The PSA Approach for the Safety Assessment of Low-Power and Shutdown States

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    Pakistans Experience in Operating

    CNP-300s and Near Term

    Deployment Scheme

    Presented by:

    M. Kamran Chughtai

    Directorate of Nuclear Power Engineering Reactor

    PAKISTAN ATOMIC ENERGY COMMISSION

    IAEA Work Shop on

    Technology Assessment of Small and Medium-sized Reactors

    (SMRs) for Near Term Deployment

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    To generate electricity in a demonstrably safe,

    reliable and cost effective manner over the long

    term, for the benefit of our society and stake

    holders, as well as to consolidate the basis for

    development of the nuclear power program in

    Pakistan.

    Mission

    12/5/2011 2

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    Pakistans Nuclear Power Programme

    At present Pakistan has three operating nuclear powerplants;

    Govt. of Pakistan has planned to enhance Nuclear

    Power generation capacity till 8800 MWe by the year2030

    In this perspective, Two are under construction

    PWRs are the preferred choice in future because of

    design & operational experience

    312/5/2011

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    Status of Nuclear Power Plants in Pakistan

    4

    NPP TYPE MWe GRID CONNECT

    KANUPP PHWR 137 1972

    CHASNUPP

    Unit-1PWR 325 2000

    CHASNUPP

    Unit-2PWR 325 2011

    CHASNUPP

    Unit-3/4PWR 325 Under construction

    12/5/2011

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    Karachi Nuclear Power Plant (KANUPP)

    512/5/2011

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    Prime contractor and designer:Canadian General Electric Company Ltd.

    Civil consultant:

    Montreal Engineering Company

    Reactor type: CANDU Pressurized Heavy Water (PHWR)

    Gross output: 137 MWe

    Net station output: 125 MWe

    Construction date: 01 Aug ,1966

    Commercial Operation date: 07 Dec, 1972

    Re-licensing date: 31 Dec, 2007

    Current Net output: 80 Mwe

    Fuel Natural Uranium

    Moderator Heavy Water

    Coolant Heavy Water

    Thermal Output 432.8 MWth

    Plant is still operational after design life extension

    Karachi Nuclear Power Plant (KANUPP)

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    Chashma Nuclear Power Plant Unit 1 & 2

    712/5/2011

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    Plant Design / Specifications

    Commercial Supplier CNNC

    Rated Thermal Power 998.6 MWth

    Gross Electrical Output 325 MWe

    Net Electrical Output 300 MWe

    Operating Pressure 15.2 MPa

    Operating Temperature 280-302oC

    Equilibrium Cycle Enrichment 3.4 w/o

    Average Burnup 32000 MWD/MTUHeat flux hot channel factor 2.70

    Nuclear Enthalpy Rise Hot Channel 1.60

    Net Efficiency ~34 %

    Core Damage Frequency (CDF) 1.5210-5 /yr8

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    Chashma Nuclear Power Plant Units: C-3/C-4

    9

    C-3

    C-4

    C-3 C-4Contract Signing 20 Nov 2008 20 Nov 2008

    Contract Effective Date 31 Mar 2010 31 Mar 2010

    Groundbreaking 5 Aug 2010 1 Apr 2011

    First Concrete Pouring 4 Mar 2011 Jan 2012

    Provisional Acceptance (as percontract)

    End 2016 End 2017

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    Long-Term Nuclear Power Plan (9 additional units of 8,325 MW by 2030)

