CANDU Nuclear Power System...UNRESTRICTED TDS1 -105 I I | CANDU NUCLEAR POWER SYSTEM I • ABSTRACT...
152
CANDU Nuclear Power System .muary 1981
Transcript of CANDU Nuclear Power System...UNRESTRICTED TDS1 -105 I I | CANDU NUCLEAR POWER SYSTEM I • ABSTRACT...
CANDU NuclearPower System.muary 1981
• UNRESTRICTED TDS1-10S
III* CANDU NUCLEAR POWER SYSTEM
IIIIII1I1II ATOMIC ENERGY OF CANADA LIMITED
ENGINEERING COMPANY
1 SHERIDAN PARK RESEARCH COMMUNITY
MISSISSAUGA, ONTARIO L5K 1B2
1F
1901 JANUARY
™ UNRESTRICTED TDS1 -105
II| CANDU NUCLEAR POWER SYSTEM
I• ABSTRACT
I This report provides a comprehensive summary of the
many components that make up a CANDU reactor. Majoremphasis is placed on the CANDU 600 MW(e) design. The
I reasons for CANDU's superior performance and the
inherent safety of the system are also discussed.
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TABLE OF CONTENTS
1. INTRODUCTION
1.1 Introduction to AECL1.2 History and Performance of CANDU1.3 Layout and Component Parts of CANDU
2. POWER SYSTEMS
2.1 Reactor Core2.2 Heat Transport Systems2.3 Overall Plant Control2.4 Core Control2.5 Reactivity Control Devices2.6 Fuel2.7 Startup, Operation and Shutdown Sequences2.8 Secondary Side Systems
3. MODERATOR AND AUXILIARY SYSTEMS
3.1 Main Moderator System3.2 Moderator Auxiliary Systems3.3 Heavy Water Management
4 . SAFETY SYSTEMS
4.1 Inherent Safety Features of CANDU4.2 Safety Design Philosophy4.3 Safety Systems Description
5. REFUELLING SYSTEM
5.1 Fuelling Machines5.2 Fuel Transfer5.3 Fuel Storage
6. SUMMARY
6.1 Advantages of CANDU6.2 Conclusion
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1.0 INTRODUCTION
1.1 INTRODUCTION TO AECL
AECL is a crown corporation of the Canadian government charged with themandate to develop nuclear power and associated industries in Canada.AECL commenced operation in 1945 with the construction of elaboratenuclear laboratories and research reactors in Chalk River, Ontario.Today, AECL is a large multi-faceted company with over 7,500 employees.The various activities of the company are illustrated in Figure 1.1-1, andthe structure of the company as it is presently constituted, is shown inFigure 1.1-2.
The company's major facilities in Canada (Figure 1.1-3), are:
. The Head Office, located in Ottawa, Ontario
. Nuclear laboratories at Chalk River, Ontario and Pinawa, Manitoba -
. Douglas Point and Gentilly-1 nuclear generating stations, located inOntario and Quebec respectively
. The Glace Bay and Port Hawkesbury heavy water plants in Nova Scotia, andthe partially completed La Prade heavy water plant in Quebec
. The Radiochemical Company, located in Ottawa, markets radioisotopes anddesigns, manufactures and markets radiation equipment
. The Chemical Company, located in Ottawa, responsible for heavy waterproduction
. The Engineering Company, located in Mississauga with a subsidiary officein Montreal, responsible for the engineering of nuclear power reactors
» AECL International located at Mississauga and elsewhere is responsiblefor overseas projects and international marketing
AECL today has world-wide experience in design, project management andconstruction of large nuclear generating stations and research reactors.AECL is also experienced in transferring CANDD expertise from Canada toparticipating countries through the establishment of local manufacturingprograms. AECL has a genuine interest in this transfer of technology andin the generation of local employment in participating countries.
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AECL *OPERATES LABORATORIES. I
PROVIDES NUCLEAR CONSULTING SERVICES.DESIGNS CANDU NUCLEAR POWER STATIONS. |BUILDS AND MARKETS NUCLEAR PLANTS.
BUILDS AND OPERATES HEAVY WATER PUNTS. }PRODUCES AND MARKETS RADIOISOTOPES.LIAISES WITH INDUSTRY AND UNIVERSITIES. I
COOPERATES WITH OTHER COUNTRIES AND AGENCIES.
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FIGURE 1.1-1
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ATOMIC ENERGY OF CANADA LIMITEDCORPORATE OFFICE
• Directs and administers the Company's activities.
• Markets CANDU nuclear reactors, components and technology.
• Effects scientific and technological exchange agreements with counterpart agencies inother countries.
• Makes available its special facilities and expertise to assist Utilities n the practical useof nuclear energy, and other Government agencies In their operation and services.
RESEARCH COMPANY
• Operates laboratories for fundamental andapplied research and engineering developmentin the nuclear field.
• Enters into co-operative research and develop-ment contracts with industry and universities.
• Makes available its special facilities and ex-pertise to assist universities in nuc ear studies
and techniques.
CHEMICAL COMPANY
• Constructs and operates Heavy Water Produc-tion Plants.
• Provides heavy water for nuclear reactor re-quirements.
• Coordinates the development of Heavy Watertechnology.
ENGINEERING COMPANY
• Designs nuclear generating stations, in co-operation with electric utlities and privateindustry.
• Provides nuclear power plant consulting ser-vices and undertakes development work Insupport ot the CANDU nuclear power plant.
* Provides nuclear steam plant equipment andmakes available its special facilities and ex-pertise to assist in developing manufacturingcapability to nuclear specifications.
RADIOCHEMICAL COMPANY
• Produces and markets radioisotopes.
• Designs, manufactures, and markets equip-ment for the utilization of radioisotopes andradiation.
AECL INTERNATIONAL
• Identifies offshore sales opportunities.
• Formulates marketing strategies in order todevelop new reactor and associatedtechnology sales overseas.
• Promotes CANDU export sales to these newlydefined markets, and to existing markets (incooperation with the organization of CANDUindustries O.C.I.)
• Represents the Chemical Company and theResearch Company in offshore sales.
• Constructs nuclear generating stations ncooperation with international electrlcautilities and local industries.
FIGURE 1.1-2 ATOMIC ENERGY OF CANADA LIMITED ORGANIZATION/RESPONSIBILITIES
1 Whitnnall Nuclear RtMarchEstablishment
2 Bme* Nuclear Power DevelopmentDouglas Point Nuclear Generating Station
3 AECL Engineering Company
AECL International
4 Picturing Nuclear Generating Station
5 DarUngton Nuclear Generating Station
6 Nuclear Power Demonstration (NPD)
7 Chalk River Nuclear Laboratories
S AECL Corporate Office
AECL International Head Office AECL Research CompanyAECL Radiochemical Company AECL Chemical Company
9 AECL Engineering Company-Montreal ;
10 Gentilly-1 Nuclear Generating Station :Gentilly-2 Nuclear Generating Station •La Prade Heavy Water Plant :
11 Point Lepreau Nuclear Generating Station :
12 Port Hawkesbury Heavy Water Plant :
13 Glace Bay Heavy Water Plant ;
FIGURE 1.1-3 NUCLEAR ENERGY ESTABLISHMENTS IN CANADA
f^^'vr*1^ (--̂ Frt-ii -IJ i
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1.2 HISTOK? AND PERFORMANCE OF CANDU
The history of CANDU is illustrated in Figure 1.2-1, and shows thegenealogy of the CANDU reactor from its inception to the current state ofthe art. Figure 1.2-2 summarizes this graph in tabular form.
The Pickering "A" nuclear generating station (Figure 1.2-3) consists offour units of 515 MW(e) electric each, for a total net output of 2060MW(e). It is situated on the outskirts of, and provides power to, thecity of Toronto, Ontario.
These units came into service during the years 1971 to 1973. Adjacent tothe Pickering "A" station is the Pickering "B" station, again comprisingfour units, which will deliver 2060 HW(e) when completed during the years1981 to 1983. The Bruce "A" nuclear generating station (Figure 1.2-4) islocated approximately 160 miles north-west of Toronto, and consists offour units of 746 MW(e) each, for a total station output of 2984 MW(e)electric, plus enough steam to feed the Bruce heavy water plants. Theseunits came into service during the years 1977 to 1979 and hence form thenewest additions to the Canadian nuclear grid. On the same site, theBruce "B" nuclear generating station is under construction with its fourunits of 769 MW(e) electric each, due in service during the years 1984 to1987.
Ontario Hydro has also commenced construction of a four unit nucleargenerating station of 850 HW(e) per unit at the Darlington siteapproximately 60 miles east of Toronto.
Currently, more than 31% of the electrical demand in the Province ofOntario is generated by nuclear power.
The Gentilly-2 (G-2) nuclear generating station (Figure 1.2-5) is a 638MW(e) net single unit station under construction adjacent to theGentilly-1 nuclear generating station, the only CANDU station with boilinglight water coolant.
The Point Lepreau generating station is under construction in the Provinceof New Brunswick. It is very similar to the Gentilly-2 station and willproduce 633 MW(e) net.
CANDU stations are also operating or under construction in othercountries. Three of these are shown in Figures 1.2-6, 7 and 8.
CANDU stations are the world leaders in availability. These facts areillustrated in Figures 1.2-9 and 1.2-10 which show extracts from 'NuclearEngineering International1 comparing the CANDU system against othercompeting reactor types.
1 1.1-2
800
700
600
POWERS 500REACTORSMW(e)
RESEARCHREACTORSMW(Ih)
Length and location of rectangles denotes• construction and commissioning period.Arrows denote flow of information.
General research, developmentand design information.
1945 1950 1955 1960 1965 1970YEARS
FIGURE 1.2-1 GENEALOGY OF CANDU REACTORS
1975 1980 1985 1990
NAME
NPD
DOUGLAS POINTPICKERING AGENTIuLY 1
KANUPPRAPP1
RAPP2
BRUCE AGENTILLY 2
POINT LEPREAUCORDOBAPICKERING B
WOLSUNG 1BRUCE B
DARLINGTONCERNAVODA
LOCATION
ONTARIOONTARIO
ONTARIOQUEBECPAKISTANINDIAINDIA
ONTARIOQUEBECNEW BRUNSWICKARGENTINA
ONTARIOKOREAONTARIO
ONTARIOROMANIA
TYPE
PHW
PHWPHW
BLWPHW
PHWPHWPHW
PHWPHW
PHWPHW
PHWPHW
PHWPHW
TOTAL
POWERMWeNET
22206515x4266
125203
203740x4640
635600516x4600756x4
850x4
600
18,208 MWe
NUCLEARDESIGNER
AECL & CGEAECL
AECLAECLCGEAECLAECL
AECLAECL
AECLAECLAECL
AECLAECL
AECLAECL
DATE OFFIRSTPOWER
196219671971/73
197119711972—
1976/79——
—
——
—
—_
FIGURE 1.2-2 CANDU POWER REACTORS
FIGURE 1.2-3 PICKERING 'A' AND 'B' 8 x 515 MW{a)
FIGURE 1.2-4 BRUCE 'A' 4 X 700 MW(e> START OF PROJECT 1970
FIGURE 1.2-5 GENTILLY-2 600MW(e)
FIGURE 1.2-6 ARGENTINA — CORDOBA CANDU 600 MW(#)
FIGURE 1.2-7 REPUBLIC OF KOREA - WOLSUNG CANDU 600 MW(e)
C
FIGURE 1.2-8 ROMANIA - CERNAVODA 4 x 600 MW(e)
20
1973 1974 1975 1976 1977 1978 1979 1980
(Courtesy of Nuclaar Engineering lnternational,1980 December)
FIGURE 1.2-9 COMPARISON OF THE PERFORMANCE OF FOUR TYPES OF NUCLEAR REACTORS
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Station
Bruce-3Stade-1Pickering-2Pickering-1Point Beach-2Pickering-4Pickering-3Prairie lsland-2Calvert Cliffs-2Connecticut YankeeBruce-4Bruce-1
Cumulative LoadFactor %
82.081.280.980.377.477.375.475.274.774.673.573.0
Type
CANDUPWRCANDUCANDUPWRCANDUCANDUPWRPWRPWRCANDUCANDU
Ref: Nuclear Engineering International Vol. 25 No. 307,1980
III FIGURE 1.2-10 CUMULATIVE LOAD FACTORS FOR REACTORS OVER 500 MW(e)
TO END OF SEPTEMBER 1960
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1.3 LAYOUT AND COMPONENT PARTS OF CANDO |
The terminology used in describing CANDU is shown in Figure 1.3-1. This _terminology is most convenient in referring to delineations in scope of Isupply. "
Figure 1.3-2 shows the Nuclear Steam Plant and Balance of Plant portions. IFigure 1.3-3 illustrates containment, which consists solely of the reactor 1building. Figure 1.3-4 shows the CANDU Nuclear Steam Supply System insidecontainment and Figure 1.3-5 shows pictorially the layout inside "tcontainment* Figure 1.3-6 illustrates the control room which is in the Jservice building outside of containment.
The site layout for.G-2 (Figure 1.3-7) shows water intake and discharge |facilities and the pumphouse in addition to the Nuclear- Steam Plant and *Balance of Plant facilities. Figure 1.3-8 shows a twin 600 MW(e) NuclearGenerating Station which is a feature of the plant. The reactor building .1layout is identical in both units and the service building layout is 1adjusted to accommodate the dual unit configuration.
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1.1-3
CANDU 900 NUCLEARGENERATING STATION (NGS)
NUCLEAR STEAMPLANT (NSP)
NUCLEAR STEAM SUPPLYSYSTEM (NSSS)
BALANCE OF NUCLEARSTEAM PLANT (BNSP)
I - REACTOR BUILDINGI - SERVICE BUILDING*~ SPENT FUEL BAY(SFB)
BALANCE OFPLANT (BOP)
, TURBINE GENERATORAND AUXILIARIES
ELECTRIC POWERSYSTEMS (BOP)
, COMMON PROCESSESAND SERVICES (BOP)
, BUILDINGS ANDSTRUCTURES
NUCLEAR CORESYSTEM (NCS)
NUCLEARCORE(NC)
BALANCE OFNUCLEAR CORESYSTEM (BNCS)
STEAM GENERATINGSYSTEM (SOS)
BALANCE OF NUCLEARSTEAM SUPPLY
SYSTEM (BNSSS)
• REACTOR• FUEL HANDLING•CONTROL CENTRE (NC) t
PRIMARY HEATTRANSPORT SYSTEMSTEAM AND WATERSYSTEMS (NSP)
I - HEAVY WATER MANAGEMENTI - ELECTRIC POWER SYSTEMS (NSSS)*- COMMON PROCESSES AND SERVICES (NSSS)
& SPENT FUEL BAY PROCESS SYSTEMS
MODERATOR SYSTEMAUXILIARY SYSTEMSEWS, EPS
FIGURE 1.3-1 COMPONENT PARTS OF THE PLANT
1 REACTOR2 FUELLING MACHINE3 STEAM GENERATOR4 DOUSING SYSTEM5 NEW FUEL HANDLING6 SPENT FUEL HANDLING
7 SPENT FUEL INSPECTION AND STORAGE8 MAINTENANCE AND INSPECTION9 CONTROL CENTRE
10 MECHANICAL WORKSHOP11 WATER TREATMENT12 TURBINE AND GENERATOR
FIGURE 1.3-2 600 MW(e) NUCLEAR GENERATING STATION
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SEPARATION MEMBRANE
FIGURE 1.3-3 REACTOR BUILDING SECTION
1 MAIN STEAM SUPPLY PIPING
2 STEAM GENERATORS
3 MAIN PRIMARY SYSTEM PUMPS
4 FEEDERS
5 CALANDRIA ASSEMBLY
6 FUEL CHANNEL ASSEMBLY
7 FUELLING MACHINE BRIDGE
8 MODERATOR CIRCULATION SYSTEM
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FIGURE 1.3-4 NUCLEAR CORE SYSTEM AND STEAM GENERATING PLANT
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DOUSING WATER TANKDOUSING WATER VALVESMODERATOR PUMPMODERATOR HEAT EXCHANGERFEEDER CABINETSREACTOR FACEREACTORREACTIVITY MECHANISM
9 HEAT TRANSPORT SYSTEM PUMP10 FUELLING MACHINE BRIDGE
FUELLING MACHINE CARRIAGEFUELLING MACHINE CATENARYFUELLING MACHINE MAINTENANCE LOCKFUELLING MACHINE MAINTENANCE LOCK DOOREND SHIELD COOLING WATER DELAY TANKVAULT COOLERPRESSURIZER
18 STEAM GENERATOR19 STEAM GENERATOR ROOM CRANE
FIGURE 1.3-5 600 MW(e) REACTOR BUILDING CUTAWAY
CONTAINMENTEMERGENCYCORECOOLING
FUELLING MACHINEAND FUEL HANDLINGCONTROL CONSOLE
MODERATOR ANDREACTOR MISCELLANEOUSSYSTEMS
PRIMARY HEATTRANSPORT SYSTEM
ANNUNCIATIONAND DIGITAL
CONTROLCOMPUTERS
STEAM GENERATOR
SYSTEM TURBINE GENERATOR
ELECTRICALDISTRIBUTION
SWITCHYARD SYSTEMS
MISCELLANEOUSAUXILIARYSYSTEMS
/ / / I / IPL14 PL15 PL16 PL17 PL18 PL19 PL20
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FIGURE 1.3-6 CONTROL CENTRE
• I ' . i '
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(0
I
o<
2oco
i5
o
= UJ g o 5 g
UJ j * u § 3 s < S 5L5
» - (NJ CO * ID <
F/M MAINTENANCE
F/M DECONTAMINATION
DECONTAMINATIONCENTRE .
