CANDU Nuclear Power System

.muary 1981 • UNRESTRICTED TDS1-10S I I I

* CANDU NUCLEAR POWER SYSTEM I I I I I I 1 I 1 I

I ATOMIC ENERGY OF CANADA LIMITED ENGINEERING COMPANY 1 SHERIDAN PARK RESEARCH COMMUNITY MISSISSAUGA, ONTARIO L5K 1B2 1901 JANUARY 1 F ™ UNRESTRICTED TDS1 -105 I I

| CANDU NUCLEAR POWER SYSTEM I

• ABSTRACT

I This report provides a comprehensive summary of the many components that make up a CANDU reactor. Major emphasis 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. I I I I I I I I I 1 I I

I TABLE OF CONTENTS I 1. INTRODUCTION 1.1 Introduction to AECL 1.2 History and Performance of CANDU I 1.3 Layout and Component Parts of CANDU I 2. POWER SYSTEMS 2.1 Reactor Core 2.2 Heat Transport Systems I 2.3 Overall Plant Control 2.4 Core Control 2.5 Reactivity Control Devices 2.6 Fuel I 2.7 Startup, Operation and Shutdown Sequences 2.8 Secondary Side Systems I 3. MODERATOR AND AUXILIARY SYSTEMS 3.1 Main Moderator System I 3.2 Moderator Auxiliary Systems 3.3 Heavy Water Management I 4. SAFETY SYSTEMS 4.1 Inherent Safety Features of CANDU 4.2 Safety Design Philosophy I 4.3 Safety Systems Description I 5. REFUELLING SYSTEM 5.1 Fuelling Machines 5.2 Fuel Transfer I 5.3 Fuel Storage 6. SUMMARY I 6.1 Advantages of CANDU I 6.2 Conclusion r • I I

I 1.0 INTRODUCTION I 1.1 INTRODUCTION TO AECL AECL is a crown corporation of the Canadian government charged with the mandate to develop nuclear power and associated industries in Canada. AECL commenced operation in 1945 with the construction of elaborate I nuclear 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, and I the structure of the company as it is presently constituted, is shown in Figure 1.1-2. I The company's major facilities in Canada (Figure 1.1-3), are: . The Head Office, located in Ottawa, Ontario

I . Nuclear laboratories at Chalk River, Ontario and Pinawa, Manitoba -

. Douglas Point and Gentilly-1 nuclear generating stations, located in I Ontario and Quebec respectively . The Glace Bay and Port Hawkesbury heavy water plants in Nova Scotia, and I the partially completed La Prade heavy water plant in Quebec . The Radiochemical Company, located in Ottawa, markets radioisotopes and I designs, manufactures and markets radiation equipment . The Chemical Company, located in Ottawa, responsible for heavy water r production . The Engineering Company, located in Mississauga with a subsidiary office in Montreal, responsible for the engineering of nuclear power reactors

» AECL International located at Mississauga and elsewhere is responsible for overseas projects and international marketing

AECL today has world-wide experience in design, project management and construction of large nuclear generating stations and research reactors. AECL is also experienced in transferring CANDD expertise from Canada to participating countries through the establishment of local manufacturing programs. AECL has a genuine interest in this transfer of technology and in the generation of local employment in participating countries. r r 1.1-1 I I 1 ] • I I 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. 1 1 I 1 I I I FIGURE 1.1-1 I I ATOMIC ENERGY OF CANADA LIMITED CORPORATE OFFICE

I • Directs and administers the Company's activities.

• Markets CANDU nuclear reactors, components and technology.

• Effects scientific and technological exchange agreements with counterpart agencies in I other countries. • Makes available its special facilities and expertise to assist Utilities n the practical use I of nuclear energy, and other Government agencies In their operation and services. I RESEARCH COMPANY ENGINEERING COMPANY • Designs nuclear generating stations, in co- • Operates laboratories for fundamental and operation with electric utlities and private applied research and engineering development industry. in the nuclear field. I • Provides nuclear power plant consulting ser- • Enters into co-operative research and develop- vices and undertakes development work In ment contracts with industry and universities. support ot the CANDU nuclear power plant. * Provides nuclear steam plant equipment and • Makes available its special facilities and ex- makes available its special facilities and ex- pertise to assist universities in nuc ear studies I pertise to assist in developing manufacturing and techniques. capability to nuclear specifications. I CHEMICAL COMPANY RADIOCHEMICAL COMPANY

• Constructs and operates Heavy Water Produc- I tion Plants. • Produces and markets radioisotopes. • Provides heavy water for nuclear reactor re- • Designs, manufactures, and markets equip- quirements. ment for the utilization of radioisotopes and radiation. I • Coordinates the development of Heavy Water technology. I AECL INTERNATIONAL

I • Identifies offshore sales opportunities. • Formulates marketing strategies in order to develop new reactor and associated technology sales overseas. • Promotes CANDU export sales to these newly I defined markets, and to existing markets (in cooperation with the organization of CANDU industries O.C.I.) • Represents the Chemical Company and the I Research Company in offshore sales. • Constructs nuclear generating stations n cooperation with international electrlca I utilities and local industries. I r FIGURE 1.1-2 ATOMIC ENERGY OF CANADA LIMITED ORGANIZATION/RESPONSIBILITIES 1 Whitnnall Nuclear RtMarch Establishment 2 Bme* Nuclear Power Development Douglas Point Nuclear Generating Station 3 AECL Engineering Company AECL International 4 Picturing Nuclear Generating Station 9 AECL Engineering Company-Montreal ; 5 DarUngton Nuclear Generating Station 10 Gentilly-1 Nuclear Generating Station : Gentilly-2 Nuclear Generating Station • 6 Nuclear Power Demonstration (NPD) La Prade Heavy Water Plant : 7 Chalk River Nuclear Laboratories 11 Point Lepreau Nuclear Generating Station : S AECL Corporate Office 12 Port Hawkesbury Heavy Water Plant : AECL International Head Office AECL Research Company AECL Radiochemical Company AECL Chemical Company 13 Glace Bay Heavy Water Plant ;

FIGURE 1.1-3 NUCLEAR ENERGY ESTABLISHMENTS IN CANADA

1 f^^'vr* ^ (--^Frt-ii -IJ i I I I 1.2 HISTOK? AND PERFORMANCE OF CANDU The history of CANDU is illustrated in Figure 1.2-1, and shows the genealogy of the CANDU reactor from its inception to the current state of I the art. Figure 1.2-2 summarizes this graph in tabular form. The Pickering "A" nuclear generating station (Figure 1.2-3) consists of I four units of 515 MW(e) electric each, for a total net output of 2060 MW(e). It is situated on the outskirts of, and provides power to, the city of Toronto, Ontario. I These units came into service during the years 1971 to 1973. Adjacent to the Pickering "A" station is the Pickering "B" station, again comprising four units, which will deliver 2060 HW(e) when completed during the years I 1981 to 1983. The Bruce "A" nuclear generating station (Figure 1.2-4) is located approximately 160 miles north-west of Toronto, and consists of four units of 746 MW(e) each, for a total station output of 2984 MW(e) I electric, plus enough steam to feed the Bruce heavy water plants. These units came into service during the years 1977 to 1979 and hence form the newest additions to the Canadian nuclear grid. On the same site, the Bruce "B" nuclear generating station is under construction with its four I units of 769 MW(e) electric each, due in service during the years 1984 to 1987. I Ontario Hydro has also commenced construction of a four unit nuclear generating station of 850 HW(e) per unit at the Darlington site I approximately 60 miles east of Toronto. Currently, more than 31% of the electrical demand in the Province of Ontario is generated by nuclear power.

I The Gentilly-2 (G-2) nuclear generating station (Figure 1.2-5) is a 638 MW(e) net single unit station under construction adjacent to the Gentilly-1 nuclear generating station, the only CANDU station with boiling I light water coolant. The Point Lepreau generating station is under construction in the Province I of New Brunswick. It is very similar to the Gentilly-2 station and will produce 633 MW(e) net.

CANDU stations are also operating or under construction in other I countries. Three of these are shown in Figures 1.2-6, 7 and 8.

CANDU stations are the world leaders in availability. These facts are I illustrated in Figures 1.2-9 and 1.2-10 which show extracts from 'Nuclear Engineering International1 comparing the CANDU system against other I competing reactor types. 1 1.1-2 800

Length and location of rectangles denotes 700 • construction and commissioning period. Arrows denote flow of information.

600

POWERS 500 REACTORS MW(e)

RESEARCH REACTORS MW(Ih)

General research, development and design information.

1945 1950 1955 1960 1965 1970 1975 1980 1985 1990 YEARS FIGURE 1.2-1 GENEALOGY OF CANDU REACTORS POWER DATE OF MWe NUCLEAR FIRST NAME LOCATION TYPE NET DESIGNER POWER

NPD ONTARIO PHW 22 AECL & CGE 1962 DOUGLAS POINT ONTARIO PHW 206 AECL 1967 PICKERING A ONTARIO PHW 515x4 AECL 1971/73 GENTIuLY 1 QUEBEC BLW 266 AECL 1971 KANUPP PAKISTAN PHW 125 CGE 1971 RAPP1 INDIA PHW 203 AECL 1972 RAPP2 INDIA PHW 203 AECL — BRUCE A ONTARIO PHW 740x4 AECL 1976/79 GENTILLY 2 QUEBEC PHW 640 AECL — POINT LEPREAU NEW BRUNSWICK PHW 635 AECL — CORDOBA ARGENTINA PHW 600 AECL — PICKERING B ONTARIO PHW 516x4 AECL — WOLSUNG 1 KOREA PHW 600 AECL — BRUCE B ONTARIO PHW 756x4 AECL — DARLINGTON ONTARIO PHW 850x4 AECL — CERNAVODA ROMANIA PHW 600 AECL _ TOTAL 18,208 MWe

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 I I I I

I Station Cumulative Load Type I Factor % Bruce-3 82.0 CANDU Stade-1 81.2 PWR I Pickering-2 80.9 CANDU Pickering-1 80.3 CANDU Point Beach-2 77.4 PWR I Pickering-4 77.3 CANDU Pickering-3 75.4 CANDU Prairie lsland-2 75.2 PWR I Calvert Cliffs-2 74.7 PWR Connecticut Yankee 74.6 PWR I Bruce-4 73.5 CANDU Bruce-1 73.0 CANDU I Ref: Nuclear Engineering International Vol. 25 No. 307,1980 I I I

I I FIGURE 1.2-10 CUMULATIVE LOAD FACTORS FOR REACTORS OVER 500 MW(e) I TO END OF SEPTEMBER 1960 i 1

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 I supply. "

Figure 1.3-2 shows the Nuclear Steam Plant and Balance of Plant portions. I Figure 1.3-3 illustrates containment, which consists solely of the reactor 1 building. Figure 1.3-4 shows the CANDU Nuclear Steam Supply System inside containment and Figure 1.3-5 shows pictorially the layout inside "t containment* Figure 1.3-6 illustrates the control room which is in the J service 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) Nuclear Generating Station which is a feature of the plant. The reactor building .1 layout is identical in both units and the service building layout is 1 adjusted to accommodate the dual unit configuration. 1 I I I I I 1

1.1-3 CANDU 900 NUCLEAR GENERATING STATION (NGS)

NUCLEAR STEAM BALANCE OF PLANT (NSP) PLANT (BOP)

, TURBINE GENERATOR AND AUXILIARIES

NUCLEAR STEAM SUPPLY BALANCE OF NUCLEAR ELECTRIC POWER SYSTEM (NSSS) STEAM PLANT (BNSP) SYSTEMS (BOP)

, COMMON PROCESSES I- REACTOR BUILDING AND SERVICES (BOP) I- SERVICE BUILDING *~ SPENT FUEL BAY(SFB) , BUILDINGS AND STRUCTURES

NUCLEAR CORE STEAM GENERATING BALANCE OF NUCLEAR SYSTEM (NCS) SYSTEM (SOS) STEAM SUPPLY SYSTEM (BNSSS)

PRIMARY HEAT I- HEAVY WATER MANAGEMENT TRANSPORT SYSTEM I- ELECTRIC POWER SYSTEMS (NSSS) STEAM AND WATER *- COMMON PROCESSES AND SERVICES (NSSS) BALANCE OF NUCLEAR SYSTEMS (NSP) & SPENT FUEL BAY PROCESS SYSTEMS NUCLEAR CORE CORE(NC) SYSTEM (BNCS)

• REACTOR MODERATOR SYSTEM • FUEL HANDLING AUXILIARY SYSTEMS •CONTROL CENTRE (NC) t EWS, EPS

FIGURE 1.3-1 COMPONENT PARTS OF THE PLANT 1 REACTOR 7 SPENT FUEL INSPECTION AND STORAGE 2 FUELLING MACHINE 8 MAINTENANCE AND INSPECTION 3 STEAM GENERATOR 9 CONTROL CENTRE 4 DOUSING SYSTEM 10 MECHANICAL WORKSHOP 5 NEW FUEL HANDLING 11 WATER TREATMENT 6 SPENT FUEL HANDLING 12 TURBINE AND GENERATOR

FIGURE 1.3-2 600 MW(e) NUCLEAR GENERATING STATION I I I I I I I I I I I I I I I

I SEPARATION MEMBRANE I T i FIGURE 1.3-3 REACTOR BUILDING SECTION I I I I I I i 0 1 1 I 1 1

1 MAIN STEAM SUPPLY PIPING I 2 STEAM GENERATORS 3 MAIN PRIMARY SYSTEM PUMPS 4 FEEDERS T 5 CALANDRIA ASSEMBLY 6 FUEL CHANNEL ASSEMBLY 7 FUELLING MACHINE BRIDGE 8 MODERATOR CIRCULATION SYSTEM

FIGURE 1.3-4 NUCLEAR CORE SYSTEM AND STEAM GENERATING PLANT I I I I I I I I I I I I I I I

DOUSING WATER TANK DOUSING WATER VALVES FUELLING MACHINE CARRIAGE I MODERATOR PUMP FUELLING MACHINE CATENARY MODERATOR HEAT EXCHANGER FUELLING MACHINE MAINTENANCE LOCK FEEDER CABINETS FUELLING MACHINE MAINTENANCE LOCK DOOR REACTOR FACE END SHIELD COOLING WATER DELAY TANK 1 REACTOR VAULT COOLER REACTIVITY MECHANISM PRESSURIZER 9 HEAT TRANSPORT SYSTEM PUMP 18 STEAM GENERATOR 10 FUELLING MACHINE BRIDGE 19 STEAM GENERATOR ROOM CRANE

FIGURE 1.3-5 600 MW(e) REACTOR BUILDING CUTAWAY EMERGENCY CONTAINMENT CORE MODERATOR AND MISCELLANEOUS REACTOR MISCELLANEOUS PRIMARY HEAT COOLING AUXILIARY SYSTEMS TRANSPORT SYSTEM SYSTEMS FUELLING MACHINE ANNUNCIATION ELECTRICAL AND FUEL HANDLING AND DIGITAL STEAM GENERATOR DISTRIBUTION CONTROL CONSOLE CONTROL SYSTEM TURBINE GENERATOR SWITCHYARD SYSTEMS COMPUTERS / / / I /I PL14 PL15 PL16 PL17 PL18 PL19 PL20 I I 1 I I i I

FIGURE 1.3-6 CONTROL CENTRE

• I ' . i ' I I I I (0 I I I o I < I 2 I oc I i I 5 I I I I o I 1 = UJ g o 5g UJ j* u § 3 s < S 5L5

»- (NJ CO * ID < F/M MAINTENANCE

F/M DECONTAMINATION

DECONTAMINATION CENTRE .