    2004 2006 2008 2010 2012 2014 2016 2018 2020 2022 2024 2026 2028 2030

    YEARS

    NPP4 K-2 1000 MW

    NPP5 K-3 1000 MW

    NPP6, 1000 MW

    NPP7 1000 MW

    NPP8 1000 MW

    NPP9 1000 MW

    NPP10 1000 MW

    NPP11 1000 MW

    NPP3 C-2 325 MW

    Long Term Planning for NPPs

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    Plant Overview

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    Reactor Coolant System

    Steam generator

    Pressurizer

    Reactor

    coolant

    pump

    Reactor

    pressure

    vessel

    Hot legCold leg

    12/5/2011 12

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    Reactor Core

    12/5/2011 13

    Parameter Value

    Number of Fuel Assemblies 121

    Equivalent Diameter of Core, m 2.486Core Average Active FuelHeight, cm

    290

    Height-to-diameter Ratio 1.1665H2O/UO2 Volume Ratio,Lattice(Cold)

    2.065

    Fuel Weight (as UO2), t 40.704

    Neutron Absorber 80%Ag,15%In5%Cd

    No. of Rod Cluster Control

    Assemblies 37No. of Burnable Poison Rods(First Core)

    576

    MaterialBorosilicate

    Glass

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    Fuel Assemblies

    12/5/2011 14

    Parameter Value

    Rod Array 1515

    Rods per Assembly 204

    Rod Pitch, mm 13.3

    Overall Transverse Dimensions,

    mm

    199.3199.3

    Number of Grids per Assembly 8

    Material of Grids GH-4169A

    Number of Guide Thimbles perAssembly

    20

    Material of Guide Thimbles 0Cr18Ni10Ti

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    Reactor Pressure Vessel Design Data

    12/5/2011 15

    Parameter Value

    Design/operating pressure, MPa 17.2/15.2

    Design temperature, 350Overall height of vessel and closure head,mm

    10366

    Thickness of insulation, minimum, mm 77Number of reactor closure head/studs 48

    Diameter of reactor closure head/studs,(minimum shank) mm

    151

    Inside diameter of flange, mm 3260Outside diameter of flange, mm 3990

    Inside diameter at shell, mm 3374Inlet nozzle inside diameter, mm 700Outlet nozzle inside diameter, mm 700Cladding thickness, minimum, mm 4Lower head thickness, minimum, mm 115Vessel belt-line thickness, minimum, mm 170

    Closure head thickness, minimum, mm 155

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    Pressurizer Design Data

    12/5/2011 16

    Parameter Value

    Design pressure, MPa 17.2

    Design temperature, oC 370

    Surge line nozzle diameter, mm 273

    Heatup rate of pressurizer using heatersonly, oC/h

    30

    Internal volume, m3 35

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    Steam Generator Design Data

    12/5/2011 17

    Parameter ValueDesign pressure, primary side, MPa 17.16

    Design pressure, secondary side, MPa 7.55

    Design temperature, primary coolantside,

    350

    Design temperature, secondary side, 320

    Design steam flow rate, t/h 1010Heat transfer surface area, m2 3088.67

    Maximum moisture carryover, wtpercent, %

    0.25

    Overall height, m 17.678

    Number of U-tubes 2977

    Number of separators 53

    U-tube nominal diameter, mm 22

    Tube wall nominal thickness, mm 1.2

    Number of manways 4

    Inside diameter of manways, mm 457

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    NPP Structures

    1812/5/2011

    All the systems and equipment of a nuclear power plant are housed inabout 30 buildings and structures of different sizes. The major plantstructures are grouped into the following:

    Nuclear Island

    Reactor Building Nuclear Auxiliary Building

    Fuel Storage Building

    Electrical Building

    Diesel Generator Building

    Conventional Island

    Turbine Generator Building

    De-mineralized Water Building

    Switch Yard

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    1912/5/2011

    Balance of Plant Liquid Radwaste Solidification

    Building

    Solid Radwaste StorageBuilding

    Low-level Radwaste StorageHouse

    Hot Laundry

    Boiler House

    Maintenance Building

    Warehouse

    Control Access

    Ultimate Heat Sink

    Essential Services Water PumpStation

    Outdoor Engineering

    NPP Structures (Cont)

