technology development, design and safety features of phwr and their operating performance
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7/29/2019 Technology Development, Design and Safety Features of PHWR and Their Operating Performance
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U.C.MuktibodhNuclear Power Corporation of India Limited
Workshop on
Technology Assessment of SMRs forNear Term Deployment
Dec 5th 9th , 2011IAEA Headquarters,
Vienna, Austria
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Lecture Outline
Technology Development
Design features of 220, 540 & 700 MWe Indian
PHWRs
Safety features of 220, 540 & 700 MWe IndianPHWRs
Operating Performance of Indian PHWRs
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Technology
Development
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Launch of Nuclear Power Program
1964
Construction
work at First
NPP Began
1948
Atomic
EnergyCommission
1954
Department of
Atomic Energy
Bhabha
Atomic
ResearchCentre
Research
Reactor
APSARA
1956
Training
School
(Nuclear
Science &Technology)
Research
Reactor
CIRUS
1957 1960
Before setting up the first NPP, we had the basic infrastructure Policy,
Knowledge Base, Research Reactors, Radiation Protection, Human Resources
and since then moving continuously
and moving responsibly
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PHWR Program
1970s
TECHNOLOGY
DEMONSTRATION
1980s
INDIGENISATION
1980s
STANDARDISATION
1990s
CONSOLIDATION
2000s
COMMERCIALISATION
ECONOMY
OF SCALE
RAPP-3&4 KAIGA-3&4
RAPS-1&2 MAPS-1&2 NAPS-1&2 KAPS-1&2 KGS-1&2 RAPP-5&6
TAPS-3&4
220 MWe
540 MWe
700 MWe Reactors
2005-2006
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Development of Nuclear materials Mining and processing of nuclear fuel Uranium and
Thorium were developed.
Fabrication of all types of fuel required for reactors
Production of Heavy Water
Back end technology of Waste Management
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Calandria- Reactor pressure vessel for PHWR
Development of Manufacturing Technologyfor Class-1 Components
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REACTOR END SHIELD
Development of ManufacturingTechnology for Class-1 Components
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Development of ManufacturingTechnology for Class-1 Components
Steam Generator
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Development of ManufacturingTechnology for Class-1 Components
Fuelling Machine
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Development of Manufacturing Technologyfor Class-1 Components
Technology development for Zr Components
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Precision component manufacturing
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Development of inspection techniquesConcurrent with manufacturing technologies, NonDestructive Examination techniques and equipment
for these techniques were developed indigenously.
Optical instruments Laser technology
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Development of Instrumentation & Control
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Back-end technology development
Densification unit for plastic waste Simultaneous incineration of low level solid waste along with organic
liquid waste
Immobilisation of spent resins in polymer matrix
Special slag cement developed as an alternate matrix for spentresin, to avoid potential hazards in using polyster styrene
Special tile holes with higher integrity and shielding developed for
storing spent SPNDs
Evaporation system developed to reduce tritium discharge to waterbody
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Other advancements in Reactor technology
Analytical capabilities Reactor core design / burn-up optimisation studies
Seismic input parameter generation & evaluation / re-
evaluation (walk-throughs & re-analysis)
Probabilistic Safety Assessment
Ageing Management techniques Coolant Channel replacement
Feeder replacement
Robust monitoring & inspection plan
Control & Instrumentation From relay-based technology to Computer-based
Full scale simulator
St t f th t T i i
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State of the art TrainingFacilities: Simulator
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Erection ofTurbo Generator
SG Erection Calandria
Erection
End ShieldErection
Improved construction methodology
Open top construction
Steam GeneratorErection
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Growth of Nuclear Reactor Technology
Research reactors to commercial power
reactors with emphasis on self reliance
Innovations
Evolutions
Improvements
Capacity
Safety
Reliability
Economics
Sustainability