ERADIATION PROTECTION

FIGURE 1.3-8 SERVICE BUILDING PLAN EL. 100 (GRADE) I I

I 2.0 POWER SYSTEMS 2.1 REACTOR CORE

I 2.1.0 General

This section of the CfiNDU Nuclear Power System presentation introduces the I principal features of the Reactor Core and its location arrangements. Particulars of the related process and control systems are covered in subsequent sections.

I The CANDU 600 MW(e) Reactor is the sixth in a series of Pressurized Heavy Water Reactor (PHWR) designs developed in Canada for the production of I electric power from natural uranium fuel. Like preceding CANDU reactors, this design incorporates a standardized, geometrical arrangement of horizontal pressure tubes which contain fuel I and circulating heavy water coolant at high pressure. These fuel channels are mounted within a cylindrical calandria, which I contains heavy water moderator in a separate low pressure system. The CANDU 600 MW(e) Reactor Core is located at the heart of the Reactor I Containment building within biological shielding. 2.1.1 Reactor Assembly I The Reactor Assembly is mounted inside a steel-lined light water filled concrete vault (Figure 2.1-2) and comprises: I . A cylindrical low-pressure calandria vessel of stainless steel construction.

. Two integral end shields (also of stainless steel construction with I carbon steel shielding balls) easjh horizontally penetrated by 380 lattice tubes.

I . 380 Zircaloy-2 calandria tubes joining the lattice tubes at each position in the lattice.

. 3S0 fuel channel assemblies mounted within these lattice sites.

. Vertical and horizontal reactivity control devices which penetrate the vault shielding to provide power sensing, control and shut-down I features.

. Connections for the heat removal recirculation of the heavy water r moderator.

E 2.1-1 I I I I I I I I I I I I I I

CALANORIA 16. EARTHQUAKE RESTRAINT CALANDRIA - SIDE TUBESHEET 17. CALANDRI A VAULT WALL 3. CALANDRIA TUBES 18. MODERATOR EXPANSION TO HEAD TANK I 4. EMBEDMENT RING 19. CURTAIN SHIELDING SLABS 5. FUELLING MACHINE - SIDE TUBESHEET 20. PRESSURE RELIEF PIPES 6. ENO SHIELD LATTICE TUBES 21. RUPTURE DISC 7. END SHIELD COOLING PIPES 22. REACTIVITY CONTROL UNIT NOZZLES S. INLET-OUTLET STRAINER 23. VIEWING PORT I 9. STEEL BALL SHIELDING 24. SHUTOFF UNIT 10. END FITTINGS 2S. ADJUSTER UNIT 11. FEEDER PIPES 26. CONTROL ABSORBER UNIT 12. MODERATOR OUTLET 27. ZONE CONTROL UNIT 13. MODERATOR INLET 2B. VERTICAL FLUX OETECTOR UNIT 14. HORIZONTAL FLUX DETECTOR UNIT 29. LIQUID INJECTION SHUTDOWN NOZZLE I IS. ION CHAMBER 30. BALL FILLING PIPE I! FIGURE 2.1-1 REACTOR ASSEMBLY I I

I . Cooling connections for the water within the end shields and calandria vault• I The transverse reactivity devices (Figure 2.1-2) are: 1) Six externally mounted ion chamber assemblies, 3 each side, which I sense low levels of neutron flux, 2) Seven Horizontal Flux Detector Assemblies which sense flux levels in I regions of the reactor core, 3) Six Poison Injector Nozzle Assemblies which provide rapid injection of neutron-absorbing gadolinium nitrate solution into the moderator I when trip sensors in Eihutdown System No. 2_ (SDS2) are actuated. Vertical penetrations from the reactor deck (Figure 2.1-3) provide access, I through thimbles, to the calandria, for positioning the other reactivity control devices. I The vertical reactivity devices (Figure 2.1-4) are: 26 Vertical Flux Detector Units I These embody flux sensors which provide inputs to: 1) The Reactor Regulating System which controls power levels in various I regions of the reactor, 2) The Reactor Protective System to actuate Shutdown System No. J. (SDS1) I in the event of excessive power indications. 3) The Flux Mapping System which is used to record local power levels to I identify zones where refuelling would be most appropriate. 6 Liquid Zone Control Units which provide a total of 14 compartments in which light water levels are varied in response to control requirements of I the Reactor Regulating System. 21 Adjuster Units which serve the dual functions of: r 1) Flux flattening during normal operation through neutron-absorption in inserted absorbers,

2) Xenon override following power reduction or shutdown, through absorber withdrawal.

28 Shut-off Units which drop neutron absorbing rods into the reactor core r to shut down the reactor when a trip is actuated in the Reactor Protective System's SDSL

2.1-2 I 1 1 1 < 1 ICL IC2, IC3 ON FAR SIDE ONLY 1 I I (SIDE'C') 1 1 1 1 ! 1 •

IC ION CHAMBER HOUSING ( • J HORIZONTAL FLUX DETECTOR (7) I (3 EACH SIDE) LI POISON INJECTOR NOZZLES (6) I I 1 FIGURE 2.1.2 TRANSVERSELY MOUNTED REACTIVITY CONTROL DEVICES ""i

STEAM GENERATOR MOUNTING POST ROOM FLOOR LEVEL ADJUSTER UNIT. SHUTOFF UNIT VERTICAL FLUX FLUX DETECTOR DETECTOR HELIUM COVER VIEWING TREADPLATE MOUNTINGo 1 LIQUID ZONE UNIT CONTROL & UNIT GAS LINES PORT POST CONTROL UNIT / POWER CABLES

1 r SHIELDING • CABLE COLLAR \H TRAYS '' *• 1 |THIOKOLP-STYROFOAM SEALANT!

CALANDRIA t • . CALANDRIA VAULT WALL'i VAULT

NITROGEN GROUT SHIMS -SEAL PLATE . VAULT WATER

' DOUBLER PLATE

FIGURE 2.1.3 REACTIVITY MECHANISM DECK - SCHEMATIC SECTIONAL VIEW E I I I • I O FUEL CHANNEL I I O ®$®&® O I I I I I 1 I ' I 5 VERTICAL FLUX DETECTOR (26) LIQUID ZONE CONTROL (6) ADJUSTER (21) ^ SOLID CONTROL ABSORBER (4) SHUTOFF UNIT (28) © VIEW PORT (2) I OVERFLOW « HELIUM BALANCE

FIGURE 2.1-4 PLAN - VERTICALLY MOUNTED REACTIVITY CONTROL DEVICES I I These vertical reactivity control devices are positioned in the calandria I in guide tubes which pass through the thimbles and the calandria tube lattice^ and are secured at the bottom of the calandria shell.

I 2.1.2 Fuel Channel Assemblies I The fuel channel assemblies (Figure 2.1-5) consist of: 1) Zirconium-niobium alloy Pressure Tubes (6.3 m long x 105 nun nominal bore x 4.16 mm minimum wall thickness) to house fuel and pressurized I D2O, 2) AISI type 403 Stainless Steel End Fittings, with type 410 stainless I steel 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 in I the end shield,

4) Garter spring Tube Spacers to support each pressure tube within its I calandria tube, 5) Bellows Assemblies to seal the annular space between each pressure I tube and its calandria tube (and between the end fittings and end shield lattice tubes),

6) Shield Plugs for every end fitting to minimize neutron leakage from I the fuel channel and (in the case of the downstream shield plug) to provide axial support to the column of 12 fuel bundles - (see Section I 2.6), 7) Removable Closure Plugs (Figure 2.1-6) to seal each end of the fuel I channels and to enable access for refuelling by the fuelling machine, 8) Feeder Connections to the Heat Transport System for the supply and removal of D2O coolant for each fuel channel.

I Fuel channels are installed at the reactor site by a highly trained crew in a closely controlled production operation. I Subsequently the feeders are installed to connect fuel channels to the t Heat Transport circuit, described in the next section. I I 2.1-3 n

FIXED END OF CHANNEL

1 CHANNEL CLOSURE 2 CLOSURE SEAL INSERT 3 FEEDER COUPLING 4 LINER TUBE 5 END FITTING BODY 6 END FITTING BEARING 7 TUBE SPACER 8 FUEL BUNDLE 9 PRESSURE TUBE 10 CALANDRIA TUBE 11 CALANDRIA SIDE TUBE SHEET 12 END SHIELD LATTICE TUBE 13 SHIELD PLUG 14 END SHIELD SHIELDING BALLS 15 FUELLING MACHINE SIDE TUBE SHEET 16 CHANNEL ANNULUS BELLOWS 17 CHANNEL POSITIONING ASSEMBLY

FIGURE 2.1-5 FUEL CHANNEL ASSEMBLY I I 1 FRONT HOUSING 2 REAR HOUSING 3 SPRING I 4 PLUNGER 5 STEM END 6 JAW 7 TOGGLE I 8 CAP SCREW 9 SEAL DISC PIN 10 SAFETY LATCH SPRING 11 SAFETY LATCH I 12 SEAL DISC 13 SPIDER I 14 STEM I I SAFETY LATCH LOCKED SAFETY LATCH UNLOCKED I VIEW 2 VIEW 3 I I 1 SECTION SHOWING THE JAWS AND SPIDER 2 SECTION SHOWING THE SAFETY MECHANISM SPRINGS. THE RAM ASSEMBLY HAS JUST CON- AND THE CAP SCREWS. THE SAFETY LATCHES 1 TACTED THE REAR HOUSING, ADVANCING THE ARE IN THEIR LOCKED POSITION PREVENTING SEAL DISC 0.9 mm. THE ACCIDENTAL DEPRESSION OF THE STEM. I I I

I 3 HERE THE LATCH RAM HAS ADVANCED 12.7 mm 4 THE LATCH RAM AND 'C RAM HAVE BOTH TO UNLOCK THE SAFETY MECHANISM BY MOVED A FURTHER 21 mm TO COMPLETELY I PUSHING THE FOUR SAFETY LATCHES INWARD. RETRACT THE FOUR JAWS. r

FIGURE 2.1-6 FUEL CHANNEL CLOSURE PLUG I 1

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 I transverse reactivity control devices. The latter are described in * greater detail in Sections 2.4 and 2.5. 1 2.2 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 calandria is filled with heavy water (DjO) which moderates or slows the fast neutrons, making a chain reaction possible. The heat of fissicn generated X within the fuel is removed by the pressurized heavy water coolant which is II pumped through the fuel channels. This hot coolant is passed through the steam generator where heat is transferred to light water to generate IB steam. 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 I bundles of natural uranium fuel located within the pressure tube. The calandria tube is in contact with the moderator. The annular space between the pressure tube and the calandria tube provides thermal I insulation between the hot heat transport system coolant and the cool I moderator. I The portions of the fuel channel assemblies external to the calandria | (Figure 2.2-3) are known as the end fittings; the end fittings have connections 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 ~f array 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 I I STEAM LINES I I Jc: I I I I I I I I

FUEL CHANNEL I ASSEMBLIES

I NATURAL URANIUM FUEL LIGHT WATER STEAM I LIGHT WATER CONDENSATE HEAVY WATER COOLANT

I HEAVY WATER MODERATOR I MODERATOR HEAT EXCHANGER

FIGURE 2.2.1 CANDU NUCLEAR STEAM SUPPLY SYSTEM 1 1 1 1 1 1 1 1 1 1 ,I;I;I;I;I|I|I;I;I;I;I;I;I; I,'!!!'!!'!!'!!'!'!!',', ,i I i i I I i I I i i I i I I in 1111111111111 I il I II I II I I I I II II IIIIIIIIIIIIIIII IIIIIIIIIIIIIIIi!i!i!i!i!i!i!i!i!i!i!i!i!i!i!Ii il I I I IilI I I I II I I I .IIIIIIIIIIIIIIII il I ilIilI I I I il I I I i 111111111111111 il IIII II I Hi il 11 I II IIIIII III I I! I I 11 i 11 i 11 i i i i 11 i 11 II II I I II I I I I II I 1 il I I I I I I I I I I il il I I I I I I I II I I Illll! .11 I ill IIIIII il ill I I I i I I i I I I I I I I I I I IIIIIIIIIIIIIIII ,11 i I I i i I i i i I i i i i i .lilllil IIIIII lilli I I I I I ilI I I I II I I II I 'i'l'l'i'!'!'!'!'!'!'!'!'!'!'!'! 1 1111 hi 11111111 II i i!J!i!!J!i!J!i!!!JjJ J IIII Hi Hi Hi i II i; 11111111111111111 .'I!'!'!!!'!'!!!!!!111111111 hi hi !H III; 11111 II 111 I 'ii'i i" ' I I i" '< I I I IIIIIIIIIII I I I I I I I I I I I I I I uli I i I IIIIIIII I I I II I I II ll I I III I I I 1111111111 IIIII' .1 I II I I I I I II II l!l!l I I I 111111111111111 1111IIII11 II i II 1111 II 1111111111 1 111111111111111 II II II 11 II Hi 11

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FIELD WELD END SHIELD

TUBE SHEET CALANDRIA TUBE

GAS ANNULUS (Between Lines)

SPACER FUEL BUNDLES

\ T PRESSURE TUBE (Inner) A SHIELD PLUG

CALANDRIA

FIGURE 2.2-3 REACTOR CORE SCHEMATIC SAFETY VALVES ££ TO SUPPRESSION TANKS

37170O0kgft 260"C ___^ 4690KN;m2 IAI 2S

2 MOlSTUilt SFPAHATOR HEHEAHRS

.P. I L.P TURBINE TURBINE CiENtRATCin

PRESSURIZER

187"c

266°C 27.U5O.OOO k3/!, 310°C ' 9990 KN/m2(AI !I.Z30KN/m2(A(

Q n

nOstfa HtAVY WATER MODERATOR FUELLING MACHINE HEAVY WATER COOLANT

STEAM

CONDENSATE

H1VFH VVATLR

MODERATOR HEAT EXCHANGEB FIGURE 2.2-4 CANDU NUCLEAR POWER SYSTEM

*r4 r-r™-E»« I I

The circuits are shown in Figure 2.2-5; each circuit contains 2 pumps, 2 I steam generators, 2 inlet headers and 2 outlet headers in a 'figure-of- eight' arrangement. Feeders connect the inlet and outlet of the fuel I channels to the inlet and outlet headers respectively. The flow through the fuel channels is bidirectional (i.e. opposite directions in adjacent channels). The fee<":>:s are sized such that the I coolant flow to each channel is proportional to channel power. The enthalpy increase of the coolant is therefore the same for each fuel channel assembly.

