U.C.Muktibodh Nuclear Power Corporation of India Limited

Workshop on

Technology Assessment of SMRs for Near Term Deployment

Dec 5th – 9th , 2011 IAEA Headquarters,

Vienna, Austria

Lecture Outline

Technology Development

Design features of 220, 540 & 700 MWe Indian

PHWRs

Safety features of 220, 540 & 700 MWe Indian

PHWRs

Operating Performance of Indian PHWRs

Technology Development

Launch of Nuclear Power Program

1964

Construction

work at First

NPP Began

1948

Atomic

Energy

Commission

1954

Department of

Atomic Energy

Bhabha

Atomic

Research

Centre

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

Madras Atomic Power Station (2 220 MWe)

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

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

Calandria- Reactor pressure vessel for PHWR

Development of Manufacturing Technology for Class-1 Components

REACTOR END SHIELD

Development of Manufacturing Technology for Class-1 Components

Development of Manufacturing Technology for Class-1 Components

Steam Generator

Development of Manufacturing Technology for Class-1 Components

Fuelling Machine

Development of Manufacturing Technology for Class-1 Components

Technology development for Zr Components

Precision component manufacturing

Development of inspection techniques

Concurrent with manufacturing technologies, Non

Destructive Examination techniques and equipment

for these techniques were developed indigenously.

Optical instruments

Laser technology

Development of Instrumentation & Control

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 spent

resin, 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 water

body

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

17

State of the art Training Facilities: Simulator

Erection of

Turbo Generator

SG Erection Calandria

Erection

End Shield Erection

Improved construction methodology

Open top construction

Steam Generator Erection

Growth of Nuclear Reactor Technology

Research reactors to commercial power

reactors with emphasis on self reliance

Innovations

Evolutions

Improvements

• Capacity

• Safety

• Reliability

• Economics

• Sustainability

Design Features of 220, 540 & 700 MWe

Indian PHWRs

Design Features (PHWR-220)

Thermal Output : 756 MWt Gross Electrical Output : 235 MWe Moderator/Coolant – Heavy Water No. of channels – 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 pressure – 4.03 MPa(a) Steam temperature – 250 deg. C Natural Uranium (UO2), 19 element Fuel Bundle 12 bundles per channel Average discharge burn-up : 6700 MWD/TeU On-power refueling

2 independent offsite power sources 3 X 100% DGs as Class-3 power supply 3 tier Emergency Power Supply (Class-3,2&1)

Main Control Room for normal operation & Backup Control Room for independent Safety System operation & monitoring of critical parameters

Plant Design Life : 40 years Core Damage Frequency : 10-5

Large Early Release Frequency : 10-6

Design Features (PHWR-540)

Thermal Power Output : 1700 MWt Gross Electrical Output : 540 MWe Moderator/Coolant – Heavy Water No. of channels – 392 Reactor Coolant Pressure – 98 MPa Reactor Coolant temp. – 304 deg. C Coolant Loops – Two (Vertically split), 4 SGs Moderator temp. – 53/76 deg. C Steam pressure – 4.17 MPa(a) Steam temperature – 253 deg. C

Natural Uranium (UO2), 37 element Fuel Bundle 13 bundles per channel Average discharge burn-up : 7500 MWD/TeU On-power refueling

2 independent offsite power sources 4 X 50% DGs as Class-3 power supply 3 tier Emergency Power Supply (Class-3,2&1)

Main Control Room for normal operation & Backup Control Room for independent Safety System operation & monitoring of critical parameters

Plant Design Life : 40 years Core Damage Frequency : 10-5

Large Early Release Frequency : 10-6

PRESSURISER

Design Features (PHWR-700)

Thermal Power Output : 2166 MWt Gross Electrical Output : 700 MWe Moderator/Coolant – Heavy Water No. of channels – 392 Reactor Coolant Pressure – 98 MPa Reactor Coolant temp. – 310 deg. C (3% partial boiling) Coolant Loops – Two (Interleaved feeders), 4 SGs Moderator temp. – 53/76 deg. C Steam pressure – 4.5 MPa(a) Steam temperature – 256 deg. C Natural Uranium (UO2), 37 element Fuel Bundle 12 bundles per channel Average discharge burn-up : 7050 MWD/TeU On-power refueling

2 independent offsite power sources 4 X 100% DGs as Class-3 power supply 3 tier Emergency Power Supply (Class-3,2&1) Alternate AC Source located at higher elevation

Main Control Room for normal operation & Backup Control Room for independent Safety System operation & monitoring of critical parameters

Plant Design Life : 40 years Core Damage Frequency : 10-5

Large Early Release Frequency : 10-6

PRESSURISER

Reactor Vessel (Calandria) inside water filled Vault

Reactor Coolant System layout to assist Natural circulation

Steam Generator

Coolant Pump PHWR-220 PHWR-540 / 700

Bi-directional On-power refueling

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 single

bundle 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

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 Element 37 Element

Fuel Transfer Scheme (PHWR-220)

SCHEMATIC OF FUEL MOVEMENT IN THE STATION

REACTOR

NEW FUEL MAGAZINE

TRANSFER MAGAZINE

FUELLINGMACHINE

FUELTRANSFERPORT

TRANSFER MAGAZINE

SHUTTLETRANSFERSTATION

CONTAINMENT WALL

SHUTTLE TRANSPORT TUBES

TRANSFER ARM

SHUTTLERECEIVING STATION

SPENT FUEL BAY

NEW FUEL LOADING TROUGH

Fuel Transfer Scheme (PHWR-540)

Fuel Transfer Scheme (PHWR-700)

MOBILE TRANSFER

MACHINE

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 for

verification and validation

Control Room (PHWR-220)

Control Room (PHWR 540)

35

Control Room (PHWR 700)

Safety Features of 220, 540 & 700 MWe

Indian PHWRs

Safety Features (PHWR 220)

Shutdown

Systems

Core Cooling

Systems

Containment

Systems

PSS SSS ECCS Double Containment

Engineered Safety

Features High pressure D2O

injection

Low pressure H2O

injection

Long term

recirculation

S.

