PHWRs

(700 MWe)

Global

Range

Capital Cost

(2008 Basis)

1700*

$/kWe

> 5000

$/kWe

Construction period

5-6 years 5-6 years

UEC $/MWh 60 80 - 100

Capital Cost of recently completed units

TAPS – 3 & 4

(540 MWe)

Rs. 58850* million

~ 1200* $/kWe

Kaiga – 3&4

(220 MWe)

Rs. 26000* million

~ 1300* $/kWe

* Cost figures pertain to the Indian domestic

context. In the international context these

figures will be location dependent.

Overall Capacity Factor (All Units) since fruition of International Cooperation

0

20

40

60

80

100

2009-10 2010-11 2011-12 2012-13

upto Oct.

Cap

acit

y Fa

cto

r (%

)

PHWR Performance in India

22

84

86

90

91

88

89

85

83

82

92

89

94

90

88

82

85

85

68

76

74

63

54

50

61

71

79

80

83

0

10

20

30

40

50

60

70

80

90

100

2000-01

2001-02

2002-03

2003-04

2004-05

2005-06

2006-07

2007-08

2008-09

2009-10

2010-11

2011-12

2012-13

2013-14

AF(%) CF(%)

712

451 415

342 320 319

RA

PS-

5

KA

PS-

2

KA

IGA

-1

KA

IGA

-3

NA

PS-

2

MA

PS-

1

Recent Continuous Operation

(Number of days)

Availability and

Capacity Factor

RAPS-5 has exceeded two years of continuous operation on August 3,

2014

Relevance of PHWRs in today’s context- Fuel versatility

UO2 + PuO2

UO2 + ThO2 (Thorium utilisation)

Nat.U + Enriched U

LWR + PHWR in tandem

23

Bundle Type Maximum burnup

(MWD/Te)

MOX-7 20000

Thorium 13000

Recycled Depleted

Uranium

9000

Natural Uranium 22000

SEU 25000

MOX 19-element Fuel Bundle

MOX Elements

Natural U Elements

Bundle Type No. of bundles

MOX-7 50

Thorium 250

SEU 50

Recycled Depleted

Uranium

Large

Comparison of behaviour of Thoria & Urania based fuels

UO2-4%PuO2 ThO2-4%PuO2 ThO2 fuel more

tolerant for clad failure!

• Inherently stable

• Single valency

• Lower diffusivity

• Better thermal conductivity

• Higher FG retention

Failed ThO2-4%PuO2 Failed UO2

800 1000 1200 1400 1600 18001.5

2.0

2.5

3.0

3.5

4.0

4.5

5.0

5.5

6.0

'k' fo

r 9

5%

TD

(W

/mK

)

Temperature (K)

(Th+4%U)MOX,BARC

(Th+4%U)MOX,Bakker

UO2

Fission Gas Bubbles & Channels in Irradiated Fuel

Higher thermal

conductivity of thoria

leads to lower operating

temperatures and lower

Fission Gas Release

No FG channels in

ThO2+4%PuO2

Burnup 4,400MWd/t Burnup15,000MWd/t

Burnup 18,400MWd/t

Fission Gas

channels in

UO2

PHWRs – Some concerns

Concerns

Small positive void coefficient

Tritium background inside containment

Opening of pressure boundary daily for on power refueling.

Zirconium alloy in pressure boundary

Solutions

Reduced lattice pitch

Burnable poison

Light water cooling Enriched fuel

High burn up fuel Reduced refueling

Proven performance of zirconium alloy as pressure boundary material

28

29

Indian Advanced Heavy Water Reactor (AHWR-Pu)

AHWR is a 300 MWe vertical pressure tube type, boiling light water cooled and heavy water moderated reactor using 233U-Th MOX and Pu-Th MOX fuel.

Design validation through extensive experimental programme. Pre-licensing safety appraisal by AERB Site selection in progress. AHWR Fuel

assembly

Major design objectives

65% of power from Th Void Coefficient negative

Several passive features

10 days grace period No radiological impact Additional Passive shutdown

system

Design life of 100 years.

Easily replaceable coolant

channels.

AHWR-Pu is a Technology demonstrator for the closed thorium fuel cycle

Bottom Tie Plate

Top Tie

Plate

Water Tube

Displacer

Rod

Fuel Pin

30

Passive Features of AHWR

• Passive core cooling by natural circulation

• Passive decay heat Removal by Isolation Condensers

• Passive shutdown system

• Passive ECCS injection by Accumulators & GDWP

• Passive Containment Coolers

31

Heat Removal Paths under Normal Operation & Shut down Condition and Passive Shut-down System

STEAM

CONDENSER

TO SEA

CORE

ISOLATION CONDENSERS(ICS)-8Nos.

FEED PUMP

HP & LP

(4Nos.)

DISCHARGE VALVE (CSDV)

DOWNCOMERTAIL PIPE

& REHEATERS

SEAWATER PUMP

DRUM

PASSIVE

RDLIQUIDPOISON

RD

CONDENSER STEAM

TURBINE

TO GDWP

EXTRACTION PUMP

VALVE

FEED HEATING& DEAERATION

HELIUM

PRESSURE

POISON

TANK

INLET HEADER

71 bar82 bar

58 bar

Pressure 70 bar

Pressure 71 bar Pressure 76.5 bar Pressure 82 bar

Steam overpressure can passively shut down reactor

32

Transients following station black-out and failure of wired shut-down systems

Peak clad temperature hardly rises even in the extreme condition of complete station blackout and failure of primary and secondary shutdown systems.