I One of the advantages of this 'figure-of-eight' arrangement is that in the event of a heat transport pump failure, the coolant flow in the circuit is maintained at approximately 70% of the normal value, thereby permitting I continued reactor operation at reduced power. The arrangement of the heat transport system within the reactor building I is illustrated in Figures 2.2-6 and 2.2-7. The steam generators, HTS pumps and headers are located above the reactor; this permits the heat transport 3ystem coolant to be drained to the header elevation for maintenance of the HTS pumps and.steam generators, and also facilitates I thermosyphoning (natural circulation) when the HTS pumps are unavailable. I 2.2.?..?. Heat Transport System Conditions Heavy water (D2O) is utilized as the reactor coolant; the principle advantage of heavy water is its low neutron absorption. The variation of saturation temperature with pressure for D?O is shown in (Figure I 2.2-8). The heat transport system operating pressure is one of the key elements in optimizing the CANDU cycle; high primary pressure permits high secondary pressures and increased unit efficiency. They also, however, 1 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 is 1 utilized, leading to an outlet header quality of up to approximately 4% at full power. Figures 2.2-9 and 2.2-10 illustrate a typical steam generator and its heat load diagram respectively. The feedwater enters the preheater section of the steam generator secondary side, is warmed to I saturation temperature, and is then evaporated to produce steam. The heavy water coolant enters the steam generator U-tubes opposite the preheater section; the heavy water vapour is condensed, and the liquid I D2O progressively cooled, until it leaves the U-bend near the preheater section entrance. The small amount of vapour in the heavy water coolant entering the steam generator increases the Log Mean Temperature Differences (LMTD) across the stem generator, and improves steam 1 generator perfomance. i

2.1-5 r STEAM GENERATOR STEAM GENERATOR

1 :"\i i PUMF 1>UMP n » i i I W T 1 u \ m

OU1rLET HEADER INLET HEADER INLET HEADER OUTLET HEA[)ER " Lr - "1 r 1 r- —| 1 I> LOOP L,

REACTOR

OUTLET LOOP OUTLET LJ - STEAM GEh ERATOR GENEFIATOF S—-s L| y INLET INLET HEADER HEADER

PUMP PUMP 1 1 1 j 1 1 1 1 i i

FIGURE 2.2-5 A HEAT TRANSPORT SYSTEM I I I I I I I I I I I I I I I I

1 STEAM GENERATORS I 2 HTS PUMPS 3 REACTOR 1 FIGURE 2.2-6 LOCATION OF HEAT TRANSPORT SYSTEM EQUIPMENT i I

1

1 1

1

OUTLET HEADER INLET HEADER FEEDERS STEAM GENERATORS END FITTINGS HEAT TRANSPORT PUMPS INSULATION CABINET

FIGURE 2.2-7 FEEDER AND HEADER ARRANGEMENT I I I I I I PRESSURE/TEMPERATURE — D2O

I 12

OPERATING POINT REACTOR OUTLET - I HEADER I I

I I 200 300 400 I TEMPERATURE CO 1 I I 1 1 I r FIGURE 248 VARIATION OF SATURATION PRESSURE WITH TEMPERATURE FOR HEAVY WATER I STEAM OUTLET 1

STEAM SEPARATION 1 SCREEN 1 STEAM DRUM 1

HIGH CAPACITY CYCLONES 1

U-TUBES I 1 SHROUD I TUBE SUPPORT PLATES I 1 PHEHEATER SECTION I I FEEDWATER INLET I I

D2OOUT I D2OIN I FIGURE 2.2-9 TYPICAL STEAM GENERATOR O

266 111 Q_

100 PERCENTAGE HEAT TRANSFERRED

Data presented Is tor AECL 600 MW(e) reactor

FIGURE 2.2-10 STEAM GENERATOR HEAT LOAD DIAGRAM I

A simplified Heat Transport System flowsheet is shown in Figure 2.2-11. I HTS parameters are summarized in Figure 2.2-12. This figure also presents • data for Douglas Point, Pickering and Bruce/ and illustrates the evaluation 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 single suction 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 I series. The seals are provided with a cool, clean flow of D2O during normal operation via the gland seal system (Figure 2.2-15). Cooling water ~l is 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 I feedwater enters the preheater section of the steam generator, which encompasses the lower portion of the cold leg of the tube bundle. The two phase light water flow rising from the U-tube region of the steam F generator is passed through cyclone separators and secondary scrubbers to -t assure that the moisture content of steam leaving the steam generator is less than 0.25%. The liquid removed from the steam is returned to the T tube sheet region of the steam generator via the annular downcomer. The X circulation ratio for CANDU steam generators is approximately 5 to 1.

Steam generators in caNDCJ nuclear steam supply system, have an excellent I operating history. The tube defect rate for CANDU systems is 0.001% per year 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 which connect to the feeders are cold drawn from the parent header material. T i Figure 2.2-18 is a photograph of an actual installation and shows the feeder 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) is controlled by 'feeding' D2O into, or 'bleeding' D2O out of, the HTS system. At power, the HTS pressure is controlled by a pressurizer \ connected to the two HTS circuits. Heat is added to the pressurizer via I electric heaters to increase pressure and is removed via steam bleed to 1 2,1-6 STEAM FLOW (TYPICAL)

STEAM GENERATOR

/ FEEDWATER FLOW (TYPICAL) r

FIGURE 2.2-11 HEAT TRANSPORT SYSTEM FLOWSHEET DOUGLAS REACTOR POINT PICKERING BRUCE GENTILLY-2

NUMBER OF ELEMENTS PER BUNDLE 19 28 37 37

OPERATING CONDITIONS

COOLANT D2O D2O D2O D2O NOMINAL INLET PRESSURE MN/mz 9.8 9.8 9.3 11.09 PRESSURE DROP/CHANNEL (CRUD FREE) kNlmd 738 565 738 758 BUNDLES/CHANNEL 10 12 13 12 MAXIMUM CHANNEL POWER MW 2.743 5.125 5.74 6.5 INLET TEMPERATURE °C 249 249 252/256* 266.4 OUTLET TEMPERATURE °C 293 293 298.9 312.3 EXIT STEAM QUALITY % 0/3.5* 2.9 MAX. MASS FLOW/CHANNEL kg/s 12.6 23.8 23.8 23.94

* INNER ZONE/OUTER ZONE

FIGURE 2.2-12 HEAT TRANSPORT SYSTEM PARAMETERS I I I I I I I I I I I

1 UPPER OIL POT COVER 17 MOTOR FLYWHEEL I 2 THRUST BEARING OIL POT 18 STATORCORE 3 RUNNER 19 ROTOR ASSEMBLY 4 THRUST BEARNG ASSEMBLY 20 LOWER GUIDE BEARING 5 DOWN THRUST BEARING 21 THRUST DISC I 6 UP THRUST BEARING 22 SPACER COUPLING 7 THRUST BEARING COOLING COILS 23 MOTOR STAND 8 BRAKE RING 24 PUMP SHAFT 9 MOTOR SHAFT 25 VAPOUR CONTAINMENT SEAL I 10 OIL LEVEL CONTROL 26 SECONDARY MECHANICAL SEAL 11 BEARING COOLING WATER PIPES 27 PRIMARY MECHANICAL SEAL 12 AIR COOLER WATER PIPES 28 PUMP BEARING 13 SURGE CABINET 29 PUMP CASE I 14 AIR SHIELD 30 CASE WEAR RING 15 AIR SHIELD 31 PUMP DISCHARGE I 16 BLOWER RINGS 32 SUCTION PIPE 1

FIGURE 2.2-13 HEAT TRANSPORT SYSTEM PUMP 1 1 1

DRAIN TO LEAKAGE COLLECTION 1 VENT TO LEAKAGE SEGMENTED COLLECTION CARBON BACK-UP SEAL 1 I 1 I 1 A TO LEAKAGE A COLLECTION 1 1 1 RESTRICTION BUSHING

PUMP SHAFT PUMP END

I 1

FIGURE Z2-14 HEAT TRANSPORT PUMP GLAND SEAL I I I I I I I I I

PUMP PUMP 3 4 I GLAND GLAND I CLASS 3 I I

D2O I ' SAMPLING L DISCHARGE D2O OF D2O SAMPLING FEED PUMPS t D2O SAMPLING I

I FIGURE 2.2-15 HEAT TRANSPORT SYSTEM PUMP GLAND SEAL COOLING SYSTEM 1

STEAM OUTLET NOZZLE 1 SECONDARY STEAM CYCLONES PRIMARY STEAM CYCLONES CHEMICAL FEED NOZZLE AND HEADER I DOWNCOMER ANNULUS REHEATER DRAINS RETURN AND EMERGENCY WATER SUPPLY NOZZLE U-BEND SUPPORTS I ' TUBE BUNDLE TUBE SUPPORT PLATE BACK-UP SUPPORTS I OBSERVATION PORT BLOWDOWN NOZZLE DIVIDER PLATE 1 I 1 1 1 1

14 D2O INLET NOZZLE 15 BASE SUPPORT I 16 D2O OUTLET NOZZLE BAFFLE PLATE 18 PREHEATER 1 19 LATERAL SUPPORTS 20 WATER LEVEL CONTROL TAPS 1 ' MANWAY FEEDWATER NOZZLE I I I

FIGURE 2.2-16 600 MW STEAM GENERATOR FIGURE 2.2-17 TYPICAL HEADER FIGURE 2.2-18 FEEDER END FITTING ARRANGEMENT PRESSURIZER RELIEF VALVES (2) DEQASSER CONDENSER ' PRESSURIZER STEAM BLEED VALVES (2) RELIEF VALVES (2)

STEAM GENERATOR

STEAM GENERATOR

-CA>-^, T HEADERS I ^k INLET OUTLET A FROM D2O STORAGE TANK AND DEGASSER CONDENSER h^ r

FIGURE 2.2-19 HEAT TRANSPORT PRESSURE AND INVENTORY CONTROL SYSTEM 1 I

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 pressure transients by receiving coolant from the heat transport system when I pressure is increasing, and by supplying coolant to the heat transport • system when pressure is decreasing. Three typical heat transport system transients are shown in Figure 2.2-20. I

Valves that discharge D2O from the heat transport system (HT relief valves, pressurizer steam bleed valves and relief valves) connect to the i* degasser condenser (Figure 2.2-19). The relief devices of the degasser If condenser are set above the normal HTS operative pressure, thereby limiting 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 IJ decay power following a reactor shutdown. Two independent shutdown cooling 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 in the heat transport system circuits, a portion of the shutdown cooling _ system flow passes from the outlet header to the inlet header via the I steam generators. The shutdown cooling system can also be operated * utilizing the heat transport system pumps; in this mode of operation, the shutdown cooling system flow bypasses the shutdown cooling system pimps. "1

This system is also effective with the heat transport system depressurized and the D2O level lowered to the elevation of the headers; this facilitates maintenance of the steam generators and HTS pumps. 1

2.2.5 Heat Transport System Purification •

The accumulation of active materials in the CANDU heat transport system is inherently very low. This is primarily due to restrictions placed on "1 materials used in the HT system (for example, very low cobalt levels are 1 permitted), and the absence of failed fuel during reactor operation (in the event fuel failures do occur, they are detected and removed). ••

To further minimize the accumulation of active deposits within the HT system, the coolant is continuously filtered and purified. The head of one heat transport system pump in each circuit is utilized to provide a T flow 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 and pressurized. 1

2.1-7 1 • III |i| |ii 1 1 1 1 1 1

12 -

, LOSS OF CLASS IV POWER 11 -

10 - z 111 a. 9 \.7 ^--^^ \^ LOSS OF ONE PUMP ^w STEPBACK TO 70% POWER

8 "

REACTOR TRIP ^"^^s^^ / % FROM FULL POWER ^ «%>>>^ y 7 -

6

• Ill III III i i i i i i 12 16 20 24 28 32 TIME-SECONDS

FIGURE 2.2-20 HEAT TRANSPORT SYSTEM TRANSIENTS STEAM GENERATOR I.STEAM GENERATOR

HEAT TRANSPORT SYSTEM

SHUTDOWN SHUTDOWN COOLING SYSTEM COOLING HEAT ISOLATION VALVE (TYPICAL) EXCHANGER

SHUTDOWN COOLING PUMP

FEEDERS (TYPI

CIRCUIT 1

FIGURE 2.2-21 SHUTDOWN COOLING SYSTEM STEAM GENERATOR STEAM GENERATOR

HEAT TRANSPORT SYSTEM

COOLER

FILTER

FEEOERS (TYP)

REACTOR ION EXCHANGE

FIGURE 2.2-22 HEAT TRANSPORT PURIFICATION SYSTEM I 1 2.3 OVERALL PLANT CONTROL I 2.3.1 Introduction The term "Overall Plant Control" is used to describe the controls that I coordinate the turbine-generator output with the power output of the reactor. These controls are described in this section, but the section also describes other features of the controls and instrumentation which I are unique to CANDU nuclear power plants. The CANDU plants use direct digital control for all major control functions. The control computers -r are a highly reliable dual computer system. A second feature of CANDU | control is the advanced control room concept which uses computer driven displays, alarms, messages, and logs to replace much of the more conventional instrumentation used in other plants. In addition to the f main control room, CANDU plants incorporate a Secondary Control Area from - which important variables can be monitored and controlled and from where the plant can be shutdown. The Secondary Control Area and its equipment ~j is 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 direct digital computer controls. However the control room designs of the T earlier plants were quite conventional. In more recent plants the control L room design has been modified to make better use of the more flexible display and message capability. Pickering 'A' which came in service in the early 1970's uses conventional instruments supplemented by a few I computer-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 I plant information in a variety of formats - graphs, bar charts, printed messages, etc.. Plants designed since Bruce 'A' have carried this concept further by using colour CRT's in place of the monochrome CRT's (see Figure I 2.3-1). 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 field terminations of wiring to be completed independently of connection to the -.. control panels and computers. !