No. Device

Neutron

Absorber

1 Primary Shutdown

System Cadmium

2 Secondary

Shutdown System

Natural

Boron

3 Liquid Poison

Injection System

Natural

Boron

Passive Vapour Suppression Pool

Primary Cont. Filtration System

Secondary Cont. Clean-up & Purge System

Primary Cont. Controlled Discharge System

RB Cooling System

GROUP-1 GROUP-2

PSS SSS

ECCS Cont. Sys.

Two Group Concept :

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

Emergency Core Cooling System (PHWR 220)

High Pressure

Injection

Long Term Re-circulation

Containment Systems (PHWR 220)

Design leakage rate through Containment : 1% volume per day

Safety Features (PHWR 540)

Shutdown

Systems

Core Cooling

Systems

Containment

Systems

SDS#1 SDS#2 ECCS Double Containment

Engineered Safety

Features High pressure H2O

injection

Long term

recirculation

S. No.

Device Neutron

Absorber

1 Shut Down System # 1 Cadmium

2 Shut Down System # 2 Gadolinium Nitrate

Passive Vapour Suppression Pool

Primary Cont. Filtration & Pump Back System

Sec. Cont. Cleanup & Purge System

Primary Cont. Controlled Discharge System

RB Cooling System

GROUP-1 GROUP-2

SDS#1 SDS#2

ECCS Cont. Sys.

Two Group Concept :

Reactor Shutdown Systems (PHWR 540)

S.

No. Device

Neutron

Absorber Features

1 Shutdown System#1

(SDS#1) Cadmium 28 Rods, Gravity driven

2 Shutdown System#2

(SDS#2)

Gadolinium

Nitrate

6 LPI perforated tubes,

Stored Energy

SDS#1

SDS#2

Emergency Core Cooling System (PHWR 540)

High Pressure Injection

Long Term

Re-circulation

Pumps :

4 X 50%

Heat Exchangers :

3 X 50%

Containment Systems (PHWR 540)

Design leakage rate through Containment : 1% volume per day

Safety Features (PHWR 700)

Shutdown

Systems

Core Cooling

Systems

Containment

Systems

SDS#1 SDS#2 ECCS Double Containment

Engineered Safety

Features High pressure H2O

injection

Long term

recirculation

S. No.

Device Neutron

Absorber

1 Shut Down System # 1 Cadmium

2 Shut Down System # 2 Gadolinium Nitrate

Containment Spray System

Sec Cont. Clean-up & Purge System

Primary Cont. Controlled Discharge System

GROUP-1 GROUP-2

SDS#1 SDS#2

ECCS Cont. Sys.

Two Group Concept :

Reactor Shutdown Systems (PHWR 700)

S.

No. Device

Neutron

Absorber Features

1 Shutdown System#1

(SDS#1) Cadmium 28 Rods, Gravity driven

2 Shutdown System#2

(SDS#2)

Gadolinium

Nitrate 6 PIU tubes, Stored Energy

SDS#1

SDS#2

High Pressure Injection TRAIN-2

Long Term Re-circulation TRAIN-1

Emergency Core Cooling System (PHWR 700)

High Pressure Injection TRAIN-1

Long Term Re-circulation TRAIN-2

Passive Decay Heat Removal System (PHWR 700)

Containment Systems (PHWR 700)

Design leakage rate through Containment : 1% volume per day No emergency counter measures anticipated after Severe Accident.

2 Trains, each train having 2 X 100% pumps and 2 X 100% Heat Exchangers

Provisions for Severe Accident Management

Independent Fire Water injection provision (Diesel driven

pumps)

Hook-up provisions for :

Steam Generators

Reactor Vessel

Calandria Vault

End Shields

Reactor Coolant System

Alternate AC Source located at higher elevation

Operating Performance of Indian PHWRs

PHWR Units in Operation

S. No.

Site/Station/Project Units Status Year of

commercial operation

Rated capacity (MWe)

1 Tarapur Atomic Power Station TAPS-3&4 Operating 2005, 2006 2 x 540

2 Rajasthan Atomic Power Station RAPS-1&2 RAPS-3&4 RAPS-5&6

Operating Operating Operating

1973, 1981 2000 2009

100, 200 2 x 220 2 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&2 KGS-3 KAIGA-4

Operating Operating Operating

2000 2008 2010

2 x 220 220 220

More than 300 reactor years of safe & reliable operation

Availability Factors of Operating Units

86 90 91

88 89 85 83 82

92 88

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*

54

Capacity Factors of Operating Units

84.89 89.66

81.1 76.29 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

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

Day

s

P -

Sv

Thank You for your attention