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DESIGN AND DEVELOPMENT OF INDIAN PRESSURIZED WATER REACTOR

R. N. Sen (Advanced Light Water Reactor, BARC, Visakhapatnam, India)

A. K. Balasubramanian, (Engineering, NPCIL, Mumbai, India)

R. S. Yadav (Reactor Projects Group, BARC, Mumbai, India)

1. Introduction

The IPWR is an indigenous project of Department of Atomic Energy (DAE) with joint

participation of Bhabha Atomic Research Centre (BARC) and Nuclear Power Corporation of

India Limited (NPCIL). In this effort it is planned to construct a 900 MWe PWR (IPWR) plant.

It is also planned to source all raw material and equipment from various industries within India.

The plant is being designed as suitable for setting up at seismic zone-IV, which is the

highest seismic excitation in the country for any plant site. Once the design is standardized the

plant can be set up at any location in the country. Identification of an IPWR site for a twin

reactor plant is initiated.

Each of the proposed twin unit reactors is designed for 900 MWe power. Initially, first

IPWR will be operated at 700 MWe using available technology of secondary side of 700 MWe

PHWR. Latest regulatory requirements and state of the art technology is followed for design of

critical equipment and structures. The IPWR is designed for 60 years of plant life with average

80% times for full power operation. Irradiation damage of RPV core belt pressure boundary is

limited by providing large water gap between core and RPV wall. Also no weld is planned in

RPV in the active core region.

Slightly enriched UO2 is used as fuel. Light water is used as coolant and moderator for

the reactor. Off-line refueling is carried out for the reactor during shut-down after specific period

(~18 months) of full power operation.

Steam Generators supply nearly dry saturated steam to the turbine. Turbine is coupled in

tandem with an electrical generator, which produces electricity. Generator voltage is stepped up

by the generator transformer, which in turn is connected to the switchyard. Generated power is

transmitted to the grid from the IPWR station at 400KV.

Reactor protection system ensures shutdown requirements through fast acting shutdown

actuation. Reactor regulating system enables automatic control of reactor power and maintains

neutron flux profile.

The IPWR design incorporates use of modern technologies for reactor safety as well as

dedicated control systems. Any off-normal situation is detected through diverse control and

protection systems and immediate corrective action is actuated. In addition to computer based

control and protection systems, separate diverse protection system with conventional hardware is

provided.

In a situation of prolonged station black out condition, a passive decay heat removal

system starts working for cooling the core to give reactor autonomy for minimum 7 days. Also

separate air-cooled DG sets are placed at higher elevation to assure power supply for core

cooling under emergency conditions. In case of core-meltdown has happened, system exist for

corium retention and confinement.

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Ultimate heat sink is atmosphere. Heat is rejected by use of draft cooling towers that will

draw make-up water from the nearby water resource at plant site.

Plant Data

• Reactor Power : 900 MWe

• Core discharge Burn-up : 46,260 MWD/T

• Primary System Design Pressure : 17.7 MPa

• Primary System Operating Pressure : 15.7 MPa

• Primary System Design temperature : 350⁰C

• Reactor Coolant inlet temperature : 292⁰C

• Reactor Coolant outlet temperature : 323⁰C

• Number of Fuel Assemblies : 151

• Refueling cycle : 1/3rd

core every 18 months

• Reactor Service life : 60 Years

2. Design Principles and Guidelines

The safety goal of protection of public from accidental release of radioactivity in IPWR

plant is achieved by adherence to the following well-established principles and guidelines.

1) Application of defense-in-depth approach, incorporating several levels of defense, by-

- Sound design, construction and operation to prevent failures and deviations from

normal operation.

- To protect and intercept incipient failures and deviation from normal operations

conditions, in order to prevent these from escalating into accidents.

- To limit the consequences of accident conditions.

- In addition to the above, for more severe events, protection of the public by making

use of ultimate safety capability of the plant and appropriate plans emergency actions.

2) Application of defense-in-depth concept for confinement of radioactive material by a

series of physical barriers, each backing the others.

3) Provision of more than one means / systems for performance of each of the three safety

functions viz. shutdown of reactor, core cooling and confinement of radioactivity.

4) Provision of redundancy in systems important to safety, having mitigation function,

including safety systems, such that at least the minimum safety function can be

performed even in the event of failure of a single active component in the system.

5) Analyzing reliability and unavailability targets of safety systems and equipment.