ERADIATIONPROTECTION
FIGURE 1.3-8 SERVICE BUILDING PLAN EL. 100 (GRADE)
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2.0 POWER SYSTEMS
2.1 REACTOR CORE
2.1.0 General
This section of the CfiNDU Nuclear Power System presentation introduces theprincipal features of the Reactor Core and its location arrangements.Particulars of the related process and control systems are covered insubsequent sections.
The CANDU 600 MW(e) Reactor is the sixth in a series of Pressurized HeavyWater Reactor (PHWR) designs developed in Canada for the production ofelectric power from natural uranium fuel.
Like preceding CANDU reactors, this design incorporates a standardized,geometrical arrangement of horizontal pressure tubes which contain fueland circulating heavy water coolant at high pressure.
These fuel channels are mounted within a cylindrical calandria, whichcontains heavy water moderator in a separate low pressure system.
The CANDU 600 MW(e) Reactor Core is located at the heart of the ReactorContainment building within biological shielding.
2.1.1 Reactor Assembly
The Reactor Assembly is mounted inside a steel-lined light water filledconcrete vault (Figure 2.1-2) and comprises:
. A cylindrical low-pressure calandria vessel of stainless steelconstruction.
. Two integral end shields (also of stainless steel construction withcarbon steel shielding balls) easjh horizontally penetrated by 380lattice tubes.
. 380 Zircaloy-2 calandria tubes joining the lattice tubes at eachposition in the lattice.
. 3S0 fuel channel assemblies mounted within these lattice sites.
. Vertical and horizontal reactivity control devices which penetrate thevault shielding to provide power sensing, control and shut-downfeatures.
. Connections for the heat removal recirculation of the heavy watermoderator.
2.1-1
3.4.5.6.7.S.9.
10.11.12.13.14.IS.
CALANORIACALANDRIA - SIDE TUBESHEETCALANDRIA TUBESEMBEDMENT RINGFUELLING MACHINE - SIDE TUBESHEETENO SHIELD LATTICE TUBESEND SHIELD COOLING PIPESINLET-OUTLET STRAINERSTEEL BALL SHIELDINGEND FITTINGSFEEDER PIPESMODERATOR OUTLETMODERATOR INLETHORIZONTAL FLUX DETECTOR UNITION CHAMBER
16.17.18.19.20.21.22.23.24.2S.26.27.2B.29.30.
EARTHQUAKE RESTRAINTCALANDRI A VAULT WALLMODERATOR EXPANSION TO HEAD TANKCURTAIN SHIELDING SLABSPRESSURE RELIEF PIPESRUPTURE DISCREACTIVITY CONTROL UNIT NOZZLESVIEWING PORTSHUTOFF UNITADJUSTER UNITCONTROL ABSORBER UNITZONE CONTROL UNITVERTICAL FLUX OETECTOR UNITLIQUID INJECTION SHUTDOWN NOZZLEBALL FILLING PIPE
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FIGURE 2.1-1 REACTOR ASSEMBLY
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r
. Cooling connections for the water within the end shields and calandriavault•
The transverse reactivity devices (Figure 2.1-2) are:
1) Six externally mounted ion chamber assemblies, 3 each side, whichsense low levels of neutron flux,
2) Seven Horizontal Flux Detector Assemblies which sense flux levels inregions of the reactor core,
3) Six Poison Injector Nozzle Assemblies which provide rapid injectionof neutron-absorbing gadolinium nitrate solution into the moderatorwhen trip sensors in Eihutdown System No. 2_ (SDS2) are actuated.
Vertical penetrations from the reactor deck (Figure 2.1-3) provide access,through thimbles, to the calandria, for positioning the other reactivitycontrol devices.
The vertical reactivity devices (Figure 2.1-4) are:
26 Vertical Flux Detector Units
These embody flux sensors which provide inputs to:
1) The Reactor Regulating System which controls power levels in variousregions of the reactor,
2) The Reactor Protective System to actuate Shutdown System No. J. (SDS1)in the event of excessive power indications.
3) The Flux Mapping System which is used to record local power levels toidentify zones where refuelling would be most appropriate.
6 Liquid Zone Control Units which provide a total of 14 compartments inwhich light water levels are varied in response to control requirements ofthe Reactor Regulating System.
21 Adjuster Units which serve the dual functions of:
1) Flux flattening during normal operation through neutron-absorption ininserted absorbers,
2) Xenon override following power reduction or shutdown, throughabsorber withdrawal.
28 Shut-off Units which drop neutron absorbing rods into the reactor coreto shut down the reactor when a trip is actuated in the Reactor ProtectiveSystem's SDSL
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I C L IC2,IC3 ON FAR SIDEONLY
(SIDE'C')
ICION CHAMBER HOUSING(3 EACH SIDE)
LIPOISON INJECTOR NOZZLES (6)
( • J HORIZONTAL FLUX DETECTOR (7)
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FIGURE 2.1.2 TRANSVERSELY MOUNTED REACTIVITY CONTROL DEVICES
""i
MOUNTING POSTSTEAM GENERATORROOM FLOOR LEVEL
TREADPLATE
CALANDRIAVAULT WALL'i
SHIMS
ADJUSTER UNIT.
FLUXDETECTORUNIT
MOUNTINGPOST
o
'' *• 1
r SHIELDING •COLLAR \ H
t • .
NITROGEN
VAULT WATER
SHUTOFF UNIT
HELIUM COVER VIEWING1 GAS LINES PORT
VERTICALFLUXDETECTOR
LIQUID ZONE UNIT CONTROL &CONTROL UNIT / POWER CABLES
1 CABLETRAYS
| T H I O K O L P - S T Y R O F O A M
SEALANT!
CALANDRIAVAULT
GROUT
-SEAL PLATE .
' DOUBLER PLATE
FIGURE 2.1.3 REACTIVITY MECHANISM DECK - SCHEMATIC SECTIONAL VIEW
O FUEL CHANNEL
O ®$®&® O
VERTICAL FLUX DETECTOR (26)
ADJUSTER (21)
SHUTOFF UNIT (28)
OVERFLOW
LIQUID ZONE CONTROL (6)
^ SOLID CONTROL ABSORBER (4)
© VIEW PORT (2)
« HELIUM BALANCE
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FIGURE 2.1-4 PLAN - VERTICALLY MOUNTED REACTIVITY CONTROL DEVICES
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These vertical reactivity control devices are positioned in the calandriain guide tubes which pass through the thimbles and the calandria tubelattice^ and are secured at the bottom of the calandria shell.
2.1.2 Fuel Channel Assemblies
The fuel channel assemblies (Figure 2.1-5) consist of:
1) Zirconium-niobium alloy Pressure Tubes (6.3 m long x 105 nun nominalbore x 4.16 mm minimum wall thickness) to house fuel and pressurizedD2O,
2) AISI type 403 Stainless Steel End Fittings, with type 410 stainlesssteel liners, to provide shielding extensions to pressure tubes,
3) Positioning assemblies for each end fitting, one of Which is locked.to locate its end fitting at the required position on its bearings inthe end shield,
4) Garter spring Tube Spacers to support each pressure tube within itscalandria tube,
5) Bellows Assemblies to seal the annular space between each pressuretube and its calandria tube (and between the end fittings and endshield lattice tubes),
6) Shield Plugs for every end fitting to minimize neutron leakage fromthe fuel channel and (in the case of the downstream shield plug) toprovide axial support to the column of 12 fuel bundles - (see Section2.6),
7) Removable Closure Plugs (Figure 2.1-6) to seal each end of the fuelchannels and to enable access for refuelling by the fuelling machine,
8) Feeder Connections to the Heat Transport System for the supply andremoval of D2O coolant for each fuel channel.
Fuel channels are installed at the reactor site by a highly trained crewin a closely controlled production operation.
Subsequently the feeders are installed to connect fuel channels to theHeat Transport circuit, described in the next section.
2.1-3
n
FIXED END OF CHANNEL
1 CHANNEL CLOSURE2 CLOSURE SEAL INSERT3 FEEDER COUPLING4 LINER TUBE5 END FITTING BODY6 END FITTING BEARING7 TUBE SPACER8 FUEL BUNDLE9 PRESSURE TUBE
10 CALANDRIA TUBE11 CALANDRIA SIDE TUBE SHEET12 END SHIELD LATTICE TUBE13 SHIELD PLUG14 END SHIELD SHIELDING BALLS15 FUELLING MACHINE SIDE TUBE SHEET16 CHANNEL ANNULUS BELLOWS17 CHANNEL POSITIONING ASSEMBLY
FIGURE 2.1-5 FUEL CHANNEL ASSEMBLY
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1 SECTION SHOWING THE JAWS AND SPIDERSPRINGS. THE RAM ASSEMBLY HAS JUST CON-TACTED THE REAR HOUSING, ADVANCING THESEAL DISC 0.9 mm.
3 HERE THE LATCH RAM HAS ADVANCED 12.7 mmTO UNLOCK THE SAFETY MECHANISM BYPUSHING THE FOUR SAFETY LATCHES INWARD.
1 FRONT HOUSING2 REAR HOUSING3 SPRING4 PLUNGER5 STEM END6 JAW7 TOGGLE8 CAP SCREW9 SEAL DISC PIN
10 SAFETY LATCH SPRING11 SAFETY LATCH12 SEAL DISC13 SPIDER14 STEM
SAFETY LATCH LOCKEDVIEW 2
SAFETY LATCH UNLOCKEDVIEW 3
2 SECTION SHOWING THE SAFETY MECHANISMAND THE CAP SCREWS. THE SAFETY LATCHESARE IN THEIR LOCKED POSITION PREVENTINGTHE ACCIDENTAL DEPRESSION OF THE STEM.
4 THE LATCH RAM AND 'C RAM HAVE BOTHMOVED A FURTHER 21 mm TO COMPLETELYRETRACT THE FOUR JAWS.
FIGURE 2.1-6 FUEL CHANNEL CLOSURE PLUG
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2.1.3 Summary g
In summary, the CANDU-600 MW(e) Reactor Core comprises an assombly of 380 _fuel channels in a calandria assembly which also includes vertical and Itransverse reactivity control devices. The latter are described in *greater detail in Sections 2.4 and 2.5.
12.2 HEAT TRANSPORT SYSTEMS
2.2.1 Fundamentals of the CANDU Nuclear Steam Supply System J
The CANDU reactor is contained within a low pressure tank called the -*calandria (Figure 2.2-1). The fuel channel assemblies run through the !calandria and contain the bundles of natural uranium fuel. The calandriais filled with heavy water (DjO) which moderates or slows the fastneutrons, making a chain reaction possible. The heat of fissicn generated Xwithin the fuel is removed by the pressurized heavy water coolant which is IIpumped through the fuel channels. This hot coolant is passed through thesteam generator where heat is transferred to light water to generate IBsteam. j|
The pressure tube forms the pressure boundary of the heat transport system <«.(Figure 2.2-2); the heavy water coolant passes through and around the Ibundles of natural uranium fuel located within the pressure tube. Thecalandria tube is in contact with the moderator. The annular spacebetween the pressure tube and the calandria tube provides thermal Iinsulation between the hot heat transport system coolant and the cool Imoderator.
IThe portions of the fuel channel assemblies external to the calandria |(Figure 2.2-3) are known as the end fittings; the end fittings haveconnections to the feeders which feed coolant into and out of the fuel •channels. I
The following sections provide further detail on the principle process
systems of the CANDU nuclear power system (Figure 2.2-4). 1
2.2.2 Heat Transport System
2.2.2.1 Arrangement !
The CANDU 600 MW(e) reactor has 380 fuel channels arranged in a square ~farray within the calandria. The heat transport system is arranged into {two circuits, one to each side of the vertical centre line of the reactor *~core, with 190 fuel channels in each circuit.
2.1-4
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Jc:
FUEL CHANNELASSEMBLIES
LIGHT WATER STEAM
LIGHT WATER CONDENSATE
HEAVY WATER COOLANT
HEAVY WATER MODERATOR
STEAM LINES
NATURAL URANIUM FUEL
MODERATOR HEAT EXCHANGER
FIGURE 2.2.1 CANDU NUCLEAR STEAM SUPPLY SYSTEM
,I;I;I;I;I|I|I;I;I;I;I;I;I;
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END SHIELD
TUBE SHEET
FIELD WELD
CALANDRIA TUBE
GAS ANNULUS(Between Lines)
SPACER FUEL BUNDLES
\
TPRESSURE TUBE
(Inner)
CALANDRIA
ASHIELD PLUG
FIGURE 2.2-3 REACTOR CORE SCHEMATIC
SAFETY VALVES ££
37170O0kgft 260"C _ _ _ ^ 4690KN;m2 IAI 2 S
TO SUPPRESSION TANKS
PRESSURIZER
310°C '9990 KN/m2(AI
Q n
266°C 27.U5O.OOO k3/!,
!I.Z30KN/m2(A(
MODERATORHEAT EXCHANGEB
187"c
.P. ITURBINE
2 MOlSTUilt SFPAHATOR HEHEAHRS
L.P TURBINECiENtRATCin
nOstfaFUELLINGMACHINE
HtAVY WATER MODERATOR
HEAVY WATER COOLANT
STEAM
CONDENSATE
H1VFH VVATLR
FIGURE 2.2-4 CANDU NUCLEAR POWER SYSTEM
*r4 r-r™-E»«
IIIIIIIIIII11II1i
r
The circuits are shown in Figure 2.2-5; each circuit contains 2 pumps, 2steam generators, 2 inlet headers and 2 outlet headers in a 'figure-of-eight' arrangement. Feeders connect the inlet and outlet of the fuelchannels to the inlet and outlet headers respectively.
The flow through the fuel channels is bidirectional (i.e. oppositedirections in adjacent channels). The fee<":>:s are sized such that thecoolant flow to each channel is proportional to channel power. Theenthalpy increase of the coolant is therefore the same for each fuelchannel assembly.
One of the advantages of this 'figure-of-eight' arrangement is that in theevent of a heat transport pump failure, the coolant flow in the circuit ismaintained at approximately 70% of the normal value, thereby permittingcontinued reactor operation at reduced power.
The arrangement of the heat transport system within the reactor buildingis illustrated in Figures 2.2-6 and 2.2-7. The steam generators, HTSpumps and headers are located above the reactor; this permits the heattransport 3ystem coolant to be drained to the header elevation formaintenance of the HTS pumps and.steam generators, and also facilitatesthermosyphoning (natural circulation) when the HTS pumps are unavailable.
2.2.?..?. Heat Transport System Conditions
Heavy water (D2O) is utilized as the reactor coolant; the principleadvantage of heavy water is its low neutron absorption. The variation ofsaturation temperature with pressure for D?O is shown in (Figure2.2-8). The heat transport system operating pressure is one of the keyelements in optimizing the CANDU cycle; high primary pressure permits highsecondary pressures and increased unit efficiency. They also, however,require thicker walled pressure tubes, and hence incur a burnup penalty.The operating pressure of the 600 MW(e) reactor (outlet header) is 10 MPa.In order to maximize unit efficiency, boiling in the core at high power isutilized, leading to an outlet header quality of up to approximately 4% atfull power. Figures 2.2-9 and 2.2-10 illustrate a typical steam generatorand its heat load diagram respectively. The feedwater enters thepreheater section of the steam generator secondary side, is warmed tosaturation temperature, and is then evaporated to produce steam. Theheavy water coolant enters the steam generator U-tubes opposite thepreheater section; the heavy water vapour is condensed, and the liquidD2O progressively cooled, until it leaves the U-bend near thepreheater section entrance. The small amount of vapour in the heavy watercoolant entering the steam generator increases the Log Mean TemperatureDifferences (LMTD) across the stem generator, and improves steamgenerator perfomance.
