Pakistan s Experience in Operating CNP-300s and Near Term Deployment Scheme

Pakistan s 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

Mission 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. 12/5/2011 2

Pakistan s Nuclear Power Programme At present Pakistan has three operating nuclear power plants; Govt. of Pakistan has planned to enhance Nuclear Power generation capacity till 8800 MWe by the year 2030 In this perspective, Two are under construction PWRs are the preferred choice in future because of design & operational experience 12/5/2011 3

Status of Nuclear Power Plants in Pakistan NPP TYPE MWe GRID CONNECT KANUPP PHWR 137 1972 CHASNUPP Unit-1 CHASNUPP Unit-2 CHASNUPP Unit-3/4 PWR 325 2000 PWR 325 2011 PWR 325 Under construction 12/5/2011 4

Karachi Nuclear Power Plant (KANUPP) 12/5/2011 5

Prime contractor and designer: Canadian General Electric Company Ltd. Civil consultant: Karachi Nuclear Power Plant (KANUPP) 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

Chashma Nuclear Power Plant Unit 1 & 2 12/5/2011 7

Plant Design / Specifications Commercial Supplier CNNC Rated Thermal Power 998.6 MW th Gross Electrical Output 325 MWe Net Electrical Output 300 MWe Operating Pressure 15.2 MPa Operating Temperature 280-302 o C Equilibrium Cycle Enrichment 3.4 w/o Average Burnup 32000 MWD/MTU Heat flux hot channel factor 2.70 Nuclear Enthalpy Rise Hot Channel 1.60 Net Efficiency ~34 % Core Damage Frequency (CDF) 1.52 10-5 /yr 8

Chashma Nuclear Power Plant Units: C-3/C-4 C-3 C-4 C-3 C-4 Contract 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 per contract) End 2016 End 2017 9

Long Term Planning for NPPs Long-Term Nuclear Power Plan (9 additional units of 8,325 MW by 2030) NPP9, 1000 MW NPP11 1000 MW NPP10 1000 MW NPP8, 1000 MW NPP7, 1000 MW NPP6, 1000 MW NPP5, K-3, 1000 MW NPP4, K-2, 1000 MW NPP3, C-2, 325 MW 2004 2006 2008 2010 2012 2014 2016 2018 2020 2022 2024 2026 2028 2030 YEARS

Plant Overview

Reactor Coolant System Steam generator Pressurizer Cold leg Hot leg Reactor coolant pump Reactor pressure vessel 12/5/2011 12

Reactor Core Parameter Value Number of Fuel Assemblies 121 Equivalent Diameter of Core, m 2.486 Core Average Active Fuel Height, cm 290 Height-to-diameter Ratio 1.1665 H2O/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 37 No. of Burnable Poison Rods (First Core) 576 Material Borosilicate Glass 12/5/2011 13

Fuel Assemblies Parameter Value Rod Array 15 15 Rods per Assembly 204 Rod Pitch, mm 13.3 Overall Transverse Dimensions, 199.3 199.3 mm Number of Grids per Assembly 8 Material of Grids GH-4169A Number of Guide Thimbles per 20 Assembly Material of Guide Thimbles 0Cr18Ni10Ti 12/5/2011 14

Reactor Pressure Vessel Design Data Parameter Value Design/operating pressure, MPa 17.2/15.2 Design temperature, 350 Overall height of vessel and closure head, 10366 mm Thickness of insulation, minimum, mm 77 Number of reactor closure head/studs 48 Diameter of reactor closure head/studs, 151 (minimum shank) mm Inside diameter of flange, mm 3260 Outside diameter of flange, mm 3990 Inside diameter at shell, mm 3374 Inlet nozzle inside diameter, mm 700 Outlet nozzle inside diameter, mm 700 Cladding thickness, minimum, mm 4 Lower head thickness, minimum, mm 115 Vessel belt-line thickness, minimum, mm 170 Closure head thickness, minimum, mm 155 12/5/2011 15

Pressurizer Design Data Parameter Value Design pressure, MPa 17.2 Design temperature, o C 370 Surge line nozzle diameter, mm 273 Heatup rate of pressurizer using heaters only, o C/h 30 Internal volume, m 3 35 12/5/2011 16

Steam Generator Design Data Parameter Value Design pressure, primary side, MPa 17.16 Design pressure, secondary side, MPa 7.55 Design temperature, primary coolant 350 side, Design temperature, secondary side, 320 Design steam flow rate, t/h 1010 Heat transfer surface area, m 2 3088.67 Maximum moisture carryover, wt 0.25 percent, % 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 12/5/2011 17