6) To prevent common cause failure provision of physical and functional separation and

independence to the extent possible (including cabling) in practical sense shall be

- Between process systems and related safety systems.

- Among systems performing the same safety function.

- Among redundant components within a system.

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7) Consideration given at all stages of design for logics and instrumentation to fail in the

safe direction.

8) Provision of periodic testability of active components in systems important to safety

having mitigation function, preferably on power.

3. Plant Layout

The layout has following main provisions:

a) The layout is based on twin unit module concept with independent operation of each unit

with only some common facilities.

b) The Reactor Building, Safety Buildings, Fuel Building and the Access Tower form the

Nuclear Building which is located on a common basement.

c) Unprotected Safety Related Buildings are kept out of Low Trajectory Turbine Missile

zone of both units of IPWR turbines.

d) An equipment of particular seismic class is housed in same seismic class structure.

e) For each Reactor unit, four numbers of Class-3 power Emergency Diesel Generators are

housed in two separate Emergency Power Supply Buildings.

f) The twin unit module in the nuclear island has been so chosen that it will be possible to

enforce single point entry in the radiation zones and follow radiation-zoning philosophy.

g) Other aspects such as accessibility, maintainability and ease of construction has also been

given due consideration in arriving at the plant layout.

h) Special provision of adequate space adjacent to the Reactor Building, Fuel Building and

Turbine Building has been made to enable maneuvering of the high capacity outdoor

mobile crane during erection and handling of heavy equipment. Reactor components like

RPV, SG etc. will be taken inside the Reactor Building through Fuel Building via main

equipment hatch.

i) Main Control Room is located in Safety Building-2 and Separate Back-up Control Room

is located in Safety Building-3.

j) The Common Utility Building houses stores, laboratories, restaurants, health physics

room, warehouses, offices related to plant operations etc.

k) The Main Administrative Building consists of administrative wing, canteens,

auditoriums, nuclear training centre, simulators etc.

l) Buildings and roads are laid out to avoid permanent obstruction by planning proper entry

from nearby road as well as adjacent buildings. Turning radii at road curves around the

main plant area are suitably provided so that heavy duty cranes can also freely move, turn

and reverse.

4. Design of Reactor Systems and Equipment

4.1 Reactor Core Design

The IPWR is designed with enriched Uranium Oxide as fuel and light water as coolant in

a hexagonal lattice arrangement. An important feature of the IPWR is the use of Integrated Fuel

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Burnable Absorber (IFBA) where Gadolinium is used as the burnable absorber in the fuel. This

helps in achieving a better power profile over the cluster. The absorber content is optimized with

the aim of maximizing the discharge burnup and achieving acceptable pin power profile in the

fuel assembly. The design aims to minimize the initial soluble boron content, achieve negative

temperature loads, monitoring capability at all stages of operation and to maximize fuel

utilization by optimizing the batch size. Also the design aims to achieve all the reactivity

coefficients to be negative.

The core has 151 fuel assemblies arranged in a hexagonal pitch. The core is designed for

a rated power of 2700 MWth (900 MWe) with acceptable linear heat generation rates. The fuel

management is governed by excess reactivity and power distribution profiles.

The reactivity control is achieved by control rods having neutron poison and chemical

shim in the form of soluble boron. The reactor shut down is achieved by control rod assemblies.

The design uses 121 fuel assemblies to house the rod control clusters made up of boron carbide

and Dysprosium Titanate. They provide reactivity control for power defect compensation

(reactivity changes due to temperature changes with power), for rapid shutdown, for reactivity

changes due to coolant temperature changes in the power range, and for reactivity changes due to

void formation. The control assemblies are grouped suitably in different working groups to

function for both control and shutdown of the reactor. Of these a few assemblies of the last

control group have partially inserted control rods to provide reactivity control. Burnup

compensation is achieved by regulating the soluble boron content. In addition IPWR has a

Liquid Boron Injection System (LBIS) for a fast injection of boron under certain shutdown

requirement conditions.

Nuclear instrumentation system consists of a system of in-core and ex-core monitoring

devices. Each fuel assembly is provided with one instrumentation tube. For in-core monitoring

Self-Powered Neutron Detectors (SPNDs) distributed in seven axial locations in a tube are used

and are provided in 50 fuel assemblies. A few detectors are located in the core baffle region to

aid in reactor start-up. The ex-core monitoring is achieved by ion chambers located in two rings

beyond the reactor pressure vessel. The first ring containing 6 detectors are used to monitor the

fuel loading. The next ring of 12 Ion Chambers will perform the protection function using

triplicated logic.