2.1-5
1
Lr OU1
:"\i i» ii I
uSTEAM GENERATOR
PUMF
rLET HEADER
T
-
OUTLET
y STEGEh
S—-s
AMERATOR
PUMP
INLET HEADER
"11
INLETHEADER
r-
LJ
LOOP
REACTOR
LOOP
— |
INLET HEADER
r1 I
L,
L|INLETHEADER
STEAM GENERATOR
1>UMP n
W1 \
OUTLET HEA[
>
m
)ER "
-
PUMP
GENEFIATOF
1 1
1 j1 11 1i i
OUTLET
FIGURE 2.2-5 A HEAT TRANSPORT SYSTEM
IIIIIIIIIIIIIIIII1i
1 STEAM GENERATORS2 HTS PUMPS3 REACTOR
FIGURE 2.2-6 LOCATION OF HEAT TRANSPORT SYSTEM EQUIPMENT
I
OUTLET HEADERINLET HEADERFEEDERSSTEAM GENERATORSEND FITTINGSHEAT TRANSPORT PUMPSINSULATION CABINET
1
11
1
FIGURE 2.2-7 FEEDER AND HEADER ARRANGEMENT
IIIIIIIIIIII1II11Ir
12
PRESSURE/TEMPERATURE — D2O
OPERATING POINTREACTOR OUTLET -HEADER
I
200
TEMPERATURE C O
300 400
FIGURE 248 VARIATION OF SATURATION PRESSURE WITH TEMPERATURE FOR HEAVY WATER
STEAM OUTLET
FEEDWATER INLET
STEAM SEPARATIONSCREEN
STEAM DRUM
HIGH CAPACITYCYCLONES
U-TUBES
SHROUD
TUBE SUPPORT PLATES
PHEHEATER SECTION
D2OOUTD2OIN
I11111I1II1IIIIII
FIGURE 2.2-9 TYPICAL STEAM GENERATOR
O
111Q_
266
PERCENTAGE HEAT TRANSFERRED
100
Data presented Is tor AECL 600 MW(e) reactor
FIGURE 2.2-10 STEAM GENERATOR HEAT LOAD DIAGRAM
IA simplified Heat Transport System flowsheet is shown in Figure 2.2-11. IHTS parameters are summarized in Figure 2.2-12. This figure also presents •data for Douglas Point, Pickering and Bruce/ and illustrates theevaluation of the CANDU system. 1
2.2.2c3 Heat Transport System Major Components -|
Each vertical centrifugal type HT pump (Figure 2.2-13) has a singlesuction and double discharge. The rotational inertia of the pump-motor _assemblies is sufficient to extend pump rundown so that coolant flow |matches the reactor power decrease following a loss of power to the pump •motors.
The pump 3eal package (Figure 2.2-14) consists of three carbon seals in Iseries. The seals are provided with a cool, clean flow of D2O duringnormal operation via the gland seal system (Figure 2.2-15). Cooling water ~lis also provided to the pump gland jacket. £
The steam generators (Figure 2.2-16) feature integral preheaters and steam r.drums. The heavy water coolant passes through the U-tube bundle. The Ifeedwater enters the preheater section of the steam generator, whichencompasses the lower portion of the cold leg of the tube bundle. The twophase light water flow rising from the U-tube region of the steam Fgenerator is passed through cyclone separators and secondary scrubbers to -tassure that the moisture content of steam leaving the steam generator isless than 0.25%. The liquid removed from the steam is returned to the Ttube sheet region of the steam generator via the annular downcomer. The Xcirculation ratio for CANDU steam generators is approximately 5 to 1.
Steam generators in caNDCJ nuclear steam supply system, have an excellent Ioperating history. The tube defect rate for CANDU systems is 0.001% peryear from 1971 to 1977, compared to a world average failure rate of 0.4%per year. I
Figure 2.2-17 shows a typical header: the nozzles on the header whichconnect to the feeders are cold drawn from the parent header material. T
iFigure 2.2-18 is a photograph of an actual installation and shows thefeeder and end fitting arrangement. -r
2.2.3 Pressure and Inventory Control System
The inventory of the Heat Transport System (HTS) (Figure 2.2-19) iscontrolled by 'feeding' D2O into, or 'bleeding' D2O out of, theHTS system. At power, the HTS pressure is controlled by a pressurizer \connected to the two HTS circuits. Heat is added to the pressurizer via Ielectric heaters to increase pressure and is removed via steam bleed to
12,1-6
STEAM FLOW(TYPICAL)
FEEDWATER FLOW(TYPICAL)
r
/
STEAM GENERATOR
FIGURE 2.2-11 HEAT TRANSPORT SYSTEM FLOWSHEET
REACTORDOUGLAS
POINT PICKERING BRUCE GENTILLY-2
NUMBER OF ELEMENTS PER BUNDLE 19 28 37 37
OPERATING CONDITIONS
COOLANTNOMINAL INLET PRESSURE MN/mz
PRESSURE DROP/CHANNEL (CRUD FREE) kNlmd
BUNDLES/CHANNELMAXIMUM CHANNEL POWER MWINLET TEMPERATURE °COUTLET TEMPERATURE °CEXIT STEAM QUALITY %MAX. MASS FLOW/CHANNEL kg/s
D2O9.8
738102.743
249293
12.6
D2O9.8
565125.125
249293
23.8
D2O9.3
738135.74
252/256*298.9
0/3.5*23.8
D2O11.09
758126.5
266.4312.3
2.923.94
* INNER ZONE/OUTER ZONE
FIGURE 2.2-12 HEAT TRANSPORT SYSTEM PARAMETERS
IIIIIIIIIIIIIIII1
1 UPPER OIL POT COVER2 THRUST BEARING OIL POT3 RUNNER4 THRUST BEARNG ASSEMBLY5 DOWN THRUST BEARING6 UP THRUST BEARING7 THRUST BEARING COOLING COILS8 BRAKE RING9 MOTOR SHAFT
10 OIL LEVEL CONTROL11 BEARING COOLING WATER PIPES12 AIR COOLER WATER PIPES13 SURGE CABINET14 AIR SHIELD15 AIR SHIELD16 BLOWER RINGS
17 MOTOR FLYWHEEL18 STATORCORE19 ROTOR ASSEMBLY20 LOWER GUIDE BEARING21 THRUST DISC22 SPACER COUPLING23 MOTOR STAND24 PUMP SHAFT25 VAPOUR CONTAINMENT SEAL26 SECONDARY MECHANICAL SEAL27 PRIMARY MECHANICAL SEAL28 PUMP BEARING29 PUMP CASE30 CASE WEAR RING31 PUMP DISCHARGE32 SUCTION PIPE
FIGURE 2.2-13 HEAT TRANSPORT SYSTEM PUMP
DRAIN TO LEAKAGECOLLECTION
SEGMENTEDCARBONBACK-UP SEAL
VENT TO LEAKAGECOLLECTION
A TO LEAKAGEA COLLECTION
RESTRICTIONBUSHING
PUMP SHAFTPUMP END
11111I1I1111
I1
FIGURE Z2-14 HEAT TRANSPORT PUMP GLAND SEAL
IIIIIIIIIIIIIILtII
PUMP3GLAND
D2OSAMPLING
PUMP4GLAND
CLASS 3
D2O' SAMPLING
D2O SAMPLING
DISCHARGEOF D2OFEED PUMPS
FIGURE 2.2-15 HEAT TRANSPORT SYSTEM PUMP GLAND SEAL COOLING SYSTEM
STEAM OUTLET NOZZLESECONDARY STEAM CYCLONESPRIMARY STEAM CYCLONESCHEMICAL FEED NOZZLE AND HEADERDOWNCOMER ANNULUSREHEATER DRAINS RETURN ANDEMERGENCY WATER SUPPLY NOZZLEU-BEND SUPPORTSTUBE BUNDLETUBE SUPPORT PLATEBACK-UP SUPPORTSOBSERVATION PORTBLOWDOWN NOZZLEDIVIDER PLATE
14 D2O INLET NOZZLE15 BASE SUPPORT16 D2O OUTLET NOZZLE
BAFFLE PLATE18 PREHEATER19 LATERAL SUPPORTS20 WATER LEVEL
CONTROL TAPSMANWAYFEEDWATER NOZZLE
11II 'I1I1111I11 'III
FIGURE 2.2-16 600 MW STEAM GENERATOR
FIGURE 2.2-17 TYPICAL HEADER
FIGURE 2.2-18 FEEDER END FITTING ARRANGEMENT
DEQASSER CONDENSERRELIEF VALVES (2)
PRESSURIZER RELIEF VALVES (2)
' PRESSURIZER STEAM BLEED VALVES (2)
FROM D2O STORAGE TANKAND DEGASSER CONDENSER
STEAM GENERATOR
STEAM GENERATOR
-CA>-^, T HEADERS I^ k INLET OUTLET A
h^ r
FIGURE 2.2-19 HEAT TRANSPORT PRESSURE AND INVENTORY CONTROL SYSTEM
2.1-7
1I
reduce pressure. The inventory control system can also provide pressure •control at low power (less that 5%) when the pressurizer may be isolated.The pressurizer also serves to limit the magnitude of HTS pressuretransients by receiving coolant from the heat transport system when Ipressure is increasing, and by supplying coolant to the heat transport •system when pressure is decreasing. Three typical heat transport systemtransients are shown in Figure 2.2-20. I
Valves that discharge D2O from the heat transport system (HT reliefvalves, pressurizer steam bleed valves and relief valves) connect to the i*degasser condenser (Figure 2.2-19). The relief devices of the degasser Ifcondenser are set above the normal HTS operative pressure, therebylimiting the discharge of D2O from the HTS in the event that any of _these discharge valves fail open. It
2.2.4 Shutdown Cooling System
The shutdown cooling system (Figure 2.2.-21) can be utilized to remove IJdecay power following a reactor shutdown. Two independent shutdowncooling system circuits are provided, one at each end of the reactor core. ••D2O is taken from the outlet header, passed through a pump and heat J|exchanger, and returned to the inlet header. Since there are no valves inthe heat transport system circuits, a portion of the shutdown cooling _system flow passes from the outlet header to the inlet header via the Isteam generators. The shutdown cooling system can also be operated *utilizing the heat transport system pumps; in this mode of operation, theshutdown cooling system flow bypasses the shutdown cooling system pimps. "1
This system is also effective with the heat transport system depressurizedand the D2O level lowered to the elevation of the headers; thisfacilitates maintenance of the steam generators and HTS pumps. 12.2.5 Heat Transport System Purification •
The accumulation of active materials in the CANDU heat transport system isinherently very low. This is primarily due to restrictions placed on "1
materials used in the HT system (for example, very low cobalt levels are 1permitted), and the absence of failed fuel during reactor operation (inthe event fuel failures do occur, they are detected and removed). ••
To further minimize the accumulation of active deposits within the HTsystem, the coolant is continuously filtered and purified. The head ofone heat transport system pump in each circuit is utilized to provide a Tflow of heat transport system coolant through the purification system -I(Figure 2.2-22). An intercooler is utilized to minimize heat losses.Flow through the filters and ion exchange columns is cold andpressurized. 1
1
z111a.
12
11
10
9
8
7
6
• I I I | i | | i i
, LOSS OF CLASS IV POWER
\ . 7 ^--^^\ ^ LOSS OF ONE PUMP
^ w STEPBACK TO 70% POWER
REACTOR TRIP ^ " ^ ^ s ^ ^ /FROM FULL POWER ^ % « % > > > ^ y
• I l l I I I I I I
1 1 1 1 1 1
-
-
-
"
-
i i i i i i12 16 20
TIME-SECONDS
24 28 32
FIGURE 2.2-20 HEAT TRANSPORT SYSTEM TRANSIENTS
STEAM GENERATOR
SHUTDOWNCOOLING HEATEXCHANGER
SHUTDOWNCOOLINGPUMP
I.STEAM GENERATOR
HEAT TRANSPORT SYSTEM
CIRCUIT 1
FEEDERS (TYPI
SHUTDOWN COOLING SYSTEMISOLATION VALVE (TYPICAL)
FIGURE 2.2-21 SHUTDOWN COOLING SYSTEM
STEAM GENERATOR STEAM GENERATOR
HEAT TRANSPORT SYSTEM
FEEOERS (TYP)
REACTOR
COOLER
FILTER
ION EXCHANGE
FIGURE 2.2-22 HEAT TRANSPORT PURIFICATION SYSTEM
2.3 OVERALL PLANT CONTROL
2.3.1 Introduction
I1III
The term "Overall Plant Control" is used to describe the controls thatcoordinate the turbine-generator output with the power output of thereactor. These controls are described in this section, but the sectionalso describes other features of the controls and instrumentation whichare unique to CANDU nuclear power plants. The CANDU plants use directdigital control for all major control functions. The control computers -rare a highly reliable dual computer system. A second feature of CANDU |control is the advanced control room concept which uses computer drivendisplays, alarms, messages, and logs to replace much of the moreconventional instrumentation used in other plants. In addition to the fmain control room, CANDU plants incorporate a Secondary Control Area from -which important variables can be monitored and controlled and from wherethe plant can be shutdown. The Secondary Control Area and its equipment ~jis seisfflically qualified and protected against other external events. {
2.3.2 Main Control Centre I
All CANDU reactors built since the mid 1960's have used centralized directdigital computer controls. However the control room designs of the Tearlier plants were quite conventional. In more recent plants the control Lroom design has been modified to make better use of the more flexibledisplay and message capability. Pickering 'A' which came in service inthe early 1970's uses conventional instruments supplemented by a fewcomputer-driven CRT's (Cathode Ray Tubes) to display messages and alarms.Bruce 'A', which was about 5 years later in design, replaces much of the •»conventional instrumentation with monochrome CRT's capable of displaying Iplant information in a variety of formats - graphs, bar charts, printedmessages, etc.. Plants designed since Bruce 'A' have carried this conceptfurther by using colour CRT's in place of the monochrome CRT's (see Figure I2.3-1). I
Most of the wiring between the plant and the control room equipment T(including the computer) is routed through a Control Distribution Frame [(CDF). This arrangement gives greater flexibility and allows fieldterminations of wiring to be completed independently of connection to the -..control panels and computers. !
2.3.3 Electrical Powerf-
Power in the plant is divided into four classes depending on the .1requiranents of the equipment being supplied. The power supply buses arefurther subdivided and separated to help meet equipment reliability 7
2.1-8
I
CONTAINMENTEMERGENCYCOKECOOLING
FUELLING MACHINEA M FUEL HANDLINGCONTROL CONNIE
SHUTDOWNSYSTEM
MODERATOR AMDREACTOR MISCELLANEOUSSYSTEMS
PRIMARYHEATTRANSPORT SYSTEM
MISCELLANEOUSAUXILIARY
f - . „
\
i „ , , \ SHUTDOWN \\ ""v \ SYSTEM \
\ \ \ N°-1 » \PL fL3 PL4 P
\ \ \ \ \
REACTOR\ REGULATING
SYSTEMI
PL4 PL5 PLB
CONTROLCOMPUTERS
PL7
AMGENERATORSYSTEM
H.V) PL11 PH2
E GENERATOR
PLI3
ELECTRICALDISTRIKITION
ITCHYARD SYSTEMSWITCHYARD SYSTEMS I 1
I /I l \TIT I I T I
LINE PRINTERS V V
FIGURE 2.3.1 TYPICAL ARRANGEMENT OF CRT DISPUYS ON MAIN CONTROL PANELS
11
requirements. Most of the control equipment is supplied by Class II Jpower. This is uninterruptable power and separate buses are used toredundant instrumentation. _
12.3.4 Dual Computer System
•High reliability of control functions is assured through the use of two •identical independent digital computers (DCCX and DCCY). Each computer iscapable of complete station control and can transfer control automatically mto the other computer on detection of a fault. Faults in either software IIor hardware are detected by a combination of internal hardware andsoftware self checking facilities plus an external "watchdog timer" or _,operations monitor. Fault detection may result in automatic reloading of IIcore memory from the disc and computer restart, or transfer of control to '*the other computer. The computers also verify the incoming data usingredundant information and rationality checks. Messages to the operator [Iidentify out-of-range data. Both computers are normally running but the IIoutputs to the plant are only connected to one computer. Switching ofoutputs is automatic when required. •
2.3.5 Plant Controls
In a nuclear power plant there are a large number of variables to becontrolled. Examples are:
. Moderator temperature
. Deaerator level
. Heat transport system pressure
. Pressurizer level
. Reactor power
. Steam generator pressure
. Steam generator level
In order to control the plant electrical output to the desired value,these variables and others must be controlled in a co-ordinated way. Some _of the variables can be controlled quite independently, and in some cases fjhave their own controllers. Most of the variables are controlled by ^interacting programs in the plant computers. In controlling theturbine-generator output, it is also necessary to control the reactor fTpower. Steam generator pressure is also closely related to this control [£
2.1-9 ff
I111I
IIIIIIIIIIII
I1
problem as is the steam generator level. These three control loops sharecommon filaments and can be considered together. Within the logic diagram,Figure 2.3-2, can be seen separate control loops for reactor power, steamgenerator pressure, and steam generator level, all controlled from themain computer system. The overall plant control scheme is similar to thatused in non-nuclear plants. It operates in two modes:
a) Normal Mode (Reactor follows plant loads)
The turbine generator load is set by the operator, and the turbinegovernor valves open to supply the necessary steam. The steam generatorpressure control program senses pressure changes due to governor valvemotion and requests variations in reactor power to maintain drum pressureconstant. For example a frequency drop due to increased grid loadingwould cause the turbine governor valves to open further. The resultingdrop in steam drum pressure would cause the pressure control program torequest an increase in reactor power. This would occur unless there werelimits on reactor power available. The reactor control system isdiscussed separately in Section 2.4.
b) Alternate Mode (Turbine follows reactor)
Reactor power is controlled to a setpoint supplied by an operator. Thesteam generator pressure control program manipulates plant loads to keepsteam drum pressure constant. This mode is used:
1) At low power when the steam drum pressure is insensitive to reactorpower;
2) During upset conditions where it may not be desirable to maneuverreactor power.