0 200 400 600 800 1000

500

550

600

650

700

10 sec delay

5 sec delay

2 sec delay

Cla

d T

em

pe

ratu

re (

K)

Time (s)

Clad Surface Temperature

Per Capita Electricity Consumption

In India 40% of households do not have access to electricity

Nuclear Power for Sustainable supply of Energy

– Indian Scenario

0

200

400

600

800

1000

2011-12 2016-17 2021-22 2026-27 2027-32

Capacity required for 8% Growth Capacity required for 9% growth

Renewables+Hydro Capacity Projection

Required Installed Capacity Projection (GW)

Meeting the balance requirement through fossil fuels (coal) would involve about 3 to 4 billion tonnes of CO2 emissions per annum

Source: Integrated Energy Policy, Report of the Expert Committee Planning Commission 2006

35

Adopting closed fuel cycle also reduces nuclear waste burden

Radiotoxicity of spent fuel is dominated by :

FPs for first 100 years.

subsequently, Pu (>90%)

After Pu removal

Minor Actinides specially Am (~ 9%)

Natural decay of spent fuel radiotoxicity

With early introduction of fast reactors using (U+Pu+Am) based fuel, long term raditoxicity of nuclear waste will be reduced.

200000 years

300 years

36

Fuel Cycle & Sustainability

“A development that meets the needs of the present without

compromising the ability of future generations to meet their own needs”

- Brundtland Commission 1987

We have an obligation towards our next generation

Important considerations for

sustainability are:

Environment

Clean concentrated energy

Resources

Fissile material

Fertile material

Neutrons

Long lived nuclear waste

Safety and security

Economy

ECONOMY

OPTIMUM REPOSITORY

MINIMUM

ENVIRONMENTAL IMPACT

RADIOTOXICITY REDUCTION

RESOURCE UTILISATION

CLOSED FUEL CYCLE

Reprocessing capacity matching with FBR fuel requirement

Vitrified HLW inventory presently stored in engineered interim storage facility

PHWR

U (Natural)

UOX Fuel

Heavy Metal

α Bearing HLW

Non α Waste (L&IL)

Indian Nuclear Fuel Cycle: Today

Volume Reduction Near Surface Disposal

Vitrification Interim Storage

Repository

PUREX

Reprocessing of Carbide Fuel

Mixed Carbide Fuel for Fast breeder Test Reactor

Mixed Oxide Fuel for Prototype Fast Breeder Reactor

Vitrification Facility, Trombay Vitrified Waste Storage Facility, Tarapur Reprocessing Facility, Tarapur

38

Performance evaluation of FBTR carbide fuel through Post Irradiation Examination

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

25 & 50 & 100 GWd/t burn-up

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

155 GWd/t - CENTRE of the fuel column

155 GWd/t – END of the fuel column

Clad Fuel-clad gap

• Excellent performance of Mixed Carbide Fuel canned in Stainless Steel Tubes upto a maximum burn up of 165 GWd/t

• Doubts raised on achieving high burn up in fast reactor fuel due to radiation damage and transmutation processes in the cladding material

There is also no news yet on PuC + UC but I would

expect it to be worse – W.B. Lewis (AECL-2584, 1966)

39

Indian Prototype Fast Breeder Reactor

2580

1000

Fig. 4 -- PFBR FUEL PIN ASSEMBLY (all dimensions in mm)

30

Bottom Blanket

Middle Plug

Bottom End Plug

710

300

10

Ø6.6

S.S.Wrapping wire

200

Spring support

Top Blanket

Fuel

5300

Top End Plug25

Spring

Core-1 (85)

Core-2 (96)

Radial Blanket (120)

SS Reflector (138)

B4C (78) -Not all shown

CSR- (9)

DSR- (3)

Core Layout

Radial Blanket (U238/Th232)

Thermal Power (MWth) : 1250 Electrical output (MW): 500 Fuel material : (U,Pu)O2 Coolant : Molten Sodium

Axial Blanket

Core-1 Core-2

Prototype Fast Breeder Reactor, Kalpakkam

• Physical progress of the Project : >95%

Induction of LWRs of Large Capacity (> 1000 MWe)

• Power Parks from LWRs (6-8 units each. Total ~ 40 GWe)

• Progressively growing equipment supply chain from within India

• Development of Indigenous LWR

• Tying up Fuel supply

VVERs at Kudankulam

LWR

FR

PUREX MA partitioning

U (enriched)

MOX Fuel

MOX Fuel

Heterogeneous Recycling

Storage ~ 100 years

Pu

Pu Am, Np

Irradiated Pins (high specific activity, short life)

PHWR U (Natural)

Pu

Evolving Fuel Cycle

U (depleted)

U ( ~ 1.2% Enriched)

Pu

Cm

FPs

Disposal

Recovery of FPs

Oxide Fuel in Fast

Reactors

• Long doubling time

• Burning MA will

Increase doubling time

Fuelling Scheme Initial fuel: Nat. U & Th Normal refuelling of U bundles (~ 7 GWd/t) Th will reside longer

233 U generation adds reactivity Compensate by replacing some U by Th

Th increases and U decreases Ultimately fully Th core

In situ breeding and burning Th

Thorium utilisation in PHWR -ADS: Once through mode

1-2 GeV Proton Accelerator

Spallation Target

Subcritical Reactor

Advantages Use of natural fuels only 140 tons U consumption during reactor life High burnup of Th ~ 100 GWd/t

Disadvantage Low keff ~0.9 and gain nearly 10 with Pb target Accelerator power ~ 60 MW for a 200 MWe ADS

Rediscovering PHWRs

Sustained energy production by effective utilisation of both fissile & fertile elements.

Versatile fuelling options

Technology for world wide users New entrants

Utilisation of reprocessed fuel from LWRs

Possibility of India-Canada co-operation in spreading PHWR technology

44

Thanks for your attention