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Design Features of220, 540 &700 MWe
Indian PHWRs
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Design Features (PHWR-220)
Thermal Outp ut : 756 MWt
Gross Electr ical Output : 235 MWeModerator/Coolant Heavy Water
No. of chann els 306
Reactor Coolant Pressure 8.5 MPa
Reactor Coolant temp. 293 deg. C
Coolant Loops Single, 4 SGs
Moderator temp . 44/65 deg. C
Steam pressur e 4.03 MPa(a)
Steam temperature 250 deg. C
Natural Uranium (UO2), 19 el em en t Fu el Bund le
12 bundles per channel
Average disc harge bur n-up : 6700 MWD/TeU
On-power refuel ing
2 independent of fs i te pow er sourc es
3 X 100% DGs as Class-3 power s upply
3 tier Emerg ency Pow er Supply (Class -3,2&1)
Main Contro l Room for no rmal operat ion &
Backup Contro l Room for independent Safety System
operat ion & monitor ing o f cr i t ica l parameters
Plant Design L ife : 40 years
Core Damage Frequency : 10-5
Large Early Release Frequenc y : 10-6
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Design Features (PHWR-540)
Thermal Power Output : 1700 MWtGross Electr ical Output : 540 MWe
Moderator/Coolant
Heavy Water
No. of chann els 392
Reactor Coolant Pressure 98 MPa
Reactor Coolant temp. 304 deg. C
Coolant Loops Two (Vert ically s plit), 4 SGs
Moderator temp . 53/76 deg. C
Steam pressur e 4.17 MPa(a)
Steam temperature 253 deg. C
Natural Uranium (UO2), 37 el em en t Fu el Bund le
13 bundles per channel
Average disc harge bur n-up : 7500 MWD/TeU
On-power refuel ing
2 independent of fs i te pow er sourc es
4 X 50% DGs as Class-3 pow er supply
3 tier Emerg ency Pow er Supply (Class -3,2&1)
Main Contro l Room for no rmal operat ion &
Backup Contro l Room for independent Safety System
operat ion & monitor ing o f cr i t ica l parameters
Plant Design L ife : 40 years
Core Damage Frequency : 10-5
Large Early Release Frequenc y : 10-6
PRESSURISER
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Design Features (PHWR-700)
Thermal Power Output : 2166 MWtGross Electr ical Output : 700 MWe
Moderator/Coolant
Heavy Water
No. of chann els 392
Reactor Coolant Pressure 98 MPa
Reactor Coolant temp. 310 deg. C (3% part ial bo il ing )
Coolant Loops Two(Inter leaved feeders), 4 SGs
Moderator temp . 53/76 deg. C
Steam pressur e 4.5 MPa(a)
Steam temperature 256 deg. C
Natural Uranium (UO2), 37 el em en t Fu el Bund le12 bundles per channel
Average disc harge bur n-up : 7050 MWD/TeU
On-power refuel ing
2 independent of fs i te pow er sourc es
4 X 100% DGs as Class-3 power s upply
3 tier Emerg ency Power Supp ly (Class-3,2&1)
Alternate AC Source located at higher elevat ion
Main Contro l Room for no rmal operat ion &
Backup Contro l Room for independent Safety System
operat ion & monitor ing o f cr i t ica l parameters
Plant Design L ife : 40 years
Core Damage Frequency : 10-5
Large Early Release Frequenc y : 10-6
PRESSURISER
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Reactor Vessel (Calandria) inside waterfilled Vault
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Reactor Coolant System layout to assist Naturalcirculation
Steam GeneratorCoolant Pump
PHWR-220PHWR-540 / 700
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Bi-directional On-power refueling
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Inherent Design Safety Features of PHWRs
Higher neutron generation time Low fissile content
Passive core cooling
Online re-fuelling and low excess reactivity in the core.
Short bundle length limits consequences in case of singlebundle failure
On power detection & removal of failed fuel.
Moderator as heat sink in the event of LOCA.
Reactor vessel surrounded by large pool of water
Reactivity Devices located in low pressure moderator : Rod
ejection ruled out
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Fuel Bundle
Fuel Bundle
End Plate
Fuel Element
Pellets
Spacers
Fuel Bundle Dia : 81.7 mm
Length : 495 mm Fuel Bundle Dia : 102.4 mm
Length : 495 mm
PHWR-220 PHWR-540 / 700
19 Element37 Element
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Fuel Transfer Scheme (PHWR-220)
REACTO
R
NEW FUEL MAGAZINE
TRANSFER MAGAZINE
FUELLING
MACHINE
FUELTRANSFER
PORT
TRANSFER MAGAZINE
SHUTTLE
TRANSFER
STATION
CONTAINMENT WALL
SHUTTLE TRANSPORT TUBES
TRANSFER ARM
SHUTTLE
RECEIVING STATION
SPENT FUEL BAY
NEW FUEL LOADING TROUGH
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Fuel Transfer Scheme (PHWR-540)
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Fuel Transfer Scheme (PHWR-700)
MOBILE TRANSFER
MACHINE
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Control & Instrumentation
Use of digital technology for alarm generation
Adoption of Computer Based Systems (CBS) for data
acquisition for major process and reactor control application.