2.3.3 Electrical Power f- Power in the plant is divided into four classes depending on the .1 requiranents of the equipment being supplied. The power supply buses are further subdivided and separated to help meet equipment reliability 7

2.1-8 EMERGENCY CONTAINMENT MODERATOR AMD COKE MISCELLANEOUS COOLING REACTOR MISCELLANEOUS PRIMARYHEAT AUXILIARY SYSTEMS TRANSPORT SYSTEM FUELLING MACHINE SHUTDOWN ELECTRICAL AM FUEL HANDLING SYSTEM AMGENERATOR DISTRIKITION CONTROL CONNIE i „,, \ SHUTDOWN \ REACTOR CONTROL SYSTEM E GENERATOR SWITCHYARITCHYARD SYSTEMSYSTEMS I 1 \ ""v \ SYSTEM \ REGULATING SYSTEM COMPUTERS N 1 I PL4 PL5 PLB f - . „ \ \ \ °- » \ PL7 H.V) PL11 PH2 PLI3 I /I l\ PL fL3 PL4 P \ \ \ \ \ \ TIT II T I

LINE PRINTERS V V

FIGURE 2.3.1 TYPICAL ARRANGEMENT OF CRT DISPUYS ON MAIN CONTROL PANELS 1 1

requirements. Most of the control equipment is supplied by Class II J power. This is uninterruptable power and separate buses are used to redundant instrumentation. 1_ 2.3.4 Dual Computer System • High reliability of control functions is assured through the use of two • identical independent digital computers (DCCX and DCCY). Each computer is capable of complete station control and can transfer control automatically m to the other computer on detection of a fault. Faults in either software II or hardware are detected by a combination of internal hardware and software self checking facilities plus an external "watchdog timer" or _, operations monitor. Fault detection may result in automatic reloading of II core memory from the disc and computer restart, or transfer of control to '* the other computer. The computers also verify the incoming data using redundant information and rationality checks. Messages to the operator [I identify out-of-range data. Both computers are normally running but the II outputs to the plant are only connected to one computer. Switching of outputs is automatic when required. •

2.3.5 Plant Controls

In a nuclear power plant there are a large number of variables to be controlled. Examples are: . Moderator temperature I . Deaerator level . Heat transport system pressure 1 . Pressurizer level 1 . Reactor power . Steam generator pressure 1 . Steam generator level

In order to control the plant electrical output to the desired value, I 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 fj have their own controllers. Most of the variables are controlled by ^ interacting programs in the plant computers. In controlling the turbine-generator output, it is also necessary to control the reactor fT power. Steam generator pressure is also closely related to this control [£ 2.1-9 ff I I

I problem as is the steam generator level. These three control loops share common filaments and can be considered together. Within the logic diagram, Figure 2.3-2, can be seen separate control loops for reactor power, steam generator pressure, and steam generator level, all controlled from the I main computer system. The overall plant control scheme is similar to that used in non-nuclear plants. It operates in two modes:

I a) Normal Mode (Reactor follows plant loads)

The turbine generator load is set by the operator, and the turbine I governor valves open to supply the necessary steam. The steam generator pressure control program senses pressure changes due to governor valve motion and requests variations in reactor power to maintain drum pressure constant. For example a frequency drop due to increased grid loading I would cause the turbine governor valves to open further. The resulting drop in steam drum pressure would cause the pressure control program to request an increase in reactor power. This would occur unless there were I limits on reactor power available. The reactor control system is discussed separately in Section 2.4. I b) Alternate Mode (Turbine follows reactor) Reactor power is controlled to a setpoint supplied by an operator. The steam generator pressure control program manipulates plant loads to keep I steam drum pressure constant. This mode is used:

1) At low power when the steam drum pressure is insensitive to reactor I power;

2) During upset conditions where it may not be desirable to maneuver I reactor power. In addition to adjusting the turbine load to accept the steam output from the reactor, the pressure control program also has access to steam discharge valves so that excess steam can be dumped directly to the condenser or to atmosphere. The Condenser Steam Discharge Valves (CSDV) typically can carry 70% or more of the steam production if necessary while I the Atmospheric Steam Discharge Valves (ASDV) are limited to about 10%. If the turbine becomes temporarily unavailable, the reactor can continue to operate by dumping most of its steam to the condenser. The atmospheric 1 steam discharge valves are used during startup when the condenser may be unavailable and temporarily during other transients.

n 2.1-10 ATMOSPHERIC STEAM DISCHARGE

GENERATOR

STEAM REACTOR GENERATORS

FLUX POWER AND RATE

REACTOR r POWER

REACTOR STEAM GENERATOR FLUX PRESSURE CONTROL CONTROL

REACTOR POWER SETPOINT TURBINE LOAD CONTROL OR UNIT POWER I REGULATOR (UPR)

DEMANDED REACTOR POWER ELECTRICAL OUTPUT (ALTERNATE MODE ONLY) SETPOINT (NORMAL MODE)

FIGURE 2.3-2 OVERALL PLANT CONTROL — BLOCK DIAGRAM

i.. •..Jin} I I I 2.4 CORE CONTROL 2.4.1 Introduction I Reactor control was mentioned as one of the sub-loops in the Overall Plant Control System. The reactor control system is one of the most important control systems. It combines hierarchies of measurement instruments and I control devices with complex computer logic to meet a number of requi rements. I The system is required to: 1) Monitor and control reactor power to satisfy station load demands. I 2) Monitor and control the three-dimensional power distibution in the reactor so that individual fuel bundles and fuel channels operate at I powers within their design specifications. 3) Monitor important plant variables and reduce reactor power at appropriate rates to keep the variables within specified limits.

I The reactor control system can be more easily understood by examining the measuring instruments, the reactivity control devices, the logic that ' relates the device operation to the measurements, the requirements of the I system, and the disturbances it is subjected to.

I 2.4.2 Reactor Power Measurement Power from the reactor is measured with combinations of:

I 1) Startup counters I 2) Ion chambers 3) Self-powered in-core flux detectors 1 4) Thermal power measurements. t 2.4.2.1 Startup counters are used only during the first criticality or for starting after a very long shutdown. They are used along with manual 1 controls to raise power above a range (7 decades below full power) where the ion chambers give useful readings to the computers. Following high power operation, heavy water reactors retain a source term which keeps the 1 ion chambers on scale even after extensive shutdowns. Startup counters

2.1-11 1 1

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 flux measurements in the range from 10"' to 1.5 times full power. The signals are provided to the computers through logarithmic amplifiers. -m Shielding in the ion chamber housings provides good discrimination against I gamma rays. The signal response is essentially prompt except at the lowest powers. Automatic startup and shutdown of the reactor can be — accomplished over the full range of ion chamber signals. Figures 2.4-1 I and 2.4-2 show the locations of ion chambers and some of the other reactor * control devices. 1 2.4.2.3

Self-powered in-core flux detectors are generally used above a few percent I of full power for flux measurements. Unlike ion chambersf they can give information about the spatial distribution of neutron flux and their _ response is essentially unaffected by dissolved poisons in the moderator. I The prime source of flux measurements is 28 prompt responding detectors at ~ 14 locations in the core. These detectors use platinum, inconel, or platinum coated inconel emitters. While their response is partially "i neutron sensitive, they also have some gamma sensitivity. The information .1 from these detectors is supplemented by 102 additional vanadium emitter type detectors. Their response is entirely due to neutrons but is not "J prompt. The detectors are usually coiled onto vertical assemblies. More | recent designs have used shorter straight detectors which are individually inserted into well-tubes. In both designs, detectors are strategically _ located throughout the core. Flux detectors (and ion chambers) are also I used to provide signals to the special safety systems - shutdown system. • However entirely separate instruments are used for those systems. I 2.4.2.4

The neutron flux measurements are calibrated against measurements of I reactor total thermal power. At high power this comes from redundant measurements of steam flow, steam pressure, steam temperature, feedwater «_ flow and feedwater temperature. At low power the steam flow measurement I is not sufficiently precise, and reactor power is calculated from -1 temperature rise measurements across the reactor. I I 2.1-12 I I I I I I I I I I I I I I

CALANDRIA "*"" 16 EARTHQUAKE RESTRAINT I CALANDRIA- SIDE TUBESHEET 17 CALANDRIA VAULT WALL CALANDRIA TUBES 18 MODERATOR EXPANSION TO HEADTANK EMBEDMENT RING 19 CURTAIN SHIELDING.SLABS 5 FUELLING MACHINE-SIDE TUBESHEET 20 PRESSURE RELIEF PIPES G END SHIELD LATTICE TUBES 21 RUPTURE DISC 1 7 END SHIELD COOLING PIPES 22 REACTIVITY CONTROL UNIT NOZZLES B INLET-OUTLET STRAINER 23 VIEWING PORT 9 STEEL BALL SHIELDING 24 SHUTOFF UNIT 10 END FITTINGS 25 ADJUSTER UNIT 11 FEEDER PIPES 26 CONTROL ABSORBE.fi UNIT 12 MODERATOR OUTLET 27 ZONE CONTROL UNIT 13 MODERATOR INLET 28 VERTICAL FLUX DETECTOR UNIT 14 HORIZONTAL FLUX DETECTOR UNIT 29 LIQUID INJECTION SHUTDOWN NOZZLE 15 ION CHAMBER 30 BALL FILLING PIPE

FIGURE 2.4.1 REACTOR ASSEMBLY I I I I I I I I

1 2 3 4 5 6 7 B 9 10 11 112 13 14 15 16 17 18 19 20 21 I

VERTICAL FLUX DETECTOR (26) SOLID CONTROL ABSORBER (4) ADJUSTER (21) LIQUID ZONE CONTROLLER (6) I SHUTOFF ROD (28) HORIZONTAL FLUX DETECTOR (7) I I I I I E ZONE CONTROL ABSORBERS ZONE CONTROL I •DETECTORS VIEW OF REACTOR FACE I! FIGURE 2.4-2 REACTIVITY MECHANISM LAYOUT I I I 2.4.3 Control Devices The regulating system (Figure 2.4-3) controls the neutron flux in the reactor (level and distribution) by adjusting a hierarchy of devices - I light water control absorbers, mechanical control absorbers/ adjusters, I moderator poison, and fuel. 2.4.3.1 I There are 14 light water zone control absorber compartments distributed throughout the reactor. These compartments are partially filled with light water - a neutron absorber in a heavy water reactor. There is a constant outflow of water from the compartments and a controlled inflow I which allows the computer to raise or lower water levels in unison or differentially. These absorbers are the primary reactivity control devices used for both bulk and spatial control of neutron flux in the I reactor. A compartment can be completely emptied or filled in a minimum time of one minute giving a reactivity change per compartment of

I 0.5 mk (i.e. ^ = 0.0005). I 2.4.3.2 The four mechanical control absorbers are normally out of the core but can be driven in at variable speeds or dropped to supplement the negative I reactivity from the light water absorbers. They are mechanically the same as shutoff rods but are functionally and physically separate and under the control of the computer system. They provide up to 6 mk of negative I reactivity when inserted for a power reduction. I 2.4.3.3 The 21 adjuster rods are normally fully inserted in interlattice positions where they contribute to flux flattening. They are not used for dynamic I control of the flux distribution but may be withdrawn in symmetrical banks and at variable speeds to provide additional reactivity. This would be required to compensate for xenon following a large power reduction or I shutdown followed by a restart. A total reactivity worth of 15 mk is • available in the adjusters and each bank can be withdrawn in a minimum I time of 1 minute. I

2.1-13 I I

INPUTS EVICED S LOW BOILER j OIGITAL COMPUTER CONTROLLER LEVEL HIGH POWER I ERROR HIGH HT PRESSURE I HT PUMP 4CONTROL FAILURE DRIVIN G OU T STEP-BACK . TO O MAN Y ADJUSTE R • TRI P SYSTEM S SE T ABSORBERS REACTOR TRIP ROUTINE CLUTCHES TT ONLY I TURBINE TRIP LOSS OF LINE t t 21 ADJUSTERS LOSS OF STATOR REGULATING ADJUSTER DRIVE COOLING PROGRAM VARIABLE I INTERLOCKS REACTIVITY SPEED DRIVE OPERATOR CONTROL DEMAND 14 ZONE I DEMAND POWER CONTROL ROUTINE VALVES BOILER PRESSURE M j I CONTROLLER 28 SHUT-OFF RODS ZONE OUT DRIVE CONTROLLER TRIP SYSTEMS SET I FAILURE t FLUX TILT CONTROL 4 CONTROL PROBLEMS ABSORBER ABSORBERS HIGH LOCAL DRIVE VARIABLE I SPEED DRIVE FLUX INTERLOCKS SET BACK HIGH BOILER MODERATOR PRESSURE I POISON LOW BOILER I \ ADDITION LEVEL HIGH SURGE TANK LEVEL I MANUAL

VANADIUM FLUX I IN CORE MAPPING DETECTOR POWER ROUTINE

ION CHAMBER I POWER :- ..•':••( f: POWER INCORE MEASUREMENT DETECTOR ALARMS AND I POWER * CALIBRATION BOILER SECONDARY SIDE MEASUREMENTS I f OPERATOR C.R.T. DISPLAY DISPLAY DISPLAYS SELECTOR PROGRAM I I! FIGURE £4.3 REACTOR REGULATING SYSTEM BLOCK DIAGRAM 7 I I I 2.4.3.4 Boron or gadolinium salts can be dissolved in, or removed from, the moderator as an additional reactivity shin. This is normally an operator I function, but under exceptional circumstances the control computer can intervene, "fine reactivity effects are approximately 9 mk/ppm for boron and 32 mk/ppm for gadolinum. Only very small concentrations are needed I and the addition and removal rates are slow. Gadolinum burns out at a rate similar to the buildup rate of xenon.

I 2.4.4 Control Logic

In each of the 14 zones of the reactor the flux is sensed by a pair of I prompt responding in-core flux detectors. Their response is calibrated against a flux map derived from the slower vanadium-emitter type of flux detectors. The flux map is normalized against thermal measurements. A I setpoint for flux in each zone is calculated by the computer from the overall plant control requirements or operator demands, and a zone level setpoint is derived. The valve lift on the input valve to each zone of the light water control absorbers is then varied dynamically to drive the I flux and level errors towards zero. If the average power error in the reactor gets too large or if the average zone level approaches its upper or lower limits, the logic then drives adjusters or mechanical control I absorbers to supplement the range of the light water control absorbers. The logic of this procedure is illustrated in Figure 2.4-4. During normal full power operation the adjusters would remain inserted and the I mechanical control absorbers withdrawn. Their positions would only change during transients such as power level changes.

I 2.4.5 Di sturbances

Changes in the net reactivity of the core occur because of fuel burn-up, I new fuel addition, or because the power output of the reactor is changed. These changes are amplified both locally and generally in the reactor by the effects of xenon poison. Xenon-135 is a neutron absorbing fission I product which decays naturally with a time constant of several hours. Because of the nature of the xenon production through the decay of Iodine-135, there will be an inital increase in xenon when power is reduced, although it will eventually return to a slightly lower I equilibrium level. Similarly there will be an initial decrease in xenon when power is increased. This is both a local and a general effect. Local changes in power will cause local changes in xenon. In the absence of I spatial control, spatial xenon induced flux oscillations would also be possible. The reactor control system must compensate for disturbances caused by fuel changes, xenon effects accompanying these changes, and r changes in operating power.

2.1-14 1

(a) 100% I I I I

-1 0 +1 +5 POWER ERROR % I (b) 100% i i i Drive Absorbers IN I u i 70% ll Drive Absorbers OUT I 4 Rods 2 Rods 2 Rods 4 Rods I

-I »-Ep -1 0 +1 +2 +4 +5 I POWER ERROR %

•-100% I O UJ 5£ 50% I

—I *-Ep I -1 0 +1 +2 +3 +4 +5 POWER ERROR % I I I

Idealized zone reaclivity rate vs. power error I I FIGURE 2.4.4 REACTIVITY LIMIT CONTROL DIAGRAM I I

For small or slow changes in power the xenon effects are small and easily I controlled. For larger changes the xenon effects impose some broad limits on power manuevering. These are discussed further in Section 2.7. I 2.4.6 Stepbacks and Setbacks The reactor control system in the plant computers includes routines which I monitor a number of plant variables for operation within acceptable limits. Reactor power is promptly reduced if the variables exceed these limits. These systems are entirely separate from the shutdown systc and serve to reduce the frequency of operation of the special safety I (shutdown) syste

If a very fast power reduction is needed the computer initiates a I "stepback". It opens clutches on the four mechancial control absorbers, allowing them to drop into the reactor, making it subcritical. The rods can be "caught" part way in by re-closing the clutches. The reactor is I then critical at a lower power. If the rods are not caught, the stepback will take the reactor towards zero power.