    Intake Structure

    Circulating Cooling Water PumpStation

    Water Treatment Plant

    Sewage Treatment PlantDrainage Structure

    Parking Area

    Fire Pump Station

    Cafeteria

    Guard House

    Hazard Cargo Storage House

    Administration Building andEmergency Centre

    Environmental Radiation

    Monitoring Hut

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    Nuclear Power plant Systems

    There are about 200 systems in CHASNUPP. These are classified, inaccordance with their functions, into following categories:

    - Reactor Core/Fuel - Nuclear Systems- I & C / Computer Systems - Conventional Systems

    - Radiation Monitoring Systems - Electrical Systems- Communication Systems - Lighting Systems- Common Systems - HVAC Systems- Miscellaneous Systems

    NI Systems 109

    CI Systems 33

    BOP Systems 52

    2012/5/2011

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    Safety Feature

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    Engineered safety Feature

    CONTAINMENT SYSTEMS

    Containment system which provides the last barrier against thepost-accident releases consists of containment structures,containment heat removal system, containment isolation system,

    and containment combustible gas control system. Containment system is designed such that for all break sizes, up

    to and including the double-ended severance of a reactor coolantpipe or secondary system pipe, the containment peak pressureremains below the design pressure, with adequate margins and

    Beyond Design Basis Accidents (BDBA), and it can be reducedto half of the design value in 24 hours by the safeguards system

    2212/5/2011

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    Engineered safety Feature

    EMERGENCY CORE COOLING SYSTEM The emergency core cooling system is called the safety injection

    system (SIS). The SIS is designed to cool the reactor core. Itprovides the capability of cooling following the initiation of thefollowing accident conditions:

    The pipe break of reactor coolant pressure boundary (including the double-endedrupture of the largest reactor coolant pipe) or inadvertent relief valve or safetyvalve opening in the reactor coolant system which would result in a dischargelarger than that could be made up by the normal makeup system.

    Rupture of a control rod drive mechanism causing a rod cluster control assemblyejection accident.

    The pipe break of secondary system (including the break of the largest pipe inthe secondary system) or inadvertent relief valve or safety valve opening in thesecondary system.

    Rupture of a steam generator tube.

    2312/5/2011

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    Engineered safety Feature

    HABITABILITY SYSTEMS

    Habitability Systems are designed to ensure that Control Roomoperators can remain inside the spaces served by the MainControl habitability Ventilation System during all normal andabnormal station conditions.

    The Habitability Systems cover all the equipment, supplies, andprocedures provided to ensure that Control Room operators areprotected from postulated releases of radioactive materials, toxicgases, smoke, and steam.

    The environments in all spaces served by the Main Control

    Habitability Ventilation System (Control Room envelope) arecontrolled within specified limits.

    The Habitability Systems are designed to support a maximum ofseven persons during normal and 30 days abnormal stationoperating conditions.

    2412/5/2011

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    Engineered safety System

    FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS Engineered Safety Feature (ESF) Filter Systems

    The following filtration systems that are required to perform the safety-related functionssubsequent to Design-Basis Accident (DBA) and BDBA (SCS only) are provided:

    Main Control Habitability Ventilation fresh Air cleaning Units

    Primary Nuclear Auxiliary Building Exhaust System:

    Fuel Storage Building Emergency Exhaust System:

    Containment Spray System

    The Containment Spray System (SCS) is designed to remove fission products, primarilyelemental iodine, from the containment atmosphere for the purpose of minimizing the

    offsite radiological consequences following the design-basis loss-of-coolant accident andBDBA. At the same time, the spray water serves to nominally reduce containmenttemperature and pressure during the injection phase

    Fission Product Control Systems

    The containment is steel-lined post tensioned pre-stressed concrete cylinder with ashallow dome. The containment is designed to withstand post-accident pressure and

    temperature and to contain the radioactive material that could be released from a loss ofintegrity of the reactor coolant pressure boundary. 2512/5/2011

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    Enhanced SafetyFeature

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    Enhance Safety Features in C-2 Design

    More than 160 design changes incorporated on the basis offeedback from C-1 and some from Qinshan-1.

