ML023040268
| ML023040268 | |
| Person / Time | |
|---|---|
| Site: | Millstone, Davis Besse, Summer, North Anna, Ginna, Diablo Canyon, Robinson, South Texas, Yankee Rowe, Zion |
| Issue date: | 09/19/2002 |
| From: | Westinghouse |
| To: | NRC/FSME |
| References | |
| -RFPFR, FOIA/PA-2002-0343 | |
| Download: ML023040268 (149) | |
Text
Westinghouse Technology Manual Chapter 17.0 Plant Operations
Westinghouse Technology Manual Plant Operations TABLE OF CONTENTS 17.0 PLANT OPERATIONS..
17-1 17.1 Introduction..................................................
17-1 17.2 Plant Heatup..
17-1 17.2.1 Initial Conditions..
17-1 17.2.2 Operations..............................................
17-1 17.3 Reactor Startup to Minimum Load...................................
17-3 17.4 Power Operations..............................................
17-4 17.5 Plant Shutdown..
17-5 Appendix 17-1 Plant Startup from Cold Shutdown...............................
17-6 LIST OF FIGURES "17-1 Solid Plant Operations...........................................
17-11 17-2 Control Bank Insertion Limits for 4-Loop Operation.......................
17-13 USNRC Technical Training Center 17-i Rev 0198
Westinghouse Technology Manual Plant Operations 17.0 PLANT OPERATIONS Learning Objectives:
- 1. Arrange the following evolutions in the proper order for a plant startup from cold shutdown:
- a.
b.ý C.
- d.
e.°
- f.
g.
Start all reactor coolant pumps, Place all engineered safety systems in an operable mode, Establish no-load Tavg, Take the reactor critical, Start a main feedwater pump, Load main generator to the grid, and Place steam generator level control system in automatic.
17.1 Introduction This chapter will briefly discuss the,basic procedures for startup, power operation, and shutdown of the pressurized water reactor described in this manual. The discussion will be general in nature and is designed to show how the systems previously discussed are utilized, during'plant operations.
'17.2. Plant Heatup 17.2.1 Initial Conditions
-The nuclear steam supply system (NSSS) is in the "cold shutdown" mode (Tavg = 120*F, pressurizer pressure = 50 - 100 psig, boron concentration sufficient to yield 10% shutdown' margin, pressurizer solid, reactor coolant pumps off). Decay heat is being removed by the residu al heat removal system (RHR) with letdown from RHR established for reactor coolant system cleanup. Pressure in the solid system (Figure" 17-1) is being maintained by adjusting charging, and letdown flow. The steam generators are in the "wet layup" condition (filled to the 100%
'level with water) and all secondary systems are secured with the exception of one circulating water pump. The main and feedwater pump turbines are on the turning gear. All pre-startup checklists have been completed.
17.2.2 Operations A pressurized water reactor may have a
- positive moderator temperature coefficient-at low
- temperatures due to the soluble poison in the moderator. To minimize the-magnitude of the positive moderator temperature coefficient or make it negative, the plant is,brought to near operating temperatures with reactor coolant pump heat before 'the reactor is made critical. To operate the reactor coolant pumps, reactor coolant system pressure must be increased to, approxi mately 400 psig to satisfy net positive suction head requirements. - (Pressure must be main tained below 425 psig while RHR is aligned to the reactor coolant system.)- When operating the reactor coolant pumps at low pressures, the reactor coolant-pump number one seal bypass
-valve must be open to ensure adequate flow to cool and lubricate the pump radial bearing.
Pressure is increased by maintaining charging flow greater than letdown flow. -When pressure is stable between 400 and 425 psig, the reactor coolant pumps are started to begin'reactor coolant "system heatup. Pressurizer heaters are energized
'to begin pressurizer heatup. 'Residual 'heat removal flow is diverted through 'the bypass line "to bypass the heat exchanger and allow heatup.
This RHR system alignment is maintained to provide adequate letdown for pressure control and to remove the excess coolant volume pro duced by expansion due to heatup. During the USNRC Technical Training Center 17-1 Rev 11198 Westinghouse Technology Manual Plant Operations Rev T0198
- USNRC Technical Training Center 17-1
Westnghuse echolog MaualPlant Operations entire heatup and pressurizer draining process, approximately one-third of the reactor coolant system volume (30,000 gallons of water) will be diverted to the holdup tanks through the chemical and volume control system.
As the reactor coolant system temperature approaches 200'F, steam generator draining is commenced through the, normal blowdown system. If reactor coolant system oxygen con centration is high, hydrazine is added through the chemical and volume control system for oxygen scavenging. Oxygen must-be in specification before exceeding 250"F.
After oxygen is within specification, a hydrogen blanket is established in the volume control tank. This is accomplished by securing the nitrogen regulator, opening the vent from the volume control tank to the waste gas header, and raising the volume control tank level to force the nitrogen to the waste gas system. After the volume control tank level has raised to approxi mately 95%, the hydrogen regulator is placed in service and the last of the nitrogen is purged to the waste gas system. Volume control tank level is allowed to return to normal with the hydrogen regulator maintaining an overpressure of approxi mately 15 - 20 psig.