For one of the Reactor Protection Systems, hardwired logics
are retained to achieve diversity
Operator interface with menu-driven screens for control action
and system information
Computer Based Systems developed and qualified in a
systematic manner with extensive documentation forverification and validation
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Control Room (PHWR-220)
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Control Room (PHWR 540)
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Control Room (PHWR 700)
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Safety Features of220, 540 &700 MWe
Indian PHWRs
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Safety Features (PHWR 220)
ShutdownSystems
Core CoolingSystems
ContainmentSystems
PSS SSSECCS Double Containment
Engin eered Safety
FeaturesHigh pressure D2O
in ject ion
Low pressure H2O
inject ion
Long term
recirculat ion
S.
No.Device
Neutron
Absorber
1Primary Shutdown
SystemCadmium
2Secondary
Shutdown System
Natural
Boron
3 Liquid PoisonInjection System
NaturalBoron
Passive VapourSuppression Pool
Primary Cont. FiltrationSystem
Secondary Cont. Clean-up
& Purge SystemPrimary Cont. Controlled
Discharge System
RB Cooling SystemGROUP-1 GROUP-2
PSS SSS
ECCS Cont. Sys.
Two Group Concept :
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Reactor Shutdown Systems (PHWR 220)
ASSEMBLY.TOP HATCH
GUIDE TUBE ASSEMBLY
STOPPER PLATE
CENTRAL BEAM
GUIDE TUBE LOCATOR
ASSEMBLY.
(PARKED OUT POSITION)ROD BOTTOM TIP
HORIZONTAL CENTRAL
PLANE OF CALANDRIA
CALANDRIA TUBE
CALANDRIA NOZZLE
GUIDE TUBE EXTENSION
SPRING ASSEMBLYINITIAL ACCELERATION
SHUT-OFF ROD ASSY.
CALANDRIA VAULT
SUPPORT SLEEVEDECK PLATE
DECK PLATE
SHIELD PLUG
DRIVE MECHANISM
STANDPIPE THIMBLE
HELIUM LINE
Primary Shutdown System
S.No. Device Absorber Features
1 Primary Shutdown System Cadmium 14 Rods, Gravity driven
2 Secondary Shutdown System Li Pentaborate 12 locations, Stored Energy
3 Liquid Poison Injection System Natural Boron Direct inj., Stored Energy
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Emergency Core Cooling System(PHWR 220)
High Pressure
Injection
Long Term Re-circulation
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Containment Systems (PHWR 220)
Design leakage rate through Containment : 1% volume per day
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Safety Features (PHWR 540)
ShutdownSystems
Core CoolingSystems
ContainmentSystems
SDS#1 SDS#2 ECCS Double Containment
Engin eered Safety
FeaturesHigh pressure H2O
in ject ion
Long term
recirculat ion
S.
No.
DeviceNeutron
Absorber1 Shut Down System # 1 Cadmium
2 Shut Down System # 2 GadoliniumNitrate
Passive VapourSuppression Pool
Primary Cont. Filtration &Pump Back System
Sec. Cont. Cleanup & PurgeSystem
Primary Cont. ControlledDischarge System
RB Cooling SystemGROUP-1 GROUP-2
SDS#1 SDS#2
ECCS Cont. Sys.
Two Group Concept :
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Reactor Shutdown Systems (PHWR 540)
S.