    Use of PSA to provide insight into safety of different aspects

    Severe Accidents and Beyond Design Basis AccidentsConsideration.

    Development of Symptom based Emergency OperatingProcedures (SEOPs) for C-1 & C-2 with the help of designer.

    12/5/2011 27

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    Major Improvements of CHASNUPP-2 over

    CHASNUPP-1

    12/5/2011 28

    DESIGN FEATURES TARGET

    The Reactor Cavity Flooding

    system

    Take water from the Refueling Water Storage Tank can

    be injected to reactor cavity up to the bottom elevation

    of lower push-pull rod of RPV in case of severe

    accident.

    Hydrogen concentration

    monitoring system

    To provide information about containment hydrogen

    volumetric concentration continuously.

    Passive hydrogen

    recombination facilities

    To operate during the event of design basis accident and

    severe accident.

    The countermeasure for

    Heterogeneous Boron

    Dilution

    To prevent spurious automatic or manual injection of

    non-borated water by Anti-Dilution Protection (ADP)

    signal.

    The loose parts monitoring

    system (LPMS)

    To detect loose parts in the primary system as early as

    possible. The sensor location of LPMS is RPV, SG and

    RCP.

    Installation of motor

    throttled valve on pressurizer

    To fulfill the function of overpressure protection and

    rapid depressurization.

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    Severe Accident Management

    Detection Instruments with their limiting capability to meet the Severe

    Accidents environment

    Wide range hydrogen concentration monitoring system

    High temperature indicator in the reactor cavity

    Prevention

    A motor throttle valve to function during abnormal conditions toprevent high pressure events, to avoid possible direct containment

    heating and containment failure at early stage Anti-dilution mechanism or interlocks to prevent inadvertent

    boron dilution in the primary system

    12/5/2011 29

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    Severe Accident Management (Cont.) Addition of a diverse diesel generator in addition to two EDGs to

    withstand SBO

    Increased design pressure of Residual Heat Removal System pipingto prevent IS-LOCA

    Mitigation

    Reactor Cavity flooding, Cooling Water Injection system toincrease possibility of in-vessel corium retention or mitigate ex-vessel molten corium concrete interaction in case of reactor vesselfailure

    Passive Hydrogen Recombination Facilities

    Strengthening the containment boundary including the penetrations

    SAMGs to be jointly developed by the designer and the utility

    12/5/2011 30

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    Hydrogen Passive Auto-catalyticRecombiners

    Hydrogen generation can be described in three phases.

    In-vessel core degradation phase, where large surfaces of metallic zirconiumare available and temperature excursion is amplified by the exothermicreaction.

    Hydrogen may also be generated during the in-vessel relocation phase

    according to the failure mode of the lower head, consisting of solidified corematerial and the lower support plate, and the availability of water in the lowerhead.

    Hydrogen will be produced by melt core-concrete interaction (MCCI) in caseof RPV failure.

    Hydrogen passive auto-catalytic recombiners (PAR) have been implementedinside the containment both for DBA and severe accidents instead of theactive hydrogen recombiner (used in C-1). The PARs will be arranged atdifferent locations inside the containment and will be able to recombinehydrogen when the concentration of hydrogen is above 2~4 % to avoid

    hydrogen detonation.3112/5/2011

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    32

    Cavity Flooding

    Increase possibility of in-vesselcorium retention

    The accident management strategy toflood the reactor cavity with refueling

    water storage tank (RWST) water andsubmerge the reactor vessel is creditedwith preventing vessel failure

    Protection of the integrity of thereactor vessel containing the molten

    corium by cooling its external surface Lower part of the reactor vessel

    should be submerged with coolingwater

    12/5/2011

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    Loose Part Monitoring System(Including Trouble AnalysisSystem)

    The principal functionalrequirements of the loose partsmonitoring system (LPMS) are todetect loose parts in the primarysystem as early as possible. Earlydetection of the loose parts canavoid or mitigate safety-related

    damage or malfunctions ofcomponents in the primarysystem so as to minimizeeconomic losses.