For power reductions which are not as urgent, a "setback" occurs rather I than a stepback. The reactor power setpoint in the computers is taken downwards at a controlled rate. The power follows through normal action of the light water control absorbers and the related devices. The setback I ends either when the variable causing it returns to limits, or when a predetermined power level is reached. The rate at which reactor power is reduced and the power level at which the setback ends may be different for I each variable. Setbacks and stepbacks override other power demands and are accompanied by I annunciation defining the out of limits variables. I 2.5 REACTIVITY CONTROL DEVICES 2.5.1 Liquid Zone Control System I The liquid zone control absorbers are the primary devices for controlling reactivity within the reactor during normal operation: Reactivity is adjusted by varying the quantity of light water, which acts as a poison, L in each of the zone control compartments (Figure 2.5-1). Fourteen zone control compartments are contained within 6 assemblies. The location of each of the zone control assemblies is shown in the reactivity mechanism layout. Figure 2.4-2. Details of the liquid zone control assembly are I provided in Figure 2.5-2.

The zone control system is illustrated in Figure 2.5-3. Light water is circulated through the zone control compartments; level in the compartments is measured by a helium 'bubbler' system, and is controlled

2.1-15 THREE COMPARTMENTS IN CENTRAL ASSEMBLIES

TWO COMPARTMENTS IN OUTER ASSEMBLIES ZONE CONTROL ASSEMBLY

LIGHT WATER (H2O)

MODERATOR

CALANDRIA

FIGURE 2.5-1 LIQUID ZONE CONTROL SYSTEM ARRANQEMENT I I I

I ZONE1 I I BELLOWS SPLIT SEAL PENETRATION TUBE RING I BEARING ZONE 2 I I I CONCRETE I I I THIMBLE TUBE NOZZLE

I LOCATOR

CALANDRIA SHELL I CALANDRIA SHELL

1 WATER 11 WATER INLET 2 HELIUM 12 BULKHEAD I 3 NUT 13 BAFFLE 4 CRUSH WASHER 14 HELIUM OUTLET 5 TERMINAL BLOCK 15 HELIUM INLET 6 SHIELD PLUG 16 WATER OUTLET 7 ZONE CONTROL TUBE 17 HELIUM BALANCE LINE 8 WATER AND HELIUM TUBES 18 KEY 9 TUBE SUPPORT 19 SPRING 10 TUBE SPRING 20 LOCATOR THREAD FIGURE 2.5-2 ZONE CONTROL UNIT I I TO OTHER COMPARTMENTS I I I

RECOMBINATION GAS BALANCE HEADER UNIT l-l I I

LEVEL HELIUM TRANSMITTER BOTTLES FOR GAS I MAKE UP I I HELIUM GAS COMPRESSOR „ TO OTHER TYPICAL COMPARTMENTS ARRANGEMENT I OF ZONE CONTROL COMPARTMENT (14 IN ALL) I FROM OTHER COMPARTMENTS H2O SUPPLY I HEADER 1 I 1

H2O CIRCULATING PUMP (3) I I FIGURE 2.5-3 LIQUID ZONE CONTROL SYSTEM I I

via control valves on the water inlets, based on a signal from the station I computer. The water is forced out of the compartments at a constant rate by the helium cover gas pressure.

I 2.5.2 Vertical Flux Detector Units

Like the Zone Control Unit already described, the Vertical Flux Detectors I are mounted beneath the Reactivity Mechanism Deck surface. However, they may be reached through access plugs in the Deck Plate (Figure 2.5-4). I Because of their long and slender design, and of their construction of low-modulus Zircaloy 2 material, these Flux Detector Units are installed in 27.5 mm guide tubes, which are tensioned to enhance their rigidity.

I Each Unit consists of a carrier and capsule tube assembly with deteccors, connectors and seal components. I The detectors themselves are self-powered Hilborn elements oi vanadium or platinum construction, in which emitter wires are separated from their I sheathing by mineral oxide insulation. These detectors provide signals which are directly proportional to fission rates in the reactor, with varying response and sensitivity according to I their material or exposure duration. I 2.5.3 adjuster Units (Figure 2.5-5) The requirements for normal insertion and controlled removal of the absorber elements, for Xenon override, demand mechanical features foe r hoisting and lowering the elements in response to the regulating system requirements.

Adjust Drive Mechanisms mounted above the reactivity deck embody a motor i driven sheave on which absorber cables are wound or unwound for absorber raising and lowering.

The absorbers consist of stainless steel tubing (of 76 mm diameter) with a shaped central shim rod.

The adjusters are arranged in seven banks, whose collective withdrawal would increase reactivity by 15 mk. This increment would provide 30 minutes Xenon override time after shutdown from steady full power i operation. Figure 2.5-6 illustrates the positioning of the vertical reactivity i control devices under normal and actuated conditions. r 2.1-16 I I TREAD PLATE - I I COVER GAS CONNECTION -eg GLASS I INSULATION I I I I I DETECTOR ASSEMBLY I CALANDRIA NOZZLE I

CALANDRIA SHELL - DETECTOR COIL I CALANDRIA TUBES I

GUIDE TUBE GUIDE TUBE LOCATOR s

CALANDRIA SHELL i II B FIGURE 2.M VERTICAL FLUX DETECTOR UNIT DRIVE MECHANISM

SHEAVE

INSERT I GUIDE TUBE 1 EXTENSION 1 I

I H2O 1

FIGURE 2.5-5 ADJUSTER UNIT 1

SOLID CONTROL & SHUT-OFF DRIVE I OPEN FOR SERVICES 1 1 I I I 1 ]

HORIZONTAL FLUX DETECTORS 1 LIQUID INJECTORS ION CHAMBERS 1.2 4 3 SIDED 1 1 1 I I I T

FIGURE 2.5.6 SCHEMATIC SECTION SHOWING POSITIONS OF REACTIVITY CONTROL DEVICES I I I 2.5.4 Mechanical Control Absorber Units and Shut-Off Rods Mechanical Control Absorber Units and Shut-off Rods (Figure 2.5-7) have virtually identical design and capability, but have different missions, I (as outlined in Section 2.1.1) in reactor control and shutdown functions. Hence, the Control Absorbers form part of the Regulating System, whereas I shut-off rods are part of the protective Shutdown System No. 1. Both units embody a clutch, cable and winch arrangement to hoist, hold or release cadmium filled absorber tubes, of 113 mm diameter and of varying I length.

Rod Ready Indicators sense the presence of the Absorbers in the poised I (fully raised) position. Guide tube perforations minimize hydraulic resistance to absorbers' entry I into the calandria. Spring-assistance provides acceleration in response to a shutoff rod clutch de-energization. Normal time for full insertion is about 1.6 I seconds.

Rotary hydraulic dampers, geared to the winch, stop the absorbers' I movement in the calandria at the appropriate level.

I 2.5.5 Horizontal Flux Detector Units (Figure 2.5-8) The construction of these assemblies is similar to that of the vertical flux detector units. However, installation details are quite different I due to their horizontal configuration, to requirements for sealing and support, and to isolating provisions for replacement, by freezing.

I As noted in 2.1.1 these assemblies form part of the protective jShut Down I JSystem No. 2_ (SDS2). 2.5.6 Liquid Injection Shutdown Units

SDS2 provides for rapid injection of gadolinium nitrate solution, by t helium displacement, from an external poison tank into the calandria via injection nozzles, into the moderator. I Figure 2.5-9 shows the installation, which includes similar features for I sealing, support and replacement as for the horizontal flux detectors.

2.1-17 I

SPRING TO REWIND DAMPER DOG PLATES — LOST MOTION LINK BETWEEN MAIN I AND DAMPER SHAFTS - ALSO ENGAGE POSITION STOPS MAIN DRIVE MOTOR AND SHAFT .^ WORM REDUCTION GEAR ELECTRO MAGNETIC CLUTCH I BETWEEN IDLER AND MAIN SHAFTS DRIVE MECHANISM POTENTIOMETER CABLE SHEAVE POSITION INDICATOR IDLER PULLEY •> I CABLE- BEVEL REDUCTION GEAR SHUTOFF HOD AND SPUR REDUCTION GEAR ON IDLER SHAFT I SCHEMATIC SHUTOFF ROD DRIVE MECHANISM I I I I I I

SPIDER ATTACHING SUPPORT ROD TO GUIDE TUBE EXTENSION SHUTOFF ROD I ACCELERATOR SPRING (ON SHUTOFF ROD ONLY) WATER SHIELD I GUIDE TUBE TENSIONING SPRING CALANDRIA NOZZLE I II I

CALANDRIA SHELL SHUTOFF ROD I 1 FIGURE 2.5-7 SHUTOFF AND MECHANICAL CONTROL ABSORBER UNITS MOUNTING BOSS ADAPTER PLATE SEAL DETECTOR SHIELD SLEEVE SEAL CLAMP PRESSURE GAUGE ASSEMBLY

THIMBLE SUPPORT

I FREEZING DoO CONNECTION „„„..,,- DETECTOR SEAL CLAMP ! JACKET THMBLE cwlf CABLE ASSEMBLY °°VfcH

GUIDE TUBE

FIGURE 2.5-8 HORIZONTAL FLUX DETECTOR UNIT SHIELDING WALL

PROTECTIVE COVER

THIMBLE

THIMBLE SUPPORT

INJECTION NOZZLE

CALANORIA

FIGURE 2.5-9 LIQUID INJECTION SHUTDOWN UNIT I I I 2.5.7 Ion Chamber Assemblies Six ion chamber units are installed in housings mounted externally on the calandria shell, in stainless steel housings (Figure 2.5-10). Lead I shielding provides attenuation of gamma radiation for enhanced discrimination of neutron flux levels. I Access tubes enable replacement of defective ion chambers on power, with suitable suplementary shielding.

Physical separation of the two groups of ion chambers reduces the I probability of local accidental damage to more than one group.

The broad sensitivity of these units is exploited by use of logarithmic I amplifiers and applied in monitoring lower power levels and rate of change I in power level. 2.6 FUEL I 2.6.1 Description of Fuel The 600 MW(e) reactor fuel bundle comprises seven component parts (Figure 2.6-1). The elements contain high-density natural UO2 in a thin I Zircaloy-4 cylindrical sheath. A thin graphite layer (CANLUB) on the inside surface of the sheath reduces the pellet/sheath interaction. End caps, resistance welded to the sheath extremities, serve a triple purpose: I (i) to provide a seal for the contents of the element, (ii) to provide effective element termination for attachment to end plates, and (iii) to provide the structural component for interfacing with the fuel handling system. Thirty-seven elements are held in a close-packed bundle I configuration by welding them to end plates. The desired separations at the transverse mid-place of the bundle are maintained by split spacers brazed to the elements. Bundle separation from the pressure tube is I ensured by bearing pads brazed near the ends and at the middle of the outer elements. The filler metal used for brazing is beryllium. r 2.6.2 Design Basis for the Fuel 2.6.2.1 Fuel Elements

The reactor nuclear design requires high neutron economy. The fuel element in therefore designed for maximum content of fissile material and l minimum content of neutron absorbing material. r

2.1-18 CALANDRIA SHELL .

SHIELDING SLEEVES BEARING

FREEZING COIL

BELLOWS

CARBON STEEL WATER SHIELD LINER

ION CHAMBER PENETRATION TUBE (3)

FIGURE 2.5-10 TYPICAL ION CHAMBER ARRANGEMENT END VIEW INSIDE PRESSURE TUBE

ZIRCALOY BEARING PADS ZIRCALOY FUEL SHEATH ZIRCALOY END CAP ZIRCALOY END SUPPORT PLATE URANIUM DIOXIDE PELLETS CANLUB GRAPHITE INTERLAYER INTER ELEMENT SPACERS PRESSURE TUBE

FIGURE 2.6-1 37-ELEMENT FUEL BUNDLE I I

The fuel is designed to operate within the power and turnup conditions I applicable to normal station operation, defined by the nominal bundle ' power envelope (curve A in Figure 2.6-2). Since a limited number of bundles can exceed this power envelope, the fuel is assessed for operation I within the reference overpower envelope (curve B in Figure 2.6-2). •

2.6.3 Fuel Performance I 2.6.3.1 General _

CANDU fuel performance has been demonstrated by wans of out-reactor * tests, irradiation testing in experimental reactors, and successful utilization in CANDU power reactors. Nevertheless, performance is M continually being improved as a result of an ongoing fuel development | program. Some principal aspects of fuel testing and performance are discussed below. I 2.6.3.2 Out-Reactor Tests _

Out-reactor tests are conducted on prototype and initial production fuel ™ bundles, to ensure the ccmpatiblity of the fuel design with the requirements stated above. Examples are endurance tests, impact tests and I pressure drop measurements. I

2.6.3.3 In-Reactor Performance •

Irradiation testing is mainly performed in test loops at AECL's experimental reactors. Over the years a significant 1bank of data and • experience on CANDU fuel behaviour has been built up.

The in-service performance of over 200,000 CANDU fuel bundles irradiated I to date has been excellent. Only 0.18% of all fuel bundles irradiated I have been defective, as on September, 1979. The introduction of CANLUB fuel in the fuel design resulted in a marked improvement of performance. m Of the approximately 140,000 CANLUB fuel bundles irradiated, only 0*07% I were defective and none of the defects were due to power boosting, but were rather latent manufacturing defects, see Figure 2.6-3. It should be _ noted that usually only one defective element is found in a defective n bundle. Therefore, if the defect statistics are reported on the basis of ** defective elements, as is customary for other reactor vendors, the defect statistics become 0.006% defective elements (all fuel) and 0.002% If defective elements (CANLUB fuel). If I 2.1-19 s 1000

900 ^ B REFERENC E OVERPOWER ENVELOPE

800

111

700 A NOM NAL DESIGN i POWER ENVELOPE

600

500

100 200 300 400

BUNDLE AVERAGE BURN-UP (MWh/kgU)

FIGURE 2.6-2 I 1 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

1 I REASONS FOR DEFECTS IN 1 CANDU POWER REACTOR FUEL NUMBER OF DEFECTS Power ramp (previous to CANLUB) 134 (The power ramp failure rate is zero I for CANLUB fuel) Incomplete end cap welds 12 1 Porous end caps 5 Handling damage 7 Fretting by debris in coolant 6 I Flew induced fretting 1 Unknown causes 16 I

FIGURE 2.6-3 I

I 2.7 STARTUP, OPERATION AND SHUTDOWN SEQUENCES

Some of the techniques of operating the plant are considered in -this I section - especially as they relate to the plant control systems.