The design intent is to use a three batch refueling scheme with an effort to flatten the

power distribution and reduce the leakage fluxes. The fuel loading pattern and the use of thermal

shield minimizes the fluence at the Reactor Pressure Vessel.

4.2 Fuel Design

Each fuel assembly of IPWR is hexagonal with 331 lattice locations containing 311 fuel

pins, 18 locations for control guide tubes, 1 instrumentation location and a central water rod.

The fuel assembly uses an average enrichment of 4.5% which is estimated to give a targeted

discharge burnup of about 50,000 MWD/T. Gadolinium is used as burnable absorber in 24 pins

optimally located in the fuel assembly to provide reactivity suppression and acceptable local

peaking factors.

The fuel design is made based on effective burning of Gadolinium. Gd is used in the

fresh reload fuel assemblies. The design envisages that there is no reactivity penalty in

subsequent cycles. The core simulations have been done iteratively with the different Gd pins

and 5% Gd in 24 pins is optimally selected for required burnup characteristics.

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Soluble boron is used in the moderator for compensating long term reactivity changes

such as fuel depletion and fission product build-up; cold to hot zero power reactivity change,

reactivity changes produced by intermediate-term fission products, such as xenon and samarium

and burnable absorber depletion. Over the operating cycle, soluble boron in the coolant is

gradually removed till the end of cycle where there is no boron. The boron removal rates are

specified with respect to the reactivity insertion criteria.

4.3 Equilibrium Core configuration

The general layout of the fuel assemblies

in the core is shown in Fig.1. The equilibrium

core has been optimised with a 3 batch

loading pattern. The equilibrium core

configuration has been designed such that

during the cycle burnup, the radial peaking

factor, overall peaking factor and the

reactivity coefficients are within design

limits. The core characteristics of IPWR are

given in Table-1.

The equilibrium core has been estimated

as having an operating cycle length of about

410 full power days. The cycle burnup is ~

14,500 MWD/T and estimated core averaged

discharge burnup is ~ 46,260 MWD/T.

The power distribution has been

optimized with respect to heat removal

aspects and thermal hydraulics of the core. Fig. 1: Equilibrium core configuration

Table-1: Core characteristics of IPWR

No. of fuel assemblies 151

No of assemblies per batch 1 / 51, 2 / 50, 3 / 50

Cycle length Full power Days 410

Cycle burnup MWD/T 14,500

Average discharge burnup MWD/T 46,260

4.4 Fuel Pins and Assembly

The fuel assembly accommodates and holds the Fuel Pins, Burnable poison pins at the

required pitch and provides adequate support. It provides coolant flow path for removal of heat

generated in the fuel pins and the guide tube for control rod and instrumentation. The fuel

assemblies are arranged in a hexagonal array. The fuel assemblies are housed in a core barrel

made of a stainless steel grid which has provision for coolant to enter from the bottom and exit

from the top of the core. The active height of the fuel assembly is ~ 3600 mm.

For making fuel pins sintered uranium dioxide (95%) pellets are used, which provide

optimum conductivity and porosity to accommodate the fission gas. This also helps in delay of

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pellet clad interaction. Zr-1%Nb is used as clad material (~46,260 MWD/T burn-up) having

lower corrosion and hydrogen pick-up properties compared to other Zr-alloys.

4.5 Reactor Pressure vessel

The Reactor Pressure Vessel (RPV) houses the heat generating reactor core and acts as a

boundary against the release of fuel fission products. It supports, locates & aligns the reactor

core and internals, and also properly locates & aligns the control rod drive mechanisms and in-

core instrumentations. The vessel is designed and manufactured to the requirements of Section

III-NB for class-1 components of the ASME Boiler and Pressure Vessel Code. The main

dimensions of reactor vessel are shown in Figs. 2 & 3. RPV shell, flanges, and upper and lower

heads are made from special quality low alloy steel forgings of grade SA-508 Class-3. This grade

of material is used because of its high strength, resistance to neutron irradiation, and good

weldability. Inside surfaces in contact with coolant are cladded with stabilized grade austenitic

stainless steel weld deposition to minimize corrosion.

Design Features:

(a) Main Nozzles: The 4 Inlet and 4 Outlet nozzles at two different elevations which reduce

the number of penetrations in shell flange forging, better separation between inlet plenum

/ outlet plenum and better layout.