When pressurizer temperature reaches satura tion temperature for the pressure being main tained (450°F for 400 psig), a pressurizer bubble is established. Reactor coolant system tempera ture is approximately 250 - 300TF. The bubble is established by maximizing letdown and minimiz ing charging flow. This will cause the pressuriz er level to decrease. System pressure will be maintained at 400 psig as the saturated pressuriz er water flashes to steam. Pressure control can now be accomplished only by heater and spray operation. Residual heat removal is maintained in service to provide an additional letdown path to minimize the time necessary to "draw a bub ble" in the pressurizer.
The main and auxiliary steam lines are warmed as steam is available during the plant heatup. Main steam isolation valves are opened initially as heatup begins.
As reactor coolant pressure continues to increase, letdown flow will also increase. The low pressure letdown valve is adjusted (closed) until the normal letdown pressure (340 psig) is achieved and then orifice isolation valves are shut as necessary to maintain letdown flow below the maximum.
Before reactor coolant system temperature reaches 350°F, the residual heat removal system is isolated from the reactor coolant system and is aligned for at-power operation (emergency core cooling system lineup).
All reactor. coolant system letdown is now through the normal letdown orifice path to the chemical and volume control system.
After the residual heat removal system is isolated from the reactor coolant system, system pressure is allowed to increase as the pressurizer temperature increases.
When pressurizer level, as read on the hot calibrated channels, indicates the no-load pro grammed setpoint, charging flow is placed in automatic. As system heatup continues, pressur izer level will try to increase due to coolant
,expansion. Pressurizer level control will com pensate by reducing charging flow.
When reactor coolant system pressure reach es 1,000 psig, the emergency core cooling system accumulator discharge valves are opened USNKC Technical Training Center 17-2 Rev 0198 Westinghouse Technology Manual Plant Oneration*
U/SNRC Technical Training Center 17-2 Rev 0198
Westinghouse Technology Manual Plant Operations and all emergency core cooling system equipment is checked for proper alignment.
After reactor coolant pump number one seal leakoff has increased to at least one gallon per minute on all reactor coolant pumps, the number one seal bypass valve is closed.
As pressure increases above P-11, the loy pressurizer pressure engineered safety feature actuation signal is automatically unblocked Pressurizer heaters and spray valves are placed ii automatic control when pressure reaches th(
normal operating value of 2235 psig.
When steam pressure is at or above 125 psig main and feed pump turbine gland seals ari established, and a condenser vacuum is drawn Condenser vacuum is established by mechanica vacuum pumps and/or steam jet air ejectors.
As reactor coolant system heatup continues the high steam flow engineered safety feature actuation signal will be automatically unblocke4 when Tavg increases above 540"F. The -stean dump system, operating in pressure contro mode, will 'dump steam to the main condense when steam pressure reaches a predeterminec setpoint (normally 1,005 psig which is saturatior pressure for the 547°F no-load reactor coolan system temperature). The steam dump systen will dissipate the excess decay and reacto coolant pump heat,and maintain Tavg approxi mately equal to 547"F. The startup feedwate]
system is used to feed the steam generators t(
maintain level at the no-load value.
Plant conditions are now as follows: norma operating temperature and pressure, reacto shutdown, normal condenser vacuum, stean dump to the condenser in the steam pressur(
turning gear, and all electrical power supplied from off-site.
The next step in the startup of the plant is to take the reactor critical.
17.3 Reactor Startup to Minimum Load v
Reactor startups are normally performed at s
no-load temperature where the moderator temper ature coefficient is at a low or negative value.
If necessary, the reactor coolant boron concentration is adjusted-to the required value prior to startup. The required value is calculated by performing -a -reactivity balance (estimated e
critical condition calculation). For a pressurized
' water reactor; a specific critical rod height is 1,
chosen and boron concentration is adjusted to a value which will produce criticality at the desired rod height. Control rods must always be with drawn abovethe rod insertion limit prior to s
criticality to ensure adequate "cocked" reactivity I
to satisfy shutdown margin'requirements.
1 *- -* -Immediately prior to reactor startup, function r - al checks are performed to ensure proper opera I
tion of the source and intermediate range nuclear I
instrumentation channels. A source and interme t
diate range channel are recorded and the "source n
range high flux at shutdown" alarm is blocked.
r 1
1, -
1 1 The shutdown -rod banks- (if not 'already r
withdrawn) are withdrawn in sequence, and o
then, the control banks-are withdrawn in manual to achieve criticality. After criticality is achieved, a positive startup rate is established, and power level in increased.- When 'power exceeds the r
source range -permissive (P-6)' setpoint, the n
source range trip is blocked and source range e
high voltage deenergized.
mode, main and feedwater pump turbines on the USNRC Technical Training Center 17-3 Rev 0198 Westinghouse Technology Manual Plant Operations USNRC Technical Training Center 17-3 Rev 0198
Westinghouse Technology Manual Plant Operations Power is then increased to 10-8 amps in the intermediate range where neutron flux is stabi lized (lev iied out) and critical data are taken.
After critical data are taken, the reactor power increase is continued until the "point of adding heat" is reached. This is the power level (about 1 % power) where the reactor is producing sensible heat.
The reactor operator hold 1% power while the turbine-driven main feedwater pump is warmed and placed in service. Feedwater supply is switched from the auxiliary feedwater system
'to the main feedwater pump.
Reactor power is increased to' about 5%
power in preparation for rolling the main turbine.
Increasing reactor power will cause the steam dump valves to open further to dissipate the excess heat.
