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TableofC

ontents1. AN INTRODUCTION TO ACR-1000 EVOLUTION ...................................................1

1.1 The ACR-1000............................................................................................................................................11.2 Design Features .........................................................................................................................................11.3 Passive Safety Features .............................................................................................................................2

2. PLANT DESIGN .............................................................................................................................................32.1 Layout: Inherently Safer and Faster to Build .......................................................................................32.2 Plant Siting ................................................................................................................................................6

2.2.1 Unit Output .........................................................................................................................62.2.2 Adaptation to Site Requirements ...................................................................................6

2.3 Nuclear Systems ........................................................................................................................................72.4 Heat Transport System and Auxiliary Systems ...................................................................................8

2.4.1 Heat Transport Pumps ....................................................................................................112.4.2 Steam Generators............................................................................................................12

2.5 Moderator System ..................................................................................................................................132.6 Reactor Assembly....................................................................................................................................14

2.6.1 Reactor Core Characteristics.......................................................................................142.6.2 Reactor Control ...............................................................................................................152.6.3 Fuel Channel Assembly ...................................................................................................15

2.7 Fuel Handling Systems............................................................................................................................162.8 Fuel .............................................................................................................................................................182.9 Safety Systems..........................................................................................................................................19

2.9.1 Shutdown Systems ...........................................................................................................192.9.2 Emergency Core Cooling (ECC) System ...................................................................202.9.3 Containment System.......................................................................................................202.9.4 Emergency Feedwater (EFW) System.........................................................................20

2.10 Essential Service Water Systems .........................................................................................................202.11 Balance of Plant (BOP) ..........................................................................................................................20

2.11.1 Turbine Generator and Auxiliaries...............................................................................202.11.2 Steam and Feedwater Systems .....................................................................................222.11.3 BOP Services.....................................................................................................................22

2.12 Instrumentation and Control ...............................................................................................................232.13 Electrical Power System ........................................................................................................................25

3. NUCLEAR SAFETY AND LICENSING .................................................................................273.1 Safety Design ............................................................................................................................................273.2 Defence-in-Depth....................................................................................................................................273.3 Inherent Safety Features ........................................................................................................................283.4 Severe Accidents......................................................................................................................................293.5 Licensing Basis..........................................................................................................................................29

4. ACR-1000 DEPLOYMENT .................................................................................................................30

5. OPERATION AND MAINTENANCE......................................................................................335.1 Consistently Better Performance........................................................................................................335.2 Enhanced Performance Features .........................................................................................................335.3 Enhanced Maintenance Features..........................................................................................................34

6. RADIOACTIVE WASTE MANAGEMENT..........................................................................38

7. DECOMMISSIONING .............................................................................................................................39

8. CONCLUSION .............................................................................................................................................40

i

ILLUSTRATIONS

Figure 1-1 Overall ACR-1000 Plant Flow Diagram ..........................................................................................1Figure 1-2 Reserve Water System........................................................................................................................3Figure 2-1 Two-Unit Plant Layout of Major Structures...................................................................................4Figure 2-2 Reactor Building....................................................................................................................................5Figure 2-3 Nuclear Systems Schematic ...............................................................................................................7Figure 2-4 Heat Transport System Flow Diagram ............................................................................................8Figure 2-5 3D View of Heat Transport System Layout ....................................................................................8Figure 2-6 3D View of Heat Transport System in Reactor Building .............................................................9Figure 2-7 Pressure and Inventory Control Flow Diagram .........................................................................10Figure 2-8 Heat Transport System Pump..........................................................................................................11Figure 2-9 Steam Generator................................................................................................................................12Figure 2-10 Moderator System Flow Diagram..................................................................................................13Figure 2-11 Reactor Assembly...............................................................................................................................14Figure 2-12 Comparison of Core Sizes...............................................................................................................15Figure 2-13 Fuel Channel........................................................................................................................................15Figure 2-14 Fuel Channel Grooves ......................................................................................................................16Figure 2-15 New Fuel and Spent Fuel Transfer Mechanisms..........................................................................16Figure 2-16 Fuelling Machine and Carriage ........................................................................................................17Figure 2-17 Spent Fuel Transfer and Storage Layout........................................................................................17Figure 2-18 Spent Fuel Transfer and Storage Pictorial.....................................................................................18Figure 2-19 CANFLEX®-ACR Fuel Bundle.........................................................................................................18Figure 2-20 SDS1 Shutoff Rods.............................................................................................................................19Figure 2-21 Turbine Generator and Auxiliaries Flow Diagram .....................................................................21Figure 2-22 Qinshan Low-Pressure Turbine Rotor ..........................................................................................22Figure 2-23 Overview of NSP Distributed Control System..........................................................................23Figure 2-24 Plant Control and Monitoring Systems.........................................................................................24Figure 2-25 SMART CANDU™ ..........................................................................................................................25Figure 2-26 Unitized Electrical Power System...................................................................................................26Figure 3-1 Barriers for Prevention of Releases...............................................................................................28Figure 3-2 Core Damage Frequencies per Year..............................................................................................29Figure 4-1 42-Month Deployment Schedule....................................................................................................30Figure 4-2 Design Engineering Applications .....................................................................................................31Figure 4-3 Module Lift Using VHL Crane .........................................................................................................31Figure 4-4 Typical Reactor Building Modules ..................................................................................................32Figure 5-1 Comparison of Gross Capacity Factors .......................................................................................33Figure 5-2 ChemAND – Performance Monitor for Plant Chemistry .......................................................34Figure 5-3 Maintenance Basis ..............................................................................................................................35Figure 5-4 Typical System Equipment Module.................................................................................................35Figure 5-5 Service Elevator..................................................................................................................................36Figure 5-6 Accessible Areas in the Reactor Building – Level 100 m .........................................................37Figure 5-7 Accessible Areas in the Reactor Building – Level 125.4 m ......................................................37Figure 6-1 Spent Fuel Storage Basket................................................................................................................38Figure 6-2 MACSTOR Fuel Transfer ..................................................................................................................38Figure 6-2 AECL’s MACSTOR System ..............................................................................................................38

ListofIllustrations

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Figure 1-1 Overall ACR-1000 Plant Flow Diagram

Introduction1.1 The ACR-1000The ACR-1000 is built to meet industry andpublic expectations for safe, reliable,environmentally friendly, low-cost nuclearpower generation. It has been developed byAECL from experience and feedback gainedin the design, construction and operation ofCANDU plants operated by ten utilitiesaround the world.

With a 60-year design life, the ACR-1000 is alight-water-cooled, heavy-water-moderatedpressure-tube reactor derived from the well-established CANDU line. It retains basicCANDU design features while incorporatinga specific set of innovative features and state-of-the-art technologies to ensure its safety,operation, performance and economics aresecond to none.

Enhanced safety features include a coredesign with a small negative coolant voidreactivity, larger thermal margins due to theuse of CANFLEX®**** fuel, and designimprovements based on insights gained fromProbabilistic Safety Analysis (PSA) performedduring the design process.

The latest design tools (CADDS) linkingmaterial management, documentation, safetyanalysis and project execution databases areused to ensure that accurate and completeconfiguration management can be readilymaintained by the plant Owner.

1.2 Design FeaturesThe ACR-1000 benefits from the provenprinciples and characteristics of CANDUdesign and the extensive knowledge base ofCANDU technology gained over manydecades of operation.