    Loose Part Monitoring System

    LPM015LPM006

    LPM004

    LPM016

    LPM01LPM01

    LPM01

    LPM01

    LPM008

    LPM009

    LPM010LPM001LPM002LPM003

    LPM007

    LPM005

    Sensor Locations

    12/5/2011 33

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    7167.6

    55

    78.2

    68.2

    85.1

    97.8

    69.9

    6660.2

    51.6

    68.9 66.4

    82

    96.4

    74.2

    51.5

    67.1

    3

    17

    611

    6 5 37

    3 3

    76.1

    54.7

    0

    25

    50

    75

    100

    2000 2001 2002 2003 2004 2005 2006 2007 2008 2009

    %

    Availability F actor C apacity F actor Unplanned Outages

    Operational Performance (C-1)

    (upto 30 September, 2009)

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    Cumulative

    (Since 15 Sep 2000)

    After RFO-5

    (Since 23 Jan 2009)

    Reactor Operation, EFPDS 2125 198

    Generation, GWh 17399 1498 GWh

    Export, GWh 16093 83%

    Availability Factor , % 72.6 86.8 %

    Capacity Factor , % 68.7 76.7

    Outages, # 74(10 Planned) 3

    Longest Continuous

    Operation

    162 Days (Nov. 17, 2005 ~ April 29, 2006)

    Operational Performance (C-1)

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    Reactor Core Analysis results

    ICFM

    Design

    CYCLE 1 CYCLE 2 CYCLE 3 CYCLE 4 CYCLE 5 CYCLE 6 CYCLE 7

    CAL MEAS. CAL MEAS. CAL MEAS. CAL MEAS. CAL MEAS. CAL MEAS. CAL MEAS.

    Cycle

    length

    (EFPD)

    485 490 325 332 375 381 403 405 400 401 375 377 377 380

    CBC HZP

    (ppm)1297 1335 1321 1298 1477 1447 1567 1508 1447 1467 1457 1453 1474 1461

    T1 Worth

    (pcm)3068 2987 1321 1229 1907 1808 1405 1366 1710 1776 1771 1801 1590 1628

    Overlap

    banks

    worth

    (pcm)

    5566 5530 2510 2235 2368 2110 2928 2765 1952 1803 1708 1654 1574 1571

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    Periodic Safety

    Review(Safe and reliable operation - Ten (10)

    years of PWRs)

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    PSR Purpose

    The PSR is a tool to carry out a systematic and comprehensivereview of the safety case at regular intervals during plant life.

    Demonstrate that the plant is as safe as originally intended

    Obtain an overall view of actual plant safety

    like ageing effects,Modifications, operating experience feedback, development intechnology, etc.

    Compare current level of safety with latest standards and state ofknow-how and identify improvements at justifiable cost

    Obtain a broad integrated view of current safety of nuclearinstallations.

    3812/5/2011

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    PSR Objectives

    Confirm that the plant is as safe as originally intended, conforms tocurrent national safety standards and practices and the licensing

    basis remains valid and identify areas where safety improvements

    can be made at justifiable cost.

    Determine if there are any structures, systems, or components that

    could limit the safe operation of the plant in the next ten (10) years.

    To ascertain the adequacy of arrangements that are in place to

    maintain plant safety.

    Fulfill PNRA requirement for renewal of Operating License fornext ten (10) years of operation.

    Provide a higher level confidence in safety at national and

    international level.