Steam generator feedwater is controlled manually through the small (4 -6 inch) bypass valves to maintain level at the program setpoint. Providing excess reactor power yields a constant steam load as the turbine is rolled. As the turbine takes more steam, the steam dump valves will modulate closed. This makes control of the reactor and steam generator levels much easier. A heater drain pump is energized at this time.
The turbine acceleration rate' is chosen, and the turbine is accelerated to synchronous speed.
With the turbine at synchronous speed, reactor power is increased to six percent so that reactor power is greater than the initial turbine load.
The turbine is synchronized with the utilities electrical grid, and thie generator output breaker is closed: The electrohydraulic control system automatically assumes five percent of full rated load. After turbine operation and other apolicable instrumentation is checked, a turbine loading rate is, chosen, and the turbine load is increased toward 15%.
As turbine load is increased, the reactor operator withdraws control rods to maintain Tavg Tref. During the load increase, the steam dump valves will shut as steam pressure-decreases.
When the valves are shut, steam dump control is shifted to Tavg control to be ready for a possible load rejection or reactor trip. Steam generator level continues to be controlled by maanual operation of the main feedwater regulating bypass valves.
When power level exceeds the setpoint of the nuclear at-powier permissive (P-10), tlhe interme diate range rod withdrawal stop and the inrterme diate and power range (low setpoint) trips are manually blocked.
At or above fifteen percent power, the rod control system and steam generator level control system are placed in the automatic mode.
17.4 Power Operations Power level is increased by selecting a desired load and load rate with the turbine electrohydraulic control system and allowing the reactor tofollow the turbine load change. As turbine load increases, Tavg will tend to decrease.
The automatic rod control system will sense this and withdraw control rods to increase reactor power.
As load'is increased to 30% power, a second condensate/booster pump is started, and main generator hydrogen pressure is increased to its maximum value (75 psig).
As load increases between thirty and fifty percent, additional circulating water, feedwater, USNRC Technical Training Center 17-4 Rev 0198 Rev 0198 Westinghouse Technology Manual Plant Operations IUSNRC Technical Training Center 17-4
Westinghouse Technology Manual Plant Operations and heater drain pumps are started. At approxi mately 35% load, reheating steam is cut into the moisture separator-reheaters.
. The single loop loss of flow permissive (P-8) enables the single loop loss of flow reactor trip when reactor power exceeds 35%.
At approximately 50% load, the third conden sate/booster pump is started, and a calorimetric (heat balance) calibration of the power range nuclear instruments is performed.
Further calorimetrics are performed at 70%
and 100% power to ensure proper calibration of the power range nuclear instrumentation.
Negative reactivity added by the power defect during the power increase is counteracted by automatic withdrawal of the control rods while the negative reactivity due to xenon and samari
-um,production is counteracted by dilution of soluble poison from the coolant.
At all time, when the reactor is critical, the control rod banks must be maintained withdrawn above their respective insertion limits (Figure 17
- 2). All shutdown banks and control banks "A" and "B" must be'fully withdrawn, and control banks "C" and "D" must be withdrawn at least as specified in Figure 17-2. Maintaining the rods above the rod insertion limit ensures sufficient available negative reactivity to achieve required shutdown margin in the event of a reactor trip.
17.5 Plant Shutdown Plant shutdown is accomplished by essential ly reversing the steps described in plant startup.
I USNRC Technical Training Center 17-S Rev O198
-Rev -0198 SPlant Operations Westinghouse Technology Manual USNRC Technical Training Center
--17-5
W n
snPlant-One-ations APPENDIX 17-1 G. Pre-startup checklists completed PLANT STARTUP FROM COLD SHUT DOWN I. INITIAL CONDITIONS A. Cold shutdown - Mode 5 Keff < 0.99 0% rated thermal power Tavg < 200*F B. Pressurizer
- 1. Temperature approximately 320"F, with a steam bubble established.
- 2. Level approximately 25% with level control in automatic.
C. RCS temperature 150 - 160"F Note: Temperature may be less than 1507F depending on decay heat load from the core.
D. RCS pressure 100 psig
- 1. Charging and RHR letdown established
- 2. RCS pressure maintained by pressurizer temperature @ 320"F
- 3. RHR system in operation E. Steam generators filled to wet-layup (100%
level indication)
F. Secondary systems shutdown. Main turbine and main feedwater pump turbines on their turning gear II. Instructions A. Heatup from cold shutdown to hot shutdown (Mode 5 to Mode 4)
- 1. Permission received from operation supervisor for startup
- 2. Verify shutdown rods withdrawn or verify sufficient shutdown margin avail ability
- 3. Verify dr establish RCP seal injection flow
- 4. Begin pressurizer heatup to increase RCS pressure.
CAUTION: Do not exceed a heatup rate of 100lF/hr on the pressurizer, lO0*F/hr'_on-the.RCS, or 320"F T between pressurizer and spray tem perature.
Use auxiliary sprays for pressurizer-RCS mixing.
- 6. Startup checklist for Technical Specifica tion requirements completed
- 7. Begin establishing steam generator water levels to 50% on narrow range indication (steam generator blowdown system).
- 8. Open main steam line isolation valves USNRC Technical Training Center 17.6 Rev 0198 Rev 0198
,* USNRC Technical Training Center Westinghouse Technology Manual Plant* Onet:ations 17-6
Westinghouse Technology Manual Plant Operations
- 9. 'If required, commence condensate cleanup
- 10. Establish condenser vacuum
- 11. Continue -pressurizer heatup to 430°F (RCS pressure 325 psig). Use the low pressure letdown control valve to main tain letdown flow. RCS pressure control is via heater and spray actuation.