In addition to adjusting the turbine load to accept the steam output fromthe reactor, the pressure control program also has access to steamdischarge valves so that excess steam can be dumped directly to thecondenser or to atmosphere. The Condenser Steam Discharge Valves (CSDV)typically can carry 70% or more of the steam production if necessary whilethe Atmospheric Steam Discharge Valves (ASDV) are limited to about 10%.If the turbine becomes temporarily unavailable, the reactor can continueto operate by dumping most of its steam to the condenser. The atmosphericsteam discharge valves are used during startup when the condenser may beunavailable and temporarily during other transients.
n 2.1-10
ATMOSPHERICSTEAM DISCHARGE
GENERATOR
r
REACTOR
FLUX POWERAND RATE
REACTORPOWER
REACTORFLUX
CONTROL
I
STEAM
GENERATORS
STEAMGENERATOR
PRESSURECONTROL
REACTORPOWER
SETPOINT TURBINE LOAD CONTROLOR UNIT POWER
REGULATOR (UPR)
DEMANDED REACTOR POWER(ALTERNATE MODE ONLY) ELECTRICAL OUTPUT
SETPOINT(NORMAL MODE)
FIGURE 2.3-2 OVERALL PLANT CONTROL — BLOCK DIAGRAM
i.. •..Jin}
IIIIIIIIIIIII1t11
2.4 CORE CONTROL
2.4.1 Introduction
Reactor control was mentioned as one of the sub-loops in the Overall PlantControl System. The reactor control system is one of the most importantcontrol systems. It combines hierarchies of measurement instruments andcontrol devices with complex computer logic to meet a number ofrequi rements.
The system is required to:
1) Monitor and control reactor power to satisfy station load demands.
2) Monitor and control the three-dimensional power distibution in thereactor so that individual fuel bundles and fuel channels operate atpowers within their design specifications.
3) Monitor important plant variables and reduce reactor power atappropriate rates to keep the variables within specified limits.
The reactor control system can be more easily understood by examining themeasuring instruments, the reactivity control devices, the logic that 'relates the device operation to the measurements, the requirements of thesystem, and the disturbances it is subjected to.
2.4.2 Reactor Power Measurement
Power from the reactor is measured with combinations of:
1) Startup counters
2) Ion chambers
3) Self-powered in-core flux detectors
4) Thermal power measurements.
2.4.2.1
Startup counters are used only during the first criticality or forstarting after a very long shutdown. They are used along with manualcontrols to raise power above a range (7 decades below full power) wherethe ion chambers give useful readings to the computers. Following highpower operation, heavy water reactors retain a source term which keeps theion chambers on scale even after extensive shutdowns. Startup counters
2.1-11
11
are therefore not normally required and are removed after startup. I
2.4.2.2
Three ion chambers mounted in the side of the reactor give neutron fluxmeasurements in the range from 10"' to 1.5 times full power. Thesignals are provided to the computers through logarithmic amplifiers. -mShielding in the ion chamber housings provides good discrimination against Igamma rays. The signal response is essentially prompt except at thelowest powers. Automatic startup and shutdown of the reactor can be —accomplished over the full range of ion chamber signals. Figures 2.4-1 Iand 2.4-2 show the locations of ion chambers and some of the other reactor *control devices.
12.4.2.3
Self-powered in-core flux detectors are generally used above a few percent Iof full power for flux measurements. Unlike ion chambersf they can giveinformation about the spatial distribution of neutron flux and their _response is essentially unaffected by dissolved poisons in the moderator. IThe prime source of flux measurements is 28 prompt responding detectors at ~14 locations in the core. These detectors use platinum, inconel, orplatinum coated inconel emitters. While their response is partially "ineutron sensitive, they also have some gamma sensitivity. The information .1from these detectors is supplemented by 102 additional vanadium emittertype detectors. Their response is entirely due to neutrons but is not "Jprompt. The detectors are usually coiled onto vertical assemblies. More |recent designs have used shorter straight detectors which are individuallyinserted into well-tubes. In both designs, detectors are strategically _located throughout the core. Flux detectors (and ion chambers) are also Iused to provide signals to the special safety systems - shutdown system. •However entirely separate instruments are used for those systems.
I2.4.2.4
The neutron flux measurements are calibrated against measurements of Ireactor total thermal power. At high power this comes from redundantmeasurements of steam flow, steam pressure, steam temperature, feedwater «_flow and feedwater temperature. At low power the steam flow measurement Iis not sufficiently precise, and reactor power is calculated from -1temperature rise measurements across the reactor.
II
2.1-12
IIIIIIIIIIIIIII1
CALANDRIA "*"" 16CALANDRIA- SIDE TUBESHEET 17CALANDRIA TUBES 18EMBEDMENT RING 19
5 FUELLING MACHINE-SIDE TUBESHEET 20G END SHIELD LATTICE TUBES 217 END SHIELD COOLING PIPES 22B INLET-OUTLET STRAINER 239 STEEL BALL SHIELDING 24
10 END FITTINGS 2511 FEEDER PIPES 2612 MODERATOR OUTLET 2713 MODERATOR INLET 2814 HORIZONTAL FLUX DETECTOR UNIT 2915 ION CHAMBER 30
EARTHQUAKE RESTRAINTCALANDRIA VAULT WALLMODERATOR EXPANSION TO HEADTANKCURTAIN SHIELDING.SLABSPRESSURE RELIEF PIPESRUPTURE DISCREACTIVITY CONTROL UNIT NOZZLESVIEWING PORTSHUTOFF UNITADJUSTER UNITCONTROL ABSORBE.fi UNITZONE CONTROL UNITVERTICAL FLUX DETECTOR UNITLIQUID INJECTION SHUTDOWN NOZZLEBALL FILLING PIPE
FIGURE 2.4.1 REACTOR ASSEMBLY
•ZONE CONTROLABSORBERS
ZONE CONTROLDETECTORS
1 2 3 4 5 6 7 B 9 10 11 112 13 14 15 16 17 18 19 20 21
VERTICAL FLUX DETECTOR (26)
ADJUSTER (21)
SHUTOFF ROD (28)
SOLID CONTROL ABSORBER (4)
LIQUID ZONE CONTROLLER (6)
HORIZONTAL FLUXDETECTOR (7)
VIEW OF REACTOR FACE
FIGURE 2.4-2 REACTIVITY MECHANISM LAYOUT
IIIIIIIIIIIIIIIEII!
IIIIIIIIIIIIIIIII
2.4.3 Control Devices
The regulating system (Figure 2.4-3) controls the neutron flux in thereactor (level and distribution) by adjusting a hierarchy of devices -light water control absorbers, mechanical control absorbers/ adjusters,moderator poison, and fuel.
2.4.3.1
There are 14 light water zone control absorber compartments distributedthroughout the reactor. These compartments are partially filled withlight water - a neutron absorber in a heavy water reactor. There is aconstant outflow of water from the compartments and a controlled inflowwhich allows the computer to raise or lower water levels in unison ordifferentially. These absorbers are the primary reactivity controldevices used for both bulk and spatial control of neutron flux in thereactor. A compartment can be completely emptied or filled in a minimumtime of one minute giving a reactivity change per compartment of
0.5 mk (i.e. ̂ = 0.0005).
2.4.3.2
The four mechanical control absorbers are normally out of the core but canbe driven in at variable speeds or dropped to supplement the negativereactivity from the light water absorbers. They are mechanically the sameas shutoff rods but are functionally and physically separate and under thecontrol of the computer system. They provide up to 6 mk of negativereactivity when inserted for a power reduction.
2.4.3.3
The 21 adjuster rods are normally fully inserted in interlattice positionswhere they contribute to flux flattening. They are not used for dynamiccontrol of the flux distribution but may be withdrawn in symmetrical banksand at variable speeds to provide additional reactivity. This would berequired to compensate for xenon following a large power reduction orshutdown followed by a restart. A total reactivity worth of 15 mk is •available in the adjusters and each bank can be withdrawn in a minimumtime of 1 minute.
2.1-13
INPUTSLOW BOILER
LEVEL
HIGH POWERERROR
HIGH HTPRESSURE
HT PUMPFAILURE
REACTOR TRIP
TURBINE TRIP
LOSS OF LINE
LOSS OF STATOR
COOLING
OPERATORDEMAND
BOILERPRESSURE
CONTROLLERM
ZONECONTROLLER
FAILURE
FLUX TILTPROBLEMS
HIGH LOCALFLUX
HIGH BOILERPRESSURE
LOW BOILERLEVEL
HIGH SURGETANK LEVEL
MANUAL
VANADIUMIN CORE
DETECTORPOWER
ION CHAMBERPOWER
INCOREDETECTOR
POWERBOILER
SECONDARYSIDE
MEASUREMENTS
OPERATORDISPLAY
SELECTOR
j OIGITAL COMPUTER CONTROLLER
STEP-BACKROUTINE
DEMAND POWERROUTINE
j
SET BACK
I \
FLUXMAPPINGROUTINE
: - . . • ' : • • ( f:
POWERMEASUREMENT
ANDCALIBRATION
*
REGULATINGPROGRAM
REACTIVITYCONTROL
fDISPLAY
PROGRAM
. T
OO
M
AN
Y
AD
JUS
TE
RD
RIV
ING
O
UT
• T
RIP
SY
ST
EM
S S
ET
D
TTt t
ADJUSTERDRIVE
INTERLOCKS
TRIP SYSTEMS SET
tCONTROL
ABSORBERDRIVE
INTERLOCKS
EVICES
4CONTROLABSORBERSCLUTCHES
ONLY
21 ADJUSTERS
VARIABLESPEED DRIVE
14 ZONECONTROLVALVES
28 SHUT-OFFRODS
OUT DRIVE
4 CONTROLABSORBERSVARIABLE
SPEED DRIVE
MODERATORPOISON
ADDITION
ALARMS
C.R.T.DISPLAYS
FIGURE £4.3 REACTOR REGULATING SYSTEM BLOCK DIAGRAM
IIIIIIIIIIIIIIIIII!7
IIIIIIIIIIIIIIIIr
2.4.3.4
Boron or gadolinium salts can be dissolved in, or removed from, themoderator as an additional reactivity shin. This is normally an operatorfunction, but under exceptional circumstances the control computer canintervene, "fine reactivity effects are approximately 9 mk/ppm for boronand 32 mk/ppm for gadolinum. Only very small concentrations are neededand the addition and removal rates are slow. Gadolinum burns out at a ratesimilar to the buildup rate of xenon.
2.4.4 Control Logic
In each of the 14 zones of the reactor the flux is sensed by a pair ofprompt responding in-core flux detectors. Their response is calibratedagainst a flux map derived from the slower vanadium-emitter type of fluxdetectors. The flux map is normalized against thermal measurements. Asetpoint for flux in each zone is calculated by the computer from theoverall plant control requirements or operator demands, and a zone levelsetpoint is derived. The valve lift on the input valve to each zone ofthe light water control absorbers is then varied dynamically to drive theflux and level errors towards zero. If the average power error in thereactor gets too large or if the average zone level approaches its upperor lower limits, the logic then drives adjusters or mechanical controlabsorbers to supplement the range of the light water control absorbers.The logic of this procedure is illustrated in Figure 2.4-4. During normalfull power operation the adjusters would remain inserted and themechanical control absorbers withdrawn. Their positions would only changeduring transients such as power level changes.
2.4.5 Di sturbances
Changes in the net reactivity of the core occur because of fuel burn-up,new fuel addition, or because the power output of the reactor is changed.These changes are amplified both locally and generally in the reactor bythe effects of xenon poison. Xenon-135 is a neutron absorbing fissionproduct which decays naturally with a time constant of several hours.Because of the nature of the xenon production through the decay ofIodine-135, there will be an inital increase in xenon when power isreduced, although it will eventually return to a slightly lowerequilibrium level. Similarly there will be an initial decrease in xenonwhen power is increased. This is both a local and a general effect. Localchanges in power will cause local changes in xenon. In the absence ofspatial control, spatial xenon induced flux oscillations would also bepossible. The reactor control system must compensate for disturbancescaused by fuel changes, xenon effects accompanying these changes, andchanges in operating power.
2.1-14
(a ) 100%
( b )
-1 0 +1
POWER ERROR %
100%
+5
uiiii
ll4 Rods
Drive Absorbers OUT
2 Rods
Drive Absorbers IN
70%
2 Rods 4 Rods
O UJ
5£
-1 0 +1
POWER ERROR %
•-100%
50%
+2 +4- I »-Ep
+5
-1 0 +1
POWER ERROR %
+2 +3 +4—I *-Ep
+5
Idealized zone reaclivity rate vs. power error
FIGURE 2.4.4 REACTIVITY LIMIT CONTROL DIAGRAM
1IIIIIIIIIIIIIIIII
IIIIIIIIIIIIIILI
For small or slow changes in power the xenon effects are small and easilycontrolled. For larger changes the xenon effects impose some broad limitson power manuevering. These are discussed further in Section 2.7.
2.4.6 Stepbacks and Setbacks
The reactor control system in the plant computers includes routines whichmonitor a number of plant variables for operation within acceptablelimits. Reactor power is promptly reduced if the variables exceed theselimits. These systems are entirely separate from the shutdown systcand serve to reduce the frequency of operation of the special safety(shutdown) syste
If a very fast power reduction is needed the computer initiates a"stepback". It opens clutches on the four mechancial control absorbers,allowing them to drop into the reactor, making it subcritical. The rodscan be "caught" part way in by re-closing the clutches. The reactor isthen critical at a lower power. If the rods are not caught, the stepbackwill take the reactor towards zero power.
For power reductions which are not as urgent, a "setback" occurs ratherthan a stepback. The reactor power setpoint in the computers is takendownwards at a controlled rate. The power follows through normal actionof the light water control absorbers and the related devices. The setbackends either when the variable causing it returns to limits, or when apredetermined power level is reached. The rate at which reactor power isreduced and the power level at which the setback ends may be different foreach variable.
Setbacks and stepbacks override other power demands and are accompanied byannunciation defining the out of limits variables.
2.5 REACTIVITY CONTROL DEVICES
2.5.1 Liquid Zone Control System
The liquid zone control absorbers are the primary devices for controllingreactivity within the reactor during normal operation: Reactivity isadjusted by varying the quantity of light water, which acts as a poison,in each of the zone control compartments (Figure 2.5-1). Fourteen zonecontrol compartments are contained within 6 assemblies. The location ofeach of the zone control assemblies is shown in the reactivity mechanismlayout. Figure 2.4-2. Details of the liquid zone control assembly areprovided in Figure 2.5-2.
The zone control system is illustrated in Figure 2.5-3. Light water iscirculated through the zone control compartments; level in thecompartments is measured by a helium 'bubbler' system, and is controlled
2.1-15
THREE COMPARTMENTS INCENTRAL ASSEMBLIES
TWO COMPARTMENTSIN OUTER ASSEMBLIES ZONE CONTROL ASSEMBLY
LIGHT WATER (H2O)
MODERATOR
CALANDRIA
FIGURE 2.5-1 LIQUID ZONE CONTROL SYSTEM ARRANQEMENT
IIIIIIIIIIIIIIII
BELLOWS
PENETRATION TUBE
BEARING
CONCRETE
THIMBLE TUBE
SPLIT SEAL
RING
CALANDRIA SHELL
ZONE1
ZONE 2
NOZZLE
LOCATOR
CALANDRIA SHELL
1 WATER2 HELIUM3 NUT4 CRUSH WASHER5 TERMINAL BLOCK6 SHIELD PLUG7 ZONE CONTROL TUBE8 WATER AND HELIUM TUBES 18 KEY9 TUBE SUPPORT 19 SPRING
10 TUBE SPRING 20 LOCATOR THREAD
11 WATER INLET12 BULKHEAD13 BAFFLE14 HELIUM OUTLET15 HELIUM INLET16 WATER OUTLET17 HELIUM BALANCE LINE
FIGURE 2.5-2 ZONE CONTROL UNIT
TO OTHERCOMPARTMENTS
GAS BALANCE HEADER
LEVELTRANSMITTER
TYPICALARRANGEMENTOF ZONE CONTROLCOMPARTMENT(14 IN ALL)
RECOMBINATIONUNIT l-l
HELIUMBOTTLESFOR GASMAKE UP
HELIUM GASCOMPRESSOR „
FROM OTHERCOMPARTMENTS
H2O CIRCULATING PUMP (3)
TO OTHERCOMPARTMENTS
H2OSUPPLYHEADER
IIIIIIIIIIIII1I1II
FIGURE 2.5-3 LIQUID ZONE CONTROL SYSTEM
IIIIIIIIIIIri
iir
via control valves on the water inlets, based on a signal from the stationcomputer. The water is forced out of the compartments at a constant rateby the helium cover gas pressure.
2.5.2 Vertical Flux Detector Units
Like the Zone Control Unit already described, the Vertical Flux Detectorsare mounted beneath the Reactivity Mechanism Deck surface. However, theymay be reached through access plugs in the Deck Plate (Figure 2.5-4).
Because of their long and slender design, and of their construction oflow-modulus Zircaloy 2 material, these Flux Detector Units are installedin 27.5 mm guide tubes, which are tensioned to enhance their rigidity.
Each Unit consists of a carrier and capsule tube assembly with deteccors,connectors and seal components.
The detectors themselves are self-powered Hilborn elements oi vanadium orplatinum construction, in which emitter wires are separated from theirsheathing by mineral oxide insulation.
These detectors provide signals which are directly proportional to fissionrates in the reactor, with varying response and sensitivity according totheir material or exposure duration.
2.5.3 adjuster Units (Figure 2.5-5)
The requirements for normal insertion and controlled removal of theabsorber elements, for Xenon override, demand mechanical features foehoisting and lowering the elements in response to the regulating systemrequirements.
Adjust Drive Mechanisms mounted above the reactivity deck embody a motordriven sheave on which absorber cables are wound or unwound for absorberraising and lowering.
The absorbers consist of stainless steel tubing (of 76 mm diameter) with ashaped central shim rod.
The adjusters are arranged in seven banks, whose collective withdrawalwould increase reactivity by 15 mk. This increment would provide 30minutes Xenon override time after shutdown from steady full poweroperation.
Figure 2.5-6 illustrates the positioning of the vertical reactivitycontrol devices under normal and actuated conditions.
2.1-16
TREAD PLATE -
GLASSINSULATION
COVER GASCONNECTION -eg
DETECTOR ASSEMBLY
CALANDRIA NOZZLE
CALANDRIA SHELL -
CALANDRIA TUBES
GUIDE TUBE
GUIDE TUBE LOCATOR
CALANDRIA SHELL
DETECTOR COIL
FIGURE 2 . M VERTICAL FLUX DETECTOR UNIT
IIIIIIIIIIIIIIsiIIB
I11II1
DRIVE MECHANISM
SHEAVE
INSERT
GUIDE TUBEEXTENSION
H2O
FIGURE 2.5-5 ADJUSTER UNIT
SOLID CONTROL &SHUT-OFF DRIVE
HORIZONTAL FLUX DETECTORSLIQUID INJECTORSION CHAMBERS 1.2 4 3SIDED
OPEN FOR SERVICES
1I11III1]
1111IIIT
FIGURE 2.5.6 SCHEMATIC SECTION SHOWING POSITIONS OFREACTIVITY CONTROL DEVICES
IIIIIIIIIIIIIItII
2.5.4 Mechanical Control Absorber Units and Shut-Off Rods
Mechanical Control Absorber Units and Shut-off Rods (Figure 2.5-7) havevirtually identical design and capability, but have different missions,(as outlined in Section 2.1.1) in reactor control and shutdown functions.
Hence, the Control Absorbers form part of the Regulating System, whereasshut-off rods are part of the protective Shutdown System No. 1.
Both units embody a clutch, cable and winch arrangement to hoist, hold orrelease cadmium filled absorber tubes, of 113 mm diameter and of varyinglength.
Rod Ready Indicators sense the presence of the Absorbers in the poised(fully raised) position.
Guide tube perforations minimize hydraulic resistance to absorbers' entryinto the calandria.
Spring-assistance provides acceleration in response to a shutoff rodclutch de-energization. Normal time for full insertion is about 1.6seconds.
Rotary hydraulic dampers, geared to the winch, stop the absorbers'movement in the calandria at the appropriate level.
2.5.5 Horizontal Flux Detector Units (Figure 2.5-8)
The construction of these assemblies is similar to that of the verticalflux detector units. However, installation details are quite differentdue to their horizontal configuration, to requirements for sealing andsupport, and to isolating provisions for replacement, by freezing.
As noted in 2.1.1 these assemblies form part of the protective jShut DownJSystem No. 2_ (SDS2).
2.5.6 Liquid Injection Shutdown Units
SDS2 provides for rapid injection of gadolinium nitrate solution, byhelium displacement, from an external poison tank into the calandria viainjection nozzles, into the moderator.
Figure 2.5-9 shows the installation, which includes similar features forsealing, support and replacement as for the horizontal flux detectors.
2.1-17
DRIVE MECHANISMCABLE SHEAVE
IDLER PULLEY •>
CABLE-
SHUTOFF HOD
SPRING TO REWINDDAMPER
DOG PLATES — LOST MOTION LINK BETWEEN MAINAND DAMPER SHAFTS - ALSO ENGAGE POSITION STOPS
MAIN DRIVE MOTOR ANDSHAFT . ^ WORM REDUCTION GEAR
ELECTRO MAGNETIC CLUTCHBETWEEN IDLER AND MAIN SHAFTS
POTENTIOMETERPOSITION INDICATOR
BEVEL REDUCTION GEARAND SPUR REDUCTION GEARON IDLER SHAFT
SCHEMATIC SHUTOFF ROD DRIVE MECHANISM
GUIDE TUBE EXTENSION
WATER SHIELD
ACCELERATOR SPRING(ON SHUTOFF ROD ONLY)
CALANDRIA NOZZLE
SPIDER ATTACHINGSUPPORT ROD TOSHUTOFF ROD
GUIDE TUBETENSIONINGSPRING
CALANDRIA SHELL
SHUTOFF ROD
IIIIIIIIIIIIIIIIII1
FIGURE 2.5-7 SHUTOFF AND MECHANICAL CONTROL ABSORBER UNITS
SHIELD SLEEVE
MOUNTING BOSS
SEALSEAL CLAMP
ADAPTER PLATE
PRESSURE GAUGEDETECTORASSEMBLY
THIMBLE SUPPORT
SEAL CLAMPI FREEZING DoO CONNECTION „„„..,,-! JACKET THMBLE cwlf
ASSEMBLY ° ° V f c H
DETECTORCABLE
GUIDE TUBE
FIGURE 2.5-8 HORIZONTAL FLUX DETECTOR UNIT
THIMBLE
THIMBLESUPPORT
SHIELDING WALL
PROTECTIVE COVER
INJECTION NOZZLE
CALANORIA
FIGURE 2.5-9 LIQUID INJECTION SHUTDOWN UNIT
IIIIIIIIIIIIIr
lr
2.5.7 Ion Chamber Assemblies
Six ion chamber units are installed in housings mounted externally on thecalandria shell, in stainless steel housings (Figure 2.5-10). Leadshielding provides attenuation of gamma radiation for enhanceddiscrimination of neutron flux levels.
Access tubes enable replacement of defective ion chambers on power, withsuitable suplementary shielding.
Physical separation of the two groups of ion chambers reduces theprobability of local accidental damage to more than one group.
The broad sensitivity of these units is exploited by use of logarithmicamplifiers and applied in monitoring lower power levels and rate of changein power level.
2.6 FUEL
2.6.1 Description of Fuel
The 600 MW(e) reactor fuel bundle comprises seven component parts (Figure2.6-1). The elements contain high-density natural UO2 in a thinZircaloy-4 cylindrical sheath. A thin graphite layer (CANLUB) on theinside surface of the sheath reduces the pellet/sheath interaction. Endcaps, resistance welded to the sheath extremities, serve a triple purpose:(i) to provide a seal for the contents of the element, (ii) to provideeffective element termination for attachment to end plates, and (iii) toprovide the structural component for interfacing with the fuel handlingsystem. Thirty-seven elements are held in a close-packed bundleconfiguration by welding them to end plates. The desired separations atthe transverse mid-place of the bundle are maintained by split spacersbrazed to the elements. Bundle separation from the pressure tube isensured by bearing pads brazed near the ends and at the middle of theouter elements. The filler metal used for brazing is beryllium.
2.6.2 Design Basis for the Fuel
2.6.2.1 Fuel Elements
The reactor nuclear design requires high neutron economy. The fuelelement in therefore designed for maximum content of fissile material andminimum content of neutron absorbing material.
2.1-18
CALANDRIA SHELL .
SHIELDING SLEEVES
BEARING
CARBON STEELWATER SHIELD LINER
FREEZING COIL
BELLOWS
ION CHAMBERPENETRATION TUBE (3)
FIGURE 2.5-10 TYPICAL ION CHAMBER ARRANGEMENT
END VIEW INSIDEPRESSURE TUBE
ZIRCALOY BEARING PADSZIRCALOY FUEL SHEATHZIRCALOY END CAPZIRCALOY END SUPPORT PLATEURANIUM DIOXIDE PELLETSCANLUB GRAPHITE INTERLAYERINTER ELEMENT SPACERSPRESSURE TUBE
FIGURE 2.6-1 37-ELEMENT FUEL BUNDLE
II
The fuel is designed to operate within the power and turnup conditions Iapplicable to normal station operation, defined by the nominal bundle 'power envelope (curve A in Figure 2.6-2). Since a limited number ofbundles can exceed this power envelope, the fuel is assessed for operation Iwithin the reference overpower envelope (curve B in Figure 2.6-2). •
I2.6.3 Fuel Performance
2.6.3.1 General _
CANDU fuel performance has been demonstrated by wa n s of out-reactor *
tests, irradiation testing in experimental reactors, and successfulutilization in CANDU power reactors. Nevertheless, performance is Mcontinually being improved as a result of an ongoing fuel development |program. Some principal aspects of fuel testing and performance arediscussed below. I2.6.3.2 Out-Reactor Tests _
Out-reactor tests are conducted on prototype and initial production fuel ™bundles, to ensure the ccmpatiblity of the fuel design with therequirements stated above. Examples are endurance tests, impact tests and Ipressure drop measurements. I
2.6.3.3 In-Reactor Performance •
Irradiation testing is mainly performed in test loops at AECL'sexperimental reactors. Over the years a significant 1experience on CANDU fuel behaviour has been built up.experimental reactors. Over the years a significant bank of data and •
The in-service performance of over 200,000 CANDU fuel bundles irradiated Ito date has been excellent. Only 0.18% of al l fuel bundles irradiated Ihave been defective, as on September, 1979. The introduction of CANLUBfuel in the fuel design resulted in a marked improvement of performance. mOf the approximately 140,000 CANLUB fuel bundles irradiated, only 0*07% Iwere defective and none of the defects were due to power boosting, butwere rather latent manufacturing defects, see Figure 2.6-3. It should be _noted that usually only one defective element i s found in a defective nbundle. Therefore, if the defect statistics are reported on the basis of **defective elements, as i s customary for other reactor vendors, the defectstatistics become 0.006% defective elements (all fuel) and 0.002% Ifdefective elements (CANLUB fuel). If
2.1-19
Is
111
i
1000
900 ^
800
700
600
500
A NOMPOW
B REFERENC
NAL DESIGNER ENVELOPE
E OVERPOWER ENVELOPE
100 200 300 400
BUNDLE AVERAGE BURN-UP (MWh/kgU)
FIGURE 2.6-2
I1
ON-POWER FUELLING 1
• Low excess reactivity (± 1 mk) 1• Short fuel bundles (50 cm)• On-power removal of defective fuel I• Constant power shape
1I
REASONS FOR DEFECTS IN 1CANDU POWER REACTOR FUEL
NUMBER OFDEFECTS
Power ramp (previous to CANLUB)
Incomplete end cap weldsPorous end capsHandling damageFretting by debris in coolantFlew induced frettingUnknown causes
FIGURE 2.6-3
134 (The power rampfailure rate is zerofor CANLUB fuel)
125761
16
I1
II
IIIIIIIIII
The heat transport system normally remains hot and pressurized following
1reactor shutdown. However if it becomes necessary to cool down the system
or warm it up again, this is done under the control of the steam generatorpressure control program in the computer system and at predetermined ratesof temperature change. The system can be pressurized using heat from the
I pumps alone, but reactor heat is also available to help achieve design
III1
2.7 STARTUP, OPERATION AND SHUTDOWN SEQUENCES
Some of the techniques of operating the plant are considered in -thissection - especially as they relate to the plant control systems.
2.7.1 Initial Startup
In Section 2.4 it was noted that special startup counters are installedtemporarily for first reactor criticality. Using these counters andmanual control the reactor can be made critical and raised in power untilthe ion chambers provide signals to the computers. Control is thenautomatic and power can be raised by the operator supplying the requiredpower and maneuvering rates to the computer. From then on the reactorremains on automatic control. The ion chambers will continue to measurereactor power from the source term for a long period after the reactor ismade subcritical. The automatic controls therefore operate even while thereactor is subcritical. The startup instrumentation and manual startingwould only be needed again following a shutdown of many weeks duration.
2.7.2 Eversafe Shutdown
If major equipment is being maintained, the moderator is heavily poisonedto ensure that the reactor remains subcritical even following xenon decayor removal of reactivity control devices.
2.7.3 Heat Transport System Warmup and Cooldown
rates.
2.7.4 Turbine RunupII There is a program available in the plant computers to run the turbine up
to speed automatically and synchronize it to the grid.
2.7.5 Sudden Shutdown or Trip
If the reactor trips or steps back to zero power, recovery to power isautomatic once the operator resets the trip and gives a power setpoint to
2.1-20
the system. Resetting a trip of Shutdown System Number One causes theshut-off rods to withdraw. This is followed by withdrawal of mechanicalcontrol absorbers and lowering of light water control absorber levelsuntil the reactor is critical. Adjusters may also withdraw in response toxenon increases to allow the reactor to become critical. Power is held ata low value until the operator requests a power increase. The request fora power increase must come within about 20 minutes of the sudden shutdownfrom full power to avoid "poisoning out". If the reactor poisons out thexenon builds up to a level that the reactor control system cannotimmediately compensate for. The xenon will decay away allowing a restartafter about 36 hours. If the reactor is restarted within the 20 minute"decision and action" time the xenon will burn out as reactor power israised. However there will be some spatial assymetries in fluxdistribution which may not allow power to be raised immediately beyondabout 70% of full power. Power can then be raised over a few hours from70% to 100%.
2.7.6 Power Maneuvering
Reactor power can be raised at logarithimic rates up to 4% ofinstantaneous power per second while power is below 25% of full power. Amaximum linear rate of 1% of full power per second is applied at higherpowers. In general, turbine-generator maximum maneuvering rates will bemore limiting than this - 0.2% of full power per second would be typical.It is not possible to lower reactor power in a controlled way over a widerange at rates much faster than 1% per second. However the stepbackfunction and the use of steam discharge valves allows turbine generatorload to be shed at any required rate. Small step increases (of the orderof 5% of full power) or decreases in load can be handled by the normalcontrol system. It is thus possible to run the plant at less than fullpower in a spinning reserve mode.
Just as a sudden trip of the reactor leads to a transient increase inxenon, other load reductions have a similar but smaller effect. If thereactor power is suddenly reduced from 100% to 60%, the ensuing xenontransient would cause adjusters to withdraw in compensation. The reactorwould not poison out from this size of power decrease and could continueto operate at 60% power indefinitely. As the xenon transient decays, theadjusters would gradually re-insert in banks. A return to full powercould begin at any time, but might require several hours if begun whileadjusters were still out and tne flux distribution distorted. The reactorcontrol system would allow the plant to follow quite closely the typicaldaily load changes that occur on most electrical grids. Actual CANDUoperating experience is limited to base load operation.
2.1-21
2.7.7 poison Prevent Operation
It has been noted that a sudden reactor shutdown from full power wouldlead to poisoning out of the reactor unless it is brought to high poweragain within a short period. However reactor power reductions tointermediate power levels are accomplished without poisoning out.Therefore the normal procedure if the turbine trips, is to reduce reactorpower to about 60% of full power. The steam is then bypassed directly tothe condenser through the Condenser Steam Discharge Valves (CSDV). Theplant can be operated indefinitely this way until the turbine isavailable. Alternately the power can be gradually lowered from the 60%level at rates that do no lead to poisoning out.
2.8 SECONDARY SIDE SYSTEMS
2.8.1 Feedwater System
Feedwater to the steam generators is provided by two 100% pumps powered byindependent Class IV buses and by an auxiliary feedwater pump (5%) backedup by the Class III power supply.
Feedwater is controlled to each of the 4 steam generators independently(Figure 2.8-1). Each feedwater control valve station consists of two 100%control valves and one small control valve (15%) for low power operation.A check valve in each feedwater line prevents loss of steam generatorinventory in the event of a feed line rupture. The reheater drains flowis returned directly to the steam generators (depending on turbinesupplier). Each steam generator is provided with H2O sampling line tofacilitate secondary side chemistry control, and a continuous blowdownsystem, to minimize the accumulation of particulate matter at thetubesheet.
2.8.2 Steam System
Steam generated within the steam generators is normally directed to theturbine. In the event that the turbine is unavailable, steam can bedischarged to one or more of the following: the turbine condenser steamdischarge valves, atmospheric steam discharge valves or main steam reliefvalves.
2.8.3 Secondary Side Heat Balance
The secondary side flowsheet and heat balance are shown in Figure 2.8-2.The turbine consists of 1 High Pressure Stage (HP) and 3 Low PressureStages (LP).
The gross electrical ouptut is dependent on local service waterconditions, and on the turbine equipment selected.
2.1-22
ISOLATING VALVES
EnISOLATING VALVES
ATMOSPHERIC STEAMDISCHARGE VALVE (TYP)
SAFETY VALVES (TYP)
MAINSTEAMFLOW
MEASUREMENT
* - STEAM GENERATORWATER SAMPLING
FEEDWATER FROM FEEDWATER PUMPS
FIGURE 2.8.1 STEAM AND FEEDWATER SYSTEM
STEAMGENERATOR
SEPARATOHS REHEATERS
n n n nJ MAI N ST fc AM I
HP TURBINE
n
INTEHCEFVALVES
RBHEATER) DRAIN PUMPS
DEAERATORCONDENSEPRESSURE
No. G HEATERS Nn5 HEATFRS
•HEATER DRAINPUMPS GLAND SEAL
No 3 HDATfcRS No 2 HRATERS No 1 HfcATERS CONDENSER
FIGURE 2.8-2 SECONDARY SIDE FLOW DIAGRAM
3.0 MODERATOR AND AUXILIARY SYSTEMS
3.1 MAIN MODERATOR SYSTEM
The moderator system (Figure 3.1-1) removes the neutron heat generatedwithin the moderator, and the heat transferred to the moderator from thefuel channels. Heavy water is utilized as a moderator due to its highmoderating ratio (Figure 3.1-2). The heavy water is circulated throughthe moderator system for cooling, for purification and for the control ofthe concentration of substances used for reactivity adjustment. Themoderator system features two 100% pumps and two 50% heat exchangers. Thepiping arrangement permits either pump to operate with either or both ofthe heat exchangers. The pumps are provided with pony motors powered byClass IV and Class III power to provide circulation in the event of a lossof Class IV power. The location of major equipment within the reactorbuilding is shown in Figure 3.1-3.
3 . 2 MODERATOR AUXILIARY SYSTEMS
The moderator auxiliary systems (Figure 3.2-1) include:
a) Moderator Cover Gas System; A helium cover gas is maintained overthe moderator in the calandria. The moderator cover gas system(Figure 3.2-2) cools the cover gas and recombines the deuterium,generated by radiolysis within the moderator, with oxygen. Thesystem consists basically of two 100% compressors and two 100%recombining units which circulate the cover gas through the calandriapressure relief ducts.
b) Moderator Purification System: The moderator is circulated throughthe moderator purification system to minimize the accumulation ofactivity within the system, and to control the concentration ofsubstances used for reactivity adjustment. Poisons are added to themoderator in small quantities when the reactor is first started andwhen new fuel is added. Large amounts of poison are also added tothe moderator if the second shutdown system is activated.
3.3 HEAVY WATER MANAGEMENT
3.3.1 General
Because of the high cost of heavy water, and to minimize releases ofactivity from the station, maximum attention is given to minimizingD2O losses. This is accomplished by lowering the number of mechanicaljoints in D2O systems, utilizing bellows sealed valves when possible.
3.1-1
TO MODERATOR COVER GAS SYSTEM
FROM MODERATOR COVER GAS SYSTEM
TO D2O SAMPLING SYSTEM
T TO AND FROM D2OSUPPLY SYSTEM
HEAT EXCHANGER HEAT EXCHANGER
FIGURE 3.1-1 MAIN MODERATOR SYSTEM
SLOWING DOWNPOWER CM-1
MODERATING RATIO
LIGHTWATER
1.35
60
HEAVYWATER
0.178
2,000
GRAPHITE
0.06
170
NEUTRON WASTAGEIN MODERATOR,COOLANT ANDCORE STRUCTURESPER FISSION(TYPICAL)
PWR
.28
BWR
.25
CANDUPHW
.15
MAGNOX
0.16
AGR
0.3
FIGURE 3.1-2 MODERATING EFFICIENCY OF HEAVY WATER
1 MODERATOR HEAT EXCHANGER2 MODERATOR PUMP3 REACTOR
FIGURE 3.1-3 LOCATION OF MAIN MODERATOR SYSTEM EQUIPMENT
DEUTERATiONDE-DEUTERATION
— — RUPTURE DISKS
OOOOOOOOOO
oooooo o ooooooo o oooo ooo o oooo oo o o oooooo o
OOOO
RECOMBINATION
CALANDRIA
D2OSUPPLY
LIQUIDPOISON
D2OCOLLECTION
FIGURE 3.2-1 MODERATOR AND AUXILIARY SYSTEMS
fUM/y \
T M A N I F O L D S T H E L | 1 BOTTLES
FUMEARRESTER(TYPICAL) CATALYTIC
RECOMBINATION UNITS
VENT TO REACTIVITY MECHANISMS
CALANDRIAHEAD TANK
FIGURE 3.2-2 MODERATOR COVER GAS SYSTEM
and by providing a system to collect heavy water from points ofanticipated leakage. Downgrading of heavy water leakages 1* reduced bylocating most H2O systems away from D2O system areas and byminimizing mechanical joints in H2O systems within the reactorbuilding. A typical D20 collection system is shown in Figure 3.3-1.Sight glasses are provided in individual and common collection lines toprovide a visual indication of liquid flow.
To reduce operating staff exposure, all collected D20 with a hightritium content is segregated from that with a low tritium content.
3.3.2 D2O Vapour Recovery Systems
Four separate vapour recovery systems are provided, each serving aseparate area in the reactor building:
a) Major areas subject to Heat Transport System leakage which areaccessible only during reactor shutdown.
b) Areas requiring frequent personnel access.
c) The area within moderator equipment enclosures that may have a hightritium content.
d) The steam generator room which is accessible during reactoroperation.
The vapour recovery equipment is located in the service building, exceptfor that serving the steam generator area, which is located in the steamgenerator area. The recovery units are of the absorption type. Recoveryof absorbed water vapour occurs in the reactivation condensers locatedabove the dryer vessels; this water is collected in a series of tanks inthe D2O management upgrading area. Recovered water is segregatedaccording to the degree of downgrading and tritium activity.
3,3.3 D20 Cleanup System
During operation of heavy water systems, small amounts of D20 escapeunintentionally by leakage, and intentionally by deutration* anddedeuteration of ion exchange resins. The D2O cleanup system removesdissolved, particulate and organic impurities from recovered D20 toproduce a product suitable for the upgraders. Two separate and almostidentical systems are provided; one for low tritium D20 and one forhigh tritium D2O.
* Deuteration is the process whereby light water present in new resinsis displaced by heavy water. Dedeuteration is the reverse process.
3.1-2
LEAKAGE INDICATORS
22 CONNECTIONS(TYPICALI
MAIN HEAT TRANSPORT
-DRAIN INDICATORS-
TOD2O STORAGETANK
- VENT INDICATORS-
TO D2O CLEANUP SYSTEM • * • * • } - (AND SAMPLING
TC FEEDPUMP SUCTION
BCONDENSER
1D2O COLLECTION
TANK
PUMPS/MOTORS AND DRAINS
FIGURE 3.3-1 D2O COLLECTION SYSTEM
3.3.4 D2O Upgrading System
The D2O upgrading systems separate solutions of D20 and HjO bydistillation. Two upgraders are provided, one for low tritium D2O andone for high tritium D2O.
The overall D2O management system is illustrated in Figure 3.3-2.
3.1-3
DEUTERATICN& DE-DEUTERATION
SYSTEMS
VAPOURRECOVERY
SYSTEM
LIQUIDRECOVERY
SYSTEM
TOMODERATOR
SYSTEM
TOHEAT
TRANSPORTSYSTEM
LD2O
CLEANUPSYSTEM
D2OUPGRADING
SYSTEM
D2O SUPPLY
SYSTEM
f
FRESHD2O
FROMHEAT
TRANSPORTSYSTEM
FROMMODERATOR
SYSTEM
FIGURE 3.3-2 D2O MANAGEMENT SYSTEM
4.0 SAFETY SYSTEMS
4.1 INHERENT SAFETY FEATURES OF CANDU
The CANDU PHW reactor design with its heavy water moderator, naturaluranium fuel and pressure tube concept has certain inherent safetycharacteristics (Figure 4.1-1) that obviate the need for a high strengthpressure vessel. Instead, the pressure boundaries are the pressure tubeswhich are considerably simpler to manufacture to the required quality.Further, experimental evidence indicates that pressure tubes will leakbefore they break since their thickness is much less than the criticalcrack length. Such leaks can be readily detected by monitoring themoisture content and the pressure in the gas annulus between the pressuretube and the calandria tube. This is done on a continuous basis. Inaddition, ultrasonic scanning devices are mounted on the fuelling machinefor periodic in-service inspection of the pressure tubes.
The pressure tube design permits the heat transport system to besubdivided into two separate coolant circuits (loops). In the case of ahypothetical loss of coolant accident, this design feature restricts theconsequences of the loss of coolant accident to just one of the loops.This simplifies the design and reduces the burden considerably on theemergency injection and the contairment system design.
All reactivity devices are located in guide tubes positioned in the lowpressure moderator environment, Figure 4.1-2. Thus, there exists nomechanism for rapid ejection of any of these reactivity devices, nor canthey drop out of the core. The maximum reactivity rates achievable bydriving all control reactivity devices together in the wrong direction isabout 0.35 mk per second and well within the design capabilities of theprotective systems.
Fuel, coolant and moderator are arranged on a square lattice with a 28.6cm pitch. This is a near optimum geometry from a reactivity standpoint,Figure 4.1-3. Even if all fuel channels were either pushed apart orbrought together for whatever reason the net reactivity increase would beat most, 1 ink; and this only for the ideal case of uniform rearrangement.This is, of course, physically impossible. For the case where one, or afew fuel channels are displaced, the net reactivity would at worst not beaffected at all or it would decrease, thereby shutting down the reactor.Also, since a lattice of natural uranium and light water cannot be madecritical in any concentration, there can be no criticality problems in thespent fuel bay of CANDU reactors.
The pressure tube design also makes on-power fuelling a possibility.On-power fuelling results in a reactor with very low reactivity controlrequirements. Typically, the reactivity decay rate in 600 MW(e) CANDU PHWreactors is about 0.4 mk per day. This is compensated by fuelling about
4.1-1
PRESSURE TUBESSeparate moderator from coolantCool, low pressure moderatorHigh pressure coolantInterstitial reactivity devicesSubdivided PHTSTubes leak before break
FIGURE 4.1-1
REACTIVITY DEVICES
In low pressure moderatorNO pressure-driven ejectionSeparate devices for control andfor safetyModest reactivity worthMaximum combined rate < 0.35 mk/s
FIGURE 4.1-2
REACTOR PHYSICS
Natural UO2 fuelD2O moderatorLow excess reactivityNear optimum geometryCriticality in spent fuel bay not possible
FIGURE 4.1-3
two channels per day. In addition, the pressure tube concept provides anexcellent opportunity for locating fuel defects and the on-power fuellingpermits the removal of defective fuel as soon as it is detected. Thishelps to keep the heat transport system essentially free from fissionproduct activity.
Finally, the separation of the moderator from the high pressure heattransport coolant allows the moderator to act under certain circumstancesas an additional heat sink for the fuel decay heat/ e.g. where one mighthypothesize a failure or impairment in the emergency core cooling systemfollowing a primary loss of coolant accident (LOCA).
4.2 SAFETY DESIGN PHILOSOPHY
4.2.1 Design Basis Considerations
The basic safety functions to be maintained following any postulated eventare as follows:
a) The ability to shut the reactor down and maintain it in a safeshutdown condition. ,
b) The. ability to remove residual and decay heat.
c) The ability to limit the release of radioactive material.
d) The ability to perform essential safety related control andmonitoring functions.
The plant design considers both common mode events and randan failures.Since the nuclear process produces heat for a considerable period afterthe reactor is shutdown, the plant design also takes into account otherfaults or events which might occur in the post-accident recovery period.
4.2.2 Defense Against Random Failures
The design objective here emphasizes defense in depth and consistsessentially of the following:
a) High level of equipment quality. All systems are designed toestablished codes and standards which demand the highest quality inmaterial and workmanship. This makes equipment failure unlikely tobegin with.
4.1-2
b) Quality Assurance. This involves strict quality control, both duringmanufacture and subsequent installation, together with continuedperiodic inspection of major components throughout the plantoperation.
c) System redundancy and fail-safe design. The important systems aredesigned with redundancy such that the loss of a single componentdoes not cause the loss of the whole system: the provision of dualcontrol computers with one of them providing a complete backup forthe other one, two pumps circulating coolant in each of the heattransport systems, two completely different systems for decay heatremoval (i.e. the shutdown cooling system, or the steam generators)and finally, redundancy in power supply based on the so-calledodd-even concept. Two completely separate power supply systems areprovided such that one half of the load for any process is suppliedfrom an odd-bus and the other half from an even-bus.
d) Regulating and Process Systems. These systems are designed tomaintain all operating, parameters within acceptable ranges undernormal operating conditions and when minor accidents occur, e.g.small leaks from the HT system, single computer failure, etc.,without resorting to the special safety systems.
e) Special Safety Systems. If important process system parameterscannot be normally controlled and exceed certain preset values,special safety systems shutdown the reactor, provide long termcooling of the fuel, and contain potential releases ofradioactivity.
4.2.3 Protection Against Common Mode Events
The basic defense against common mode events (Figure 4.2-1), is throughthe use of superior equipment and separation (by distance and/or barriers)of reliable systems, structures and components (Figure 4.2-2). The depthof protection, based on the anticipated rate of occurence of the commonmode events, also guarantees that the common mode events underconsideration cannot disable the systems required to shutdown the reactorand to remove residual heat, i.e. the basic safety functions have to bemaintained. Some of the common mode events considered are: man inducedevents such as fires and missiles, natural phenomena such as earthquakesand floods, human errors arising from design and operation and cascadingof cross-link effects such as effects of pipe whip, environment producedby postulated events, etc. (Figure 4.2-3).
One of the important elements in the defense against common mode events isthe two group separation philosophy (Figure 4.2-4). All safety relatedsystems in the nuclear plant are divided into two groups. These groups
4.1-3
PROTECTION AGAINSTCOMMON MODE EVENTS
• Siting consideration• High quality design, manufacture,
operation• Qualification (hardening)• Duplication + diversity
•Two group approach
FIGURE 4.2-1
FIGURE 4.2-2 TYPICAL 600 MW(a) PUNT UYOUT
DEFENCE AGAINSTCOMMON MODE EVENTS
COMMON MODE EVENTS• Man induced - fires, missiles ...• Natural phenomena-earthquakes, flood• Human error -design, operation ..• Cascading - pipe whip,
harsh environment
FIGURE 4.2-3
TWO GROUP APPROACHEssential SAFETY FUNCTIONS
• Shutdown reactor• Cool the fuel• Monitor plant
• Design objective to use two groups• NOT a licensing requirement
FIGURE 4.2-4
are separated so that, within the limits of design, no directional orlocalised common mode events can disable more than one group* Inherent inthis philosophy is the premise that the reactor building is by design animpenetrable barrier to such common mode events. The systems in eachqroup must be able to carry out the basic safety functions.
a) Shutdown the reactor and maintain its shutdown.
b) Remove the decay heat.
c) Supply the necessary information for post-accident monitoring topermit the operator to assess the state of the nuclear steam supplysystem.
In group 1 these functions are performed respectively by shutdown system#1, the normal electrical and water supplies and monitoring from the maincontrol room. In group 2, the corresponding systems are shutdown system#2, the emergency power and water supplies and monitoring from thesecondary control area (Figure 4.2-5).
Group 1 Safety Support Systems
These systems (the normal electric and water supply systems) support theoperation of one of the special safety systems. Because of the relianceon these systems for both normal plant operation and continuing operationof special safety systems, special measures are taken in their design toassure reliability.
Group 2 Safety Support Systems
As part of group 2, two safety support systems are provided. They are theemergency water supplies and the emergency power supply systems. They donot perform any function for normal plant operation but are required toprovide an alternative water supply and electrical power supply duringcertain accident conditions. These alternate supplies are locatedsufficiently remote from the water and electric power supplies of group 1to ensure defense against common mode incidents.
Grouping Layout
The functional and physical independence of the two groups ensure that nocommon mode incident can disable the required systems of both groups.There is no unobstructed straightline path between redundant elements ofthe two groups above ground. Where there are no suitable obstructions,one of the elements is embedded in a suitable reinforcement. The group 1control area is the main control room which is located on the third floorof the service building. The group 2 control area is a secondary controlarea which is located on the side of the reactor building remote from the
4.1-4
TWO GROUP CONCEPT
FUNCTION GROUP 1 GROUP 2
Shutdown SDS1 SDS2
Fuel cooling Normal electrical Emergency powerand water supplies and water supplies
Plant monitoring Main control room Secondary controlarea
FIGURE 4.2-5
main control centre. Cables for group 2 exit from the reactor building ona different side than those for group 1. Control is normally exercisedfrom the main control centre, but under emergency condition*, control forshutdown and decay heat removal is also available from the secondarycontrol area.
The criteria for separation and independence between the special safetysystems belonging to the two groups are as follows:
1) Physical Independence. There must be no sharing of system componentsbetween group 1 and group 2 systems. There must also be no sharingof routes for the wiring and tubing of systems in different groups.
2) Functional and Conceptual Independence. When two special safety *systems are designed to perform the same protective function, the twosystems"must have conceptually different senses, instrumentation andactuators whenever practical. Where possible, similar components ofthe two systems must be supplied by different manufacturers. Where achoice is possible, such components must employ different principlesof operation.
The best known common mode event is, of course, an earthquake. A briefoutline of the design principles used in protecting against earthquakeswill conclude this section on defense against common mode events.
1) Following an earthquake the reactor control system and the reactorshutdown system must either remain functional or fail safe.
2) Sufficient systems required for core cooling (decay heat removal)must remain functional.
3) The earthquake should not cause a breach in the heat transport systempressure boundary.
4) The containment building and associated systems remain functionalduring and following an earthquake.
5) Structures and systems not directly required for nuclear safetyreasons are designed so that their failure or dislocation is eitherunlikely, (systems are qualified), or do not effect the safeoperation of any safety related systems required during or after theearthquake.
4.1-5
4.3 SAFETY SYSTEMS DESCRIPTION
This section describes those systems which are provided solely to performa safety function and have no function in the normal production ofelectrical power. As noted earlier, these systems consist of thecontainment system, the emergency core cooling system, shutdown system #1,shutdown system #2, the emergency water supply system and the emergencypower supply system (Figure 4.3-1).
4.3.1 Containment
The containment system (Figure 4.3-2) consists of a prestressed, post-tensioned concrete containment structure with an epoxy liner, energy sinksconsisting of an automatically initiated dousing system and building aircoolers, a filtered air discharge system, access airlocks and anautomatically initiated containment isolation system.
The dousing tank is located in the dome of the reactor building and holdswater for emergency dousing and emergency core cooling. About 500 cubicmeters of water are reserved for emergency core cooling. The totalcapacity of the tank is about 2600 cubic meters. Dousing valves controlthe flow'of water to six independent dousing spray header units locatedradially below the tank. Each spray unit has two butterfly valves in adowncomer between the tank and the spray header (Figure 4.3-3). Thedesign dousing flow rate is about 4500 kg/s and this flowrate can beprovided by any four of the six downcomers. With all six downcotnersoperating, the total spray flow is about 6800 kg/s.
4.3.1.1 Operation
Under normal operation conditions, the pressure within containment isslightly less than atmospheric and the containment ventilation dampers areall open.
For very small heat transport system leaks, the building coolers in thecontainment condense any steam that is discharged, the building pressureremains at atmospheric pressure and there may be some additional outflowof dried air through the ventilation system. For larger breaks, thebuilding pressure rises and at an overpressure of about 3.4 kPa,containment pressure sensors initiate total containment closure. Thecontainment will also be automatically isolated in the event of a highradioactivity signal which may occur following a large loss of coolantaccident. The containment pressure continues to rise and the dousingsystem starts to operate automatically at an overpressure of 13.8 kPaDepending on the break size, there is either continuous or cyclicoperation of the dousing valves, with the valves opening at 13.8 kPa and
4.1-6
SPECIAL SAFETY SYSTEMS
Emergency water supply systemEmergency power supply systemShutdown system 1Shutdown system 2Emergency core cooling systemContainment system
FIGURE 4.3-1
CONTAINMENT SYSTEM
Designed for maximum size LOCASmall breaks — air coolersLarge breaks — dousing systemFiltered discharge systemMulti-unit station •vacuum building
FIGURE 4.3-2
DOUSING WATERSUPPLY
DOUSINGSPRAY HEADER
MAIN PRIMARYSYSTEM PUMPS
FIGURE 4.3-3 SINGLE UNIT CONTAINMENT
closing at 6.9 kPa. When the dousing system overtakes the pressuretransient, the pressure begins to fall, and the building depressurizes toabout atmospheric pressure by condensation on the building walls andcooling by the air coolers. Initial containment atmosphere cleanup can beperformed by the D2O vapour recovery dryers with long term purgingachieved by discharging air through the dryers and the reactor buildingventilation system filters before release to the atmosphere.
4.3.2 Emergency Core Cooling (ECC) System
The emergency core cooling system (Figure 4.3-4) is composed of threestages: high pressure, medium pressure, low pressure. The high pressurestage uses pressurized nitrogen to inject water into the reactor core fromwater tanks located outside the reactor building, the medium pressurestage supplies water from the dousing tank. When this water supply isdepleted, the low pressure stage recovers water that has collected in thereactor building sump and pumps it back into the reactor core via theemergency cooling heat exchanger and the emergency cooling recoverypumps.
The high pressure injection stage consists of one nitrogen gas tank andtwo water tanks. The gas tank normally operates at a pressure between 4.1MPa and 5.5 MPa, whereas the water tanks operate slightly aboveatmospheric pressure. The recovery pumps are two 100% pumps. Each pumpis supplied by Class III power and by the emergency power supply system.The heat exchanger in the recovery pump discharge line is designed tomaintain the emergency cooling flow at about 50°C at entry to the heattransport system.
Operation. The emergency core cooling system is triggered automaticallywhen the heat transport pressure reaches 5.5 HPa. The following actionstake place:
a) All gas isolation valves, the high pressure injection valves, and theD2O isolating valves are opened. This will open rupture discs inthe injection lines and permit the flow of high pressure, water fromthe injection tanks to all reactor headers of the failed and theunfailed loops (Figure 4.3-5).
b) The main steam safety valves on the steam generators are opened torapidly cool down the boilers and provide an additional heat sink.This is the main heat sink for small loss of coolant accidents.
c) Valves in all lines interconnecting the two heat transport loops areclosed. This will confine the consequences of the loss of coolantaccident to just the loop containing the hypothesized break.Sufficient coolant is available during the high pressure injectionphase for at least 2.5 minutes.
4.1-7
EMERGENCY CORECOOLING SYSTEM
All-points injectionReactor is low-point in systemThree stage injection
High pressure — external tanksIntermediate — dousing tank waterLow pressure — building sump
NOT "LAST DEFENCE" FOR LOCA
FIGURE 4.3-4
FROM DOUSING TANK
GAS ISOLATION VALVES
•APV81
PV82 | ^ ^
OECC GAS TANK
ECCWATERTANKS
~^^£
ijl'lj
III
til
KKKHr
ECCHEATEXCHANGER I
PV8
MP TEST VALVES
MP INJECTIONVALVES
| Cxj—-W-V5 V3
I X}—-W-V6 V4
RIH = REACTOR INLET HEADERROH = REACTOR OUTLET HEADERMP = MEDIUM PRESSUREHP = HIGH PRESSURE
ECCPUMPS
LOOP 1 LOOP 2 LOOP 1
TO PRIMARY HEAT TRANSPORT SYSTEM
FSGURE 4.3-5 EMERGENCY CORE COOLING HIGH PRESSURE STAGE OPERATION:2.5 TO 30 MINUTES DURATION. ASSUMED LOSS OF COOLANT IN LOOP 1
The ECC dousing tank suction valves to the ECC recovery pumps will also beopened automatically on the loss of coolant signal and one of the ECCpumps will be started when these valves are opened. If this pump fails tostart (as indicated by a low pump discharge pressure), the standby ECCpump will be started automatically. No operator action is required tostart the recovery pumps or to open the valves to supply dousing tankwater to the pumps for the medium pressure emergency core phase. Thisphase will continue to supply water for at least 15 minutes following thelargest break in the heat transport system. It will last longer forsmaller breaks (Figure 4.3-6).
As the dousing tank water depletes, the operator opens valves in therecovery line from the reactor building sump, then closes the valves inthe line from the dousing tank and opens the cooling water valves tosupply service water to the ECC heat exchanger. The mixture of heattransport coolant and water from the high pressure and dousing tanks ispumped from the sump in the reactor building back to the heat transportsystem via the heat exchanger (Figure 4.3-7). For large breaks, the ECCrecovery heat exchanger is the main heat sink. For small breaks, thesteam generators continue to be the main heat sink.
4.3.3 Shutdown System #1
The shutdown system #1 (Figure 4.3-8) is the primary method of rapidlyterminating any reactor power increase or reducing reactor power whencertain parameters exceed preset values. This is accomplished by therelease of 28 cadmium rods which fall under gravity from the top of thereactor. Figure 4.3-9. This gravity drop is accomplished by de-energizingdirect current clutches which normally hold the shutoff rods out of thecore. The shutoff rod units are divided into two banks of fourteen. Eachbank is supplied with dual 90 volt DC power supply for the clutches. Eachclutch coil is held energized by the contact of the separate relay.
The design philosophy is based on triplicating the measurement of each ofthe variables that can initiate reactor shutdown. Protective action isinitiated when any two of the three measurements exceed their presetvalues. The selection of variables is such that where practicable, thereare at least two different sensing parameters for the specific processfailure being protected against. Examples of trip parameters on shutdownsystem #1 are high neutron power, high rate log neutron power, high heattransport pressure, high reactor building pressure, low steam generatorlevel, low pressurizer level.
A partial drop test facility is provided to allow the operation of eachshutoff unit to be checked during reactor operation. The shutoff unithousings are located on the reactivity mechanism deck which permitsregulated, one unit at a time, access to the clutches, motors,potentiometers, gear boxes, and winches for removal or for maintenance onpower.
4.1-8
FROM DOUSING TANK
GAS ISOLATION VALVES
PV8
MP TEST VALVES
MP INJECTIONVALVES
RIH = REACTOR INLET HEADERROH = REACTOR OUTLET HEADERMP = MEDIUM PRESSUREHP = HIGH PRESSURE
LOOP 1 LOOP 2 LOOP 1
TO PRIMARY HEAT TRANSPORT SYSTEM
FIGURE 4.3-6 EMERGENCY CORE COOLING MEDIUM PRESSURE STAGE GENERATION:13 MINUTES MINIMUM. ASSUMED LOSS OF COOLANT IN LOOP 1
HP INJECTION VALVES
HP TEST VALVES
SHUTDOWN SYSTEM 1
Independent of regulating system28 Cd shutoff unitsDevices inserted from topInstrumentation on top or verticalPrimary + alternate trip parameterfor each process failureIndependent of SDS2
FIGURE 4.3-8
Separately channelled Class I and Class II power supplies are provided foreach channel of shutdown system #1. The logic is arranged so that anyloss of power to a channel results in a channel trip. The direct currentclutches energized by rectified Class II power will release if power isdisrupted. This will shut down the reactor.
The static negative reactivity worth of the 28 shutoff rods is about 80ink.
4.3.4 Shutdown System #2
The second method of quickly terminating any reactor power increase orrapidly reducing reactor power is by the injection of concentratedgadolinium nitrate poison solution into the moderator through sixhorizontal nozzles, Figures 4.3-9 and 4.3-10. A vessel containing highpressure helium supplies the source of energy for this rapid injection.This vessel is connected through six quick opening valves to a heliumheader which supplies the poison tanks. The quick opening valves are"air-to-close, spring-to-open" design so that loss of instrument airinitiates automatic poison injection. Each of the poison tanks containsgadolinium nitrate solution at a concentration of about 8,000 parts permillion. The six zircaloy injection nozzles penetrate the calandriahorizontally and at right angles to the fuel channels. Holes are drilledinto the nozzle along its length to form four rows of jets whichfacilitate complete dispersion of the poison into the moderator (Figure4.3-11).
Each poison tank contains a floating polyethylene ball. When an injectionis initiated the helium driving gas transfers the poison to the calandriaand the ball is driven to the tank bottom. In the bottom position, theball seats at the poison tank outlet and prevents the release of a largevolume of helium into the calandria.
as with shutdown system #1 all initiating variables are triplicated andprotective action is initiated by any two of these measurements exceedingpreset values.
The eventual negative reactivity from the poison injection system is inexcess of 300 mk.
4.3.5 Emergency Water Supply System
This system (ref. Figure 4.3-12) is designed to provide an alternatesource of water to cater for:
a) A design basis earthquake.
4.1-9
SHUTOFF RODGUIDE TUBE
SHUTOFF ROD(TYPICAL)
LIQUIDPOISON PIPE(TYPICAL)
CALANDRIA
MODERATOR
LIQUID POISONNOZZLE
CALANDRIATUBE
FIGURE 4.3-9 SHUTDOWN SYSTEMS: SHUTOFF RODS AND LIQUID "POISON" INJECTION
SHUTDOWN SYSTEM 2
Independent of regulating system6 Gd injection nozzlesDevices located horizontallyInstrumentation on side or horizontalPrimary + aitemate trip parameterfor each process failureIndependent of SDS1
FIGURE 4.3-10
HELIUM VENT LINESTO EXHAUST
FROMHELIUM
H.P. HELIUM SUPPLY TANK
GADOLIUM NITRATEIN HEAVY WATER " "
-----
:--¥i-r-;:--|
-f
PRESSURE BALANCE LINE
HELIUM COVER GAS
ISOLATION BALL VALVE y(NORMALLY OPEN)
OOOOOOOOOOOOO
6 NOZZLES
HEAVY WATER MODERATOR
POISON-MODERATOR INTERFACE
CALANDRIA
FIGURE 4.3-11 SCHEMATIC OF SECOND SHUTDOWN SYSTEM
CONNECTION FORADDITIONAL PUMPFOR TWO UNITOPERATION
V9
3114-VI TO V4
I I I I FROMREHEATERDRAIN
FIGURE 4.3-12 EMERGENCY WATER SUPPLY
b) Loss of Class XV and Class III power
c) Loss of coolant accident followed (24 hours or later) by a sitedesign earthquake*
Two electrically driven pumps are located remote from the group 1equipment- and pump house. The pump suction is taken from a sump connectedto a good quality water supply. The emergency water system connects tothe heat transport system, the steam generators and the emergency corecooling heat exchange. Since the emergency water supply system is notrequired immediately following the three events listed above, systemoperation consists of manually starting the pumps and then operating thehand switches to open the appropriate motorized isolating valves to supplywater to the required loads.
4.3.6 Emergency Power Supply System
This system is designed to provide an alternate source of power to caterfor the events already described in the previous paragraph. The emergencypower supply system supplies the necessary power to the emergency watersystem pumps and valves, the emergency core cooling pumps and certainemergency core cooling valves, and power to the group 2 safety and controlsystems for operator control of the station from the secondary controlarea. This system is qualified to the design basis earthquake.
4.1-10
5.0 REFUELLING SYSTEM
S.0.1 Intr oduct ion
CANDU reactors rely on semi-continuous on-power refuelling to enable closecontrol of core reactivity and efficient utilization of its naturaluranium fuel.
The operation is carried out by an automated fuel handling system (Figure5.0-1) which utilizes features and equipment applied in the verysuccessful Pickering "A" Nuclear Generating Station.
The full power refuelling requirements for CANDU 600 MW(e) reactorsinvolve replacing about 110 fuel bundles per week. Using the referenceprocedure of replacing eight bundles per fuel channel, this entailsrefuelling 14 fuel channels per week.
In a.n eight bundle refuelling sequence (Figure 5.0-2), closure plugs areremoved from a channel, and stored within the magazines of the fuellingmachines coupled to each end of the channel.' Four pairs of new fuelbundles are then inserted from the upstream fuelling machine, displacingeight spent fuel bundles into the downstream machine. This process istermed "Flow Assisted Fuelling" (FAF Mode).
The overall refuelling operation of a CANDU 600 HW(e) reactor unit (Figure5.0-3) comprises:
1) Loading new fuel bundles, in pairs, into a fuelling machine.
2) Coupling this machine onto the upstream end of the channel to berefuelled and coupling a second machine onto the downstream end ofthat channel
3) Displacing spent fuel into the downstream fuelling machine
4) Disengaging the downstream fuelling machine, and moving it to thespent fuel port
5) Discharging the spent fuel through the spent fuel port into anelevator, which lowers the bundles into the fuel Discharge Bay to anunderwater conveyor '
6) Transferring the spent fuel from the conveyor to storage trays forstacking in the Spent Fuel Storage Bays.
This automated fuelling.cycle takes approximately two and a half hours,which nominally requires system operations for 35 hours per week.Miscellaneous check out procedures take up to one hour per day.
5.1-1
1 FUELLING MACHINE BRIDGE STRUCTURE2 FUELLING MACHINE GUIDE COLUMN3 BRIDGE SUPPORTA BALL SCREW ASSEMBLIES5 FUELLING MACHINE HEADS6 FUELLING MACHINE CARRIAGE TROLLEY7 ROLLING SHIELD
FIGURE 5.0-1 FUELLING MACHINE USED IN PICKERINGAND CANDU 600 MW(e) PHWR'S
DISCHARGE MACHINE
1B
RAM ©WITHDRAWS CLOSURE PLUG ©RAM ® WITHDRAWS CLOSURE PLUG ©
RAM ® CONNECTS TO SHIELD PLUG ©AND WITHDRAWS IT TO MAGAZINE
2B
RAM ©CONNECTS TO SHIELD PLUG©
3A
V'EW FUEL BUNDLES ROTATED INTOCHARGING POSITION. TWO BUNDLES AT A TIME
3B
4A
RAM ©CHARGES NEWACTIVITIES 3A and 4A REPEATED 4 TIMESUNTIL 8 NEW BUNDLES HAVE BEEN INSERTED.THE FUEL COLUMN IS ADVANCED.
4B
SHIELD PLUG ©WITHDRAWNSIDE STOPS (3)COME IN
SB
USED FUEL BUNDLES WITHDRAWN TOMAGAZINE, TWO BUNDLES AT A TIME
SIDE STOPS® HOLD FUEL COLUMN AND ROTATINGMAGAZINE SWINGS USED FUEL BUNDLES OUT OFTHE WAY. ACTIVITIES SB AND SB REPEATED 4 TIMES
NT BUNDLES HAVE BEEN DISCHARGED
RAM® REPLACES SHIELD PLUG© RAM ® REPLACES SHIELD PLUG @
8A
RAM ©REPLACES CLOSURE PLUG© END SHIELD
* NEW
END SHIELD H A M (©REPLACESCLOSURE PLUG®
FIGURE 5.0-2 8-BUNDLE CHANGING SEQUENCE IN A CANDU 600 MW{«) PHWR
NEW FUELSTORAGE ROOM
NEW FUELLOADING AREA
SPENT FUELDISCHARGE ROOM
SPENT FUELSTORAGE BAY
CANNEDFAILED FUEL
STORAGE TRAYS
FIGURE 5.0-3 FUEL HANDLING SYSTEM SCHEMATIC
Accordingly, operation on a regular two shift/ five day basis providesabout 100% availability margin for any operational interruptions, checksor delays.
5.1 FUELLING MACHINES
The major elements of the refuelling system are a pair of identicalunshielded fuelling machines which operate at both ends of the reactor andbring new fuel from the New Fuel Ports to the reactor, and carryspent fuel to the Spent Fuel Ports (Figure 5.1-1).
Fuelling machines are mounted in carriages which move from maintenancearea tracks onto bridges, mounted on columns in the fuelling machinevaults.
These raise and position the fuelling machines at each end of the fuelchannel end fittings, to form a sealed connection before starting the fuelreplacement sequence.
The fuelling machine (Figure 5.1-2), has a 12 station rotating magazine,snout assembly (which locks onto the channel) and a ram assembly which isused for removal, storage and replacement of fuel channel plugs, and forinsertion of fuel. The closure plugs are engaged, unlocked and withdrawninto a storage chamber in the magazine; similar motions are used inwithdrawal of shield plugs.
New fuel is pushed by the upstream ram into the pressure tube, where theheavy water flow brings it into contact with the installed fuel string.
In central channels the hydraulic forces are sufficient to move the entirefuel string along the fuel channel to displace pairs of bundles into thedownstream fuelling machine. (FAF Mode).
A FARE tool (Flow Assisting Ram Extension) is inserted to provideadditional hydraulic flow resistance in refuelling outer channels whichhave less flow (corresponding to their lower power levels). This istermed FARE mode refuelling.
As each pair of fuel bundles enter the downstream magazine, separators areinserted to limit motion of the following bundles, to advance the twodownstream bundles and to enable free rotation of the magazine.
After completion of the refuelling sequence the shield and closure plugsare reinstalled. The fuelling machine is sealed by installation of asnout plug, the space between the closures is drained, and the fuellingmachines disengage from the fuel channel.
5.1-2
NEW FUEL \ EMERGENCY ACCESS PLUGSCATENARIES
SPENT FUEL PORT. PORT
REHEARSAL CHANNEL \ T.V. CAMERA
SERVICE PORTS
S3JFIGURE 5.1-1 FUELLING MACHINE VAI,LT AND MAINTENANCE ROOM
A SNOUT PLUGB FUEL JC CHANNEL CLOSURE t0 FUEL 1E GUOE SLEEVE (INSERTION TOOL IF FUEL I
VKW PROM THE REAR
SHIELD PLUGADAPTERCHANNEL CLOSURE (SPARE)FUELSHIELD PLUG (SPARE)FUEL
1234S«78S
10111213141516
1718
ANTENNA PLATEANTENNA SWITCHCLAMPING LEVER ARMCAM BLOCKWEDGE SEGMENTCLAMPING BARRELSEALSNOUT PROBELOCK RINGSCREW AND GEARCENTRE SUPPORTSEPARATOR ASSEMBLYFUEL STOPSCLAMPING PISTONRACKSNOUT EMERGENCYLOCK ASSEMBLYLOCK PISTONMAGAZINE END COVER
1«2021222324252S27282»30
31
32333436
FRONT RETAINING PLATEWEIR30" GRAYLOC CLAMPGRAYLOC SEAL RINGMAGAZINE HOUSINGMAGAZINE DMVE SHAFTREAR RETAINING PLATEBALANCE SHAFT SEALFERGUSON INDEXING DRIVEFLOW SHIELDRAM HEADMAGAZINE POSITIONPOTENTIOMETERSMAGAZINE EMERGENCYDRIVE GEARBOXMAGAZINE DRIVE MOTOR10" GRAYLOC CLAMPRAM HOUSINGEDUCTOR
FIGURE 5.1-2 FUELLING MACHINE HEAD
Subsequently thfi Fuotlinq machines are lowered and driven into themaintenance area to discharge spent fuel and to load new fuel.
The catenaries and catenary trolleys (Figure 5.1-3} provide electrical,water and oil hydraulic connections for fuelling machine controls andcooling, and the drive supplies to the mobile fuelling machines, from thestatic control stations.
5.2 FUEL TRANSFER
5.2.1 Mew Fuel Loading
New fuel is received and stored in the Service Building in a New Fuel Roomwhich can accommodate a complete reactor's inventory of new fuel. This isequivalent to about nine months' refuelling supply at 80% capacityfactor.
Fuel is transferred by pallet and lift truck to the new fuel loading area.Two New Fuel Transfer Mechanisms are installed there for inserting fuelinto fuelling machines in either maintenance area (Figure 5.2-1).
Fuel is hoisted from its pallet, inspected and loaded into a loadingtrough. Pairs of fuel bundles are pushed into the fuel loading magazineunder semi-automatic control. Subsequently they are transferred, by amotor driven ram under fully automatic control, into vacant magazinepositions in the waiting fuelling machine.
An airlock gate valve in the transfer port minimizes the transfer of anycontamination from the fuelling machine or maintenance room.
At all times except during fuel loading the new fuel port houses a shieldplug to reduce any radiation into the fuel loading room.
r;.2.2 Spent Fuel Discharge
One spent fuel port is mounted in each maintenance room for transfer ofspent fuel to the fuel storage bays (Figure 5.2-2).
As these ports constitute physical penetrations of the Reactor ContainmentBuilding, they each embody pairs of isolating valves. Drains, sprays andpressure relief devices are also provided for containment protection.
After the fuelling machine has been positioned and coupled to thespent fuel port, its heavy water level is lowered and its snout plugremoved and stored. Its magazine is then rotated for dry transfer of onepair of fuel bundles, through the port, onto the transfer mechanism forlowering into the fuel transfer canal.
5.1-3
TROLLEY TRACK
- MOTIONCATENARY LOOP
FUELLING MACHINEHEAD
FUELLING MACHINECARRIAGE
FUELLING MACHINEMAINTENANCE LOCK TRACK
CATENARY LOOP
HOSE AND CABLECARIER
CATENARY TROLLEY
TROLLEY DRIVEUNIT
HOSE AND CABLECARRIER TRACK
NORTH ('A') ASSEMBLY SHOWN
FIGURE 5.1-3 FUELLING MACHINE SUPPORT AND CATENARY SYSTEMS
NORTH CA1) NEW FUEL TRANSFER MECHANISMSOUTH CC1) NEW FUEL TRANSFER MECHANISMNEW FUEL PORTSCONTROL PANELSBUNDLE INSPECTION TABLEBUNDLE LOADING TROUGHS1/4 TON JIB CRANE AND AIR HOIST2-TON JIB CRANES AND ELECTRIC HOISTSREMOVABLE PLATFORMNEW FUEL PALLETSFUELLING MACHINE MAINTENANCE LOCK CRANE RAILSERVICE PORT ENCLOSURESIRRADIATED FUEL HANDLING SOUTH LADLE DRIVEFUELLING MACHINE TRANSPORT CART GUIDESLIQUID INJECTION SHUTDOWN SYSTEM ENCLOSURE15-TON CRANE
FIGURE 5.2-1 600 MW(e) NEW FUEL TRANSFER EQUIPMENT
1o3456789
1011121314
END FITTINGBALL VALVESELEVATING LADLE HOISTSELEVATING LADLE DRIVE (IN NEW FUEL ROOM)ELEVATING LADLESMAIN ELEVATOR RAILSGUIDE RAILSFUEL POSITIONING ASSEMBLIESLOWER RAIL SUPPORTAUXILIARIESSPRAY HEADERSREMOVABLE PLATFORMSFUEL TRANSFER EQUIPMENTDEFECTED FUEL CANNING EQUIPMENT
FIGURE 5.2-2 600 MW(>) IRRAOIATEO FUEL DISCHARGE EQUIPMENT
5.3 FUEL STORAGE
Fuel is routinely transported by underwater conveyor to the reception bayfor loading onto trays and for subsequent storage (Figure 5.3-1).
If there should be an indication of a defective fuel bundle, the suspectbundle is transferred to a carousel which collects any fission gas bubblesreleased. After a suitable period of cooling the defective fuel iscanned, then transferred to temporary storage.
The fuel storage bay capacity is sufficient to accommodate at least tenyear's output of spent fuel under normal conditions.
5.1-4
1 SPENT FUEL DISCHARGE EQUIPMENT(REF. ONLY)
2 TRANSFER RACK DETECTIONSWITCH LEVER
3 DISCHARGE BAY CONVEYOR4 TRANSFER CANAL CONVEYOR5 TRANSFER CART
6 CONVEYOR DRIVES7 RECEPTION BAY8 TRANSFER RACK9 TRANSFER RACK HANDLING TOOL
10 RACK HANDLING TOOLSTORAGE BRACKET
11 2-TON RECEPTION BAY CRANE12 SINGLE RACK STAND-OFF13 EMPTY RACKS ON TRIPLE RACK
STAND-OFF14 STORAGE TRAY STAND15 PARTIALLY FILLED TRAY ON STAND16 BUNDLE LIFTING TOOL17 FULL STORAGE TRAYS18 STORAGE TRAY CONVEYOR19 CONVEYOR DRIVE20 STORAGE TRA.V LIFTING TOOL21 SPENT FUEL STORAGE BAY22 EMPTY STORAGE TRAYS23 DEFECTED FUEL TRANSFER EQUIPMENT
(REF. ONLY)24 DEFECTED FUEL STORAGE BAY25 DEFECTED FUEL BAY ISOLATION
VALVE (REF. ONLY)26 ISOLATION VALVE DRIVE (REF. ONLY)
FIGURE 5.3-1 600 MW(«) FUEL STORAGE BAY EQUIPMENT
6.0 SUMMARY
6.1 ADVANTAGES OF CANDU
. CANDU is a proven technical product, a product that has put Canada aheadof all the countries of the Western World when the achievements of allthermal reactors are considered (Figure 6-1).
. CANDU is a conserver of uranium supplies, its once through fuel cycleuses 15% less natural uranium than light water reactor (LWR) fuelcycles. This fuel savings increases to 38% if 1.2% enriched uraniumfuel is used.
. CANDU is a flexible system, it can be adapted to advanced fuel cyclesemploying other fissile or fertile materials: thorium and plutonium.
. From a safety point of view the containment of the core environment inmany small diameter pressure tubes is preferable to one very large andheavy pressure vessel. The pressure tubes have been proven to exhibitthe "leak before break" characteristic that is so important in safetyconsiderations. The presence of heavy water is readily detectable andserves notice of any leak in the system. As pointed out previously theCANDU system uses a defence in depth approach with redundant safetysystems.
. On-power refuelling, a unique feature of the CANDU, permits theimmediate correction of low reactivity areas in the reactor core.Another advantage of this feature is its ability to quickly remove anyfailed fuel from the core while the reactor continues to operate.Shutdowns are only required for reactor maintenance.
. CANDU energy costs are competitive. The difference between coal(representative of fossil fuels) and nuclear costs in the Province ofOntario in 1979 is shown in Figure 6.2. These numbers have been broughtto a common base for direct comparison and .clearly show the CANDUadvantage.
6.2 CONCLUSION
This presentation provides a technical summary of the many systems thattogether make up a CANDU reactor, and at the same time outlines thereasons for CANDU1s superior position among the power reactors of theworld today.
While focusing primarily on the CANDU 600 MW(e), AECL is able to offercustomers larger reactor units (up to 950 MW(e)), if required.
6.1-1
WORLD POWER REACTOR LIFETIME PERFORMANCE
123456789
10
CANADAW. GERMANYCANADACANADA
CANADACANADACANADAUSAUSASWEDEN
Pickering-2Stade-1Pickering-1Bruce 4Bruce 3Pickering-4Pickering-3Point Beach 2Connecticut YankeeBarsebaeck2
542 MW662 MW542 MW791 MW
791 MW542 MW542 MW524 MW602 MW600 MW
84.5%83.5%83.3%78.5%
78.2%77.6%77.5%77.4%75.4%74.5%
CUMULATIVE LOAD FACTORS FOR REACTORS OVER 500 MW(e)TO END OF SEPTEMBER 1980
Station
Bruce-3Stacto-1Picfcerlnjj-2Pickwing-1Point BMCh-2Picker! nc-4Pick»ring-3Prairie ltland-2CaKwrt CllfU-2Connecticut YankeeBruce-4Bruce-1
Cumulative LoadFactor %
82.081.280.980.377.477.375.475.274.774.673.573.0
Type
CANDUPWRCANDUCANDUPWRCANDUCANDUPWRPWRPWRCANDUCANDU
REF: NUCLEAR ENGINEERING INTERNATIONAL VOL 25, NO. 307, 1980
Country
CanadaEuropeUSAJapanUKFranceW. Germany
Annual loadfactor%
70.1156.3556.7648.4151.7756.6350.49
Number and sizeof reactors
10 (5818 MWe)53 (38500.3 MWe)68 (54684 MWe)20 (13852 MWe)22 (7949.3 MWe)10 (6429 MWe)10 (14299 MWe)
Cumulative loadfactor %
64.9056.4854.7452.4053.2451.8754.95
Number and sizeof reactors
10 (5818 MWe)57 (42284.3 MWe)68 (54658 MWe)22 (15117 MWe)22 (7949.3 MWe)12 (8343 MWe)11 (15199 MWe)
Source: Nuclear Engineering International (March 1980)
FIGURE 6-1
COAL
FUEL
17.06
CAPITAL COST,OPERATION ANDMAINTENANCE
10.19
NUCLEAR
FUEL
1.7
CAPITAL COST,OPERATION,MAINTENANCEAND HEAVYWATER UPKEEP
12.8
27 .27 COSTftnfcWh 14.S
1979 BREAKDOWN OF UNIT ENERGY COSTS (ONTARIO HYDRO FIGURES)
FIGURE 6-2 COMPARISON OF COSTS — COAL AND NUCLEAR
