Pressurised Heavy Water 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 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
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 nuclear power 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 Neutrons 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 Nuclear Fuel 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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