I 2.7.1 Initial Startup In Section 2.4 it was noted that special startup counters are installed temporarily for first reactor criticality. Using these counters and I manual control the reactor can be made critical and raised in power until the ion chambers provide signals to the computers. Control is then automatic and power can be raised by the operator supplying the required I power and maneuvering rates to the computer. From then on the reactor remains on automatic control. The ion chambers will continue to measure reactor power from the source term for a long period after the reactor is I made subcritical. The automatic controls therefore operate even while the reactor is subcritical. The startup instrumentation and manual starting I would only be needed again following a shutdown of many weeks duration. 2.7.2 Eversafe Shutdown

I If major equipment is being maintained, the moderator is heavily poisoned to ensure that the reactor remains subcritical even following xenon decay I or removal of reactivity control devices. 2.7.3 Heat Transport System Warmup and Cooldown

The heat transport system normally remains hot and pressurized following 1 reactor 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 generator pressure control program in the computer system and at predetermined rates of temperature change. The system can be pressurized using heat from the I ratespumps. alone, but reactor heat is also available to help achieve design

I 2.7.4 Turbine Runup I There is a program available in the plant computers to run the turbine up to speed automatically and synchronize it to the grid. I 2.7.5 Sudden Shutdown or Trip If the reactor trips or steps back to zero power, recovery to power is II automatic once the operator resets the trip and gives a power setpoint to 1 2.1-20 the system. Resetting a trip of Shutdown System Number One causes the shut-off rods to withdraw. This is followed by withdrawal of mechanical control absorbers and lowering of light water control absorber levels until the reactor is critical. Adjusters may also withdraw in response to xenon increases to allow the reactor to become critical. Power is held at a low value until the operator requests a power increase. The request for a power increase must come within about 20 minutes of the sudden shutdown from full power to avoid "poisoning out". If the reactor poisons out the xenon builds up to a level that the reactor control system cannot immediately compensate for. The xenon will decay away allowing a restart after about 36 hours. If the reactor is restarted within the 20 minute "decision and action" time the xenon will burn out as reactor power is raised. However there will be some spatial assymetries in flux distribution which may not allow power to be raised immediately beyond about 70% of full power. Power can then be raised over a few hours from 70% to 100%.

2.7.6 Power Maneuvering

Reactor power can be raised at logarithimic rates up to 4% of instantaneous power per second while power is below 25% of full power. A maximum linear rate of 1% of full power per second is applied at higher powers. In general, turbine-generator maximum maneuvering rates will be more 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 wide range at rates much faster than 1% per second. However the stepback function and the use of steam discharge valves allows turbine generator load to be shed at any required rate. Small step increases (of the order of 5% of full power) or decreases in load can be handled by the normal control system. It is thus possible to run the plant at less than full power in a spinning reserve mode.

Just as a sudden trip of the reactor leads to a transient increase in xenon, other load reductions have a similar but smaller effect. If the reactor power is suddenly reduced from 100% to 60%, the ensuing xenon transient would cause adjusters to withdraw in compensation. The reactor would not poison out from this size of power decrease and could continue to operate at 60% power indefinitely. As the xenon transient decays, the adjusters would gradually re-insert in banks. A return to full power could begin at any time, but might require several hours if begun while adjusters were still out and tne flux distribution distorted. The reactor control system would allow the plant to follow quite closely the typical daily load changes that occur on most electrical grids. Actual CANDU operating 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 would lead to poisoning out of the reactor unless it is brought to high power again within a short period. However reactor power reductions to intermediate power levels are accomplished without poisoning out. Therefore the normal procedure if the turbine trips, is to reduce reactor power to about 60% of full power. The steam is then bypassed directly to the condenser through the Condenser Steam Discharge Valves (CSDV). The plant can be operated indefinitely this way until the turbine is available. 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 by independent Class IV buses and by an auxiliary feedwater pump (5%) backed up 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 generator inventory in the event of a feed line rupture. The reheater drains flow is returned directly to the steam generators (depending on turbine supplier). Each steam generator is provided with H2O sampling line to facilitate secondary side chemistry control, and a continuous blowdown system, to minimize the accumulation of particulate matter at the tubesheet.

2.8.2 Steam System

Steam generated within the steam generators is normally directed to the turbine. In the event that the turbine is unavailable, steam can be discharged to one or more of the following: the turbine condenser steam discharge valves, atmospheric steam discharge valves or main steam relief valves.

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 Pressure Stages (LP).

The gross electrical ouptut is dependent on local service water conditions, and on the turbine equipment selected.

2.1-22 ISOLATING VALVES ISOLATING VALVES En

ATMOSPHERIC STEAM DISCHARGE VALVE (TYP)

SAFETY VALVES (TYP)

MAIN STEAM FLOW MEASUREMENT

*- STEAM GENERATOR WATER SAMPLING

FEEDWATER FROM FEEDWATER PUMPS

FIGURE 2.8.1 STEAM AND FEEDWATER SYSTEM SEPARATOHS REHEATERS n n n n J MAI N ST fc AM I INTEHCEF VALVES

RBHEATER ) DRAIN PUMPS

STEAM GENERATOR nHP TURBINE

CONDENSE DEAERATOR PRESSURE

•HEATER DRAIN PUMPS GLAND SEAL No. G HEATERS Nn5 HEATFRS 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 generated within the moderator, and the heat transferred to the moderator from the fuel channels. Heavy water is utilized as a moderator due to its high moderating ratio (Figure 3.1-2). The heavy water is circulated through the moderator system for cooling, for purification and for the control of the concentration of substances used for reactivity adjustment. The moderator system features two 100% pumps and two 50% heat exchangers. The piping arrangement permits either pump to operate with either or both of the heat exchangers. The pumps are provided with pony motors powered by Class IV and Class III power to provide circulation in the event of a loss of Class IV power. The location of major equipment within the reactor building 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 over the 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. The system consists basically of two 100% compressors and two 100% recombining units which circulate the cover gas through the calandria pressure relief ducts.

b) Moderator Purification System: The moderator is circulated through the moderator purification system to minimize the accumulation of activity within the system, and to control the concentration of substances used for reactivity adjustment. Poisons are added to the moderator in small quantities when the reactor is first started and when new fuel is added. Large amounts of poison are also added to the 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 of activity from the station, maximum attention is given to minimizing D2O losses. This is accomplished by lowering the number of mechanical joints 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 D2O SUPPLY SYSTEM

HEAT EXCHANGER HEAT EXCHANGER

FIGURE 3.1-1 MAIN MODERATOR SYSTEM LIGHT HEAVY GRAPHITE WATER WATER

SLOWING DOWN 1.35 0.178 0.06 POWER CM-1

MODERATING RATIO 60 2,000 170

PWR CANDU MAGNOX NEUTRON WASTAGE PHW IN MODERATOR, .28 0.16 COOLANT AND .15 CORE STRUCTURES PER FISSION BWR AGR (TYPICAL) .25 0.3

FIGURE 3.1-2 MODERATING EFFICIENCY OF HEAVY WATER 1 MODERATOR HEAT EXCHANGER 2 MODERATOR PUMP 3 REACTOR

FIGURE 3.1-3 LOCATION OF MAIN MODERATOR SYSTEM EQUIPMENT DEUTERATiON — — RUPTURE DISKS DE-DEUTERATION

RECOMBINATION

OOOO OOOOOO oooooo o o oooooo o o ooo ooo o o CALANDRIA ooo oo o o o ooooo o OOOO

D2O LIQUID D2O SUPPLY POISON COLLECTION

FIGURE 3.2-1 MODERATOR AND AUXILIARY SYSTEMS f UM/y \

MANIFOLDS T THEL|1 BOTTLES

FUME ARRESTER (TYPICAL) CATALYTIC RECOMBINATION UNITS

VENT TO REACTIVITY MECHANISMS

CALANDRIA HEAD TANK

FIGURE 3.2-2 MODERATOR COVER GAS SYSTEM and by providing a system to collect heavy water from points of anticipated leakage. Downgrading of heavy water leakages 1* reduced by locating most H2O systems away from D2O system areas and by minimizing mechanical joints in H2O systems within the reactor building. A typical D20 collection system is shown in Figure 3.3-1. Sight glasses are provided in individual and common collection lines to provide a visual indication of liquid flow.

To reduce operating staff exposure, all collected D20 with a high tritium 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 a separate area in the reactor building: a) Major areas subject to Heat Transport System leakage which are accessible only during reactor shutdown. b) Areas requiring frequent personnel access. c) The area within moderator equipment enclosures that may have a high tritium content. d) The steam generator room which is accessible during reactor operation.

The vapour recovery equipment is located in the service building, except for that serving the steam generator area, which is located in the steam generator area. The recovery units are of the absorption type. Recovery of absorbed water vapour occurs in the reactivation condensers located above the dryer vessels; this water is collected in a series of tanks in the D2O management upgrading area. Recovered water is segregated according to the degree of downgrading and tritium activity.

3,3.3 D20 Cleanup System

During operation of heavy water systems, small amounts of D20 escape unintentionally by leakage, and intentionally by deutration* and dedeuteration of ion exchange resins. The D2O cleanup system removes dissolved, particulate and organic impurities from recovered D20 to produce a product suitable for the upgraders. Two separate and almost identical systems are provided; one for low tritium D20 and one for high tritium D2O.

* Deuteration is the process whereby light water present in new resins is displaced by heavy water. Dedeuteration is the reverse process.

3.1-2 LEAKAGE INDICATORS MAIN HEAT TRANSPORT

22 CONNECTIONS PUMPS/MOTORS AND DRAINS (TYPICALI

-DRAIN INDICATORS-

TOD2O STORAGE TANK

- VENT INDICATORS-

CONDENSEBR 1

D2O COLLECTION TANK

TO D2O CLEANUP SYSTEM •*•*•}-( AND SAMPLING

TC FEED PUMP SUCTION

FIGURE 3.3-1 D2O COLLECTION SYSTEM 3.3.4 D2O Upgrading System

The D2O upgrading systems separate solutions of D20 and HjO by distillation. Two upgraders are provided, one for low tritium D2O and one for high tritium D2O.

The overall D2O management system is illustrated in Figure 3.3-2.

3.1-3 TO DEUTERATICN VAPOUR LIQUID TO HEAT & DE-DEUTERATION RECOVERY RECOVERY MODERATOR TRANSPORT SYSTEMS SYSTEM SYSTEM SYSTEM SYSTEM L

D2O D2O FRESH CLEANUP D2O SUPPLY UPGRADING D2O SYSTEM SYSTEM SYSTEM

FROfM FROM HEAT MODERATOR TRANSPORT SYSTEM 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, natural uranium fuel and pressure tube concept has certain inherent safety characteristics (Figure 4.1-1) that obviate the need for a high strength pressure vessel. Instead, the pressure boundaries are the pressure tubes which are considerably simpler to manufacture to the required quality. Further, experimental evidence indicates that pressure tubes will leak before they break since their thickness is much less than the critical crack length. Such leaks can be readily detected by monitoring the moisture content and the pressure in the gas annulus between the pressure tube and the calandria tube. This is done on a continuous basis. In addition, ultrasonic scanning devices are mounted on the fuelling machine for periodic in-service inspection of the pressure tubes.

The pressure tube design permits the heat transport system to be subdivided into two separate coolant circuits (loops). In the case of a hypothetical loss of coolant accident, this design feature restricts the consequences of the loss of coolant accident to just one of the loops. This simplifies the design and reduces the burden considerably on the emergency injection and the contairment system design.

All reactivity devices are located in guide tubes positioned in the low pressure moderator environment, Figure 4.1-2. Thus, there exists no mechanism for rapid ejection of any of these reactivity devices, nor can they drop out of the core. The maximum reactivity rates achievable by driving all control reactivity devices together in the wrong direction is about 0.35 mk per second and well within the design capabilities of the protective systems.

Fuel, coolant and moderator are arranged on a square lattice with a 28.6 cm pitch. This is a near optimum geometry from a reactivity standpoint, Figure 4.1-3. Even if all fuel channels were either pushed apart or brought together for whatever reason the net reactivity increase would be at 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 a few fuel channels are displaced, the net reactivity would at worst not be affected at all or it would decrease, thereby shutting down the reactor. Also, since a lattice of natural uranium and light water cannot be made critical in any concentration, there can be no criticality problems in the spent 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 control requirements. Typically, the reactivity decay rate in 600 MW(e) CANDU PHW reactors is about 0.4 mk per day. This is compensated by fuelling about

4.1-1 PRESSURE TUBES Separate moderator from coolant Cool, low pressure moderator High pressure coolant Interstitial reactivity devices Subdivided PHTS Tubes leak before break

FIGURE 4.1-1

REACTIVITY DEVICES In low pressure moderator NO pressure-driven ejection Separate devices for control and for safety Modest reactivity worth Maximum combined rate < 0.35 mk/s

FIGURE 4.1-2 REACTOR PHYSICS

Natural UO2 fuel D2O moderator Low excess reactivity Near optimum geometry Criticality in spent fuel bay not possible

FIGURE 4.1-3 two channels per day. In addition, the pressure tube concept provides an excellent opportunity for locating fuel defects and the on-power fuelling permits the removal of defective fuel as soon as it is detected. This helps to keep the heat transport system essentially free from fission product activity.

Finally, the separation of the moderator from the high pressure heat transport coolant allows the moderator to act under certain circumstances as an additional heat sink for the fuel decay heat/ e.g. where one might hypothesize a failure or impairment in the emergency core cooling system following 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 event are as follows:

a) The ability to shut the reactor down and maintain it in a safe shutdown 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 and monitoring functions.

The plant design considers both common mode events and randan failures. Since the nuclear process produces heat for a considerable period after the reactor is shutdown, the plant design also takes into account other faults 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 consists essentially of the following: a) High level of equipment quality. All systems are designed to established codes and standards which demand the highest quality in material and workmanship. This makes equipment failure unlikely to begin with.

4.1-2 b) Quality Assurance. This involves strict quality control, both during manufacture and subsequent installation, together with continued periodic inspection of major components throughout the plant operation.

c) System redundancy and fail-safe design. The important systems are designed with redundancy such that the loss of a single component does not cause the loss of the whole system: the provision of dual control computers with one of them providing a complete backup for the other one, two pumps circulating coolant in each of the heat transport systems, two completely different systems for decay heat removal (i.e. the shutdown cooling system, or the steam generators) and finally, redundancy in power supply based on the so-called odd-even concept. Two completely separate power supply systems are provided such that one half of the load for any process is supplied from an odd-bus and the other half from an even-bus.

d) Regulating and Process Systems. These systems are designed to maintain all operating, parameters within acceptable ranges under normal 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 parameters cannot be normally controlled and exceed certain preset values, special safety systems shutdown the reactor, provide long term cooling of the fuel, and contain potential releases of radioactivity.

4.2.3 Protection Against Common Mode Events

The basic defense against common mode events (Figure 4.2-1), is through the use of superior equipment and separation (by distance and/or barriers) of reliable systems, structures and components (Figure 4.2-2). The depth of protection, based on the anticipated rate of occurence of the common mode events, also guarantees that the common mode events under consideration cannot disable the systems required to shutdown the reactor and to remove residual heat, i.e. the basic safety functions have to be maintained. Some of the common mode events considered are: man induced events such as fires and missiles, natural phenomena such as earthquakes and floods, human errors arising from design and operation and cascading of cross-link effects such as effects of pipe whip, environment produced by postulated events, etc. (Figure 4.2-3).