NPP Structures All the systems and equipment of a nuclear power plant are housed in about 30 buildings and structures of different sizes. The major plant structures are grouped into the following: Nuclear Island Reactor Building Nuclear Auxiliary Building Fuel Storage Building Electrical Building Diesel Generator Building Conventional Island 12/5/2011 Turbine Generator Building De-mineralized Water Building Switch Yard 18

NPP Structures (Cont) Balance of Plant Liquid Radwaste Solidification Building Solid Radwaste Storage Building Low-level Radwaste Storage House Hot Laundry Boiler House Maintenance Building Warehouse Control Access Ultimate Heat Sink Essential Services Water Pump Station Outdoor Engineering 12/5/2011 Intake Structure Circulating Cooling Water Pump Station Water Treatment Plant Sewage Treatment Plant Drainage Structure Parking Area Fire Pump Station Cafeteria Guard House Hazard Cargo Storage House Administration Building and Emergency Centre Environmental Radiation Monitoring Hut 19

Nuclear Power plant Systems There are about 200 systems in CHASNUPP. These are classified, in accordance 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 12/5/2011 20

Safety Feature

Engineered safety Feature CONTAINMENT SYSTEMS Containment system which provides the last barrier against the post-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 coolant pipe or secondary system pipe, the containment peak pressure remains below the design pressure, with adequate margins and Beyond Design Basis Accidents (BDBA), and it can be reduced to half of the design value in 24 hours by the safeguards system 12/5/2011 22

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. It provides the capability of cooling following the initiation of the following accident conditions: The pipe break of reactor coolant pressure boundary (including the double-ended rupture of the largest reactor coolant pipe) or inadvertent relief valve or safety valve opening in the reactor coolant system which would result in a discharge larger than that could be made up by the normal makeup system. Rupture of a control rod drive mechanism causing a rod cluster control assembly ejection accident. The pipe break of secondary system (including the break of the largest pipe in the secondary system) or inadvertent relief valve or safety valve opening in the secondary system. Rupture of a steam generator tube. 12/5/2011 23

Engineered safety Feature HABITABILITY SYSTEMS Habitability Systems are designed to ensure that Control Room operators can remain inside the spaces served by the Main Control habitability Ventilation System during all normal and abnormal station conditions. The Habitability Systems cover all the equipment, supplies, and procedures provided to ensure that Control Room operators are protected from postulated releases of radioactive materials, toxic gases, smoke, and steam. The environments in all spaces served by the Main Control Habitability Ventilation System (Control Room envelope) are controlled within specified limits. The Habitability Systems are designed to support a maximum of seven persons during normal and 30 days abnormal station operating conditions. 12/5/2011 24

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 functions subsequent 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, primarily elemental iodine, from the containment atmosphere for the purpose of minimizing the offsite radiological consequences following the design-basis loss-of-coolant accident and BDBA. At the same time, the spray water serves to nominally reduce containment temperature and pressure during the injection phase Fission Product Control Systems The containment is steel-lined post tensioned pre-stressed concrete cylinder with a shallow 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 of integrity of the reactor coolant pressure boundary. 25 12/5/2011

Enhanced Safety Feature

Enhance Safety Features in C-2 Design More than 160 design changes incorporated on the basis of feedback 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 Accidents Consideration. Development of Symptom based Emergency Operating Procedures (SEOPs) for C-1 & C-2 with the help of designer. 12/5/2011 27

Major Improvements of CHASNUPP-2 over CHASNUPP-1 DESIGN FEATURES The Reactor Cavity Flooding system Hydrogen concentration monitoring system Passive hydrogen recombination facilities The countermeasure for Heterogeneous Boron Dilution The loose parts monitoring system (LPMS) Installation of motor throttled valve on pressurizer TARGET 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. To provide information about containment hydrogen volumetric concentration continuously. To operate during the event of design basis accident and severe accident. To prevent spurious automatic or manual injection of non-borated water by Anti-Dilution Protection (ADP) signal. To detect loose parts in the primary system as early as possible. The sensor location of LPMS is RPV, SG and RCP. To fulfill the function of overpressure protection and rapid depressurization. 12/5/2011 28

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 to prevent 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

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 piping to prevent IS-LOCA Mitigation Reactor Cavity flooding, Cooling Water Injection system to increase possibility of in-vessel corium retention or mitigate exvessel molten corium concrete interaction in case of reactor vessel failure 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