No.Device
Neutron
AbsorberFeatures
1Shutdown System#1
(SDS#1)Cadmium 28 Rods, Gravity driven
2Shutdown System#2
(SDS#2)
Gadolinium
Nitrate
6 LPI perforated tubes,
Stored Energy
SDS#1
SDS#2
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Emergency Core Cooling System(PHWR 540)
High Pressure Injection
Long Term
Re-circulation
Pumps :
4 X 50%
Heat Exchangers :
3 X 50%
C i S
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Containment Systems (PHWR 540)
Design leakage rate through Containment : 1% volume per day
S f t F t
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Safety Features (PHWR 700)
ShutdownSystems
Core CoolingSystems
ContainmentSystems
SDS#1 SDS#2 ECCS Double Containment
Engin eered Safety
FeaturesHigh pressure H2Oin ject ion
Lon g term
recirculat ion
S.
No.
DeviceNeutron
Absorber1 Shut Down System # 1 Cadmium
2 Shut Down System # 2 GadoliniumNitrate
Containment Spray System
Sec Cont. Clean-up & PurgeSystem
Primary Cont. ControlledDischarge System
GROUP-1 GROUP-2
SDS#1 SDS#2
ECCS Cont. Sys.
Two Group Concept :
Reactor Shutdown Systems (PHWR 700)
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Reactor Shutdown Systems (PHWR 700)
S.
No.Device
Neutron
AbsorberFeatures
1Shutdown System#1
(SDS#1)Cadmium 28 Rods, Gravity driven
2Shutdown System#2
(SDS#2)
Gadolinium
Nitrate6 PIU tubes, Stored Energy
SDS#1
SDS#2
Emergency Core Cooling System
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High PressureInjectionTRAIN-2
Long Term Re-circulationTRAIN-1
Emergency Core Cooling System(PHWR 700)
High PressureInjectionTRAIN-1
Long Term Re-circulationTRAIN-2
P i D H t R l S t
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Passive Decay Heat Removal System(PHWR 700)
C t i t S t (PHWR 700)
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Containment Systems (PHWR 700)
Design leakage rate through Containment : 1% volume per dayNo emergency counter measures anticipated after Severe Accident.
2 Trains, each trainhaving 2 X 100% pumpsand 2 X 100% HeatExchangers
P i i f S A id t M t
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Provisions for Severe Accident Management
Independent Fire Water injection provision (Diesel drivenpumps)
Hook-up provisions for :
Steam Generators
Reactor Vessel
Calandria Vault
End Shields
Reactor Coolant System
Alternate AC Source located at higher elevation
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OperatingPerformance of
Indian PHWRs
PHWR U it i O ti
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PHWR Units in Operation
S.
No. Site/Station/Project Units Status
Year of
commercialoperation
Rated
capacity(MWe)
1 Tarapur Atomic Power Station TAPS-3&4 Operating 2005, 2006 2 x 540
2 Rajasthan Atomic Power Station RAPS-1&2RAPS-3&4RAPS-5&6
OperatingOperatingOperating
1973, 198120002009
100, 2002 x 2202 x 220
3 Madras Atomic Power Station MAPS-1&2 Operating 1984, 1986 2 x 220
4 Narora Atomic Power Station NAPS-1&2 Operating 1991, 1992 2 x 220
5 Kakrapar Atomic Power Station KAPS-1&2 Operating 1993, 1995 2 x 220
6 Kaiga Atomic Power Station KGS-1&2KGS-3KAIGA-4
OperatingOperatingOperating
200020082010
2 x 220220220
More than 300 reactor years of safe & reliable operation
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Availability Factors of Operating Units86 90 91 88 89 85 83 82
9288
0
10
20
30
40
50
60
70
80
90
100
2001-02 2002-03 2003-04 2004-05 2005-06 2006-07 2007-08 2008-09 2009-10 2010-11*
Capacity Factors of Operating Units
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54
Capacity Factors of Operating Units
84.89 89.66
81.176.29 74.4
63.04
53.72
49.61
60.8
71.37
0
10
20
30
40
50
60
70
80
90
100
2001-02 2002-03 2003-04 2004-05 2005-06 2006-07 2007-08 2008-09 2009-10 2010-11
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Longest Continuous Reactor Operation
289
590
404 394
346
432
250
371
414 407
486
529
0
100
200
300
400
500
600
700
TAPS-1 TAPS-2 RAPS-3 RAPS-4 MAPS-1 MAPS-2 NAPS-1 NAPS-2 KAPS-1 KAPS-2 KGS-1 KGS-2
Days
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P
-Sv
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Thank You for your attention
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