Design, Safety and Operability Performances of 220 MWe, 540 MWe & 700 MWe PHWRs in India

7/29/2019 Design, Safety and Operability Performances of 220 MWe, 540 MWe & 700 MWe PHWRs in India

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U.C.MuktibodhNPCIL, India

Inter-regional workshop on

Advanced Nuclear Reactor Technology for

Near Term Deployment

July 4th – 8th , 2011

IAEA Headquarters,

Vienna, Austria



Design, Safety and Operability performances of 220 MWe, 540 MWe

and 700 MWe PHWRs in India - U.C.Muktibodh, NPCIL 2

Lecture Outline

Introduction to Indian PHWR designs Nuclear Steam Supply Systems

Safety Systems and ESFs

Turbine Generator Systems

Electrical, Control & Instrumentation Systems

Safety concepts

Security (Physical protection aspects)

Performance of Indian PHWRs In light of events at Fukushima

Design certification & near term plans



3



4

Developed 540 Mwe PHWRs

& operating experiencegained; Unit size scaled up

to 700 MWe

Developed Front End &

Back End Technologies of

Complete Fuel Cycle

Established Comprehensive

Indigenous Capabilities for

Designing, Equipment

Manufacturing, Constructing,Commissioning, O&M of 220

MWe PHWRs

FIRST STAGE

Design for 500 MWePFBR developed.

Construction

commenced in 2004

SECOND STAGE Fast Breeder Test Reactor

(40 MWt ) operational

THIRD STAGE Experimental reactor with U233 fuel

in operation.

Thorium Bundles in PHWRs.

AHWR-300 MWe being developed in

BARC. Technology Demonstration

for electricity generation from

Thourium. Bridge between the I & III

Stages to be Launched in X plan

Indian Nuclear

Power Programme:

Current Status

Current Status




and 700 MWe PHWRs in India - U.C.Muktibodh, NPCIL

The Indian PHWR

18 PHWR units in operation - installed capacity of 4360MWe

4 units of 700 MWe under construction.

More PHWRs of 700 MWe capacity planned in other

states of India.Design of PHWRs in India has evolved over the years to

grow into a robust and a proven model





Power Reactors in India

S.No. Site/Station/Project Units Type Status

Year of

commercialoperation

Rated

capacity(MWe)

1 Tarapur Maharashtra Site (TMS) TAPS-1&2TAPS-3&4

BWRPHWR

OperatingOperating

19692005, 2006

2 X 1602 x 540

2 Rawatbhata Rajasthan Site(RRS)

RAPS-1&2RAPS-3&4

RAPS-5&6RAPP-7&8

PHWR

PHWR

PHWR

PHWR

OperatingOperating

OperatingUnder construction

1973, 19812000

20092016

100, 2002 x 220

2 x 2202 x 700

3 Madras Atomic Power Station MAPS-1&2 PHWR Operating 1984, 1986 2 x 220

4 Narora Atomic Power Station NAPS-1&2 PHWR Operating 1991, 1992 2 x 220

5 Kakrapar Atomic Power Station KAPS-1&2KAPP-3&4

PHWR

PHWR

OperatingUnder construction

1993, 19952015

2 x 2202 x 700

6 Kaiga Atomic Power Station KGS-1&2KGS-3KGS-4

PHWR

PHWR

PHWR

OperatingOperatingOperating

200020082010

2 x 220220220

7 Kudankulam Atomic Power Project

KKNPP-1&2KKNPP-3&4

LWRLWR

Under constructionUnder construction

20112017

2 x 10002 x 1000





Evolution of the Indian PHWR

RAPS-1 – AECL

RAPS-2 – Indigenous efforts

MAPS – Changes due to site conditions

NAPS-1&2 - Major modifications incorporated, design upgraded in

line with the internationally evolving safety standards and to cater to

the seismic environment at the site. KAPS-1&2, KAIGA-1&2 and RAPS-3&4 saw further improvements

leading to standardisat ion in design and layout for 220 MWe

PHWRs.

2 units of 220 MWe each were constructed subsequently at KAIGA-

3&4 and RAPP-5&6. 2 units of 540 MWe have been constructed at TAPS-3&4 with

minimum import content.

4 units of 700 MWe are under construction and many more planned.





Evolution of the Indian PHWR

Indian PHWR design has come a long way right fromRAPS to current 700 MWe units.

Enormous amount of construction, operation &

maintenance experience and adoption of state of the art

technology for engineering and analysis has contributed

in development of a proven, robust, safe and reliable

model of the Indian PHWR, which will fulfil the energy

requirements of the country to a large extent.

Indian PHWR is a combination of inherent and

engineered safety features, incorporating defence indepth through active and/or passive means to cope with

DBA.





Introduction to Indian PHWR design

Horizontal reactor vessel – Calandria

Pressure tube concept (306/392 channels)

Natural Uranium fuelled (Fuel pins; Bundles)

Heavy water cooled and moderated

Calandria surrounded by water enclosed in a

concrete structure – Calandria Vault

On-power refueling

Double containment

Suppression Pool (220 MWe, 540 MWe)









Pressurised Heavy Water Reactor

11





Summarised Technical Data

Description 220 MWe 540 MWe 700 MWe

Reactor Thermal Output 755 MWth 1730 MWth 2166 MWth

Power Plant Output (Gross) 235 MWe 540 MWe 700 MWe

Power Plant Output (Net) 210 MWe 485 MWe 630 MWePower Plant Efficiency (Net) 27.8% 28.08 % 29.08 %

Plant design life 40 yrs. 40 yrs. 40 yrs.