Hydrogen Passive Auto-catalytic Recombiners Hydrogen generation can be described in three phases. In-vessel core degradation phase, where large surfaces of metallic zirconium are available and temperature excursion is amplified by the exothermic reaction. Hydrogen may also be generated during the in-vessel relocation phase according to the failure mode of the lower head, consisting of solidified core material and the lower support plate, and the availability of water in the lower head. Hydrogen will be produced by melt core-concrete interaction (MCCI) in case of RPV failure. Hydrogen passive auto-catalytic recombiners (PAR) have been implemented inside the containment both for DBA and severe accidents instead of the active hydrogen recombiner (used in C-1). The PARs will be arranged at different locations inside the containment and will be able to recombine hydrogen when the concentration of hydrogen is above 2~4 % to avoid hydrogen detonation. 12/5/2011 31

Cavity Flooding Increase possibility of in-vessel corium retention The accident management strategy to flood the reactor cavity with refueling water storage tank (RWST) water and submerge the reactor vessel is credited with preventing vessel failure Protection of the integrity of the reactor vessel containing the molten corium by cooling its external surface Lower part of the reactor vessel should be submerged with cooling water 12/5/2011 32

Loose Part Monitoring System Loose Part Monitoring System (Including Trouble Analysis System) LPM016 The principal functional requirements of the loose parts monitoring system (LPMS) are to detect loose parts in the primary system as early as possible. Early detection of the loose parts can avoid or mitigate safety-related damage or malfunctions of components in the primary system so as to minimize economic losses. LPM007 LPM008 LPM009 LPM010 LPM001 LPM002 LPM003 LPM004 LPM005 LPM006 Sensor Locations LPM015 LPM01 LPM012 LPM013 LPM014 12/5/2011 33

% Operational Performance (C-1) (upto 30 September, 2009) 100 75 50 71 66 67.6 60.2 55 51.6 78.2 68.2 68.9 66.4 85.1 82 97.8 96.4 74.2 69.9 76.1 54.7 67.1 51.5 25 0 17 11 3 6 6 5 7 3 3 3 2000 2001 2002 2003 2004 2005 2006 2007 2008 2009 Availability F actor C apacity F actor Unplanned O utag es

Operational Performance (C-1) 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)

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) CBC HZP (ppm) T1 Worth (pcm) Overlap banks worth (pcm) 485 490 325 332 375 381 403 405 400 401 375 377 377 380 1297 1335 1321 1298 1477 1447 1567 1508 1447 1467 1457 1453 1474 1461 3068 2987 1321 1229 1907 1808 1405 1366 1710 1776 1771 1801 1590 1628 5566 5530 2510 2235 2368 2110 2928 2765 1952 1803 1708 1654 1574 1571

Periodic Safety Review (Safe and reliable operation - Ten (10) years of PWRs)

PSR Purpose The PSR is a tool to carry out a systematic and comprehensive review 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 in technology, etc. Compare current level of safety with latest standards and state of know-how and identify improvements at justifiable cost Obtain a broad integrated view of current safety of nuclear installations. 12/5/2011 38

PSR Objectives Confirm that the plant is as safe as originally intended, conforms to current 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 for next 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. 12/5/2011 39

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) from current 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 be made Ranking of issues for which corrective actions to be implemented Approval of corrective action program schedule from PNRA Implementation of corrective actions 40 12/5/2011

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 findings 12/5/2011 10. Organization and administration 11. Procedures 12. The human factor 13. Emergency planning 14. Radiological impact on the environment 41

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 coolant pump shaft seizure due to change of control rod drop time by considering the effects of earthquake and uncertainties and scram reactivity worth. 12/5/2011 42

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 12/5/2011 43

Human Factors PSR Corrective Action Plan 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 to all plant personnel on annual basis. Equipment Qualification Environmental monitoring program for qualified equipment may be developed. Procedure to control list of qualified equipment may be developed. Procedure(s) for analysis of the effects of equipment failures on equipment qualification and appropriate corrective actions and/or safety improvements to maintain equipment qualification may be developed. 12/5/2011 44

Ageing PSR Corrective Action Plan Ageing Management Program (AMP) may be developed. Potential ageing degradation that may affect the safety functions of SSCs may be documented. AMP training may be imparted to relevant personnel. Requirement for monitoring of physical condition of AMP SSCs, actual safety margins, and any features that would limit service 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 during RFOs may be developed. Quarterly administrative targets for discharge limits may be 12/5/2011 established. 45

Capabilities Safe and reliable operation - Ten (10) years of PWRs and forty (40) years of CANDU Design, analysis and engineering of systems and components of Nuclear Island, 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 and beyond 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, periodic safety review etc. Structural analysis/design of reactor core and internals and Plant ageing management of NPPs. 46

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