Proven CANDU strengths• Modular, horizontal fuel channel core• Separate low-temperature and pressure

moderator• Reactor vault filled with light water

surrounding the core• On-power refuelling• Two independent passively driven, safety

shutdown systems• Reactor building access for on-power

maintenance

1. An Introduction to ACR-1000 Evolution

1

ACR innovations• A more compact core design, which reduces heavy

water inventory and results in lower costs andreduced emissions

• Use of light water as reactor coolant, resulting inreduction of systems for heavy water coolant cleanupand recovery and simplification of containmentatmosphere cleanup systems

• Improved fuel burn-up through the use of lowenriched uranium (LEU) fuel, contained in advancedCANFLEX®-ACR fuel bundles

• Efficient means for burning other fuel types such asmixed oxides (MOX) and thorium fuels

• Increased fuel safety margins• Improved plant thermal efficiency through use of

higher pressures and higher temperatures in thecoolant and steam supply systems

• Enhanced accident resistance and core damageprevention features

• Improved performance through use of SMARTCANDU™ advanced operational and maintenanceinformation systems and provision of designed-inmaintenance features such as lifting devices, platformsand laydown areas

• Approximately 60% reduction in spent fuel quantitiescompared to current operating CANDU plants

Significant design simplifications• Steel-lined containment designed for all design basis

events• Sharing of long-term emergency cooling and shutdown

cooling safety functions• Use of light water coolant enabling a simplified

Emergency Coolant Injection (ECI) system, whichreplaces large motor-operated,safety-qualified injectionvalves with passive check valves

• Reduced inspections through selection of feedermaterials for increasing resistance to flow-assistedcorrosion (FAC) and robust fuel channel designmargins

• Reduced on-line and start-up time with computerizedtesting of major safety systems and automaticcalibration of in-core detector control signals

• Fuelling machine simplification with electric driveseliminating complex pneumatic systems. Thisaccelerates the on-line fuelling process, reducesmaintenance and speeds pressure tube in-serviceinspection

• Faster movement of personnel, without risk ofairborne contamination spread, through use ofventilation systems that allow main airlock doors to beopen during an outage

• Maintenance-based design providing required spaceallocation, reduction in temporary scaffolds and hoists,and provision for built-in electrical, water and airsupplies for on-power and normal shutdownmaintenance

• Reduction in number of sensors due to permittedsharing between systems

These technical improvements, along withadvancements in project engineering, manufacturing,and construction, result in significantly reduced capitalcost and construction schedule, while enhancing theinherent safety of the ACR-1000 design.

1.3 Passive Safety Features

The ACR-1000 design includes a number of “passive”safety features, some of which are designimprovements over the already robust systems inexisting CANDU plants. Examples of optimizedfeatures include:

• Two independent passively driven shutdown systems,each of which is capable of safely shutting down thereactor

• Increased safety margins with negative reactivitycoefficients

• Passive emergency coolant injection operation• Cool, low-pressure moderator serving as a passive

heat sink for decay heat from fuel channels in severeaccident situations

• Large concrete reactor vault, surrounding the core inthe calandria vessel and containing a large volume oflight water to further slow down or arrest severecore damage progression by providing a second,passive, core heat sink

• Elevated reserve water tank (RWT) in upper level of thecontainment building to deliver passive make-up coolingwater by gravity to heat transport system, steamgenerators, moderator and the calandria vault. Thisdelays progression of severe accidents and provideseven more time for mitigating actions by the operator

• Passive, robust, seismically-qualified containmentconsisting of:- Thickened pre-stressed concrete structure designed

to withstand aircraft crashes- Leak-tight inner steel liner to reduce potential

leakages- Passive spray system from elevated reserve water

tank to reduce reactor building pressures in theevent of a severe accident

- Passive Hydrogen Recombiner2

PlantD

esign

2. Plant Design

2.1 Layout: Inherently Saferand Faster to Build

Designed for efficient operation, increasedsafety and easier and faster maintenance, theplant is laid out to provide separation bydistance, elevations and the use of barriersfor safety-related structures, systems andcomponents. Each corner of the reactorauxiliary building houses redundant safetyequipment in a four-quadrant configuration.

Security and physical protection have beenaddressed to ensure that the response topotential common and abnormal events,such as fires, aircraft crashes and malevolentacts meets latest criteria.

The plant layout is also designed to achieve theshortest practical construction schedule whilesupporting easier maintenance practices.Buildings are arranged to minimizeinterferences during construction, withallowance for on-site fabrication of moduleassemblies. Through the use of open-topconstruction, provisions exist for flexibleequipment installation sequences.

The footprint of the two-unit plant has beenminimized with the adoption of common areasfor the main control room, service andmaintenance buildings. A single-unit plant can beadapted from the two-unit layout with nosignificant changes to the reference design. Theplant is designed for an exclusion zone of 500 metre radius. The size of the power blockfor a 2-unit ACR-1000 station is 48,700 m2*(actual area).

Figure 1-2 Reserve Water System

* Power block consists of 2 reactor buildings, 2 reactor auxiliary buildings, 2 turbine buildings, 1 service building, 1 maincontrol building, 1 maintenance building, 1 crane hall, 2 secondary control buildings and four diesel generator buildings.

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Major buildings and structures oftwo-unit plant

• Reactor Buildings (2)• Reactor Auxiliary Buildings (2)• Turbine Buildings (2)• Main Control Building• Secondary Control Buildings (2)• Maintenance and Service Buildings• Condenser Cooling Water Pumphouse• Essential Service Water Pumphouse• Main Switchyard

Reactor BuildingStrengthened over previous CANDU designs, the pre-stressed concrete reactor building is seismically-qualified and tornado-proofed. The concrete outer wallhas an inner steel liner that will achieve significantlyreduced leak rates in the event of an accident.

An isolation system ensures “button-up” in case ofaccidents.

The entire structure, including concrete internalstructures, is supported on a reinforced concrete baseslab to ensure a fully enclosed boundary forenvironmental protection and biological shielding.

During reactor operation, internal shielding permitspersonnel access to an environment that istemperature-controlled for personal comfort.Airlocks are designed as routine entry/exit doors.

Containment structure perimeter walls are separatefrom internal structures, so as to eliminate anyinterdependence and to provide flexibility inconstruction.

The reactor building is the principal component of thecontainment system.

Figure 2-1 Two-Unit Plant Layout of Major Structures

CANDU 6 ACR-1000Containment StructureType Pre-stressed Pre-stressed

concrete / epoxy liner concrete / steel linerRB inside diameter 41.4 m 56.5 mRB containment wall thickness 1.07 m 1.8 mBuilding height (base slab to top of dome) 51.2 m 74.0 m

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Reactor Auxiliary BuildingThe reactor auxiliary building is a multi-level,reinforced concrete and steel structure that isseismically-qualified and tornado protected. Itsurrounds the reactor building and accommodates theumbilicals that run between the principal structures,the electrical systems, and the spent fuel bay andassociated fuel-handling facilities. It also houses thelong-term cooling (LTC) pumps and heat exchangers,the spent fuel bay cooling and purification systempumps and heat exchangers, the essential coolingwater pumps, heat exchangers and valve stations, andthe essential service water valve stations. Safety andisolation valves for the main steam lines are housed in a seismically-qualified concrete structure on top ofthe building.

Turbine BuildingThe turbine building is located to one side of thereactor auxiliary building, so that turbine shaftalignment is perpendicular to the reactor building.This is also an optimum location for access to themain control room, the piping and cable tray runs toand from the reactor auxiliary building, and thecondenser cooling water ducts to and from the mainpumphouse. Access routes are provided between theturbine building and the reactor auxiliary building.

The turbine building houses the turbine generatorand its auxiliary systems: condenser, condensate andfeedwater systems, the building heating plant, and anycompressed gas required for the balance of plant(BOP). Blow-out panels in the walls and roof serveto relieve internal pressure in the event of a steam-line break.

Figure 2-2 Reactor Building

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Main Control BuildingSeismically-qualified and tornado-protected, the maincontrol building is a multi-level structure locatedbetween the two units. It has a superstructure of steeland reinforced concrete and reinforced concretesubstructure. It contains the main control room(MCR) and associated control and electricalequipment for the two units. Each side of the MCRhas dedicated panels, computers, displays andoperator consoles with separation of cabling andequipment for each unit.

Secondary Control BuildingEach unit has a completely separate secondary controlbuilding (SCB) with sufficient control and monitoringequipment to shut down the unit, initiate requiredcooling and ensure a safe, maintained shutdown stateshould the MCR become uninhabitable or non-functional. The SCB is located so that the MCR andSCB cannot be simultaneously rendered inoperabledue to any design basis event. SCB human-systeminterface components are similar to those in the MCRso as to minimize human error should the operatorrelocate from one area to the other.

Maintenance and Service BuildingsThe seismically-qualified maintenance and servicebuildings are located between the two-unit ACR-1000plant. They house all conventional plant services,including radioactive waste handling facilities, heavywater management systems, common services, centralstores, central active/non-active maintenance shops,and change rooms for staff. They are multi-levelstructures with a reinforced concrete substructureand braced steel-frame superstructure.

Condenser Cooling Water PumphouseThe condenser cooling water (CCW) pumphouse hasa reinforced concrete substructure and braced steel-frame superstructure. It contains the CCW pumps,plant water system pumps, screen wash pumps, trashracks, screens, and chlorination equipment, if required.Together with related intake and outfall structures, thepumphouse serves the two-unit ACR-1000 plant,housing separate CCW and plant water systems withadequate separation for each unit. Sites with limitedcooling water availability can use cooling towersinstead of the conventional CCW system.