    Maintain and upgrade knowledge base for the plant.3912/5/2011

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    Review Strategy

    Compilation of changes in Standard Review Plan (NUREG-0800) Identification of issues from Pakistan Nuclear Regulatory Authority

    (PNRA) routine inspections

    Identification of issues from QA Audit findings and surveillance

    Identification of issues from external reviews

    Review of each safety factor and gap identification (issues) fromcurrent regulations, codes, standards, and current SRP

    Compilation of issues master list and associated corrective actions

    Short listing and risk assessment of issues where changes cannot bemade

    Ranking of issues for which corrective actions to be implemented

    Approval of corrective action program schedule from PNRA

    Implementation of corrective actions4012/5/2011

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    Safety Factors

    1. Plant design

    2. Actual condition of Systems,Structures and Components(SSCs)

    3. Equipment qualification

    4. Ageing

    5. Deterministic safety analysis

    6. Probabilistic safety analysis

    7. Hazard analysis

    8. Safety performance

    9. Use of experience from other

    plants and research findings41

    10. Organization andadministration

    11. Procedures

    12. The human factor

    13. Emergency planning

    14. Radiological impact on theenvironment

    12/5/2011

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    PSR Corrective Action Plan

    Plant Design

    Operating procedures changes (inclusion of new steps/configurations)may be analyzed by design group.

    Severe Accident Management Guideline (SAMGs) development.

    Severe accident analysis may be carried out based on C-2 FSAR.

    Deterministic Safety Analysis

    Loose Parts Monitoring System (LPMS)

    Re-analysis may be carried out for the accident of reactor coolantpump shaft seizure due to change of control rod drop time by

    considering the effects of earthquake and uncertainties and scramreactivity worth.

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    PSR Corrective Action Plan

    Probabilistic Safety Analysis

    Low Power analysis may be considered in PSA Level-1 Plus.

    Shutdown analysis may be considered in PSA Level-1 Plus.

    Internal flood analysis may be considered in PSA Level-1 Plus.

    Internal fire analysis may be considered in PSA Level-1 Plus..

    Hazard Analysis

    Fire Hazard Analysis (FSAR Chapter-9) may be updated.

    Pipe Whip analysis described in FSAR Chapter-3 may be

    updated.

    Containment analysis against aircraft crash

    4312/5/2011

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    PSR Corrective Action Plan

    Human Factors Develop overall plant level procedure for assessing and

    monitoring the health and fitness of plant employees.

    A comprehensive lecture on safety culture may be delivered toall plant personnel on annual basis.

    Equipment Qualification

    Environmental monitoring program for qualified equipmentmay be developed.

    Procedure to control list of qualified equipment may be

    developed.

    Procedure(s) for analysis of the effects of equipment failures onequipment qualification and appropriate corrective actionsand/or safety improvements to maintain equipmentqualification may be developed.

    4412/5/2011

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    PSR Corrective Action Plan

    Ageing

    Ageing Management Program (AMP) may be developed.

    Potential ageing degradation that may affect the safetyfunctions of SSCs may be documented.

    AMP training may be imparted to relevant personnel.

    Requirement for monitoring of physical condition of AMPSSCs, actual safety margins, and any features that would limitservice life may be included in AMP.

    AMP software & tools may be acquired.

    Radiological Impact on the Environment

    Tritium monitoring may be established.

    Procedure for estimation of liquid/gaseous/solid waste duringRFOs may be developed.

    Quarterly administrative targets for discharge limits may beestablished. 4512/5/2011

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    46

    Safe and reliable operation - Ten (10) years of PWRs and forty (40) years ofCANDU

    Design, analysis and engineering of systems and components of NuclearIsland, Conventional Island and Balance of Plant of NPPs.

    Design and analysis of nuclear reactor core and fuel, fuel management,thermal hydraulics, safety analysis, accident analysis (design basis andbeyond design basis accidents), shielding design and licensing.

    Site selection and evaluation, design, analysis of buildings and structures,geotechnical Investigations, and environmental impact assessment of NPPs.

    Technical support to operating NPPs for safety significant design changes,

    criticality and startup, fuel management & operational core analysis, periodicsafety review etc.

    Structural analysis/design of reactor core and internals and Plant ageingmanagement of NPPs.

    Capabilities

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    THANK YOU