One of the important elements in the defense against common mode events is the two group separation philosophy (Figure 4.2-4). All safety related systems in the nuclear plant are divided into two groups. These groups

4.1-3 PROTECTION AGAINST COMMON 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 AGAINST COMMON 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 APPROACH Essential 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 or localised common mode events can disable more than one group* Inherent in this philosophy is the premise that the reactor building is by design an impenetrable barrier to such common mode events. The systems in each qroup 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 to permit the operator to assess the state of the nuclear steam supply system.

In group 1 these functions are performed respectively by shutdown system #1, the normal electrical and water supplies and monitoring from the main control room. In group 2, the corresponding systems are shutdown system #2, the emergency power and water supplies and monitoring from the secondary control area (Figure 4.2-5).

Group 1 Safety Support Systems

These systems (the normal electric and water supply systems) support the operation of one of the special safety systems. Because of the reliance on these systems for both normal plant operation and continuing operation of special safety systems, special measures are taken in their design to assure reliability.

Group 2 Safety Support Systems

As part of group 2, two safety support systems are provided. They are the emergency water supplies and the emergency power supply systems. They do not perform any function for normal plant operation but are required to provide an alternative water supply and electrical power supply during certain accident conditions. These alternate supplies are located sufficiently remote from the water and electric power supplies of group 1 to ensure defense against common mode incidents.

Grouping Layout

The functional and physical independence of the two groups ensure that no common mode incident can disable the required systems of both groups. There is no unobstructed straightline path between redundant elements of the two groups above ground. Where there are no suitable obstructions, one of the elements is embedded in a suitable reinforcement. The group 1 control area is the main control room which is located on the third floor of the service building. The group 2 control area is a secondary control area 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 power and water supplies and water supplies Plant monitoring Main control room Secondary control area

FIGURE 4.2-5 main control centre. Cables for group 2 exit from the reactor building on a different side than those for group 1. Control is normally exercised from the main control centre, but under emergency condition*, control for shutdown and decay heat removal is also available from the secondary control area.

The criteria for separation and independence between the special safety systems belonging to the two groups are as follows:

1) Physical Independence. There must be no sharing of system components between group 1 and group 2 systems. There must also be no sharing of 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 two systems"must have conceptually different senses, instrumentation and actuators whenever practical. Where possible, similar components of the two systems must be supplied by different manufacturers. Where a choice is possible, such components must employ different principles of operation.

The best known common mode event is, of course, an earthquake. A brief outline of the design principles used in protecting against earthquakes will conclude this section on defense against common mode events.

1) Following an earthquake the reactor control system and the reactor shutdown 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 system pressure boundary.

4) The containment building and associated systems remain functional during and following an earthquake.

5) Structures and systems not directly required for nuclear safety reasons are designed so that their failure or dislocation is either unlikely, (systems are qualified), or do not effect the safe operation of any safety related systems required during or after the earthquake.

4.1-5 4.3 SAFETY SYSTEMS DESCRIPTION

This section describes those systems which are provided solely to perform a safety function and have no function in the normal production of electrical power. As noted earlier, these systems consist of the containment system, the emergency core cooling system, shutdown system #1, shutdown system #2, the emergency water supply system and the emergency power 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 sinks consisting of an automatically initiated dousing system and building air coolers, a filtered air discharge system, access airlocks and an automatically initiated containment isolation system.

The dousing tank is located in the dome of the reactor building and holds water for emergency dousing and emergency core cooling. About 500 cubic meters of water are reserved for emergency core cooling. The total capacity of the tank is about 2600 cubic meters. Dousing valves control the flow'of water to six independent dousing spray header units located radially below the tank. Each spray unit has two butterfly valves in a downcomer between the tank and the spray header (Figure 4.3-3). The design dousing flow rate is about 4500 kg/s and this flowrate can be provided by any four of the six downcomers. With all six downcotners operating, the total spray flow is about 6800 kg/s.

4.3.1.1 Operation

Under normal operation conditions, the pressure within containment is slightly less than atmospheric and the containment ventilation dampers are all open.

For very small heat transport system leaks, the building coolers in the containment condense any steam that is discharged, the building pressure remains at atmospheric pressure and there may be some additional outflow of dried air through the ventilation system. For larger breaks, the building pressure rises and at an overpressure of about 3.4 kPa, containment pressure sensors initiate total containment closure. The containment will also be automatically isolated in the event of a high radioactivity signal which may occur following a large loss of coolant accident. The containment pressure continues to rise and the dousing system starts to operate automatically at an overpressure of 13.8 kPa Depending on the break size, there is either continuous or cyclic operation of the dousing valves, with the valves opening at 13.8 kPa and

4.1-6 SPECIAL SAFETY SYSTEMS

Emergency water supply system Emergency power supply system Shutdown system 1 Shutdown system 2 Emergency core cooling system Containment system

FIGURE 4.3-1

CONTAINMENT SYSTEM

Designed for maximum size LOCA Small breaks — air coolers Large breaks — dousing system Filtered discharge system Multi-unit station •vacuum building

FIGURE 4.3-2 DOUSING WATER SUPPLY

DOUSING SPRAY HEADER

MAIN PRIMARY SYSTEM PUMPS

FIGURE 4.3-3 SINGLE UNIT CONTAINMENT closing at 6.9 kPa. When the dousing system overtakes the pressure transient, the pressure begins to fall, and the building depressurizes to about atmospheric pressure by condensation on the building walls and cooling by the air coolers. Initial containment atmosphere cleanup can be performed by the D2O vapour recovery dryers with long term purging achieved by discharging air through the dryers and the reactor building ventilation 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 three stages: high pressure, medium pressure, low pressure. The high pressure stage uses pressurized nitrogen to inject water into the reactor core from water tanks located outside the reactor building, the medium pressure stage supplies water from the dousing tank. When this water supply is depleted, the low pressure stage recovers water that has collected in the reactor building sump and pumps it back into the reactor core via the emergency cooling heat exchanger and the emergency cooling recovery pumps.

The high pressure injection stage consists of one nitrogen gas tank and two water tanks. The gas tank normally operates at a pressure between 4.1 MPa and 5.5 MPa, whereas the water tanks operate slightly above atmospheric pressure. The recovery pumps are two 100% pumps. Each pump is supplied by Class III power and by the emergency power supply system. The heat exchanger in the recovery pump discharge line is designed to maintain the emergency cooling flow at about 50°C at entry to the heat transport system.

Operation. The emergency core cooling system is triggered automatically when the heat transport pressure reaches 5.5 HPa. The following actions take place: a) All gas isolation valves, the high pressure injection valves, and the D2O isolating valves are opened. This will open rupture discs in the injection lines and permit the flow of high pressure, water from the injection tanks to all reactor headers of the failed and the unfailed loops (Figure 4.3-5). b) The main steam safety valves on the steam generators are opened to rapidly 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 are closed. This will confine the consequences of the loss of coolant accident to just the loop containing the hypothesized break. Sufficient coolant is available during the high pressure injection phase for at least 2.5 minutes.

4.1-7 EMERGENCY CORE COOLING SYSTEM

All-points injection Reactor is low-point in system Three stage injection High pressure — external tanks Intermediate — dousing tank water Low pressure — building sump NOT "LAST DEFENCE" FOR LOCA

FIGURE 4.3-4 FROM DOUSING TANK

GAS ISOLATION VALVES •A PV81 PV8 PV82 |^^ MP TEST VALVES

MP INJECTION VALVES ECC ijl'l j ti l II I WATER TANKS KKKHr O ~^^£ ECC GAS TANK

ECC HEAT EXCHANGER I

| Cxj—-W- V5 V3

I X}—-W- V6 V4 ECC PUMPS RIH = REACTOR INLET HEADER ROH = REACTOR OUTLET HEADER LOOP 1 LOOP 2 LOOP 1 MP = MEDIUM PRESSURE HP = HIGH PRESSURE 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 be opened automatically on the loss of coolant signal and one of the ECC pumps will be started when these valves are opened. If this pump fails to start (as indicated by a low pump discharge pressure), the standby ECC pump will be started automatically. No operator action is required to start the recovery pumps or to open the valves to supply dousing tank water to the pumps for the medium pressure emergency core phase. This phase will continue to supply water for at least 15 minutes following the largest break in the heat transport system. It will last longer for smaller breaks (Figure 4.3-6).

As the dousing tank water depletes, the operator opens valves in the recovery line from the reactor building sump, then closes the valves in the line from the dousing tank and opens the cooling water valves to supply service water to the ECC heat exchanger. The mixture of heat transport coolant and water from the high pressure and dousing tanks is pumped from the sump in the reactor building back to the heat transport system via the heat exchanger (Figure 4.3-7). For large breaks, the ECC recovery heat exchanger is the main heat sink. For small breaks, the steam 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 rapidly terminating any reactor power increase or reducing reactor power when certain parameters exceed preset values. This is accomplished by the release of 28 cadmium rods which fall under gravity from the top of the reactor. Figure 4.3-9. This gravity drop is accomplished by de-energizing direct current clutches which normally hold the shutoff rods out of the core. The shutoff rod units are divided into two banks of fourteen. Each bank is supplied with dual 90 volt DC power supply for the clutches. Each clutch coil is held energized by the contact of the separate relay.

The design philosophy is based on triplicating the measurement of each of the variables that can initiate reactor shutdown. Protective action is initiated when any two of the three measurements exceed their preset values. The selection of variables is such that where practicable, there are at least two different sensing parameters for the specific process failure being protected against. Examples of trip parameters on shutdown system #1 are high neutron power, high rate log neutron power, high heat transport pressure, high reactor building pressure, low steam generator level, low pressurizer level.

A partial drop test facility is provided to allow the operation of each shutoff unit to be checked during reactor operation. The shutoff unit housings are located on the reactivity mechanism deck which permits regulated, one unit at a time, access to the clutches, motors, potentiometers, gear boxes, and winches for removal or for maintenance on power.

4.1-8 FROM DOUSING TANK

GAS ISOLATION VALVES

PV8 MP TEST VALVES

MP INJECTION VALVES

RIH = REACTOR INLET HEADER LOOP 1 LOOP 2 LOOP 1 ROH = REACTOR OUTLET HEADER MP = MEDIUM PRESSURE TO PRIMARY HEAT TRANSPORT SYSTEM HP = HIGH PRESSURE

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 system 28 Cd shutoff units Devices inserted from top Instrumentation on top or vertical Primary + alternate trip parameter for each process failure Independent of SDS2

FIGURE 4.3-8 Separately channelled Class I and Class II power supplies are provided for each channel of shutdown system #1. The logic is arranged so that any loss of power to a channel results in a channel trip. The direct current clutches energized by rectified Class II power will release if power is disrupted. This will shut down the reactor.

The static negative reactivity worth of the 28 shutoff rods is about 80 ink.

4.3.4 Shutdown System #2

The second method of quickly terminating any reactor power increase or rapidly reducing reactor power is by the injection of concentrated gadolinium nitrate poison solution into the moderator through six horizontal nozzles, Figures 4.3-9 and 4.3-10. A vessel containing high pressure helium supplies the source of energy for this rapid injection. This vessel is connected through six quick opening valves to a helium header which supplies the poison tanks. The quick opening valves are "air-to-close, spring-to-open" design so that loss of instrument air initiates automatic poison injection. Each of the poison tanks contains gadolinium nitrate solution at a concentration of about 8,000 parts per million. The six zircaloy injection nozzles penetrate the calandria horizontally and at right angles to the fuel channels. Holes are drilled into the nozzle along its length to form four rows of jets which facilitate complete dispersion of the poison into the moderator (Figure 4.3-11).

Each poison tank contains a floating polyethylene ball. When an injection is initiated the helium driving gas transfers the poison to the calandria and the ball is driven to the tank bottom. In the bottom position, the ball seats at the poison tank outlet and prevents the release of a large volume of helium into the calandria. as with shutdown system #1 all initiating variables are triplicated and protective action is initiated by any two of these measurements exceeding preset values.

The eventual negative reactivity from the poison injection system is in excess of 300 mk.

4.3.5 Emergency Water Supply System

This system (ref. Figure 4.3-12) is designed to provide an alternate source of water to cater for: a) A design basis earthquake.

4.1-9 CALANDRIA SHUTOFF ROD GUIDE TUBE

MODERATOR

SHUTOFF ROD (TYPICAL)

LIQUID POISON NOZZLE

LIQUID POISON PIPE (TYPICAL)

CALANDRIA TUBE

FIGURE 4.3-9 SHUTDOWN SYSTEMS: SHUTOFF RODS AND LIQUID "POISON" INJECTION SHUTDOWN SYSTEM 2

Independent of regulating system 6 Gd injection nozzles Devices located horizontally Instrumentation on side or horizontal Primary + aitemate trip parameter for each process failure Independent of SDS1

FIGURE 4.3-10 HELIUM VENT LINES TO EXHAUST

PRESSURE BALANCE LINE

HELIUM COVER GAS

FROM HELIUM

-----

:--¥i-r-;:--|

GADOLIUM NITRATE IN HEAVY WATER ""

-f OOOOOOOOOOOOO H.P. HELIUM SUPPLY TANK 6 NOZZLES

HEAVY WATER MODERATOR CALANDRIA ISOLATION BALL VALVE y (NORMALLY OPEN) POISON-MODERATOR INTERFACE

FIGURE 4.3-11 SCHEMATIC OF SECOND SHUTDOWN SYSTEM CONNECTION FOR ADDITIONAL PUMP FOR TWO UNIT 3114- OPERATION VI TO V4 V9 I I I I FROM REHEATER DRAIN

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 site design earthquake*

Two electrically driven pumps are located remote from the group 1 equipment- and pump house. The pump suction is taken from a sump connected to a good quality water supply. The emergency water system connects to the heat transport system, the steam generators and the emergency core cooling heat exchange. Since the emergency water supply system is not required immediately following the three events listed above, system operation consists of manually starting the pumps and then operating the hand switches to open the appropriate motorized isolating valves to supply water to the required loads.

4.3.6 Emergency Power Supply System

This system is designed to provide an alternate source of power to cater for the events already described in the previous paragraph. The emergency power supply system supplies the necessary power to the emergency water system pumps and valves, the emergency core cooling pumps and certain emergency core cooling valves, and power to the group 2 safety and control systems for operator control of the station from the secondary control area. 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 close control of core reactivity and efficient utilization of its natural uranium fuel.

The operation is carried out by an automated fuel handling system (Figure 5.0-1) which utilizes features and equipment applied in the very successful Pickering "A" Nuclear Generating Station.

The full power refuelling requirements for CANDU 600 MW(e) reactors involve replacing about 110 fuel bundles per week. Using the reference procedure of replacing eight bundles per fuel channel, this entails refuelling 14 fuel channels per week.

In a.n eight bundle refuelling sequence (Figure 5.0-2), closure plugs are removed from a channel, and stored within the magazines of the fuelling machines coupled to each end of the channel.' Four pairs of new fuel bundles are then inserted from the upstream fuelling machine, displacing eight spent fuel bundles into the downstream machine. This process is termed "Flow Assisted Fuelling" (FAF Mode).

The overall refuelling operation of a CANDU 600 HW(e) reactor unit (Figure 5.0-3) comprises:

1) Loading new fuel bundles, in pairs, into a fuelling machine.