(b) No weld in core belt shell. This helps in achieving lesser degradation of the material and

therefore longer life.

(c) Surveillance coupons at mid-core plane near vessel inner diameter.

(d) Provision for In-Service-Inspection of full vessel from inside.

(e) Water gap between core to vessel inner diameter is adjusted with stainless steel baffle,

core barrel, thermal shield and water in-between to optimize the EOL fluence (E > 1

MeV) maximum to 2 x 1019

n/cm2

for 60 years life (Calendar years) with 80% of times at

full power.

Pressure-Temperature Limits:

Safe working Pressure -Temperature limits are established to satisfy the requirements of

ASME Section III – Appendix-G using procedures of NUREG-0800 and 10 CFR Part 50

Appendix G to protect RPV against brittle fracture under all system hydrostatic and operational

conditions.

Fig. 2: Cross sectional view of Reactor Pressure Vessel at core belt region (Thermal Shields not shown)

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Fig. 3: Reactor Pressure Vessel

4.6 Primary Heat Transport (PHT) System

The PHT system transfers the heat generated in the reactor core to the Steam Generator

(SG) wherein the heat is used to make steam in the secondary system to be used for generating

electrical power in the turbine-generator. The schematic diagram of primary heat transport

system is shown in Fig. 4. It consists of four Primary Coolant Circulating Loops (PCCL)

connected to RPV. Each loop has one Primary Coolant Pump (PCP), one U-tube steam

generator, associated piping and connections with other systems. The equipment in each loop of

PHT system is connected through primary piping. In addition, there are piping connections to

other components and with various reactor auxiliary systems.

A flow of ~19,200 m3/hr of primary coolant at a temperature of 292°C is delivered by

each pump into the reactor core. The coolant passes through the core, where it gets heated up to

323°C. The heated water leaves from RPV and enter into SG. In SG, coolant heat is transferred

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to feed water to generate steam. Primary coolant from SG after losing heat is pumped to core

inlet by PCP. The SGs are located above the core elevation to establish core flow by natural

circulation for decay heat removal in the event of PCPs being not available.

The PHT system is designed for following conditions:

Flow through reactor : 76,800 m3/hr

Design power : 2700MWth

Design pressure : 17.7 MPa

Design temperature : 350°C

The temperature and pressure in the PHT system do not exceed the maximum operating

temperature and pressure limit during normal and Anticipated Operational Occurrences (AOO).

The PHT system is protected against overpressure through pressuriser safety relief valve.

Fig. 4: Schematic of Primary Heat Transport system

4.7 Residual Heat Removal System (RHRS)

The RHRS transfers heat from the PHT system to the service cooling water system

through the Component Cooling Water System (CCWS) to reduce the temperature of the reactor

coolant during normal plant shutdown and cool down conditions and during refueling or

maintenance of PHT system and its components. The RHRS is designed to ensure that the

reactor core decay heat is safely removed from the reactor using four independent subsystems.

Each subsystem includes one RHR pump, one RHR heat exchanger, associated piping, valves,

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and instrumentation necessary for operational control. Any two of the four subsystems have a

100% capability for safe shutdown heat removal.

4.8 Chemical and Volume Control System (CVCS)

This is a major reactor auxiliary system of PHT system. The functions of the system are

to maintain required water inventory in PHT system as required by the Pressuriser level

controller, to control the concentration of boric acid in the primary coolant and to provide

purification of coolant through purification ion exchangers. Besides these function, the system is

also used to add chemicals in primary coolant, to give seal injection water to seals of PCPs and

to give auxiliary spray to pressuriser. It consist of a Regenerative Heat Exchanger (RHX), Non-

Regenerative Heat Exchanger (NRHX), filters, ion exchangers, two charging pumps, Volume

Control Tank (VCT), piping and valves. Hydrogen atmosphere is maintained in VCT to control

the concentration of H2 in the PHT coolant. CVCS consists of various subsystems for reactor

coolant storage, boron recycling and boric acid storage.

During normal operation nominal flow is maintained as required for purification.

However, the flow varies during boric acid addition or removal from primary coolant for power

control.