Essential Service Water (ESW) PumphouseThe essential service water (ESW) pumphousecontains the ESW pumps. It has a reinforced concretesubstructure, braced steel-frame superstructure and isseismically and tornado-qualified.

Main SwitchyardThe switchyard is designed to allow flexible operationfor power output, switching and maintenance. Abreaker-and-a-half design with single voltage isproposed for high reliability. Each ACR-1000 unit willhave at least four bays of power inputs/outputs fromthe main transformers and grid system, with optionsto add more as future plant and grid requirementsmay dictate.

2.2 Plant Siting

2.2.1 Unit Output

Each unit of the ACR-1000 two-unit integrated plantdesign has a nominal gross electrical output of 1165MWe. Output can be optimized by adjusting theturbine/condenser design to suit any site coolingwater conditions.

2.2.2 Adaptation to Site Requirements

The ACR-1000 can accommodate a wide range ofgeotechnical and meteorological data and conditions.Some of these flexible design features include:

• Cooling water systems for all nuclear steamrequirements, saltwater or freshwater.Conventional cooling towers can also be used

• Cooling water temperatures from typical cold to typicalwarm sites. A generic set of reference conditions hasbeen developed for potential ACR-1000 sites

• Tornado protection as required. The design basistornado (DBT) is defined by a maximum wind speedof 483 km/h. DBT for the ACR-1000 is selected tosatisfy tornado design requirements for NorthAmerican sites and potential sites overseas

6

• Plant exclusion zone capability of only 500 m radius.All unauthorized persons are restricted from thiszone. Larger zones may be selected where desired

• Design basis earthquake (DBE) protection of up to0.3 g acceleration. This is the maximum groundmotion of potentially severe quakes, with lowprobability of being exceeded during the life of theplant

2.3 Nuclear Systems

ACR-1000 nuclear systems are located in the reactorbuilding and the reactor auxiliary building. Thesebuildings are robust and shielded where necessary toensure all radioactive substances are always secure.Systems include:

• Heat transport system with light water coolant in atwo-loop, figure-eight configuration with four steamgenerators, four heat transport pumps, four reactoroutlet headers and four reactor inlet headers. Thisconfiguration is standard on all CANDU 6 reactorsand the larger 935 MWe Darlington NuclearGenerating Station (NGS) CANDU design

• Heavy water moderator system• Reactor assembly consisting of calandria vessel

installed in concrete vault• Fuel handling system consisting of two fuelling machine

heads, each mounted on a fuelling machine bridge andsupported by columns, located at each end of thereactor. The fuelling machines have been simplified toenhance maintainability and accelerate pressure tubein-service inspection

• Two independent shutdown systems, emergency corecooling (ECC) system, containment system andassociated safety support systems

Figure 2-3 Nuclear Systems Schematic

7

The ACR-1000 heat transport system (HTS) circulatespressurized light water coolant through the reactorfuel channels to remove heat produced by nuclearfission in the core. The use of light water coolant is adesign simplification allowing for the reduction ofsystems for cleanup and recovery. It also simplifiescontainment atmosphere cleanup systems.

The ACR-1000 HTS consists of 520 reactor fuelchannels with associated corrosion-resistant stainlesssteel feeders, four inlet headers, four outlet headersand interconnecting piping. The system includes foursteam generators and four electrically-driven heattransport pumps in a two-loop, figure-eightconfiguration. Headers, steam generators and pumpsare all located above the reactor.

Figure 2-4 Heat Transport System Flow Diagram

Figure 2-5 3D View of HeatTransport System Layout

2.4 Heat Transport System and Auxiliary Systems

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Table 2-1 Heat Transport System Design Data

CANDU 6 Darlington ACR-1000Reactor outlet header pressure [MPa (g)] 9.9 9.9 11.1Reactor outlet header temperature [ºC] 310 310 319Reactor inlet header pressure [MPa (g)] 11.2 11.3 12.5Reactor inlet header temperature [ºC] 260 267 275Single channel flow (maximum) [kg/s] 28 27.4 28

Figure 2-6 3D View of Heat Transport System in Reactor Building

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Pressure and Inventory Control System

The ACR-1000 heat transport pressure and inventorycontrol system consists of pressurizer, pumps, feedand bleed valves and a coolant storage tank. Thissystem provides:

• Pressure and inventory control for each heattransport system loop

• Overpressure protection• Controlled degassing flow

Light water in the pressurizer is heated electrically topressurize the vapour space above the liquid. Thevolume of the vapour space is designed to cushionpressure transients, without allowing excessively highor low pressures in the heat transport system.

The pressurizer also accommodates change in reactorcoolant volume from zero power to full power. Thispermits reactor power to be increased or decreasedrapidly, without imposing severe demand on thecoolant feed and bleed components of the system.

When the reactor is at power, pressure is controlledby the pressurizer; heat is added with the electricheaters to increase pressure, and removed by sprayingcold water via the reactor inlet headers to reducepressure. The coolant inventory is adjusted by thefeed-and-bleed circuit. Pressure can also becontrolled by the feed-and-bleed circuit with thepressurizer isolated at low reactor power and whenthe reactor is shut down. The feed-and-bleed circuitis designed to accommodate the changes in coolantvolume that take place during heat-up and cool-down.

Figure 2-7 Pressure and Inventory Control Flow Diagram

PRESSURIZERRELIEF VALVES

10

2.4.1 Heat Transport Pumps

The ACR-1000 heat transport pumps are anenhanced, larger version of the double-dischargedesign used in the CANDU 6 and Darlingtonreactors.

The ACR-1000 retains the CANDU mechanicalmulti-seal design, which allows for easy replacement.Backup seal cooling extends pump survivability, evenduring accident conditions, if service water is lost.

Figure 2-8 Heat Transport System Pump

Table 2-2 Heat Transport Pump Data

CANDU 6 Darlington ACR-1000Number 4 4 4Rated flow [L/s] 2228 3240 4300Motor rating [MWe] 6.7 9.6 10.0

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Figure 2-9 Steam Generator

2.4.2 Steam Generators

The ACR-1000 steam generators are similar to theCANDU 6 and Darlington designs, except for thelarger physical size. For the ACR-1000, steamgenerator tubing diameter is increased to takeadvantage of the change to light water coolant.

ACR-1000 tubing is made of Incoloy-800, a materialwith proven operating performance and service atCANDU 6 and Darlington stations. Steam wetness atthe steam nozzle has been reduced to 0.1%, based onlatest steam separator technology, leading toimproved turbine cycle economics.

Table 2-3 Steam Generator Design Data

Steam Generators CANDU 6 Darlington ACR-1000Number 4 4 4Type Vertical U-tube / Vertical U-tube / Vertical U-tube /

integral pre-heater integral pre-heater integral pre-heaterNominal tube diameter [mm] 15.9 (5/8”) 15.9 (5/8”) 17.5 (11/16”)Steam temperature (nominal) [ºC] 260 265 275.5Steam quality 0.9975 0.9975 0.999Steam pressure [MPa (g)] 4.6 5.0 5.9

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Figure 2-10 Moderator System Flow Diagram

2.5 Moderator System

The ACR-1000 moderator is a low-pressure, low-temperature system that isfully independent of the heat transportsystem. It consists of pumps and heatexchangers that circulate heavy watermoderator (D2O) through the calandriavessel and remove heat generated withinthe moderator during reactor operation.Heavy water acts as both moderator andreflector for the neutron flux in the core.

Inlet and outlet nozzles are located at thetop of the calandria vessel to preventinadvertent draining and are oriented toensure uniform moderator temperaturedistribution inside the calandria.

The ACR-1000 moderator system alsofulfills a safety function that is unique toACR/CANDU. It serves as a backup heatsink in the event of loss of fuel cooling viathe heat transport system, therebymitigating core damage consequences. Heatexchangers are provided with seismically-qualified cooling water and standby power.

Another safety improvement in the ACR-1000 is the connection to the reservewater tank. It provides additional passivegravity-fed inventory to the calandria vessel,extends core cooling and delays severeaccident event progression.

Table 2-4 Heavy Water Inventory Design Data

CANDU 6 Darlington ACR-1000Moderator System [Mg D2O] 265 312 250Heat Transport System [Mg D2O] 192 280 0Total [Mg D2O] 457 592 250

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2.6 Reactor Assembly

The ACR-1000 reactor assembly consists of thehorizontal, cylindrical, low-pressure calandria and end-shield assembly. This enclosed assembly contains theheavy water moderator and the 520 fuel channelassemblies. The reactor is supported within a concrete,light-water-filled calandria vault. Fuel is enclosed in thefuel channels that pass through the end shields. Each fuelchannel permits access for on-line fuelling operationwhile the reactor is at power.

The ability to replace fuel as required for maintainingpower means minimal “excess” reactivity in the core atall times, an inherent safety feature. This feature alsocontributes to operational flexibility for improved outageplanning, since fixed cycle times are not required andprompt removal of defect bundles can be accomplishedwithout shutdown.

2.6.1 Reactor Core Characteristics

The ACR-1000 reactor core offers the followingdistinctive advantages:

• Compact size due to smaller fuel channel latticepitch than CANDU, resulting in reduced heavy waterrequirements

• Use of light water as coolant

• Negative coolant void reactivity• Simplified reactor control through negative feedback

in reactor power• Flattened axial and radial profiles to optimize

channel thermal power output

The physical size of the ACR-1000 core, whileproducing greater power output, is similar to that ofthe CANDU 6.

Figure 2-11 Reactor Assembly14

2.6.2 Reactor Control

The neutronic coupling in the compact ACR-1000core and negative power coefficient ensure corestability. All harmonic modes, including the first axialmode, are stable at all power levels under nominaloperating conditions. Stable reactor physicscharacteristics allow simpler control mechanismdesign.

Mechanical zonal control units provide primarycontrol in the ACR-1000. Each zone control assemblyconsists of two independently movable segments. On-power refuelling and zone-control actions provideday-to-day reactivity control. The reactor regulatingsystem also includes control absorber units, physicallysimilar to the mechanical shutoff rods that can be usedto reduce power if larger reductions are required.

2.6.3 Fuel Channel Assembly

The ACR-1000 fuel channelassembly consists of a zirconium-niobium (Zr-2.5%Nb) pressuretube, centred in a zircaloycalandria tube. The pressure tubeis roll-expanded into stainless steelend fittings at each end.

Table 2-5 Reactor Core Design Data

CANDU 6 Darlington ACR-1000ReactorOutput [MWth] 2064 2657 3187Coolant Pressurized D2O Pressurized D2O Pressurized Light WaterModerator D2O D2O D2OCalandria diameter [m] 7.6 8.5 7.5Fuel channel Horizontal Zr 2.5wt%Nb Horizontal Zr 2.5wt%Nb Horizontal Zr 2.5wt%Nb

alloy pressure tubes with alloy pressure tubes with alloy pressure tubes withmodified 403 SS end-fittings modified 403 SS end-fittings modified 403 SS end-fittings

Fuel channels 380 480 520Lattice pitch (mm) 286 286 240 Pressure tube wall thickness (mm) 4 4 6.5

Figure 2-12 Comparison of Core Sizes

Figure 2-13 Fuel Channel 15

Each pressure tube isthermally insulated from thelow-temperature moderatorby the annulus gas betweenthe pressure tube and thecalandria tube. Fixedspacers, positioned along thelength of the pressure tube,maintain annular space andprevent contact betweenthe two tubes. Each end-fitting holds a liner tube, afuel support plug and achannel closure. Reactorcoolant flows throughadjacent fuel channels inopposite directions. TheACR-1000 calandria tubehas been thickenedcompared to the CANDUdesign to ensure it canwithstand a pressure tuberupture.

The ACR-1000 is designed for 60 years of reactoroperation with provision for mid-life refurbishment,including replacement of fuel channels. Special designfeatures, such as additional rolled joint grooves, areprovided in the end-fittings to facilitate pressure tubereplacement.

2.7 Fuel Handling Systems

The ACR-1000 fuel handling systems consist of:

• New fuel handling and storage system• Fuelling machines and their supports• Spent fuel handling and storage

Figure 2-15 New Fuel and Spent Fuel Transfer Mechanisms

Figure 2-14 Fuel Channel Grooves

16

The new ACR-1000 fuel handling and storage systemincludes the storage of the new low enriched uranium(LEU) fuel and the supply of the two fuelling machines tomaintain full-power operation. The need for operatoraccess to the reactor building is minimized with all newfuel storage, inspection and fuelling machine loadingbeing performed from an accessible area in the reactorauxiliary building.

Evolved from the CANDU 6 design, the simplified ACR-1000 fuel handling machines incorporate significantadvances. Key design improvements include replacingwater and oil hydraulic drives with electric drives, alarger capacity magazine and a mechanical ram withabsolute resolvers for position feedback. Further designsimplifications include change to light water operation,with heavy water eliminated from the fuel handlingsystems. These changes, along with built-in redundancy,will result in improved system performance, extendedin-service periods and reduced maintenancerequirements, including accelerated de-fuelling forpressure tube in-service inspection.

Two fuelling machines are located on opposite sides ofthe reactor and mounted on bridges supported bycolumns. The normal refuelling operation is an eight-bundle shift, in the direction of coolant flow, in whichspent bundles are removed from the outlet end of a fuelchannel,while fresh bundles are inserted at the inlet end.

The ACR-1000 transfer and storage system handlesspent fuel from the time it is discharged from thefuelling machine to the time it is moved to the storagebay in the reactor auxiliary building.

Once spent fuel is discharged, the transfer system usesrecirculating water, which also cools the fuel, to pushit through a pipe to receiving bays. The system thenunloads the fuel from its magazine and moves it inbaskets to the storage bay through a shielded tunnel.In the storage bay, spent fuel baskets are stacked in

frames with capacity for atleast 10 years operation. Astorage bay bridge andhandling tools permitmanipulation of spent fueland containers. Baskets arealso suitable for directtransfer to dry fuel storage,which can be provided atOwner request—for anadditional 50 or beyond.

The entire fuelling andspent fuel unloadingprocess is automated andcarried out from thestation control room.

Figure 2-16 Fuelling Machine and Carriage

Figure 2-17 Spent Fuel Transferand Storage Layout

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2.8 Fuel

The ACR-1000 uses the 43-element CANFLEX®-ACR fuel bundle design.

The centre element contains neutron absorbers, while the remainingelements contain U-235 enriched UO2 pellets. A burnable absorber is usedin some of the elements that contain enriched pellets to optimize thepower rating of the fuel. The neutron absorbers of the centre element areused for management of coolant void reactivity. A very thin layer ofCANLUB covers the inside surface of the fuel cladding to enhance fuelperformance.

The ACR inherent feature for operating with neutron absorbers makes it ideally suited to burn other fuel types such as mixed oxides (MOX) and thorium.

CANDU 6 Darlington ACR-1000

Fuel Natural UO2 Natural UO2 Low enriched UO2

Fuel burn-up [MWd/te U] 7,500 7,791 20,000

Fuel bundle assembly 37 element 37 element 43-element

CANFLEX®-ACR

Bundles per fuel channel 12 13 12

Figure 2-18 Spent Fuel Transfer and Storage Pictorial

18

Figure 2-19 CANFLEX®-ACR Fuel Bundle

2.9 Safety Systems

ACR-1000 safety systems are designed to mitigate theconsequences of plant process failures, ensuringreactor shutdown, removal of decay heat andprevention of radioactive releases.

Design follows the traditional CANDU practice ofproviding:

• Shutdown System 1, (SDS1)• Shutdown System 2, (SDS2)• Emergency Core Cooling (ECC) System• Containment System• Emergency Feedwater System

SDS1, SDS2, the ECC and containment systems meethigh reliability requirements that have beenestablished during system design and verified byreliability analysis.

Safety support systems are also provided to ensurereliable electrical power, cooling water and instrumentair supplies to the safety systems. Eight nuclear steamplant (NSP) standby generators are provided for thetwo units. Four NSP standby generators are “pre-assigned” to specific distribution buses in therespective unit. Two additional BOP standbygenerators provide backup to the NSP for postulatedstation blackout events.

Safety systems and their support services are designedto perform their safety functions with a high degree ofreliability. This is achieved through the use ofredundancy, diversity, separation, testability, theapplication of appropriate quality assurance standards,and the use of stringent technical specifications,including seismic and environmental qualification foraccident conditions.

2.9.1 Shutdown Systems

The ACR-1000 incorporates two passive, fast-acting,fully capable, diverse and separate shutdown systems,which are physically and functionally independent ofeach other.

SDS1 consists of mechanical shutoff rods that drop bygravity into the core when a trip signal de-energizesthe clutches that hold the shutoff rods out of the core.The design of the shutoff rods is based on the proven

CANDU 6 design. The in-core portion of the shutoffrods has been designed to accommodate the smallerACR-1000 core lattice pitch.

SDS2 injects a concentrated solution of gadoliniumnitrate into the low-pressure moderator to quicklyrender the core sub-critical. The gadolinium nitratesolution is dispersed uniformly with pressurized gas,maximizing shutdown effectiveness.

The reactor can be put into a guaranteed shutdownstate (GSS) using a rod-based system. Designsimplifications have been provided to achieve this.

Figure 2-20 SDS1 Shutoff Rods

19

2.9.2 Emergency Core Cooling (ECC) System

The ACR-1000 emergency core cooling (ECC) systemconsists of two subsystems:

• Passive emergency coolant injection (ECI) system:The ECI system has accumulator tanks that willsupply high-pressure water to the HTS and refill thefuel channels in the short term after a loss of coolantaccident (LOCA)

During normal operation, the ECI system is poised todetect any LOCA that results in a depletion of HTSinventory to such an extent that make-up by normalmeans is not assured. When the HTS pressure dropsbelow the pressure of the ECI accumulator tanks,water is injected into the heat transport system.

Valves on the ECI interconnect lines between thereactor outlet headers (ROH) open upon detectionof a LOCA to assist in establishing a sustainablecooling flow path.

In addition, core makeup tanks (CMTs) providepassive makeup to the intact HTS loop following aLOCA and prevent voiding for secondary sidedepressurization events.

• Long-term cooling (LTC) system:The LTC system provides long-term recirculation andrecovery. It is used for cooling of the reactor afterpostulated transients, including LOCA, and duringmaintenance.

LTC is initiated automatically when HTS is sufficientlydepressurized, at which time the LTC system beginsoperation in long-term recovery mode.

2.9.3 Containment System

The ACR-1000 containment system forms acontinuous, pressure-retaining envelope around thereactor core and the heat transport system. Thisprevents releases of radioactive material to the externalenvironment.

The containment boundary consists of a steel-lined,pre-stressed concrete reactor building, access airlocksand a containment isolation system. The containmentdesign ensures a low leakage rate. Hydrogen control isprovided in the reactor building by passive autocatalyticrecombiners and igniters to limit the hydrogen contentto below deflagration limit within the containment,following a core damage accident.

Finally, the provision of a spray system connected to theelevated reserve water tank (RWT) will reduce reactorbuilding pressures, if required, in the event of severeaccidents.

Heat removal from the containment atmosphere is alsonormally provided by the operation of local air coolers,which are suitably located in various compartments ofthe reactor building, to reduce pressure and furtherreduce leakage over a longer period following an event.

2.9.4 Emergency Feedwater (EFW) System

The emergency heat removal function is accomplishedby the EFW system. The system provides anindependent supply of feedwater to the steamgenerators to remove decay and sensible heat to cooldown the reactor following a total loss of the main andemergency feedwater systems.

The emergency feedwater system consists ofemergency feedwater pumps driven by normal Class IVpower and backed up by standby Class III electricalpower. These pumps provide emergency feedwater tothe steam generators at a rate sufficient to removedecay heat from the reactor core following a designbasis event. Emergency feedwater is supplied from thereserve feedwater tank. All the components and valvesof the system are seismically-qualified and are located inthe seismically-qualified reactor building and reactorauxiliary building.

2.10 Essential Service Water Systems

The ACR-1000 adopts a four-division concept foressential service water systems. All divisions arephysically separate, redundant and equipment in each isidentical. Systems are sized to ensure that, underaccident conditions, two divisions are capable ofhandling plant safety shutdown heat loads.

2.11 Balance of Plant (BOP)

The balance of plant (BOP) comprises the turbinebuilding, steam turbine, generator, condenser, and thefeedwater heating system with associated auxiliary andelectrical equipment. The BOP also includes the watertreatment facility, auxiliary steam facilities, condensercooling water pumphouse and/or cooling towers, andassociated equipment to provide all conventionalservices to the plant.20

Turbine Generator CANDU 6 Darlington ACR-1000Steam Turbine Type Hitachi impulse-type, Tandem-compound Impulse-type

tandem-compound tandem-compoundSteam Turbine One double-flow One double-flow One double-flow Composition high-pressure cylinder high-pressure cylinder high-pressure cylinder

Net to turbine (MWth) 2060 2650 3180Gross/Net electrical output* (nominal) [MWe] 728/666 935/881 1165/1085Turbine Generator Efficiency** 35.3% 35.3% ~36.6%Steam temperature at main stop valve [ºC] 258 263 273Final feedwater temperature [ºC] 187 177 217Condenser Vacuum [kPa (a)] 4.9 4.2 4.9

CANDU 6 data quoted is based on the Qinshan Phase III CANDU 6 design.* Approximate values: electrical output is dependent on site conditions.

** Motor-driven feedwater pump, CANDU 6 and ACR-1000 outputs are based on reference cooling water temperature of 18.8°C.Darlington output is based on reference cooling water temperature of 11°C.

Figure 2-21 Turbine Generator and Auxiliaries Flow Diagram

21

2.11.1 Turbine Generator and Auxiliaries

The turbine generator system and the condensate andfeedwater systems are based on conventional designs.They meet the design requirements specified by theNSP designer to assure the performance and integrityof the nuclear steam plant. These includerequirements for materials (i.e., titanium condensertubes, absence of copper alloys in the feed train),chemistry control, feed train reliability, feedwaterinventory and turbine bypass capability.

In the event of station blackouts, the reactors aredesigned to stay at power for the duration of theevent with the turbine generators disconnected fromthe grid. In this mode of operation, power is onlysupplied to internal auxiliaries as needed for the safeoperation of the plant.

The BOP is capable of daily and weekly powermanoeuvring to as low as 50%.

Figure 2-22 Qinshan Low-Pressure Turbine Rotor

CANDU plants have operated successfully usingNorth American, European and Japanese turbinegenerators with fresh water and seawater condensercooling water.

2.11.2 Steam and Feedwater Systems

The ACR-1000 main steam system supplies the steamfrom the steam generators in the reactor building tothe turbine through the steam balance header. Thefeedwater system takes hot, pressurized feedwaterfrom the feedwater train in the turbine building anddischarges it into the pre-heater section of the steamgenerators. The system maintains the required steamgenerator level by controlling feedwater flow.

The condenser steam discharge valves (CSDVs) aredesigned to discharge up to 100% of steam flowdirectly to the condenser. This feature provides foroperational flexibility in support of load followingoperation in conjunction with overall reactor control.It also provides a backup safety function for fuelcooling, via steam generator cooling, by making use ofthe large inventory in the condenser.

The safety functions of overpressure protection andcooling of the steam generator secondary side areprovided by main steam safety valves (MSSVs). Inaddition, main steam isolation valves (MSIVs) can beused to prevent releases from containment in the eventof steam generator tube leaks to the secondary side.

2.11.3 BOP Services

Conventional plant services include potable watersupply, heating, ventilation, air conditioning,chlorination (if required), fire protection, compressedgases and electric power systems.

Service Water SystemsThe balance of plant (BOP) water systems providecooling water, de-mineralized water and domesticwater to plant users. The systems consist of thecondenser cooling water (CCW), plant water system,water treatment facility and chlorination systems.

Heating, Ventilation and Cooling SystemsHeating, ventilation, air conditioning and chilled water(from the chilled water system) are supplied to plantbuildings to ensure a suitable environment forpersonnel and equipment during winter and summer.The building heating plant provides the steam and hotwater demands of the entire plant. Steam extractedfrom the turbine is used as the normal building heatingsource. Dedicated, separate ventilation systems areprovided for the main control building and secondarycontrol building.

Fire Protection SystemWater supply for the main fire protection systemcomes from a fresh water source via a buried pipecircuit. The main system provides fire protection forthe entire station (i.e., both NSP and BOP).

The fire protection system also includes standpipe andfire hose systems, portable fire extinguishers for firesuppression, and a fire detection and alarm systemcovering all plant buildings and areas.

22

Fire-resistant barriers are provided for mitigationpurposes, where necessary, to isolate and localize firehazards and to prevent spread of fire to otherequipment and areas. The four-quadrant layout in thereactor auxiliary building provides maximumseparation of redundant safety equipment for addedfire protection.

2.12 Instrumentation and Control

The ACR-1000 unit control and monitoring systemsapply modern distributed control, display and networkcommunication technologies. Safety system logic andcontrol are based on four-channel architecture toprovide fault tolerance protection and to minimizespurious reactor trips. This results in enhancedmonitoring capability and contributes to loweroperating and capital costs due to:

• Reduction in the number of instrumentation andcontrol components, leading to improved reliabilityand reduced maintenance and construction costs

• Design simplification through permitted sharing ofsystems, enabling the reduction in the number ofsensors

• Increased automation, thus reducing frequency ofoperator error

• Improved information and data communicationssystems that provide detailed information on unitoperational state, enabling early detection anddiagnosis of faults and improving timely preventiveequipment maintenance, thereby reducingunplanned plant outages

Most control functions are performed by a state-of-the art distributed control system (DCS) that usessmall, programmable digital controller modules inplace of a single central computer. The controllerscommunicate with one another by means of datahighways, which use reliable, high-security datatransmission methods. Manual control commands tobe executed by the DCS are entered by the operatorsvia the plant display system.

Control CentreThe ACR-1000 plant control centre enables operatingstaff to monitor, control and effectively operate theunits in both normal and abnormal modes.

A computerized plant display system (PDS) is used forall plant control and monitoring. Integrated computertechnology is used throughout the controls, displays,panels and consoles. These link operating procedures,testing requirements and configuration managementto achieve high plant performance and enhancedoperator effectiveness.

Figure 2-23 Overview of NSP Distributed

Control System

23

The control centre information system includes anadvanced alarm annunciation capability, based on theCANDU annunciation message list system (CAMLS)implemented on the Qinshan units. It conveys up-to-date unit information through fault and status displays.The control centre information system also includesan alarm interrogation application that allowsoperations staff to view fault and status display and tointerrogate alarm history from any of the controlcentre panels or console workstations. The controlcentre information system includes on-lineprocedures for operator support.

Each unit has a completely separate secondary controlbuilding (SCB) to control and monitor equipmentrequired to shut down the unit, initiate the required

fuel cooling, and monitor equipment and plant state toensure the unit remains in a safe shutdown should themain control room (MCR) become unavailable.

The ACR-1000 will also provide an integrated packageof software tools and work processes aimed at plantperformance optimization throughout its life cycle.SMART CANDU technologies use the AECLknowledge base and plant data to predict, prevent andenhance operations. The SMART CANDU suite of tools includes ChemAND and other superiorengineering tools.

Figure 2-24 Plant Control and Monitoring Systems

24

2.13 Electrical Power System

The electrical power system consists of connectionsto the off-site grid, main turbine generator, associatedmain output system, on-site standby diesel generators,battery power supplies, uninterruptible powersupplies (UPS) and the distribution equipment.Essential standby generators, batteries, UPS and theequipment distributing power from these sources areseismically and environmentally-qualified. Thisequipment is provided in a four-bus configuration,whichimproves reliability, allows for on-power maintenanceand minimizes potential for spurious trips.

The electrical distribution system (EDS) supplieselectrical power to all process and instrumentationand control loads within the unit. The EDS is dividedinto four classes of power based on availability: Class Iis delivered from batteries, Class II from UPS, Class IIIfrom standby generators and Class IV from the maingenerator or grid.

In a two-unit ACR-1000 plant, each unit has adedicated electrical distribution system with inter-unitties only in the Class III distribution system. Fourseismically-qualified, essential standby generators areprovided for each unit. Two additional standbygenerators are provided to support station operation,including ‘blackouts.’

High Capacity Factors&

Long Life

AECLKnowledge

Base

Plant Data

Prediction,Prevention,Enhanced

Operations

SMART CANDUTechnologies

Figure 2-25 SMART CANDU

CAMLSIntelligent Annunciation Message List System that assistsoperators in coping with events such as blackouts.ChemANDHealth monitor for plant chemistry. Predicts future performanceof components, determines maintenance requirements and optimal operating conditions.ThermANDHealth monitor for heat transfer systems and components.Ensures optimal margins and maximum power output.MIMCMaintenance Information Management Control system that linkshealth monitor data to the plant work management system.

25

Figure 2-26 Unitized Electrical Power System

26

3.1 Safety Design

Nuclear safety requires that the radioactiveproducts from the nuclear fission process becontained, both within the plant systems forworker protection and outside the plantstructure to protect the public. This isachieved at all times by:

• Controlling the reactor power, and ifnecessary, shutting the reactor down

• Removing reactor heat, including decay heatfollowing shutdown, in order to preventheat up of fuel

• Containing radioactive products that arenormally produced and contained withinthe fuel

• Monitoring the plant to ensure that theabove functions are being carried out, and ifnot, ensuring that mitigating actions arebeing taken

These nuclear safety functions are carried outto a high degree of reliability by applying thefollowing principles:

• The use of high-quality components andinstallations

• Maximizing the use of inherent safetyfeatures of the ACR-1000

• Implementing multiple defence-in-depthbarriers for prevention of radioactiverelease

• Providing enhanced features to mitigate andreduce consequences of design basis eventsand severe accidents

The implementation of these safety measuresis provided by safety systems, safety supportsystems, systems important to safety androbust buildings and structures that meethigh standards for diversity, reliability andprotection against common-mode eventssuch as seismic occurrences, fires, floodingand unauthorized acts.

3.2 Defence-in-Depth

The ACR-1000 is based on the CANDUprinciple of defence-in-depth by providing thefollowing multiple, diverse barriers foraccident prevention and mitigation ofconsequences:

• High-quality process systems toaccommodate plant transients and tominimize the likelihood of accidents

• Reliable safety systems for reactorshutdown, emergency core cooling,containment, and emergency heat removal(emergency feedwater)

• Reliable safety support systems to provideservices to the safety systems and othermitigating systems

• Backup systems for heat sinks and essentialcontrols

• Passive heat sinks to increase resistanceagainst both design basis events and severeaccidents

The ACR-1000 has at least sevenbarriers:1) Fuel sheath which contains the

radioactive material2) Heat transport system, including pressure

tubes3) Calandria tubes designed to withstand a

pressure tube rupture4) Cool, low-pressure moderator5) Cool, low-pressure reactor vault6) Reserve water system7) Steel-lined, concrete containment structure

The design of the safety systems follow thedesign principles of separation, diversity andreliability. High degrees of redundancy withinsystems are provided to ensure the safetyfunctions can be carried out, even whensystems or components are impaired.Protection against seismic, flooding and fireevents is also provided, ensuring highlyreliable and effective mitigation of postulatedevents, including severe accidents.

3. Nuclear Safety and LicensingN

uclearSafety

andLicensing

27

3.3 Inherent Safety Features

The ACR-1000 maintains the traditional CANDUinherent safety characteristics:

• Heavy water moderator, which is very efficient inslowing down neutrons, resulting in a fission processwhich is more than an order of magnitude slowerthan LWRs. Reactor control and shutdown areinherently easier to perform

• On-power refuelling, which reduces the ‘excess’reactivity as required. Reactor characteristics areconstant and no additional measures, such as boronaddition to the coolant (and its radioactiveremoval), are needed

• Natural circulation capability in the reactor coolantsystem, which can cope with transients due to lossof forced flow

• Reactivity control devices. These are in the low-pressure moderator, do not penetrate the reactorcoolant pressure boundary and therefore cannotbe ejected

• Moderator backup heat sink, which maintains corecoolability for loss-of-coolant accidents, even whencombined with the unavailability of emergency corecooling

• Negative power reactivity coefficient, which makesreactor power more stable and easier to control

• Small negative full-core void reactivity offering agood balance of nuclear protection between loss-of-coolant accidents and fast cool-down accidents

• Very flat and stable flux across the core minimizingdemand on the reactor control system

• Larger safety and operating margins due to the useof CANFLEX-ACR fuel, with lower element ratingand higher critical heat flux limits

Figure 3-1 Barriers for Prevention of Releases

28

3.4 Severe Accidents

A severe accident is one in which the fuel is notcooled within the heat transport system. TheACR/CANDU design principle is to prevent severeaccidents and to mitigate severe accident events, inaddition to minimizing their consequences. This isachieved by providing a number of design measures:

• Normal heat removal systems• Heat removal systems using emergency feedwater

system• Passive emergency feedwater supply from reserve

water system• Emergency core cooling• Passive emergency heat transport system make-up

from reserve water system• Heat removal using moderator systems• Passive thermal capacity of moderator• Passive emergency moderator heat sink make-up

from reserve water system• Heat removal by reactor vault water• Passive thermal capacity of reactor vault water• Passive emergency reactor vault heat sink make-up

from reserve water system• Passive containment cooling via spray• Severe accident management monitoring capabilities

Severe accident management, in addition to providingmultiple mechanisms for fuel cooling and barriers torelease, also includes mitigating measures withincontainment. In addition to the robust, concrete outerand inner steel liners, which by themselves canwithstand the largest pipe breaks, containment is alsoprovided by:

• Passive, hydrogen recombiners and igniters that willlimit the hydrogen content to below the deflagrationlimit

• A spray system to reduce the build-up of containmentpressure and reduce leakages

• Highly reliable local air coolers that can be used forcontainment heat removal

PSA studies estimate that the summed frequency ofinternal initiating events leading to reactor coredamage during at-power operation is only 3.4 x 10-7

for the ACR-700 and is expected to be better for theACR-1000. This exceeds EPRI requirements byapproximately two orders of magnitude and iscomparable to latest LWR designs. This marginal valueis comprised of probabilities of seven dominantinitiating events, all of which are relatively small.

3.5 Licensing Basis

The ACR-1000 builds on the successful CANDU trackrecord of accommodating regulatory requirements ofoffshore jurisdictions in various host countries (China,South Korea, Romania, Argentina) while retaining thestandard nuclear platform.

The ACR-1000 is designed to meet regulatoryrequirements in Canada and other countries:

• The ACR-1000 is an evolutionary, enhanced designbased on current regulations. Future licensability inCanada and abroad will be based on this experience

• ACR-1000 design meets the requirements ofapplicable IAEA Safety Series documents for nuclearpower reactors

• The design meets the Canadian and internationalrequirements for nuclear plant siting

• International codes and standards, as they apply tothe ACR-1000 design, have been incorporated.ACR-1000 has benefited from the extensive reviewof US NRC requirements—both its writtenregulations and via dialogue

Figure 3-2 Core Damage Frequencies per Year29

The feedback gained from AECL’s pastconstruction projects, associated withimprovements and optimization of keyproject elements, results in an optimum 42-month schedule (nth Unit) from firstContainment Concrete to fuel load.Deployment of the ACR-1000 requires thecoordination and timely delivery of keyproject elements, including: licensingprograms, environmental assessments, designengineering, procurement, construction andcommissioning start-up programs.

Design Engineering: Prior to a projectcontract, a series of activities are executed toensure design readiness and a seamlesstransition to the procurement andconstruction phases. Preliminary design andresearch and development programs areexecuted in parallel with the environmentalassessment and licensing programs, ensuringcontinuous improvement and plantconfiguration is maintained. The final designprogram ensures plant reliability, equipmentand component maintainability andconstructability requirements are maximizedto the fullest extent.

Licensing: The ACR-1000 builds on thesuccessful CANDU track record ofaccommodating requirements of offshorejurisdictions in various host countries whileretaining the standard nuclear platform.Licensing programs are executed andcoordinated with the engineering designprograms and environmental assessment, andare structured in a manner to supportregulatory process requirements.

Configuration Management: The ACR-1000 makes use of the latest computertechnology for managing the complete plantconfiguration from design to constructionand finally, turnover to the Owner. State-of-the-art electronic drafting tools areintegrated with material management, wiringand device design, and other technologyapplications.

Project Management: The ACR-1000project management structure provides fullyintegrated project management solutions.Performance management programs areexecuted from project concept, through a project readiness mode, and finally project

closeout. The projectmanagement frameworkconsists of three keyelements:

• Total projectexecution planning

• Critical decisionframework to controleach phase of theproject

• Comprehensive riskmanagement program

AC

R-1000

Deploym

ent

4. ACR-1000 Deployment

,

Figure 4-1 42-Month Deployment Schedule (Nominal)

30

Procurement: Standardized procurement andsupply processes are implemented to support time,cost and performance benefits to the project,including benefits such as efficiency through varietycontrol (standardization), economy in manufacturingand servicing, and avoidance of repetitive effort inproducing new specifications and processes for eachprocurement.

Construction Programs: Constructabilityprograms are implemented to ensure simplification,maximized concurrent construction, increasedconstruction productivity, minimized constructionrework, decreased construction equipment costs,minimized unscheduled activities, and reduced capitalcosts and construction risk.

Construction Strategy: The main elements of theACR-1000 construction strategy are:• Open-top construction method using a very-

heavy-lift crane• Concurrent construction• Modularization and prefabrication• Use of advanced technologies to minimize

interferences.

The open-top/vertical installationconstruction method enables animproved logic that reduces costs whilereducing the schedule risk. The internalstructure of the reactor building isinitially built as vertical walls withoutfloors. Major modules, including thefloors, are then installed in parallel.

Commissioning: The commissioningand plant start-up programs for theACR-1000 are being developed withinput received from design staff and plantoperations staff. Identification of keydesign parameters that requireconfirmation to meet overall systemobjectives are reviewed to ensurecommissioning plans can be produced tocheck those identified parameters. Inaddition, acceptance criteria will bedeveloped between the designer andexperienced commissioning technicalstaff.

Test programs will be defined as part of the overallplan, including:• Preoperational tests• Fuel loading, initial criticality, and low power tests• Power tests• Test run and performance tests

Figure 4-3 Module Lift Using VHL Crane

Figure 4-2 Design Engineering Applications

31

Figure 4-4 Typical Reactor Building Modules

32

Operation

andM

aintenance5.1 Consistently Better

Performance

The lifetime capacity factor for the ACR-1000is expected to be greater than 90% over theoperating life of 60 years. The year-to-yearexpected capacity factor is 95%. Theseexpectations are based on the proven trackrecord of CANDU 6s, which havecollectively surpassed the U.S. PWR/BWR

Gross Capacity Factor (GCF) with acombined average of 92.4% in 2006. Theseresults are consistently better than LWRsaround the world.The ACR-1000 has made a number ofimprovements to achieve these incrementalperformance targets.

5.2 Enhanced PerformanceFeatures

Incorporation of feedback from operatingreactors (both CANDU and other designs) isan integral feature of the design process.Various new features and maintenanceimprovement opportunities have beenincorporated to enhance operatingperformance throughout station life.

Major enhancements include:

• Use of improved material and plantchemistry specifications, based on operatingexperience from CANDU plants. Forexample, life-limiting components such asHTS feeders and headers have been replacedwith stainless steel to limit the effect offeeder corrosion

• Implementation of advanced computercontrol and interaction systems formonitoring, display, diagnostics andannunciation. These include ergonomicoperator consoles, touch displays, largecolored screens, smart communications forimproved operator awareness and plantstatus through modern human-factorsengineering

• Providing integrated SMART CANDUmodules for annunciation, on-linemonitoring of systems and components, andproviding a predictive maintenance capability

5. Operation and Maintenance

COGnizant Volume 12, Issue 6, 2007, 2006 U.S. and world data based on Q4 results (courtesy of NEI).The graph is for comparison of trends only.

Figure 5-1 Comparison of Gross Capacity Factors

CANDU 6

CANDU multi-unitUS PWR/BWR

World -all

95%

90%

85%

80%

75%

70%

65%

60%

Gro

ssC

apac

ity

Fact

or

1999 2000 2001 2002 2003 2004 2005 2007Q4 2006

CANDU 6/PHWR Performance Trends (1999 - 2006)Reference: CANDU Owners Group Newsletter

33

• Enhancing power maneuvering capability:- Load-following the grid provides up to 2.5% power

variation, while operating at 97.5%- Daily load-cycling capability includes rapid load

reduction from steady state 100% poweroperation to 75%, and periodic load reductionfrom 100% to 60% and as low as 50% whenrequired (e.g. , weekends)

- Use of LEU fuel and light water coolant hasresulted in a lower xenon load followingreactor power reduction compared toCANDU. This simplifies reactor operation andmakes the ACR-1000 inherently moreresponsive

• Ensuring station blackout capability for return to fullpower on restoration of electrical grid. The ACR-1000 has the capability to continue operation of houseload without a grid connection,enabling a rapid returnto full power upon reconnection

5.3 Enhanced Maintenance Features

The lifetime capacity factor of a plant is impacted by thenumber and duration of maintenance outages. Thetraditional ‘annual’ outage of up to one month forcurrently operating CANDU plants has been improvedto a ‘major’ outage of only 21 days every three years forthe ACR-1000. A number of enhancements to achievethese objectives have been incorporated.

• A maintenance-based design strategy has beenimplemented. The program incorporates lessonslearned and ensures maintainability of systems andcomponents. It will define the improvements made tomaintenance programs for earlier designs. The newprogram is based on the SMART CANDU technology.It will identify and take mitigating actions, if required,to ensure plant states are diagnosed and maintainedwithin their design performance limits. This will leadto improved preventive maintenance and reducedforced outages at a rate of less than five days/year.Only the best available equipment for criticalcomponents will be used

Figure 5-2 ChemAND – Performance Monitor for Plant Chemistry

34

Figure 5-4 Typical System Equipment Module

Figure 5-3 Maintenance Basis

35

• Plant layout has been improved by providing generousspace, laydown areas, good lighting, and use ofpermanent walkways and platforms to minimize needfor temporary scaffolding. Provision for electrical,water and air supplies are built-in for on-power andnormal shutdown maintenance

• Effective use of on-power reactor building accessibilityand on-power maintenance of four-division designsafety systems will minimize the amount ofmaintenance that must be performed during shutdown

• Computerized testing of major safety systems andautomatic calibration of in-core detector controlsignals reduce both on-line testing and start-uptesting time

• More durable materials and robust design marginssimplify fuel channel inspections

• Shielding in radiologically-controlled areas has beenincreased. This feature, along with reduced tritiumreleases due to use of light water coolant, will resultin enhanced radiological protection to furtherreduce worker exposure and occupational dose.Dose to an individual station staff member isexpected to be less than 50 mSv in any single year

• The design for planned outages every three years isaccomplished by selection of equipment and systemdesign. It is based on probabilistic safety evaluationsusing three-year outage intervals

Figure 5-5 Service Elevator

Full height service elevatorplaced in close proximity to the auxiliary airlock, improvingO&M access to all floors with-in the reactor building

36

Figure 5-6 Accessible Areas in the Reactor Building – Level 100 m

The plant layouts above show the accessible areas in the plant, enhanced for ease of operation and maintenance.

Figure 5-7 Accessible Areas in the Reactor Building – Level 125.4 m

37

The waste management systems for theACR-1000 will minimize the radiologicalexposure to operating staff and the public.Exposures for workers from the plant aremonitored and controlled to ensure they arewithin the limits recommended by theInternational Commission on RadiologicalProtection. The systems for the ACR-1000have been proven over many years at otherCANDU sites. They provide for the collection,transfer and storage of all radioactive gases,liquid and solid, including spent fuel and wastesgenerated within the plant:

• Gaseous radioactive waste gases, vapoursor airborne particulates are monitored andfiltered. Active gases are treated by the off-gas management system (OGMS) with anabsorber bed. Any tritium releases fromisolated moderator areas are collected by avapour recovery system and stored on site

• Liquid radioactive wastes are stored inconcrete tanks located in the maintenancebuilding. Any liquid requiring removal ofradioactivity, including spills, is treated usingcartridge filters and ion-exchange resins

• Solid radioactive wastes can be classified byfive main groups: spent fuel, spent ion-exchange resins, spent filter cartridges,compactable and non-compactable solids.Each type is processed and moved, usingspecially designed transporting devices, ifnecessary. After processing, wastes arecollected and prepared for on-site storageby the utility or for transport off-site

AECL has developed the MACSTOR®****(Modular Air-Cooled Storage) system forsafe, above-ground storage of spent fuel.MACSTOR has been developed from morethan 30 years of experience.

Figure 6-1 Spent Fuel StorageBasket

MACSTOR saves up to one-third of the spacerequired for comparable systems, requiresless manpower, has low operating andconstruction costs, and permits easy fuelretrieval.

With highly efficient heat-rejection andshielding capabilities, it is constructed usingmultiple barriers to provide radiationshielding for operators and the public, whilebeing appropriately qualified and equippedwith monitoring facilities.

Figure 6-2 MACSTOR Fuel Transfer

Figure 6-3 AECL’s MACSTOR System

6. Radioactive Waste Management

****MACSTOR® is a registered trademark of Atomic Energy of Canada Limited (AECL).

38

Radioactive

Waste

Managem

ent

Decom

missioning

AECL, through its membership in theOECD/NEA co-operative programme ondecommissioning, has adopted a three-stagedecommissioning strategy:

1) Placement of the station into a static state.This dormancy state is a modified IAEAStage I concept such that:

• Buildings around the reactor building are decommissioned for alternate use

• The reactor building is isolated and sealed

• The plant is monitored to ensure its dormant state

2) IAEA Stage II dormancy period, assumed tobe 40 years or more, depending on theOwner’s plans

3) IAEA Stage III final decommissioning tounrestricted use of the land

As an evolution of CANDU 6, the enhancedACR-1000 design features a number ofsystems that have been simplified and/oroptimized; some have also been eliminated.Thus, the amount of materials to bedecommissioned is less than CANDU 6.Some examples are:• Reduction of heavy water by elimination or

downsizing of heavy-water-related systems• Reactor core size reduction• Consideration of alternative structural

material yielding less cumulative radioactivityat end of life

• Civil structure size reductions

ACR-1000 design features that assist inmaintenance and inspection during thelifetime of the reactor will also facilitatedecommissioning. For example, the division ofthe reactor building into separatecompartments, with proper isolation andshielding, allows the segregation ofcontaminated from non-contaminatedsystems, facilitating efficient dismantling,removal and disposal.

AECL has decommissioned three prototypeNuclear Power Plants and one researchreactor to a static state. It hasdecommissioned at least one facility to IAEA’sStage III. AECL has also participated indecommissioning plans of facilities in Japan,the U.S. and elsewhere.

AECL has all the experience and facilities required to support Ownerdecommissioning plans.

7. Decommissioning

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EvolutionCapitalizing on the proven features ofCANDU technology, AECL has designed theevolutionary ACR-1000 to be cost-competitive with all forms of energy,including other nuclear technologies, whileachieving higher safety and performancestandards consistent with customerexpectations.

Proven CANDU Features• Heavy water moderator and horizontal

fuel channel design• Series of parallel pressure tubes—rather

than a single pressure vessel—allowingsimpler manufacturing and reduced cost

• Two independent, passive, fast-acting safetyshutdown systems and a unique inherentemergency cooling capability

• On-power fuelling for flexible outageplanning and minimal ‘excess’ reactivityburden

• Multiple heat removal systems to preventand mitigate severe accidents

ACR-1000 Innovations• Extended fuel life through use of low

enriched uranium fuel• Reduced heavy water inventory by

approximately 60% of traditional CANDUreactors, by use of light water coolant andreduced lattice

• Compact, highly stable reactor core design• Reduced spent fuel volume• Improved thermal efficiency through

optimized, higher-pressure steam turbines• Modular, prefabricated structures and

systems• Advanced construction techniques• Quadrant-based safety and heat sink

system layout design for improved on-power maintenance and testing, additionalredundancy in actuating signals for tripchannels, reduced risk of spurious trips andoverall increased reliability

• Enhanced safety design including addition ofreserve water system for passive accidentmitigation

• Improved power manoeuvrability withlower inherent xenon load after shutdownthan traditional CANDU

• Improved design for maintainability andoperability

• Design validated by exhaustive proof-testing

• Comprehensive Risk Management Program

The ACR-1000 meets customer expectationsfor safe, reliable and economicallycompetitive power production. It is thepreferred choice… based on a wealth ofexperience, technical excellence andinnovations in engineering.

Conclusion

8. Conclusion

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Company Profile

Atomic Energy of Canada Limited (AECL) is a full-service nuclear technology

company that supplies products and services to nuclear utilities around the

world. AECL is the designer and builder of CANDU® reactors, including the

Advanced CANDU Reactor® (ACR-1000®), the Enhanced CANDU 6™ (EC6™)

and the CANDU 6, one of the world's top performing reactors. AECL's over

5,000 full-time employees deliver cutting-edge world class nuclear services

including: research and development, design and engineering, construction

management, specialized technology, life extension, waste management and

decommissioning in support of CANDU reactor products.

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ACR-1000® Advanced CANDU Reactor®

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2251 Speakman DriveMississauga, OntarioCanada L5K 1B2

Tel: 905-823-9060

www.aecl.ca

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