- 12. Start the reactor coolant pumps. After five minutes running, sample the RCS for chemistry specifications. Partially open pressurizer sprays for mixing.
- 13. Stop residual heat removal system pumps
- 14. Allow RCS temperature to increase to 200"F
- 15. When RCS temperature reaches 200*F,,
-determine that primary system water
"- - chemistry is within specifications
- 16. When condensate chemistry is within specifications as determined by chemical"
, lab, 'align condensate and feedwater Ssystem to normal configuration.
- 17. Verify control rod drive cooling fans on before RCS temperature reaches 160'F
- 18. Terminate residual heat removal letdown' to chemical and volume control prior to
-_'.exceeding 350"F and 425 psig.
B. Heatup from Hot Shutdown to Hot Standby-(Mode 4-to-Mode 5)
- 1. Startup checklist for Technical Specifica tion requirements completed
- 2. Complete emergency core cooling system master checklist
- 3. As the RCS pressure increases, maintain letdown flow 120 gpm by increasing the setting of the low pressure letdown control valve, and by closing the letdown orifice isolation valves as necessary.
- 4. Prior to reaching 1;000 psig in the RCS, open each of the 'cold leg -accumulator isolation valves. -Remove each valve's power supply.
- 5. When RCP no. 1 seal leAkoff is > 1 gpm, or RCS pressure > 1,500 psig, close RCP seal bypass return valve. Verify no.
1 -seal leakoff remains > 1 gpm.
- 6. When RCS pressure reaches 1,970 psig, verify pressurizer low-pressure safety injection logic auto reset.
- 7. When Tayg exceeds 540°F, verify steam line safety injection logic auto-reset.
- 8. The steam dump control system is in pressure control mode (set at 1,005 psig)
- to maintain RCS temperature at 5470F.
- 9. Place RCS pressure control in automatic to maintain 2235 psig.
- 10. Establish hot standby conditions of 540 547"F Tavg.' -*
C. Heatup from-Hot Standby 'to Power
" :Operations. (Mode 3 to Mode 1)
- 1. Administrative permission to take the reactor critical has beenobtained.*
USNRC Technical Training Center 17-7 Rev U19
Plant Operations Westinghouse Technology Manual 17-7 Rev 0198
ý USNRC Technical Training Center
Westinghouse Technology Manual Plant Operations
- 2. Notify system dispatcher of unit startup and approximate time the generator will be tied on to the system.
- 3. Notify onsite personnel of reactor startup over P/A system.
- 4. If shutdown banks have not been with drawn, complete a shutdown margin calculation (assuming SD banks out) and if desired SD margin will exist, withdraw the shutdown banks to the fully with drawn position.
Note: Nuclear instrumentation shall be monitored very closely in anticipation of unplanned reactivity rate of change.
- 5. Calculate the estimated critical boron concentration for the desired critical control bank rod position (normally 150 steps on Bank D).
- 6. If necessary, conduct a boron concentra tion change to the estimated critical boron concentration. Equalize boron concentra tion between the reactor coolant loops and the pressurizer by turning on pressurizer backup heaters.
Note: Nuclear instrumentation shall be monitored very closely in anticipation of unplanned reactivity rate of change.
Note: Block the source - range high flux level at. shutdown alarm at both source range panels.
- 7. Withdraw the control bank rods in manual and take the reactor critical.
- a. Block source range trip at P-6
- b. Record critical data at 10-8 amps
- 8. If the control bank height at criticality is below the minimum insertion limits for the 0 percent power conditions.
- a. Re-insert all control bank rods to the bottom of the core.,
- b. Recalculate the estimated critical boron concentration
- c. Borate to the new estimated critical boron concentration
- d. Withdraw the control bank rods in manual and take the reactor critical
- 9. Withdraw rods to bring reactor power to approximately 1% on power range indicators and select the highest power range channel to be recorded on NR-45.
- 10. Start a main feedwater pump at 1% power and maintain steam generator levels at 50 percent narrow range level indication during secondary plant startup by throt tling the feedwater bypass regulating valves and operating the master feedwater pump speed controller and the individual steam generator feedwater pump control station in auto.
CA UTION: Coordinate all steam generator steam removal and significanit feedwater changes with the 'reactor panel operator while rod control is in manual USNRC Technical Training Center 17-8 Rev 0198
-1 USNRC Technical Training Center 17-8 Rev 0198
Westinq house Technolopv Manual PatOeain 11.Turbine has been on turning gear at least one hour
- 12. Increase reactor power by manual adjust ment of the control bank until the steam dump is bypassing steam flow equivalent to 8 percent nuclear power.
13.Verify the unit auxiliary and startup transformer cooling systems are aligned for automatic operation.
- 14. Start the turbine, bring it up to speed, and connect the generator to the grid. Trans fer station power from the startup trans former to the unit auxiliary transformer.
- 15. Increase generator load at the desired rate, while maintaining Tavg by manual rod control.
- 16. Transfer feedwater flow from bypass valves to the main feed regulating valves.
Maintain programmed level during this process.
- 17. When reactor power increases above 10 percent, ensure the nuclear at-power
-permissive (P-10) light comes on and the turbine at-power permissive (P-13) and at-power permissive (P-7) lights clear.