4.9 Fuel pool cooling and purification system

The fuel pool cooling and purification system removes the decay heat generated by the

spent fuel clusters; maintains uniform temperature in the storage bay water (32°C to 35°C),

filters out the suspended impurities for good visibility of bay water and provides adequate water

shielding for working personnel. The fuel storage pool is located inside reactor building and the

Fuel Storage Bay (FSB) is located inside fuel building, outside the reactor containment. These

two bays are connected through fuel transfer tube. FSB is provided for under water storage of

spent fuel and new fuel. Storage capacity of ~ 10 years is provided. Fuel clusters are stored

vertically in rack. Water shielding of 10 m above the rack in fuel-stored condition is maintained.

Minimum water shielding of 5 m (approx.) is maintained during movement of fuel cluster to

maintain minimum acceptable radiation level.

5. Reactor Safety System

Reactor safety system reduces the consequences of postulated accidents. Ultimate aim of

the system is to ensure public safety in the unlikely event of an accidental release of radioactivity

from the Reactor Coolant System (RCS). The engineered safety system will automatically act to

limit, control, and terminate unplanned events, while maintaining the radiation exposure to the

public well below the applicable regulatory limits and guidelines.

5.1 Reactor Shut-down Systems

IPWR has independent quick and reliable fail safe shut down system. Reactor shut down

system is designed to immediately terminate the nuclear chain reaction and maintain the sub-

criticality for prolonged period of time. IPWR has the following reactor shutdown systems:

a) Reactor Control and Shut-off Rod Drive Mechanism (CSRDM) - These are a series of

solid absorber rods that can be quickly inserted into the core (through control rod guide

tubes) to absorb neutrons and rapidly terminate the nuclear reaction. The magnetic jack

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type control and shut off mechanism is used for controlling and shutting down the

reactor. It can hold the absorber at fixed position or release it fast for quick shutdown.

b) Liquid Boron Injection System (LBIS) - This is an alternate mode to shut down the

reactor in Anticipated Transient Without Scram (ATWS) scenario. The nuclear reaction

is stopped by directly injecting a liquid poison (boric acid solution) into the core that

absorbs neutrons. This system is very effective during power excursion accident like Loss

Of Regulation Accident (LORA). The system is designed to automatically introduce

sufficient negative reactivity in few seconds on detection of ATWS event.

5.2 Emergency Core Cooling System (ECCS)

The primary function of the ECCS is to deliver emergency core cooling in the event

when primary coolant system is accidentally depressurized (i.e., a LOCA). The defense-in-depth

philosophy in terms of functional diversity, passive safety system, redundancy, independent

safety trains and fail safe design are incorporated. The system has four physically separated and

independent channels. The design criteria considers full spectrum of break sizes including a

break equivalent in size to the double ended guillotine break (DEGB) of the largest pipe in the

reactor coolant system. Flow Restriction Orifices (FROs) are used in Pressurizer surge line,

Purification line and other auxiliary system pipe lines to minimize blow down. The hydro

accumulator’s volume is selected on the basis that it should flood entire core at least 10 times.

The ECCS water is injected into core through cold leg and hot leg with 2x100% dedicated direct

pipe lines in each legs.

5.3 Passive Decay Heat Removal System (PDHRS)

This system removes decay heat from core continuously by natural recirculation of the

inventory on secondary side of SGs in the event of total loss of power supply. Four number

independent loops work simultaneously, each consisting of one water filled expansion tank of

50m3 capacity and two vertical emergency heat exchangers, connected to individual SG’s. The 8

numbers of emergency heat exchangers are kept submerged in two cooling water tanks each of

5000m3 capacity. The steam from SG is passed through water cooled HXs and condensate

returns back to SG. Steam pressure in expansion tanks is used to push water to SG feed line. The

water in the cooling water tanks acts as a heat sink. The capacity of the cooling water tanks is

designed in such a manner to enable passive cooling for a period of minimum seven (7) days

from the initiation of the event.

5.4 Core catcher (Molten Pool Retention System)

In IPWR design a corium retention system is provided. The in-vessel retention system

has a water flooding provision outside RPV following an accident requiring cooling of core for

removal of decay heat. In case a core melt happens, the core heat is removed through RPV body

by formation of steam of the flooded water outside.