Primary Coolant material Heavy Water Heavy Water Heavy Water

Secondary Coolant material Light Water Light Water Light Water Moderator material Heavy Water Heavy Water Heavy Water

General Plant Data





Summarised Technical Data (Contd.)


SAFETY GOALS

Core Damage Frequency 10-5 / Year 10-5 / Year 10-5 / Year

Large Early Release Frequency 10-6

/ Year 10-6

/ Year 10-6

/ Year Occupational radiation exposure 20 mSv/Yr * 20 mSv/Yr * 20 mSv/Yr *

Operator action time (in minutes) 30 30 30

* : Average in any 5 consecutive years







Steam flow rate nominal

conditions

1330 t/h 3078 t/h 3874 t/h

Steam pressure/temperature 4.03 MPA(a)

250.6 deg C

4.17 MPa(a)

/252.8 oC

4.5 MPA(a)

256.3 deg C

Feed Water flow rate at nominal

conditions

1261.7 t/h 3078 t/h 3874 t/h

Feed Water temperature 171 deg.C 180 deg.C 180 deg.C

Nuclear Steam Supply System







Primary coolant flow rate 13.26x106 kg/h 28.1x106 kg/h 28.9 x106 kg/h

Reactor operating pressure 87 kg/cm2g (nom.) 100 kg/cm2g

(nom.)

100 kg/cm2g

(nom.)

Mean temp. rise across core 44 deg. C 44 deg. C 44 deg. C

Average linear heat rate 28.6 kW/m 40.1 kW/m 50.2 kW/m

Fuel material Natural UO2 Natural UO2 Natural UO2

Fuel clad material Zircaloy - 4 Zircaloy - 4 Zircaloy - 4

Fuel assembly 19 elements 37 elements 37 elements

Average discharge burn up 6700 MWd/T 7500 MWd/T 7050 MWd/T

Reactor Core & Coolant System







Inner diameter of Calandria 5996 mm 7800 mm 7800 mm

Wall thickness 25 mm 32 mm 32 mm

Base material Austenitic SS

304 L

Austenitic

SS 304 L

Austenitic

SS 304 L

No of Coolant channels 306 392 392

Lattice pitch 22.86 sq

lattice

28.6 sq

lattice

28.6 sq

lattice

ID of coolant channel 82.6 mm 103.4 mm 103.4 mm

Core length 5.085 m 5.940 m 5.940 m

Coolant channel material Zr – 2.5% Nb Zr – 2.5% Nb Zr – 2.5% Nb

Reactor Pressure Vessel







Type Mushroom type

with integrated

steam drum and

preheater

Mushroom type

with integrated

steam drum

Mushroom type

with integrated

steam drum

Number of SGs 4 4 4

SG tube material Incoloy 800 Incoloy 800 Incoloy 800

Type of coolant pump Vertical,

Centrifugal,

single stage.

Vertical,

Centrifugal,

single stage.

Vertical,

Centrifugal,

single stage.

Pressurizer volume -- 30 Cub M 45 Cub M

Steam Generator, Coolant pumps & Pressuriser







Double Containment Prestressed ICW,

Reinforced OCW

Prestressed ICW,

Reinforced OCW

Prestressed ICW,

Reinforced OCW

PC dimensions 42.56 m dia

55.35 m ht

49.5 m dia

50.1 m ht

49.5 m dia

53.1 m ht

PC Design pressure 0.27 MPa (a) 0.24 MPa (a) 0.26 MPa(a)

Design leak rate of

Containment System

1 % V / day 1 % V / day 1 % V / day

Containment







Residual Heat

Removal (Active /

Passive systems)

Active- SD cooling

system

Passive – Thermo

syphon through SGs

Active- SD cooling

system

Passive – Thermo

syphon through SGs

Active- SD cooling

system

Passive – Thermo

syphon through SGs

Safety Injection

(Active / Passive

systems)

ECCS ECCS ECCS

Core Cooling Systems







Type of Turbine Tandem

compounded,

Horizontal

impulse reactiontype.

Tandem

compounded,

Horizontal


Tandem

compounded,

Horizontal


Number of turbine

sections

1 HP + 1 LP 1 HP + 2 LP 1 HP + 3 LP

Generator Direct coupled,

Hydrogen cooledrotor

Direct coupled,


Direct coupled,


Turbine & Generator





Inherent Safety Features of Indian PHWRs

Higher neutron generation timeLow fissile content

Short bundle length limits consequences in case of

single bundle failure.

Passive core cooling feature.

On power detection of failed fuel.

Online fuelling and low excess reactivity in the core.

Moderator as heat sink.Reactivity Devices located in low pressure

moderator : Rod ejection ruled out



22





NSSS – PHT System

PHT System has been designed to ensure adequate

cooling of reactor core under all operational states and

during and following all postulated off normal conditions.

Normal Operation By Primary Coolant Pumps (PCPs)

Loss of power to

PCPs

Initially by pump flywheel inertia and later by

thermo syphoning.

Shut down By shut down cooling pumps and heat

exchangers which are independent of SGs

Loss Of Coolant Accident (LOCA)

By Emergency Core Cooling System (ECCS)consisting of high pressure D2O / H2O injection

from accumulators and low pressure long term

recirculation by pumps.









PHT System features

Different feeder sizes & orificing

Controlled pressure at ROH

Over pressure relief to PHT pressure boundary

Feed & Bleed / Pressuriser

Assist natural circulation – Layout Small leak handling capability

Online purification & filtration

Accessibility during shutdown

Header level control for maintenance of SGs, PCPs etc.

