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7/29/2019 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
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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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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
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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.
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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.
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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.
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Major Improvements of CHASNUPP-2 over
CHASNUPP-1
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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
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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
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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
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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
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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.
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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
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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
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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.
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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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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