- 18. Manually block the intermediate range reactor trip and the power range low setpoint reactor trip after P-10 has been actuated.
- 19. When turbine power has increased above 15 percent, and Tavg equals Tref, transfer reactor control system to automatic.
- 20. After rod control is placed in automatic, check steam pressure less than steam dump set point and steam dump valves full closed, then transfer steam dump to Tavg mode.
- 21. Above 15 percent power, transfer steam generator feedwater regulating valve control to auto when level is at setpoint and steam flow equals feed flow.
- 22. Continue turbine load increase to 100%
- a. Start secondary system components as required during power escalation.
Additional components would include items such as condensate pumps, heater drain pumps, feedwater pumps, and condenser circulating water pumps.
- b. Maintain rate of load increase within plant design limits.
These limits would include the loading limits imposed upon the main turbine and the limits imposed by boron dilution rates.
USNRC Technical Training Center 17-9 Key UIY
USNRC Technical Training Center Plant Operations Rtev 0198 17-9
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Westinghouse Technology Manual Chapter 18.0 Overview and Comparision of U.S. Commercial Nuclear Power Plants
NUREG/CR-564C SAIC-89/1541.
Overview and Comparison of U.S. Commercial Nuclear Power Plants Nuclear Power Plant System Sourcebook Manuscript Completed: August 1990 Date Published: September 1990 Prepared by P. Lobner, C. Donahoe, C. Cavallin M. Rubin, NRC Technical Manager Science Applications International Corporation 10210 Campus Point Drive San Diego, CA 92121 Prepared for Division of Systems Technology Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 NRC FIN D1763
TABLE OF CONTENTS Section Page
- 2.
General Comparative Data for U.S. Commercial Nuclear Power Plants....................................................................
2-1
- 3.
Pressurized Water Reactors (PWR) System Overview.................... 3-1 3.1 Introduction to the Pressurized Water Reactor..................... 3-1 3.2 PWR Primary System................................................
3-1 3.3 Reactor Core and Fuel Assemblies..................................
3-2 3.4 Reactivity Control Systems..........................................
3-3 3.5 Heat Transfer Systems for Power Operation.......................
3-3 3.6 Heat Transfer Systems for Shutdown Cooling at High RCS Pressure...................................................
3-4 3.7 Heat Transfer-Systems for Shutdown Cooling at Low RCS Pressure....................................................
3-4 3.8 RCS Overpressure Protection System...................... :....... 3-4 3.9 Emergency Core Cooling Systems..................................
3-5 3.9.1 ECCS Injection Phase......................................
3-5 3.9.2 ECCS Recirculation Phase................................ 3-6 3.9.3 High-Pressure Feed-and-Bleed Cooling.................. 3-6 3.10 Containment and Containment Auxiliary Systems................ 3-7 3.10.1 Large, Dry Containment.................................... 3-7 3.10.2 Subatmospheric Containment.............................. 3-7 3.10.3 Ice Condenser Containment................................
3-7 3.10.4 Containment Auxiliary Systems........................... 3-7 3.11 Component Cooling Systems........................................
3-8 3.12 Safety System Actuation..............................................
3-9 3.13 Onsite Electric Power System........................................
3-9
- 4.
Westinghouse Pressurized Water Reactors (PPWRs)....................... 4-1 4.1 Westinghouse PWR Overview...................................... 4-1 4.2 2-Loop Westinghouse PWRs........................................
4-17 4.3 3-Loop Westinghouse PWRs........................................ 4-24 4.4 4-Loop Westinghouse PWRs........................................ 4-36 4.5 Westinghouse PWR Comparative Data............................. 4-55 8/90 V
LIST OF FIGURES Figure P=
1.3-1 Key to Symbols in Fluid System Drawings........................... 1-14 1.3-2 Key to Symbols in Electrical System Drawings............................
1-16 3.2-1 Westinghouse 2-Loop PWR NSSS.........................................
3-12 3.2-2 Westinghouse 3-Loop PWR NSSS.......................................... 3-13 3.2-3 Westinghouse 4-Loop PWR NSSS..................................
3-14 3.9-1 PWR Coolant Injection and Heat Transport Paths During a Large LOCA-ECCS Injection Phase.................................................
3-23 3.9-2 PWR Coolant Injection and Heat Transport Paths During Post-Transient, High-Pressure Feed-and-Bleed Cooling.................. 3-24 3.10-1 Distribution of PWR Containment Types...................................
3-25 3.10-2 Yankee-Rowe Large, Dry Containment (Steel Sphere).................... 3-26 3.10-3 Davis-Besse Large, Dry Containment (Steel Cylinder with Concrete Shield Building)................................................................
3-27 3.10-4 Diablo Canyon Large, Dry Containment (Reinforced Concrete with Steel Liner)......................................................................
3-28 3.10-5 Zion Large, Dry Containment (Post-Tensioned Concrete with Steel Liner)......................................................................
3-29 3.10-6 Millstone 3 Subatmospheric Containment (Reinforced Concrete with Steel Liner)................................................................
3-30 8/90 viii
LIST OF FIGURES (Continued)
FizurePare 4.1-12 Distribution of Containment Types for Westinghouse Reactors..........
4-16 4.2-1 Westinghouse 2-Loop NSSS.................................................
4-19 4.2-2 Section Views of the Ginna Large, Dry Containment......................