2) Coupling this machine onto the upstream end of the channel to be refuelled and coupling a second machine onto the downstream end of that channel

3) Displacing spent fuel into the downstream fuelling machine

4) Disengaging the downstream fuelling machine, and moving it to the spent fuel port

5) Discharging the spent fuel through the spent fuel port into an elevator, which lowers the bundles into the fuel Discharge Bay to an underwater conveyor '

6) Transferring the spent fuel from the conveyor to storage trays for stacking 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 STRUCTURE 2 FUELLING MACHINE GUIDE COLUMN 3 BRIDGE SUPPORT A BALL SCREW ASSEMBLIES 5 FUELLING MACHINE HEADS 6 FUELLING MACHINE CARRIAGE TROLLEY 7 ROLLING SHIELD

FIGURE 5.0-1 FUELLING MACHINE USED IN PICKERING AND CANDU 600 MW(e) PHWR'S DISCHARGE MACHINE

1B

RAM ® WITHDRAWS CLOSURE PLUG © RAM ©WITHDRAWS CLOSURE PLUG ©

2B

RAM ® CONNECTS TO SHIELD PLUG © RAM ©CONNECTS TO SHIELD PLUG© AND WITHDRAWS IT TO MAGAZINE

3A 3B

V'EW FUEL BUNDLES ROTATED INTO CHARGING POSITION. TWO BUNDLES AT A TIME

4A 4B

RAM ©CHARGES NEW SHIELD PLUG ©WITHDRAWN ACTIVITIES 3A and 4A REPEATED 4 TIMES SIDE STOPS (3)COME IN UNTIL 8 NEW BUNDLES HAVE BEEN INSERTED. THE FUEL COLUMN IS ADVANCED.

SB

USED FUEL BUNDLES WITHDRAWN TO MAGAZINE, TWO BUNDLES AT A TIME

SIDE STOPS® HOLD FUEL COLUMN AND ROTATING MAGAZINE SWINGS USED FUEL BUNDLES OUT OF THE 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 END SHIELD HAM (©REPLACESCLOSURE PLUG®

* NEW

FIGURE 5.0-2 8-BUNDLE CHANGING SEQUENCE IN A CANDU 600 MW{«) PHWR NEW FUEL STORAGE ROOM

NEW FUEL LOADING AREA

SPENT FUEL DISCHARGE ROOM

SPENT FUEL STORAGE BAY CANNED FAILED FUEL STORAGE TRAYS

FIGURE 5.0-3 FUEL HANDLING SYSTEM SCHEMATIC Accordingly, operation on a regular two shift/ five day basis provides about 100% availability margin for any operational interruptions, checks or delays.

5.1 FUELLING MACHINES

The major elements of the refuelling system are a pair of identical unshielded fuelling machines which operate at both ends of the reactor and bring new fuel from the New Fuel Ports to the reactor, and carry spent fuel to the Spent Fuel Ports (Figure 5.1-1).

Fuelling machines are mounted in carriages which move from maintenance area tracks onto bridges, mounted on columns in the fuelling machine vaults.

These raise and position the fuelling machines at each end of the fuel channel end fittings, to form a sealed connection before starting the fuel replacement 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 is used for removal, storage and replacement of fuel channel plugs, and for insertion of fuel. The closure plugs are engaged, unlocked and withdrawn into a storage chamber in the magazine; similar motions are used in withdrawal of shield plugs.

New fuel is pushed by the upstream ram into the pressure tube, where the heavy water flow brings it into contact with the installed fuel string.

In central channels the hydraulic forces are sufficient to move the entire fuel string along the fuel channel to displace pairs of bundles into the downstream fuelling machine. (FAF Mode).

A FARE tool (Flow Assisting Ram Extension) is inserted to provide additional hydraulic flow resistance in refuelling outer channels which have less flow (corresponding to their lower power levels). This is termed FARE mode refuelling.

As each pair of fuel bundles enter the downstream magazine, separators are inserted to limit motion of the following bundles, to advance the two downstream bundles and to enable free rotation of the magazine.

After completion of the refuelling sequence the shield and closure plugs are reinstalled. The fuelling machine is sealed by installation of a snout plug, the space between the closures is drained, and the fuelling machines disengage from the fuel channel.

5.1-2 CATENARIES NEW FUEL \ EMERGENCY ACCESS PLUGS . PORT SPENT FUEL PORT REHEARSAL CHANNEL \ T.V. CAMERA SERVICE PORTS

S3 J

FIGURE 5.1-1 FUELLING MACHINE VAI,LT AND MAINTENANCE ROOM 1 ANTENNA PLATE 1« FRONT RETAINING PLATE 2 ANTENNA SWITCH 20 WEIR 3 CLAMPING LEVER ARM 21 30" GRAYLOC CLAMP 4 CAM BLOCK 22 GRAYLOC SEAL RING S WEDGE SEGMENT 23 MAGAZINE HOUSING « CLAMPING BARREL 24 MAGAZINE DMVE SHAFT 7 SEAL 25 REAR RETAINING PLATE 8 SNOUT PROBE 2S BALANCE SHAFT SEAL S LOCK RING 27 FERGUSON INDEXING DRIVE 10 SCREW AND GEAR 28 FLOW SHIELD 11 CENTRE SUPPORT 2» RAM HEAD 12 SEPARATOR ASSEMBLY 30 MAGAZINE POSITION 13 FUEL STOPS POTENTIOMETERS A SNOUT PLUG SHIELD PLUG 14 CLAMPING PISTON 31 MAGAZINE EMERGENCY B FUEL J ADAPTER 15 RACK DRIVE GEARBOX C CHANNEL CLOSURE t CHANNEL CLOSURE (SPARE) 16 SNOUT EMERGENCY 32 MAGAZINE DRIVE MOTOR 0 FUEL 1 FUEL LOCK ASSEMBLY 33 10" GRAYLOC CLAMP E GUOE SLEEVE (INSERTION TOOL I SHIELD PLUG (SPARE) 17 LOCK PISTON 34 RAM HOUSING F FUEL I FUEL 18 MAGAZINE END COVER 36 EDUCTOR

VKW PROM THE REAR

FIGURE 5.1-2 FUELLING MACHINE HEAD Subsequently thfi Fuotlinq machines are lowered and driven into the maintenance 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 and cooling, and the drive supplies to the mobile fuelling machines, from the static 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 Room which can accommodate a complete reactor's inventory of new fuel. This is equivalent to about nine months' refuelling supply at 80% capacity factor.

Fuel is transferred by pallet and lift truck to the new fuel loading area. Two New Fuel Transfer Mechanisms are installed there for inserting fuel into fuelling machines in either maintenance area (Figure 5.2-1).

Fuel is hoisted from its pallet, inspected and loaded into a loading trough. Pairs of fuel bundles are pushed into the fuel loading magazine under semi-automatic control. Subsequently they are transferred, by a motor driven ram under fully automatic control, into vacant magazine positions in the waiting fuelling machine.

An airlock gate valve in the transfer port minimizes the transfer of any contamination from the fuelling machine or maintenance room.

At all times except during fuel loading the new fuel port houses a shield plug 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 of spent fuel to the fuel storage bays (Figure 5.2-2).

As these ports constitute physical penetrations of the Reactor Containment Building, they each embody pairs of isolating valves. Drains, sprays and pressure relief devices are also provided for containment protection.

After the fuelling machine has been positioned and coupled to the spent fuel port, its heavy water level is lowered and its snout plug removed and stored. Its magazine is then rotated for dry transfer of one pair of fuel bundles, through the port, onto the transfer mechanism for lowering into the fuel transfer canal.

5.1-3 FUELLING MACHINE CARRIAGE

TROLLEY TRACK FUELLING MACHINE MAINTENANCE LOCK TRACK

HOSE AND CABLE CARIER

CATENARY TROLLEY

- MOTION CATENARY LOOP

FUELLING MACHINE HEAD

TROLLEY DRIVE UNIT

HOSE AND CABLE CARRIER TRACK

CATENARY LOOP

NORTH ('A') ASSEMBLY SHOWN

FIGURE 5.1-3 FUELLING MACHINE SUPPORT AND CATENARY SYSTEMS NORTH CA1) NEW FUEL TRANSFER MECHANISM SOUTH CC1) NEW FUEL TRANSFER MECHANISM NEW FUEL PORTS CONTROL PANELS BUNDLE INSPECTION TABLE BUNDLE LOADING TROUGHS 1/4 TON JIB CRANE AND AIR HOIST 2-TON JIB CRANES AND ELECTRIC HOISTS REMOVABLE PLATFORM NEW FUEL PALLETS FUELLING MACHINE MAINTENANCE LOCK CRANE RAIL SERVICE PORT ENCLOSURES IRRADIATED FUEL HANDLING SOUTH LADLE DRIVE FUELLING MACHINE TRANSPORT CART GUIDES LIQUID INJECTION SHUTDOWN SYSTEM ENCLOSURE 15-TON CRANE

FIGURE 5.2-1 600 MW(e) NEW FUEL TRANSFER EQUIPMENT 1 END FITTING o BALL VALVES 3 ELEVATING LADLE HOISTS 4 ELEVATING LADLE DRIVE (IN NEW FUEL ROOM) 5 ELEVATING LADLES 6 MAIN ELEVATOR RAILS 7 GUIDE RAILS 8 FUEL POSITIONING ASSEMBLIES 9 LOWER RAIL SUPPORT 10 AUXILIARIES 11 SPRAY HEADERS 12 REMOVABLE PLATFORMS 13 FUEL TRANSFER EQUIPMENT 14 DEFECTED 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 bay for loading onto trays and for subsequent storage (Figure 5.3-1).

If there should be an indication of a defective fuel bundle, the suspect bundle is transferred to a carousel which collects any fission gas bubbles released. After a suitable period of cooling the defective fuel is canned, then transferred to temporary storage.

The fuel storage bay capacity is sufficient to accommodate at least ten year's output of spent fuel under normal conditions.

5.1-4 6 CONVEYOR DRIVES 7 RECEPTION BAY 8 TRANSFER RACK 9 TRANSFER RACK HANDLING TOOL 10 RACK HANDLING TOOL STORAGE BRACKET 11 2-TON RECEPTION BAY CRANE 12 SINGLE RACK STAND-OFF 13 EMPTY RACKS ON TRIPLE RACK STAND-OFF 14 STORAGE TRAY STAND 15 PARTIALLY FILLED TRAY ON STAND 16 BUNDLE LIFTING TOOL 17 FULL STORAGE TRAYS 18 STORAGE TRAY CONVEYOR 19 CONVEYOR DRIVE 20 STORAGE TRA.V LIFTING TOOL 21 SPENT FUEL STORAGE BAY 1 SPENT FUEL DISCHARGE EQUIPMENT 22 EMPTY STORAGE TRAYS (REF. ONLY) 23 DEFECTED FUEL TRANSFER EQUIPMENT 2 TRANSFER RACK DETECTION (REF. ONLY) SWITCH LEVER 24 DEFECTED FUEL STORAGE BAY 3 DISCHARGE BAY CONVEYOR 25 DEFECTED FUEL BAY ISOLATION 4 TRANSFER CANAL CONVEYOR VALVE (REF. ONLY) 5 TRANSFER CART 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 ahead of all the countries of the Western World when the achievements of all thermal reactors are considered (Figure 6-1).

. CANDU is a conserver of uranium supplies, its once through fuel cycle uses 15% less natural uranium than light water reactor (LWR) fuel cycles. This fuel savings increases to 38% if 1.2% enriched uranium fuel is used.

. CANDU is a flexible system, it can be adapted to advanced fuel cycles employing other fissile or fertile materials: thorium and plutonium.

. From a safety point of view the containment of the core environment in many small diameter pressure tubes is preferable to one very large and heavy pressure vessel. The pressure tubes have been proven to exhibit the "leak before break" characteristic that is so important in safety considerations. The presence of heavy water is readily detectable and serves notice of any leak in the system. As pointed out previously the CANDU system uses a defence in depth approach with redundant safety systems.

. On-power refuelling, a unique feature of the CANDU, permits the immediate correction of low reactivity areas in the reactor core. Another advantage of this feature is its ability to quickly remove any failed 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 of Ontario in 1979 is shown in Figure 6.2. These numbers have been brought to a common base for direct comparison and .clearly show the CANDU advantage.

6.2 CONCLUSION

This presentation provides a technical summary of the many systems that together make up a CANDU reactor, and at the same time outlines the reasons for CANDU1s superior position among the power reactors of the world today.

While focusing primarily on the CANDU 600 MW(e), AECL is able to offer customers larger reactor units (up to 950 MW(e)), if required.

6.1-1 WORLD POWER REACTOR LIFETIME PERFORMANCE

1 CANADA Pickering-2 542 MW 84.5% 2 W. GERMANY Stade-1 662 MW 83.5% 3 CANADA Pickering-1 542 MW 83.3% 4 CANADA Bruce 4 791 MW 78.5% 5 CANADA Bruce 3 791 MW 78.2% 6 CANADA Pickering-4 542 MW 77.6% 7 CANADA Pickering-3 542 MW 77.5% 8 USA Point Beach 2 524 MW 77.4% 9 USA Connecticut Yankee 602 MW 75.4% 10 SWEDEN Barsebaeck2 600 MW 74.5%

CUMULATIVE LOAD FACTORS FOR REACTORS OVER 500 MW(e) TO END OF SEPTEMBER 1980

Station Cumulative Load Type Factor %

Bruce-3 82.0 CANDU Stacto-1 81.2 PWR Picfcerlnjj-2 80.9 CANDU Pickwing-1 80.3 CANDU Point BMCh-2 77.4 PWR Picker! nc-4 77.3 CANDU Pick»ring-3 75.4 CANDU Prairie ltland-2 75.2 PWR CaKwrt CllfU-2 74.7 PWR Connecticut Yankee 74.6 PWR Bruce-4 73.5 CANDU Bruce-1 73.0 CANDU

REF: NUCLEAR ENGINEERING INTERNATIONAL VOL 25, NO. 307, 1980

Country Annual load Number and size Cumulative load Number and size factor% of reactors factor % of reactors

Canada 70.11 10 (5818 MWe) 64.90 10 (5818 MWe) Europe 56.35 53 (38500.3 MWe) 56.48 57 (42284.3 MWe) USA 56.76 68 (54684 MWe) 54.74 68 (54658 MWe) Japan 48.41 20 (13852 MWe) 52.40 22 (15117 MWe) UK 51.77 22 (7949.3 MWe) 53.24 22 (7949.3 MWe) France 56.63 10 (6429 MWe) 51.87 12 (8343 MWe) W. Germany 50.49 10 (14299 MWe) 54.95 11 (15199 MWe)

Source: Nuclear Engineering International (March 1980)

FIGURE 6-1 COAL

FUEL 17.06

NUCLEAR

FUEL 1.7

CAPITAL COST, OPERATION, MAINTENANCE AND HEAVY WATER UPKEEP CAPITAL COST, OPERATION AND 12.8 MAINTENANCE 10.19

27 .27 COSTftnfcWh 14.S

1979 BREAKDOWN OF UNIT ENERGY COSTS (ONTARIO HYDRO FIGURES)

FIGURE 6-2 COMPARISON OF COSTS — COAL AND NUCLEAR