6. Instrumentation and Control

The function of the instrumentation and control system in IPWR is to provide capability

to control, monitor and regulate the plant systems during normal operation of the plant and

protect the reactor against unsafe plant operation. The I&C systems also provides initiating

signals to actuate safety functions which are assigned to mitigate the consequences of accident

conditions and ensure safe shut down of the reactor. The design includes concepts on defense-in-

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depth and Diversity, Single Failure Criterion, Redundancy, Physical separation Isolation and

Functional Independence, Equipment Qualification and Protection, Unified Architecture, Self-

Diagnostics Functions and Human-System Interface, etc. The overall I&C are such that all the

safety functions are normally achievable from Main Control Room (MCR). In case MCR is

inhabitable then all the safety functions can be carried out from Back-up Control room (BCR).

The MCR is designed to perform centralized monitoring and control of I&C systems that are

necessary for use during normal plant operation, AOOs, and PIEs. HSIS are also designed to

reduce the potential for human errors and to allow easy operation by adopting sound Human

Engineering principles with adequate intelligence in its design. Independent sensor and analogue

systems are provided for assuring reactor trip when design trip signal is exceeded.

7. Radiological Protection

Accessible areas in the station are divided into areas according to the radiation potential.

The Accessible and Shutdown Areas are clearly distinguished. The Shutdown Areas have doors

and interlock system to ensure that inadvertent personnel entries in these areas are avoided. Entry

to other locations of high radiation potential is restricted by administrative control. In order to

minimise contamination and also to control its spread, the entire plant area is divided into four

distinct zones of different contamination potential. Each zone is clearly demarcated and provided

with inter-zonal check points equipped with contamination monitors. The personnel movement to the

plant is arranged from zones of lower to higher contamination potential for entry and vice versa for

exit. At the final exit point (zone 2 to zone 1 interface) automatic portal surveillance monitor is

installed.

For demonstrating compliance with prescribed limits and providing information on

changes in radiation levels, radiation monitoring system is provided in the plant. Monitoring

includes individual, workplace and environmental monitoring. Individual external and internal

monitoring for the occupational workers will be carried out on a routine basis. Workplace

monitoring will be done for external radiation, airborne contamination and surface

contamination. Monitoring for protection of the public includes monitoring of effluents from the

plant as well as environmental monitoring during normal operation and accident conditions.

The on-site / off-site radiation emergency planning is incorporated to ensure adequate

preparedness for protection of the plant personnel and members of the public from significant

radiation exposures in the unlikely event of an accident. The probability of an accident though

brought to a very low level by incorporating defense-in-depth approach, however, can never be

reduced to absolute zero and therefore this residual risk is sought to be mitigated by appropriate

siting criteria and implementing suitable arrangements for emergency planning and

preparedness.

8. Safety Classification

In IPWR, the Structures, Systems, and Components (SSCs) are graded and classified

according to the role played by them in the measures to control radiological hazards. Based on

the classification, their design requirements are accordingly established without compromising

the overall safety objective. This is achieved by identifying the different safety functions

performed by individual SSCs in terms of their role in achieving the safety objective. These

safety functions are then grouped and ranked into safety classes, taking into consideration the

consequences of failure of the safety function performed by SSC and the probability of

occurrence of a failure.

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9. Nuclear Security

Nuclear Security in IPWR will incorporate comprehensive technical, operating and

administrative measures which in association with nuclear safety measures will address any

possible potential threats of theft and sabotage by an external as well as insider adversary.

IPWR is provided with comprehensive Integrated Physical Protection System (IPPS)

backed up by strong administrative and operational measures which will be designed to address

site specific design basis threats (DBT). The security system of the plant will be operational at all

the time and for all credible scenarios. It is designed to perform the functions of detection, delay,

assessment and response as applicable to IPWR. The system provides adequate protection

following a multilayered architecture providing defense-in-depth, balance protection, graded

approach, tolerance to component failures using diversity and redundancy. Technical measures

for IPWR nuclear security includes multi-layered security arrangement whereby the most vital

and important areas / facilities are placed in the innermost layer.

10. Acknowledgement

The design and development of IPWR plant is a joint work of a group of engineers and

scientists working at Bhabha Atomic Research Centre (BARC) and Nuclear Power Corporation

of India Limited (NPCIL). The author acknowledges the effort put up by all these engineers and

scientists for their valuable contributions.

11. References

11.1 Conceptual design report of Indian PWR (IPWR) by R N Sen et. el. – Report No.

BARC/IPWR/RPG/09/2013/Rev.0, Sept-2013.

11.2 Detailed Project Report on 900 MWe Indian PWR (IPWR) by R N Sen et. el. – Report

No.: BARC-NPCIL/IPWR/10/2014/Rev.0, July-2014.