4-20 4.2-3 Plan View of the Ginna Large, Dry Containment Below the Elevation of the Operating Floor..............................................
4-22 4.2-4 Plan View of the Ginna Large, Dry Containment, Above the I
Operating Floor.................................................................
4-23 4.3-1 Westinghouse 3-Loop NSSS.................................................
4-27 4.3-2 Section View of the t-LB. Robinson Large, Dry Containment (1-D Post-Tensioned Concrete).............................................. 4-28 4.3-3 Plan View of the H.B. Robinson Large, Dry Containment (1-D Post-Tensioned Concrete)...............................................
4-29 4.3-4 Section Views of the Summer Large, Dry Containment (3-D Post-Tensioned Concrete)...............................................
4-30 4.3-5 Plan View of the Summer Large, Dry Containment (3-D Post-Tensioned Concrete)...............................................
4-32 4.3-6 Section Views of the North Anna Subatmospheric Containment......... 4-33 4.3-7 Plan View of the North Anna Subatmospheric Containment..............
4-35 4.4-1 Westinghouse 4-Loop NSSS.................................................
4-39 4.4-2 General Arrangement of the Yankee-Rowe Core........................... 4-40 4.4-3 General Arrangement of a 193 Fuel Assembly Core....................... 4-41 4.4-4 Section View of the Yankee-Rowe Large, Dry Containment (Steel Sphere)...................................................................
4-42 4.4-5 Plan View of the Yankee-Rowe Large, Dry Containment (Steel Sphere)...................................................................
4-43 4.4-6 Section Views of the Diablo Canyon Large, Dry Containment (Reinforced Concrete).........................................................
4-44 4.4-7 Plan View of the Diablo Canyon Large, Dry Containment (Reinforced Concrete).........................................................
4-45 8/90 X
LIST OF FIGURES (Continued) 4.4-8 Section View of South Texas Large, Dry Containment (3-D Post-Tensioned Concrete)...............................................
4-46 4.4-9 Plan View of South Texas Large, Dry Containment (3-D Post-Tensioned Concrete)...............................................
4-47 4.4-10 Section View of the Millstone 3 Subatmosifheric Containment........... 4-49
LIST OF TABLES Table Pare 2-1 General Plant Data - Sorted by Plant Name...................................
2-2 2-4 General Reactor Site.Data........................................................
2-20 2-5 Summary of General Licensing Data - Sorted by Plant Name.............. 2-25 3.1-1 Summary of PWR Systems.....................................................
3-11 4-1.1 General Characteristics of Westinghouse Steam Generators................ 4-2 4.5-1 Design Parameters for Representative Westinghouse PWRs...............
4-56 4.5-2 Comparison of Westinghouse PWR Vessel and Core Parameters.......... 4-58 4.5-3 Westinghouse PWR System Comparison - RCS, AFW, Charging and HPSI..............................................................
4-60 4.5-4 Comparison of Westinghouse PWR Pressurizer Relief Capacity...........
4-63 4.5-5 Comparison of Westinghouse PWR Containments.......................... 4-66 4.5-6 Comparison of Westinghouse PWR Backup Electric Power Systems.....
4-68 4.5-7 Comparison of Westinghouse PWR Power Conversion Systems.......... 4-70 8/90 xix
Table 2-1. General Plant Data - Sorted by Plant Name Reactor Plant City State Utility Reactor NSSS Architect/ Coro Power Net Electrical MW.
Rating Typo Vendor Engineer MWI Output MW.
MDC or DER ANO-1I Russellville AR Arkansas Power & Light Co.
PWR B&W Bechtel 2568 836 MCC ANO-2 Russellville AR Arkansas Power & Light Co PWR C-E Bechtel 2815 858 MDC Beaver Valley I Shipplngport PA Duquesne Light Co PWR W
Stone &
2652 810 MDC Webster Beaver Valley 2 Shlppingport PA DuquesneLight Co PWR W
Stone &
2652 833 MDC I _
Webster Bellefonte I Scottsboro AL Tennessee Valley Authority PWR B&W TVA 3413 1 213 DER "Belletonte 2 Scottsboro AL Tennessee Valley Authority PWR B&W TVA 3413 1213 DIR Big Rock Point Charlevoix Mi Consrniers Power Co.
BWR GE Bechtel 240 69 MDC Braidwood 1 Braldwood IL Commonwealth Edison Co.
PWR W
Sargent 3411 1120 MOC
& Lundy Braidwood 2 Braidwood IL Commorwealth Edison Co.
PWR W
Sargent 3411 1120 MDC
_.......A Lundy Browns Ferry 1 Decatur AL Tennessee Valley Authority owR GE TA 3293 1065 MOC Browns Ferry 2 Decatur AL Tennessee Valley Authority BWR GE TVA 3293 1065 MDC Browns Ferry 3 Decatur AL Tennessee Valley Authority BWR GE TVA 3293 1065 MDC Brunswick I Southport NC Carolina Power & Light Co.
OWn GE UE & C 2436 790 MDC Brunswick 2 Southport NC Carolina Power A Light Co.
BWR GE UE & C 2436 790 MDC Byron I Byron IL Comrorwealth Edison Co.
PWR W
Sargent 3411 1105 MDC
& Lundy......
Byron 2 Byron IL Commonwealth Edison Co.
PWR W
Sargent 3411 1105 MDC
& Lundy_
Callaway Fulton MO Union Electnc Co.
PWR W
Bechtel 3585 1145 MDC Calvert Chlfs t Lusby MD Baltimore Gas & Electric Co.
PWR C.E Bechtel 2700 825 M[DC Calvert Chilts 2 Lusby MD Baltimore Gas & Electric Co.
PWR C-E Bechtel 2700 825 MDC Catawba I Clover SC Duke Power Co.
PWR -
W Duke Power 3411 1129 MDC I
Co.
Catawtb 2 Clover SC Duke Power Co.
PWR W
Duke Power 3411 1129 MUG CO.
Clinton 1 Clinton IL Illinois Power Co.
_A Lundy Comancho Peak I Glen Rose TX Texas Utilities Electric Co.
PWR W
Gibbs &
3425 1150 D.II Peak__
2_ Glen Rose T X
~
Utilitis Electric Co Gibbs 3425 1150Hill CorminchiePeai 2 GlenRose TX Texus Uhtiltes Electric Co.
PWR W
Gibbs 6 3425 1 150 tIP I-ill_____
tOj
Table 2.1.
General Plant Data - Sorted by+Plant Name (Continued)
Reactor Plant City State Utility Reactor NSSS Archltectl Core Power Net Electrical MWo Rating Type Vendor EngIneer MWI Output UWe MDC or DER Cooper Brownvllle NE Nebraska Public Power District BWR GE Burns &
2381 764 Moc I
Pow_
Crystal River 3 Red Level FL Flonda Power Corp..
PWR B&W Gilbert 2544 821 MIX DC.Cook
.I Bridgman MI IndianafMtchigan Power Co.
PWR W
AEP 3250 1020 MDC OC.Cook2 B.r-Orldgman MI IndianatMichigan Power Co.
PWR W
AEP 3411 1080 MDC Davis-Besse Oak Harbor OH Toledo Edion Co.
PWR B&W Bechtel 2772 860 MDC Diablo Canyon I Avila Beech CA Pacifc Gas A Electnc Co.
PWR W
Pacific Gas & 3338 1073 MDC Electric Dablo Canyon 2 Avila Beach CA Padfic Gas A Electric Co.
PWR W
Pacific Gas & 3411 1087 MDC Electric Dresden 2 Morris IL Commorwealti Edson Co.
BWR GE Sargent 2527 772 MoC S....
& Lundy Dresden 3 Morris IL Commovwealth Edson Co.
BWR GE Sargent 2527 773 MoC II
& Lundy Duane Arnold Palo I A Iowa Electic Light & Power Co.
1658 515 MDC Farey 1
.Dothan AL Alsbama Power Co.
PWR W
Bechtel 2652 813 MDC Farley 2 Dothan AL Alabama Power Co.
PWR W
Bechtel 2652 823 MD.;
Fermi 2 Newport Mill DetoltEdisonCo.
8WR GE Detroit 3292 1093 MD, Edison Fitzpatrick Scribe NY New York Power Authority
2436 778 MIX SI Webster Fort Calhoun I Fort Calhoun NE Ornahe Public Power District PWR C-E Gibbs &
1500 478 MDC Hill Fort SL Vran Plattevllle CO Public Services Company of Colorado
.HRiM GA Sargent 842 330 MDC
& Lundy Geria, -
Ontario NY Rochester Gas & Electric Corp.
PWR W
Glilbert-1520 470 MMC Grand Gulfl Port Gibson MS System Energy Resources. Inc BWR GE
.- Bechtel 3833 1142 MDC Gand Gulf 2 Port Gibson MS System Energy Resources, In.
BWR GE Bechtel 3833 1250 DER HaddamNed Ibddent Nack CT Conneticut Yankee Atomic Power Co.
PWR W
Stone A 1825 569 MDC Webster Hiatch I Baxley GA Georgia Power Co.
- ++
+< +
+
Bechtel Hatch 2 Baxley GA Georgia Power Co.
2436 768 MDX Bechtel Hope Creek I Salem NJ Public Services Electrc & Gas Co.
Indian Point 2 Indian Point NY Como*
riodaEdionCo.
PWR W
UE&C 2758 849 Mt' Ls)
U)
Table 2-1. General Plant Data - Sorted by Plant Name (Continued)
Reacto, Plant Icily S81819 jUstitty Indian Point
-3 LavalleI LaSalle 2, Limerick I Limerick 2 Maine Yie McGuire I McGuire 2 Millstone Millstone 2 Monicllton3 NeMonilelloin Nine Milo Point1 NrtiAnenie io~
NoruthArms 2 M
Dortw" a I1 Oconee S
Oconee2 3 OyswCreek Si Oystere Cseo1 F
PaloVades SI Palo Verde 1 W
Palo~
Ved 2
W Indian Point N
Pottalow Pottstown Cornelius Waterlord_
Waterford Aonticello IN_ INolther States Power Co.
"Commonwealt Edson Co.
Pladelphia Power a Uoht Co.
P--adelphia Power £ Ught Co.
MieYankeeo Atoiuc Power Co.
Duke Power Co.
Duke Oe.
Northeast Utilities Northeast utilities
- cribs tineral tinstal ut Haven Fninterburg intersburg wY Vlra P
ower Co Virginia Power Co.
Duke Power Co.
Duke Power Co.
uk Poe Co.
NJ GUtlNuclear Corp.
MI ConSumners Pwer Co.
AZ Arizona Public Service Co.
A rLzona Public Service Co.
Reactor Teo PWR PWR BWYR 1WR
CT,I BWR PWR PWR PWR PWR PWRR--W*-
USSS Vendor W
W GE GE GE Bechtel 3293 GE JBechtel 13293 -
Archlteg Engineer Pioneer Sargent
& Lund Sargent A L und Core Powe MWI 3323 33293 C-E
- -E GE GE GE_
W W
08W B8W PWR C-E PWR C. -E PWR C-E Stonel A Webster Duke Power Ca Duke Power Cca EMoaw Stone &
Webster Bechtel Niagael Mohawk Stone _&
Webster SMovie -&,
Webster Stone £,
Webster Duke/
Bechtel Duke/
Be8;tel 3800 1221 2630 3411 34-11 27011 2700ý 3117 1675 0 1850ý3 33233 28 9 3 2893 2568 2566 t4j r Net
- Electric, 965 503 1036 810 11290 1129 863 71142 5386 91 5 8 46ý 846::::
4-'
Table 2-1.
General Plant Data - Sorted by Plant Name (Continued) 4' Reactor Plant 1 City State Utility Reactor NSSS Architect/
Core Power Not Electrical MW.
Raling Type Vendor Engineer MWI Output MWe MUC or DER Palo Verde 3 Wintersburg AZ Arizona Public Service Co.
PWR C.E Bechtel 3800 1221.
MDC Peach Bottom 2 Peach Bottom PA Philadelphia Power & Light Co.
Peach Bottom 3 Peach Bottom PA Philadelphia Power A Light Co.
BWR GE Bechtel 3293 1035 MDC Perry 1 North Perry OH The Cleveland Electric Illuminating Co.
BWR GE Gilbert 3579 1205 MDC Perry 2 North Perrya 1
OH Cleveland Electric Iluminating Co.
BWR GE Gilbert 3579 1205 DEIn Pilgrim I Plymouth MA Boston Edson Co.
BWR GE Bechtel 1998 570 MDC Poit Beech I TroCreeks WI eWicornsin Electric Power Co.
PWR W
Bechtel 1518 485 MDC Pont Beech 2 Two Creeks WI, Wiaconsin Electric Power Co.
PWR W
Bechtel 1510 485 MDC Prairie Island I Red Wing MN Northern States Power Co.
PWR W
Pioneer 1650 503 MLD Prairie Island 2 RedWing M
MN NorthernStates PowelCo.
PWR W
Pioneer 1650 503 MDC QuadCites 1 Cord" IL Commonwealth Edisonflowa-Ilkinols Gas a Electric BWR
& Lundy.
OuddCities2 Cordova IL Commonwealth Edlsonilowa-llhnols Gas A Electric BWR GE Sargent 4 2511 769 MIX,,
RanchoSew Clay Station
- CA Sacramento Municipal Utility District PlWR
&W
,Bechtel 2772 873 MDC River Bend 1 St. Franciaville LA Gulf States Utlittes Co.
2894 936 MDC
-I I.Webster Robinson 2 Hartsville SC Carolina Power a Light Co.
PWR W
E*-"m.
2300 665 MDO SalemI Salem N
, Public Services Electric &GasCo.
PWR W
Pacific Gas& 3411 1106 MDC S,,
Electric Salem 2 Salem NJ Publc Services Electric
& Gas Co.
PWR W
Pacific Gas & 3411 1106 MDC SElectric SenOnofre I SanClemenle CA
,Southern Calilornis EdisovnSan Diego Gas A Electric PWR W
Bechtel.,
- 1347, 436 MDC SanOnofre2 SanClemente CA Southem Calfontoiad EdoSan Diego Gas A& Elecic PWR C-E Bechtel 3390 1070 LAjC SanOnofre3 SanClemerfe CA Southern Californma Edison/San Diego Gas A Electric PWR C-E Bechtel 3390 1080 MDC SeabrookI Seabrok' NH Newl-HampshweYareUnt PWR W
LUE&C, 3411 1150 MDC Sequoyhl I Soddy-Daisy T N Tennessee Valley Authority PWR W
TVA 3411 1148 MDC Sequoyah 2 Soddy-Daisy TN Tennessee Valley Authority PWR W
TVA 3411 1148 Sheaon Harris I New Hill NC Carolina Power & Light Co PWR W
Ebasco 2775 860 1MDC" in' S..
Table 2-1.
General Plant Data - Sorted by.Plant Name (Continued)
Reactor Plant South Texas I City Brookhaven
)alhacios South Texas 2 P a-_c lo-TX SL Luci I St. Lucle 2 Hlulchin*on Island Hutchinson Istand Parr Stale NY TX-FL FL Surry I
"..id i..L Nk 1
A I
utility Long Island Lighting Co.
Houston LUghting & Power CO.
,Houslon Lighting & Power Co.
Florida Power & Ught Co.
Florida Power & Light Co.
-_o, Caroýia Electric & Gas Co.
virginia Power Co.
Reactor Type 8Wfl PKR PWfl PWR PWR PWR INSSS Vendor W
W C-E C-E W
Surry 2 TGravel Nock
,7,,,,, p.,...
r M
andI B-orwlck Pennsytvania Power & Light Co.
I'WH kmmmu~f
- i.
i I
I W
UI 1
P-ensnlvania Power A Light Co.
hrte Mile Island I I Lonidonderr, