Pressurised Reactor (PHWR) Technology -It’s relevance today

Srikumar Banerjee Homi Bhabha Chair Professor Bhabha Atomic Research Centre Mumbai W.B. Lewis Lecture Pacific Basic Nuclear Conference , Vancouver, Canada August 25, 2014

Bhabha-Lewis Friendship

Shared Views • Importance of 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

U233 U235 Total Fissile Uranium W.B. Lewis

(1908-1987) Fissile uranium / Total uranium (%)uranium / uraniumTotalFissile 3 Burn up GWd/t Indian three stage programme Thorium in the centre stage

Homi Jehangir Bhabha PHWR FBTR AHWR (1909-1966) 300 GWe-Year Th U fueled Nat. U Electricity 42000 Th PHWRs GWe-Year Dep. U Electricity Pu Fueled 155000 Fast Breeders GWe-Year Pu U233 U233 Fueled Electricity Reactors Pu U233

Power generation primarily by PHWR Expanding power programme Thorium utilisation for Building fissile inventory for stage 2 Building U233 inventory Sustainable power programme Stage 1 Stage 2 Stage 3 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 Neutron economy is directly related to fissile deeper burning of Pu. 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 Natural Uranium Slightly Enriched Uranium “If I had to pick one person who UO2 + PuO2 MOX contributed most to the success of the Canadian 233 program it would be W.B. Lewis” – J.L. Th-U Gray (Nuclear Journal of Canada ) 6 Best Fissile Utilisation Per Ton of Mined Uranium

PHWR , 1000 MWe, 7000 MWd/T Fast Reactor

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

Annual Fuel Requirement 1.25 Tons F.P. for PHWR and PWR for Storage 1000 Mwe 165 Tons Dep U 1000 MWe 23 Kg, MA PWR 250 Kg Pu 190 Tons N.U. 23.6 Tons 25 Tons 4% En U R.U. 45000 MWd/T 1.15 Tons F.P. Enrichment 6000 MWd/T of natural U Plant Pressurised Heavy Water Reactor Deployment Worldwide

No. of Country PHWRs

Canada 19 India 18 South Korea 4 China 2 Argentina 1+2 Romania 2 Pakistan 1 Total 49 Tarapur 540 MWe 8 Douglas Point to Rajasthan

Douglas Point Reactor, First Criticality : 15 November, 1966

RAPS-1, First Criticality : August 11, 1972

9 Evolution of PHWR Technology in India for 220 MWe Units  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 10 for 220 Mwe

 RAPS (Same as DPGS) (1Rod of Drive 2) Moderator dumping Mechanism  NAPS 220MWe onwards Shut off rod Liquid Poison Two independent Shut Down Guide Tube Tube Systems . Mechanical Shut off Rods . Liquid Shut off Rods CALANDRIA

Poison Tank

Dump tank

11 Control & Shut Down Systems

(2 of 2) ALPAS: Automatic Liquid Poison Addition System GRAB : Gravity Addition of Boron LPIS : Liquid Poison Injection System

GRAB

CALANDRIA

LPIS

ALPAS 12 Shut Down & Control Systems for 540 & 700 MWe PHWRs

2 Independent Shut Down Independent Control of 14 Systems Zones Evolution of PHWR technology in India – Seismic design

UN-WELDED CALANDRIA-ENDSHIELDS ASSEMBLY FUELLING MACHINE WITH COLUMNS

WELDED CALANDRIA-ENDSHIELDS ASSEMBLY FUELLING MACHINE WITH BRIDGE & COLUMNS

14 Evolution of PHWR technology in India – Containment

Suppression pool

RAPS- Single Containment NAPS- Double Containment Evolution of PHWR technology in India – Containment Penetration for Steam Generator replacement

Suppression pool

Double Containment standardised for all 220MWe and 540 MWe Evolution of PHWR technology in India – Containment

Containment pressure Lined containment suppression by Spray

Passive decay heat removal

Why PHWRs are relevant today ? Large thermal inertia in • Inherent Safety Beyond Design Basis Scenario • Economy in Small Size • Uninterrupted Operation  Pressure tube ballooning • Choice of Fuel Cycle Options  Moderator as heat sink

Moderator Time for operator action 260 Te

Moderator 20 hrs.

Calandria vault 57 hrs.

water Calandria Vault Data for 540 MWe Water 680 Te

24 Performance & Economics of Indian Pressurised Heavy Water Reactors (PHWR) Indian Global PHWRs Range (700 MWe) Capital Cost 1700* > 5000 (2008 Basis) $/kWe $/kWe Construction 5-6 years 5-6 years period UEC $/MWh 60 80 - 100

Capital Cost of recently completed units Overall Capacity Factor (All Units) since fruition of International Cooperation TAPS – 3 & 4 Rs. 58850* million (540 MWe) ~ 1200* $/kWe 100 80 Kaiga – 3&4 Rs. 26000* million 60 (220 MWe) ~ 1300* $/kWe 40

20 * Cost figures pertain to the Indian domestic Capacity(%)Factor context. In the international context these 0 2009-10 2010-11 2011-12 2012-13 figures will be location dependent. upto Oct. PHWR Performance in India

AF(%) CF(%)

94

100

92

91

90 90

89 89

88 88

86

85 85 85

84

83 83

90

82 82

80

79

76

80 74

71

68 63

Availability and 70

61

Capacity Factor 60 54 50 50 40 30 20 10 0 2000- 2001-2002-2003-2004- 2005- 2006- 2007- 2008- 2009- 2010- 2011- 2012- 2013- 01 02 03 04 05 06 07 08 09 10 11 12 13 14

712 Recent Continuous Operation (Number of days)

451 415

342 320 319

RAPS-5 has exceeded two years of

continuous operation on August 3,

2 2

5 5 2 1

1 1 3

-

- - -

- -

22 2014

KAPS

RAPS

NAPS

MAPS

KAIGA KAIGA Relevance of PHWRs in today’s context- Fuel versatility

Bundle Type Maximum burnup  UO2 + PuO2 (MWD/Te) MOX-7 20000  UO2 + ThO2 (Thorium utilisation) Thorium 13000 Recycled Depleted 9000  Nat.U + Enriched U Uranium  LWR + PHWR in tandem Natural Uranium 22000 SEU 25000

MOX Elements Bundle Type No. of bundles

MOX-7 50 Thorium 250 Natural U SEU 50 Elements Recycled Depleted Large Uranium MOX 19-element Fuel Bundle 23 Comparison of behaviour of Thoria & Urania based fuels

• Inherently stable • Single valency • Lower diffusivity • Better thermal conductivity • Higher FG retention

ThO -4%PuO 2 2 ThO2 fuel more UO2-4%PuO2 tolerant for clad failure!

6.0 (Th+4%U)MOX,BARC 5.5 (Th+4%U)MOX,Bakker UO 5.0 2 4.5 4.0 3.5 3.0 2.5

'k' for 95% TD (W/mK) TD 95% 'k'for 2.0 1.5 800 1000 1200 1400 1600 1800 Failed UO Failed ThO2-4%PuO2 Temperature (K) 2 Fission Gas Bubbles & Channels in Irradiated Fuel

Fission Gas channels in

UO2

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

Higher thermal conductivity of thoria No FG channels in leads to lower operating ThO2+4%PuO2 temperatures and lower Fission Gas Release

Burnup 18,400MWd/t PHWRs – Some concerns

Concerns Solutions  Small positive void coefficient  Reduced lattice pitch  Tritium background inside  Burnable poison containment  Light water cooling  Opening of pressure boundary . Enriched fuel daily for on power refueling.  High burn up fuel  Zirconium alloy in pressure . Reduced refueling boundary  Proven performance of zirconium alloy as pressure boundary material

28 Indian Advanced Heavy Water Reactor (AHWR-Pu)

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

Major design objectives

 65% of power from Th

 Void Coefficient negative Top Tie

PlateDisplacer Water Tube Rod

 Several passive features  10 days grace period Fuel Pin

 No radiological impact Design validation through  Additional Passive shutdown extensive experimental system programme.

Pre-licensing safety  Design life of 100 years. appraisal by AERB  Easily replaceable coolant Bottom Tie Plate Site selection in progress. channels. AHWR Fuel assembly AHWR-Pu is a Technology demonstrator for the closed thorium fuel cycle 29 Passive Features of AHWR

• Passive core cooling by natural circulation • Passive shutdown system • Passive decay heat Removal by Isolation • Passive ECCS injection by Accumulators & Condensers GDWP • Passive Containment Coolers 30

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

TO GDWP

82 bar CONDENSER STEAM 71 bar DISCHARGE VALVE (CSDV) Steam ISOLATION CONDENSERS (ICS)-8Nos. overpressure HP & LP TURBINE can passively & REHEATERS shut down reactor STEAM PressurePressure DRUM (4Nos.) 8276.57170 barbar bar TO SEA HELIUM FEED PUMP SEAWATER PUMP PRESSURE

RD CONDENSER POISON DOWNCOMER TANK TAIL PIPE

FEED HEATING INLET HEADER & DEAERATION

PASSIVE VALVE 58 bar CORE

RD LIQUID POISON EXTRACTION PUMP

31 Transients following station black-out and failure of wired shut-down systems 700 Clad Surface Temperature

10 sec delay Peak clad temperature 650 5 sec delay hardly rises even in the 2 sec delay extreme condition of complete station blackout and failure of primary and 600

secondary shutdown systems.

Clad TemperatureClad(K) 550

500 0 200 400 600 800 1000 Time (s)

32 Per Capita Electricity Consumption

In India 40% of households do not have access to electricity Nuclear Power for Sustainable supply of Energy – Indian Scenario

Required Installed Capacity Projection (GW)

1000

800 Meeting the balance requirement through 600 fossil fuels (coal) would involve about 3

400 to 4 billion tonnes of CO2 emissions per annum 200

0 2011-12 2016-17 2021-22 2026-27 2027-32

Capacity required for 8% Growth Capacity required for 9% growth Renewables+Hydro Capacity Projection

Source: Integrated Energy Policy, Report of the Expert Committee Planning Commission 2006 Adopting closed fuel cycle also reduces nuclear waste burden

Radiotoxicity of spent Natural decay of spent fuel fuel is dominated by : radiotoxicity FPs for first 100 years. 200000 years subsequently, Pu (>90%)

After Pu removal

Minor Actinides specially Am (~ 9%)

300 years

With early introduction of fast reactors using (U+Pu+Am) based fuel, long term raditoxicity of nuclear waste will be reduced35 . 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

Important considerations for sustainability are: CLOSED FUEL CYCLE  Environment  Clean concentrated energy  Resources  Fissile material

 Fertile material MINIMUM ENVIRONMENTAL  ECONOMY IMPACT RESOURCE UTILISATION  Long lived nuclear waste  Safety and security OPTIMUM REPOSITORY RADIOTOXICITY  Economy REDUCTION

We have an obligation towards our next generation 36 Indian Cycle: Today

U (Natural) Reprocessing of Carbide Fuel

UOX PHWR Fuel PUREX Mixed Carbide Fuel for Fast breeder Test Reactor Non α Waste Heavy Metal

(L&IL) Mixed Oxide Fuel for α Bearing Prototype Fast Breeder Reactor HLW

Volume Reduction Vitrification Repository Near Surface Disposal Interim Storage

Reprocessing Facility, Tarapur Vitrification Facility, Trombay Vitrified Waste Storage Facility, Tarapur Reprocessing capacity matching with FBR fuel requirement

Vitrified HLW inventory presently stored in engineered interim storage facility Performance evaluation of FBTR carbide fuel through Post Irradiation Examination

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

Clad Fuel-clad gap

Micrographs of fuel pin cross section at the centre of fuel column after 25 & 50 & 100 GWd/t burn-up 155 GWd/t - CENTRE of the fuel column • 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 155 GWd/t – END expect it to be worse – W.B. Lewis (AECL38 -2584, 1966) of the fuel column Indian Prototype Fast Breeder Reactor

Top End Plug Core-1 (85) 25

Core-2 (96)

Radial Blanket (120) 200 SS Reflector (138) Spring

B4C (78) -Not all shown

5 Spring support CSR- (9)

DSR- (3)

Top Blanket 300

Core Layout S.S.Wrapping wire Core-1

Fuel

Core-2 Ø6.6 1000 Radial Blanket 2580 (U238/Th232) Axial Blanket

Bottom Blanket 300

10 Middle Plug 710

Thermal Power (MWth) : 1250 Electrical output (MW): 500 30 Bottom End Plug

Fuel material : (U,Pu)O2 Coolant : Molten Sodium Fig. 4 -- PFBR FUEL PIN ASSEMBLY (all dimensions39 in mm) 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 Evolving Fuel Cycle

U (Natural) PHWR U ( ~ 1.2% Enriched) Disposal

U (enriched) LWR MOX Fuel U (depleted) Recovery of FPs Pu FPs PUREX MA partitioning MOX Fuel FR

Cm Heterogeneous Recycling Storage ~ 100 years Am, Np Pu Irradiated Pins Pu (high specific activity, short life) Pu Oxide Fuel in Fast Reactors • Long doubling time • Burning MA will Increase doubling time Thorium utilisation in PHWR -ADS: Once through mode

Advantages Spallation Use of natural fuels only Target 140 tons U consumption during reactor life 1-2 GeV High burnup of Th ~ 100 GWd/t Proton Disadvantage Accelerator Low keff ~0.9 and gain nearly 10 with Pb target Subcritical Accelerator power ~ 60 MW for a Reactor 200 MWe ADS

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 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

Thanks for your attention

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