Variable / constant pressure program for SG pressure

control





PHT System – Features (Contd.)

Total elimination of valves in PHT System General reduction in the number of components has

helped to decongest the layout in the pump room

Better maintenance approachability, less

maintenance and lesser dose uptake

Two loop concept

Minimize the inventory loss

Minimize core positive void coefficient

Minimize the enthalpy release to the containmentunder loss of coolant accident





PHT System – Features (Contd.)

In 540 MWe PHWR, a pressurizer was introducedfor pressure control, while feed and bleed is

retained for inventory control.

In 700MWe units, interleaving of feeders has

been adoptedReduction in void coefficient

Minimise the reactor over-power during a LOCA

Passive Decay Heat Removal System (PDHRS)has been introduced for the first time in 700 MWe

units to ensure removal of decay heat in Station

Blackout condition



STEAM GENERATOR

540 MWe





Reactor Core and Fuel design

Heavy Water moderator and coolant

Natural Uranium Dioxide – Fuel

The reactor is an integral assembly of two end shields

and a Calandria with the latter being submerged in the

water filled vault.

Fuel bundles are contained in 306 / 392 Zr-2.5%Nb

pressure tubes, arranged in a square lattice.

At each end, the pressure tubes are rolled in AISI 403

modified stainless steel end fittings, which penetrate the

end shields and extend into the fuelling machine vaults

so as to facilitate on power fuelling





Calandria

Calandria is a horizontal vessel containing the coolant

channel assemblies, moderator and internal components

of various shutdown mechanisms and reactivity control

devices.

The Calandria structure is fabricated from Austenitic

stainless steel type 304 L.

The design, fabrication, inspection and testing is in

accordance with ASME Section III NB





End Shield

The end shield is a cylindrical box whose ends are closed

by the Calandria side tube sheet (CSTS) and the fueling

side tube sheet FSTS).

The box is pierced by 306 / 392 Stainless Steel Lattice tube

arranged in a square lattice. The box is filled with water andcarbon steel balls in the ration 47:53.

The End shields are designed, fabricated and tested as

class II components according to the ASME section III NC.





Calandria and End Shields



Design, Safety and Operability performances of 220 MWe, 540 MWeand 700 MWe PHWRs in India - U.C.Muktibodh, NPCIL

Reactor inside vault




Fuel Bundle




37-element fuel bundle (540 MWe/700 MWe)




Residual Heat Removal & Auxiliary Cooling

Shut down cooling system.

End shield cooling system.

Calandria Vault Cooling system.

Spent Fuel Storage Bay cooling system.

Active process water system.

Service Water system.




Fuel Handling Systems

Fuel Handling System has seen major evolutions in carriage design,

fuel transfer system and controls in progressive stations.

In RAPS/MAPS, FM head is supported on a mobile type carriage,

which moves on rails.

In the standardised design of 220 MWe PHWR, a seismically qualified

fixed column & moving bridge was introduced, which is better suited

for high intensity seismic events. Transit equipment called Transfer Magazine was introduced in Fuel transfer system in place of Air Lock

and Transfer Arm used in RAPS / MAPS. This facilitates the parallel

simultaneous operation of refueling by FMs on the reactor and

transferring of irradiated fuel from the Transfer Magazine to the storage

pool through Shuttle Transport Tube . While earlier fuel handling controls employed hard wired system, for

standard PHWRs, computerized control system has been provided.

This has resulted in flexibility and better man –machine interface.




Fuel Handling Systems (Contd.)

540 MWe PHWR, basic design of FT system is

similar to that in 220 MWe PHWRs. Additional

features :

FM design with rack & pinion based ram assembly.

Separate calibration and maintenance facility to

test various sub-assemblies like ram assembly,

separators, B-ram drive, various process devices and

control equipment was introduced . This is specifically

meant for performance testing after major maintenance.




Fuel Handling Systems (Contd.)

FH system in 700 MWe PHWRs adopts the features of

220 & 540 MWe PHWRs. Based on operating

experience, Fuel Transfer system is based on a unique

concept of Mobile Transfer Machine, which receives

spent fuel from the Fuelling Machine & discharges it to

the spent fuel bay. Accordingly, shuttle transport systemhas been eliminated.

There is a single Spent Fuel Storage Bay (SFSB) in 220

MWe PHWRs, whereas for the later 540 & 700 MWe

units, separate SFSB for each unit has beenincorporated.




FM BRIDGE & CARRIAGE




FM Head




MOBILE TRANSFER MACHINE

FT

PORT

SWING TABLE

Z-

FRAME

INDEXER

SUPPORT FRAME

CYLINDER

CABLE DRAG CHAIN

BALL VALVE

NEW FUEL PORT



43




Reactor Shutdown Systems

Concept of two diverse and independent shut down

systems was introduced in standard 220 MWe reactors.

Primary Shutdown System (PSS) using gravity fall of

cadmium absorber elements, and

Secondary Shutdown System (SSS) injecting liquid

poison column in vertical tubes located inside the core.

Each of these systems is independently capable of

terminating all conceivable fast reactivity transients from

any state of the reactor. The reactivity transients

considered include those from a large loss of coolantaccident, which result in the fastest reactivity addition rate

in a PHWR due to coolant voiding in the core.




Reactor Shutdown Systems (Contd.)

The design in 220 MWe units has additional system

called Liquid Poison Injection System (LPIS) to

augment the negative reactivity worth to take care of

xenon decay during long term shut down.

For 540 Mwe / 700 MWe PHWRs, each of the two shutdown systems have adequate worth for long-term

shutdown. These systems are :

SDS#1 : Cadmium rods that fall under gravity

SDS#2 : Direct injection of poison in moderator inside

calandria




Emergency Core Cooling

In earlier versions of Indian PHWRs (RAPS/MAPS), the cold

heavy water available in the moderator system is used for

emergency injection into the PHT system and long term

recirculation for postulated LOCA. Subsequently ECCS for

these plants were upgraded by retrofitting high pressure

injection system.For standard 220 MWe units, the ECCS was modified to

delink the system from moderator system and it includes

High pressure heavy water injection

Intermediate pressure light water injection Low pressure long-term recirculation

All actions up to and including the establishment of long-

term recirculation from suppression pool are automatic.




Emergency Core Cooling (Contd.)

In 540 MWe PHWR, the high pressure injection is from light

water accumulators. A simple scheme of injecting light water into all reactor headers followed by low pressure long term

recirculation has been adopted.

In 700 MWe PHWR, there are two trains of ECCS injection

and long term recirculation 2x100% trains of high pressure light water injection system

followed by

2x100% trains of low pressure long term recirculation system by

ECCS pumps.

The equipment/components of two trains are located diversely toavoid any common cause failure. Pump suction of each train has

got multiple strainers protected by coarse screen to assure

continued long term recirculation flow from ECCS sump.




Reactor Containment Systems

The containment system of Indian PHWRs has seen several

modifications in successive projects. The first Indian PHWR (RAPS) reactors have a s ingle con tainment

envelope of reinforced concrete cylindrical section and a pre-

stressed concrete dome. In RAPS, pressure suppression in the

containment is provided by a dousing system where water stored in

a very large tank, at the topmost floor of the containment establishesa curtain of water in the path of releasing steam during postulated

loss of coolant accident to limit the building pressure.

In MAPS and all standard 220/540 MWe reactors a pressure

supp ress ion sy stem is used where the released steam –air mixture

is led to a large body of water (suppression pool) stored at thebottom of the containment.

In MAPS, a part ial double containment was used with primary

containment of pre-stressed concrete, and secondary of rubble

masonry.




Reactor Containment System (Contd.)

For standard 220 MWe/ 540 MWe/ 700 MWe PHWRs,

full double containment design has been adopted with

primary containment of pre stressed concrete and

Secondary containment of Reinforced concrete.

In 700 MWe PHWRs, a containment spray cooling

system is provided for dual functions of fission product

mop up and depressurisation of containment to reduce

ground level releases in place of pool-based vapour

suppression system.

The primary containment for 700 MWe units is lined with

steel for achieving better leak tightness




50

Evolution of Indian

PHWR Containments




Engineered Safety Features

Reactor Building Cooling Systems.

Primary Containment Filtration and pump back

system (for 220 MWe and 540 MWe PHWRs)

Secondary Containment Recirculation and

Purge system.

Primary containment Controlled Discharge

system.



52




TG & Auxiliaries

Steam turbines for PHWRs are configured as follows :

One single flow high pressure (HP) turbine and one double flow low

pressure (LP) turbine tandem compounded and coupled to 2-pole

generator for 220 MWe PHWRs.

One single flow high pressure (HP) turbine and two double flow low


generator for 540 MWe PHWRs One double flow high pressure (HP) turbine and three double flow low


generator for 700 MWe PHWRs

As the steam at the exhaust of HP turbine is around 12% wet, it is routed

through moisture separator and re-heater before it is led to LP turbines.

The TG sets for the latest plants are designed to operate on

continuous basis in the band of 47.5-51.5 Hz as against 48-51

Hz to improve availability of the units considering prevailing

variation in grid frequency.




TG & Auxiliaries (Contd.)

Integral drain arrangement is provided in the turbine

casing to remove the excess moisture

Redundant turbine over speed protection systems

provided

Non-return valves are provided in almost all extractionlines to close automatically

The generator is provided with hydrogen cooled rotor,

stator core & overhang and water cooled stator

conductors




TURBINE GENERATOR – 220 MWe



56




C&I in Indian PHWRs

Control and instrumentation is an area of nuclear reactor design,

which has undergone major changes during past 60 years.

Reactor control and protection system of RAPS-1&2 and MAPS-

1&2 was made using conventional discrete analog circuits.

Control room computer system was another step towards

centralization of acquisition of control room data with improved

human –machine interface. The introduction of digital systems

(computers) did away with the problems of analog circuit design.

For KAPS-1&2, more steps were taken towards utilizing the

flexibility of computer-based systems. The boldest of the stepswas to computerize the alarm and trip contact generation

function. The system was called Programmable Digital

Comparator System.




C&I in Indian PHWRs (Contd.)

Computer Based Systems were further evolved and used for

standard 220 MWe, 540 MWe and 700 MWe PHWRs for process and reactor control applications. For one of the

Reactor Protection Systems and other safety systems, hard-

wired logics are retained.

In 700 MWe PHWRs, Computer Based Systems are beingdesigned in clusters. This will ensure uniform architecture for

systems in a cluster.

Operator interface with menu driven screens for control action

and system information were also introduced.

Upgradation of old projects was started in decade of 1990sand lot of old systems were re-engineered with computer-

based systems.

R P i S




Reactor Protection Systems

In order to protect the plant against „common mode‟

incidents such as fires that could affect many safetysystems at the same time, the I&C of safety systems are

located in distinct, physically separate rooms / panels in

control building (control equipment room).

Each of the safety systems viz. reactor shutdownsystems, emergency core cooling system, containment

isolation system are provided with a triplicate channel

philosophy with 2 out of 3 coincidence logic. This permits

one channel to be tested without affecting normal plantoperation. It also allows one faulty channel to be put in a

safe state. It facilitates inter-channel comparison among

the signals

C t l R




Control Room

The main control rooms (MCR) have conventional control

room design, which has evolved from plant to plant withincremental improvements based on the plant design and

technology available.

Hybrid control rooms, wherein computer based operator

information displays and parameter selection and settingsfacilities have been provided.

The normal operations are mostly carried out from discrete

hardwired controls at the panels.

The operator consoles located in the center of these control

rooms provide facilities for detailed presentation of data invarious formats and also provide capabilities for changing

operational parameters using Visual Display Unit (VDU)

consoles of individual systems.

C l R




Control Room (Contd.)

The control panels are organized in a plant system-wise

manner, with systems associated with variousassociated functions suitably located nearby for ease of

operations.

Control Room is located in the seismically qualified

control building (CB).

CB location is such that it is not affected from internal

missiles from turbine.

Unitized Control Room concept is followed.

C t l R





The MCR panels are placed behind the OICs so that the

operator can view the indications & annunciations on theMCR panels from his seat.

The MCR panels and OICs of both the units are arranged in

“L” shape having linear image with the other unit. The use of

computerized systems has reduced the density of components on Main Control Room panels.

A separate Back up Control Room (BCR) has been provided

for each unit. Essential safety functions can be carried out

from BCR to bring the unit under safe cold shut down state in

case of unavailability of MCR.

BCR has been back-fitted in older units.

C t l R





In 700 MWe, compact computerized Main Control Room

(MCR) is provided.

MCR comprises of sitting console based operations with

computerized operating procedure, advance fault

diagnostics and intelligent alarm system.

Plant over view panels and limited hardwired back upcontrol panels are provided.

Separate Back up Control Room (BCR) is provided for

each unit.

Essential safety functions can be carried out from BCR

to bring the unit under cold shut down state in case of

unavailability of MCR.

KGS 2 Control Room




KGS-2 Control Room



TAPS -4 Control Room

KAPP-3&4 MCR




KAPP 3&4 MCR

66

El t i l S t




Electrical Systems

Offsite Power System

400 kV and

220 kV switchyards,

400kV and 220kV grids.

Start up power for each reactor unit is derived from 220kV switchyard through one start-up transformer (SUT)

having two secondary windings. Two main cum transfer

switching is adopted for 220 kV switchyard.

Station Auxiliary Power Supply System (SAPSS)

Normal Power Supplies and

Emergency Power supplies

El t i P S li




Electric Power Supplies

Normal Power Supply

This system has two voltage levels at

6.6 kV, 3 phase supply and

415 V, 3 phase supply.

Emergency Power SupplyEmergency Power Supply System consists of three tier

power supply classes i.e.

Class III,

Class-II and

Class-I power supplies.

These Power Supplies feed all the safety / safety related system

loads of the unit and also some of the non-safety system loads.

C t l P S li




Control Power Supplies

Main Control Power Supply (MCPS) system

Class-I 48 VDC MCPS system (220 & 540 MWe)

Class-I 24 VDC / 220 VDC MCPS System (700 MWe)

Class-II 240 VAC MCPS system

Supplementary Control Power Supply (SCPS)system

Class-I 48 VDC SCPS system (220 & 540 MWe)

Class-I 24 VDC / 220 VDC SCPS System (700 MWe)Class-II 240 VAC SCPS system

Operating Modes




Operating Modes

Normal operation

Operation within specified operational limits and

conditions.

Hot shutdown state

Shutdown state of the reactor with primary coolanttemperature (inlet to reactor) and pressure close to

normal operating condition and the primary coolant

pumps (PCPs) running is defined as hot shutdown state.

Operating Modes




Operating Modes (Contd.)

Cold shutdown state

Reactor maintained sub-critical with specified sub-

criticality margin and temperature of the PHT system at

inlet to the core is less than 550C.

Guaranteed shutdown state(GSS)

A specified shutdown state of the reactor with sufficiently

large reactivity shutdown margin, established by the

addition of liquid poison into the moderator to provide

positive assurance that an inadvertent increase inreactivity by withdrawal of all other reactivity devices

cannot lead to criticality.

Standard Fuel Cycle




Standard Fuel Cycle

PHWRs use 'Natural' uranium in dioxide form as fuel.

Closed end fuel cycle - based on availability of fissile and

fertile fuel resources in the country.

The spent fuel bundles from PHWRs are reprocessed and the

depleted uranium and plutonium is planned to be used in fast

breeder reactors. A small quantity of reprocessed depleted uranium is also

recycled in PHWRs.

The Front-End of this cycle like mineral exploration, mining

and processing of ore and fuel fabrication; and back end of the cycle, which includes fuel reprocessing, re-fabrication and

nuclear waste management are carried out by different units

of Department of Atomic Energy (DAE), Government of India.

Alternative Fuel Options





To achieve fuel burn up beyond 15000 MWd/TeU, fissile

materials like slightly enriched uranium and Mixed Oxide fuelelements have been studied for use in 220 MWe PHWRs.

Studies showed that burn-up can be increased up to 30000

MWd/ TeU.

Studies on reactor physics characteristics like reactor control,shut down margin, fuel, systems thermal-hydraulics and

material compatibility have been carried out for each fuel type

before taking up actual loading in the reactor.

Reprocessed depleted uranium dioxide fuel bundles, Slightly

Enriched Uranium Bundles (SEU), MOX bundles and thorium

dioxide bundles were designed, developed and successfully

irradiated in different 220 MWe reactors.





Alternative Fuel Options (Contd.)

Thorium bundles and reprocessed depleted uranium dioxide

bundles were used for flux flattening in the initial core suchthat the reactor can be operated at rated full power in the

initial phase.

MOX-7 bundle design evolved is a 19-element cluster, with

inner seven elements having MOX pellets consisting of plutonium dioxide mixed in natural uranium dioxide and outer

12 elements having only natural uranium dioxide pellets.



75

Defence in Depth




Defence in Depth

Multiple barriers to radioactivity release

Prevention – is priority and Mitigation – if

deviations or accidentt happens

To achieve defence in depth - SSCs at first four

levels Elaborate emergency plans at the fifth level

Defence in Depth (C td )




Defence in Depth (Contd.)

Prevention Seismic design of SSCs

Consideration of possible flood mechanisms to decide safe

grade elevation

Consideration of internal flood in design

Quality requirements for design, fabrication, plantconstruction and operation

Safety Classification of SSCs

Assignment of SSCs at various levels of defence in depth

Environmental qualification

Redundancy and diversity at system and function level

Physical and functional separation between process and

safety systems






Prevention (contd.) Single failure criteria

In service inspection and regular surveillance

Operation within defined limits in technical specificationsfor operation

Licensing and periodic relicensing of operating personnel

Single point vulnerability assessment

Balanced design, quantification of reliability, absence of cliff edge effects, permissible down time of equipment etc.are overseen by PSA

Adherence to normal operating procedures Regular and multi tier review by utility and regulators

Operating experience feedback






Mitigation – AOOs and DBAs Control systems and protection systems

Safety systems

Multiple means to achieve three fundamental safety

functions

Passive features and systems Reliable safety support services

Procedures for off normal and emergency conditions

Rehearsing these procedures on plant simulators

Safety analysis to show compliance to acceptable dose

limits and PSA targets

Regular and multi tier review by utility and regulators

Operating experience feedback






Mitigation – BDBAs and severe accidents

Available plant systems

Severe accident management provisions

Severe accident management guidelines

Hydrogen management

Defence in Depth (Contd )





Emergency Plans and Preparedness

Hierarchy of emergency plans (Plant, Site and Offsite)

These plans are regularly rehearsed with a defined

frequency

Participation of local government authorities inemergency exercises

Participation of nearby population in public awareness

programme and emergency exercises

Licensing




Licensing

Identified stages (Siting, Construction, Operation)

As per requirements specified by regulators in

codes and guides

Periodic safety review at every 10 years

Review at the time of renewal of license at every 5years

Internal review within utility

Multi tier review by regulators

Licensing Approach - Projects




Licensing Approach - Projects

The regulatory body adopts a multi-tier review process for safety

review and assessment of NPP First level of review and assessment

Site evaluation Committee (SEC),

Project Design Safety Committee (PDSC)

Civil Engineering Safety Committee (CESC).

These Committees as a body are comprised of experts in variousaspects of NPP safety.

The next level of review is conducted through an Advisory

Committee on Project Safety Review (ACPSR)

Members drawn from the regulatory body, reputed national laboratories

and academic institutions.

Representation from other governmental organizations and ministries.

After considering the recommendations of ACPSR and the first level

committee, the regulatory board decides on the authorization.

Licensing




Licensing

Safety Review Process for Nuclear Power Projects

APPLICATION FOR

SITE CONSENT

APPLICATION FOR

CONSTRUCTION CONSENT

LEGEND:SEC - SITE EVALUATION COMMITTEE

PDSC - PROJECT DESIGN SAFETY COMMITTEE

ACPSR- ADVISORY COMMITTEE FOR PROJECT

SAFETY REVIEW

Governing Document “Regulation of

Nuclear and Radiation Facilities”

(AERB-SC-G)

• Siting

• Construction

• Commissioning

• First Criticality / Tests

• Power Operation (In stages)

Consenting Stages:REGULATORY

BOARD

NPC-SRC HQ

ACPSR

PDSC SEC

R

e g u l a t o r y B o d y

U t i l i t y

Licensing Approach - Stations




Licensing Approach - Stations

First tier of safety review is carried out by the „Unit Safety

Committee‟Representatives from the regulatory body

Representatives from NPP under review and

Experts in various aspects of nuclear technology

drawn from different institutions. The second-tier of safety review of Indian NPPs is by

Safety Review committee for Operating Plants

(SARCOP), which is Apex body to decide on the matters

of nuclear safety pertaining to NPPs. The third-tier is the regulatory board, which based on the

recommendations of SARCOP, considers the major

safety issues pertaining to operation of NPPs.

Licensing




Licensing

Safety Review Process for Operating NPPs

Regulatory Board

EXPERT GROUPS

-REACTOR PHYSICS

-REACTOR CHEMISTRY

-CONTROL & INST.

-ISI

-COOLANT CHANNEL SAFETY

REGULATORY STAFF

-INPUTS

-INTERNAL REVIEW

-CONTINUITY

-CO-ORDINATION

-ENFORCEMENT

-FOLLOW-UP

-INSPECTION

Safety Review

Committee for

Operating Plants

(SARCOP)

Unit Safety

Committee

NPC –SRC HQ Station Operation Review

Committee (SORC)

R e g u l a t o

r y B o d y

U t i l i t y

Seismic Design Considerations




Seismic Design Considerations

The seismic design is incorporated by classifying SSC under three

categories SSE Category : SSE category incorporates all systems, components,

instruments and structures conforming to safety classes 1, 2 and 3 and

are designed for the maximum seismic ground motion potential at site

(i.e. SSE) obtained through appropriate seismic evaluations based on

regional and local geology, seismology and soil characteristics.

OBE Category : All systems, components, instruments and structures

which are to remain functional for continued operation of the plant

without undue risk fall under OBE category and the design basis is a

lower level seismic ground motion than SSE which may reasonably be

expected during the plant life. Exceeding of OBE level seismic event

requires a shutdown of the plant and carry out detailed inspection of entire plant prior to startup.

General : Seismic design by relevant Indian standards

Safety Systems to cope with Severe Accidents




Safety Systems to cope with Severe Accidents

An accident sequence involving loss of coolant with failure

of emergency core cooling can lead to a severe accidentwith failure of maintaining moderator and Calandria Vault

water heat sink. Fire water injection into SGs.

Fire water injection into Calandria.

Fire water injection into Calandria Vault.

Fire water injection into End Shields.

Provision for passive/active mixing of containment atmosphere to

limit hydrogen concentration.

Fire water back-up is provided to moderator heat exchangers andECCS heat exchangers.

Provision for manual interconnection of Class-III emergency power

supplies between units (700 MWe).

Safety Assessment




Safety Assessment

A comprehensive safety analysis by rigorous

deterministic and complementary probabilistic methodsis carried out covering the following plant states

Normal operational modes of plant

Anticipated operational occurrences

Design basis accidents

During combination of events leading to beyond

design basis scenarios including severe accidents

The deterministic safety analysis is available up to the

severe accident and is being utilized in conjunction with

probabilistic safety assessment in preparation of severe

accident management programme

Emergency Planning




Emergency Planning

In accordance with different degrees of severity

of the potential consequences, emergency

situations are graded as:

1. Plant emergency

2. Site emergency and3. Off-site emergency.

Emergency Measures




Emergency Measures

The emergency measures consist of the following

Notification Assessment action during Emergency

Corrective actions

Protective measures (countermeasures)

Contamination control measures

Infrastructure for Emergency Response

Plant Control Room

Emergency Control Centre

Communication System Assessment Facilities

Protective Facilities



92

Security & Physical Protection




Security & Physical Protection

Security systems have under gone several changes based on

changing threat perceptions and the technologicaldevelopments.

NPCIL has established the environment to create and foster

characteristics and attitudes in organization and individuals so

that physical protection issues receive attention as warranted

by their significance. A multi pronged approach is in place toensure security of the country‟s NPPs, which includes the

following :

Screening/ongoing intelligence about employees

Physical Protection System National Security Force

Defence coverage

Regulatory Frame work



94

Plant Operation




Plant Operation

The commencement of operation of a Nuclear Power

Plant (NPP) begins with approach to the first criticalityof the station.

Before the start of commissioning activities, the station

prepares a comprehensive programme for the

commissioning of plant components and submits thesame for review and acceptance of Regulatory body.

The Operation and Maintenance (O&M) department at

the station prepares the Technical specification for

operation in consultation with the plant designers before

the approach to first criticality, based on the inputs from

the design and safety analysis.

Plant Operation (Contd.)





Once the commissioning activities are completed, the entire

plant is handed over for regular operation andmaintenance, to the O&M department which already exists

at the Site.

To ensure a high degree of quality in operation, all

operation persons who are at or above the position of Assistant Shift Charge Engineer (ASCE) are qualified

graduate engineers who are trained and licensed as per

the licensing procedures approved by Regulatory body.

All activities including surveillance testing are performedwith approved procedures to minimize errors due to human

factors.






The station has a well defined organization chart. The chart

clearly defines the lines of responsibility and authority toensure smooth operation as well as safety during start up,normal and abnormal operations.

Station Director is the Chief of Station O&M managementat site. He has the overall responsibility for the safeoperation of the plant and in implementing all relevantpolicies and radiation protection rules and other instructionsand procedures laid down by the operating organization for plant management, and the statutory / regulatoryrequirements.

The performance of operating Indian PHWRs hasimproved significantly and an overall availability factor of greater than 90% has been achieved.

Reliability & Availability Targets




Reliability & Availability Targets

Contributing factors :

Successful and proven technology

Extensive use of vast experience available from

Indian PHWR and similar plants elsewhere

High degree of automation to minimize human error

Functionally and physically independent safety

systems

Basic safety functions carried out by multiple means

Online testing and maintenance of a protectionchannel without affecting reactor operation

Use of fire-retardant materials

Reliability & Availability Targets (Contd.)




y y g ( )

Contributing factors :

Fuel reliability

Elimination of human error

Extensive training

Earthquake resistant design

Adequate defences have been built in the design against

flooding, externally or internally generated missiles, fire, etc.

Highly reliable safety systems with very low unavailability

targets

Defence in Depth Periodic testing and inspection of active components in safety

systems - Online





y y g ( )

Designed for an average annual availability factor of

greater than 90%, averaged over the life of the plant Targets for different types of outages are planned

accordingly.

Indian PHWRs are normally designed to have one planned

biennial shut down for about one month duration

The maintenance programme is put in place to ensure that

Safety Status of the Plant is not adversely affected due to

aging, deterioration, degradation or defects of plant structures,

systems or components since commencement of operation

and their functional reliability is maintained in accordance with the

design assumptions and intent over the operational life span

of the plant





y y g ( )

Preventive maintenance schedule for systems,

structures and components

In Service Inspection (ISI) programme

Performance Review Programme to identify and rectify

gradual degradation, chronic deficiencies, potentialproblem areas or causes

Capacity Factors of Operating Units





p y p g

102

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



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

D a y s





Collective Dose/Unit in Indian NPPs for Routine O&M activities

(Older units)





0

0.5

1

1.5

2

2.5

3

3.5

4

4.5

5

2007 2008 2009

P e r s

o n S i e v e r t

(Older units)

TAPS-1&2 RAPS-2 MAPS NAPS

Collective Dose/Unit in Indian NPPs for Routine O&M activities

(Ne nits)





0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.6

2007 2008 2009

P e r s o

n S i e v e r t

(New units)

KAPS KGS-1&2 RAPS-3&4 TAPS-3&4 KGS-3





P

- S v

Construction Management





g

Setting up of nuclear power projects in India in about 5 years has

been demonstrated with the help of developments in construction

technology, mechanization, parallel civil works and equipment

erection, computerized project monitoring and accounting systems

Further reduction of construction time is being aimed at

The design of 700 MWe is being carried out in an Integrated

Engineering Environment using state of art tools like 3D plantmodeling, Prodok etc.

The Plant Design, Construction, Operation and Maintenance

organizations together develop a detailed overall Project Master

Plan prior to the start of construction. This encompasses :

Design

Procurement

Construction

Commissioning activities up to the commercial operation

Construction Management (Contd.)





All schedules are regularly reviewed and monitored to check for

compliance with the overall project plan and to identify anydeviation requiring corrective action.

The project is monitored using quantitative methods appropriate

to the particular activity. Schedules are maintained using modern

technology (Primavera software, etc.) and methods, and updated

as work progresses to realistically reflect the actual work status

Regular interaction between the construction engineers and the

design engineers as well as interdisciplinary design reviews are

periodically carried out to successfully implement the

constructability requirements at the design stage itself. Standardized component sizes, types and installation details are

provided to improve productivity and reduce material inventories



111

Plant Layout considerations





y

The main plant layout of Indian PHWRs has been

developed on the basis of twin unit concept.

The principal features of plant layout

Concept of independent operation of each unit. Only

some of the common facilities are shared for reasonsof economy.

The buildings have been grouped according to their

seismic classification in consonance with the

classification of the system/ equipment contained.

Plant Layout considerations (Contd.)





Mirror images in equipment layout are avoided to the

maximum extent possible for O & M convenience A separate Control Building has been provided as a

common facility. However, the control room and control

equipment rooms located in this building are provided to

cater for unitized operation.

A separate backup Control Room has been provided for

each unit

Emergency power supply systems such as Diesel

Generators, UPS systems and Batteries are separately

housed in safety related structures, for each unit.

Plant Layout considerations (Contd.)





Proper access control measures are provided by

means of Central Alarm Station (CAS), physicalprotection fencing and manned gates.

The two unit module in the nuclear island has been so

chosen that it is possible to :

Enforce single point entry in the radiation zones. Follow radiation zoning philosophy without undue

inconvenience to the operating personnel.

Plant layout – 220 MWe PHWR





Plant layout – 540 MWe PHWR









Plant

Layout(700 MWe)







119

Fukushima Accident - Implications





In depth review by utility and regulators

Extreme natural events

Loss of on site power sources and supply

Loss of water storage and supply

Safety of spent fuel

More than one unit getting affected

Severe accident management

Issues related to emergency handling

Fukushima Accident – Implications (Contd.)





Extreme natural events

Confirmation of design basis

Evaluating margins

Strengthening as required (like water proofing,

raising levels, additional shore protection measures)

More than one unit getting affected

Revision of procedures

Training and defining roles and responsibilities

Mitigation provisions to take into account this situation






Safety of spent fuel

Evaluating existing capabilities (before fuel isexposed following loss of cooling)

External water make up provisions

Loss of power and water sources/ supplies Alternate power sources

Augmenting on site water storage

Arranging water from nearby sources

„Hook up‟ schemes to various systems






Severe accident management

Enhancing severe accident management programmewith post Fukushima recommended provisions

Incorporation of severe accident management

guidelines

Hydrogen management

Containment safety

Issues related to emergency handling

Accessibility and communication enhancementFurther improving emergency preparedness



124

Regulatory review





All 220 MWe and 540 MWe units are operating

successfully.

Conceptual design review of 700 MWe PHWRs

completed and detailed design review is in

progress.Regulatory consent for construction of KAPP

3&4 obtained and construction is in progress.

Regulatory consent for construction of RAPP

7&8 is expected shortly.

Deployment schedule





Sr

No.Project Units

Expected criticality /

Remarks

1. Kakrapar Atomic Power Project 2 X 700 MWeKAPP-3: December 2014

KAPP-4: June 2015

2. Rajasthan Atomic Power Project 2 X 700 MWeRAPP-7: December 2015

RAPP-8: June 2016

3. Haryana Atomic Power Project 4 X 700 MWe Approved in principle

4. MP Atomic Power Project 2 X 700 MWe Approved in principle



Design, Safety and Operability Performances of 220 MWe, 540 MWe & 700 MWe PHWRs in India

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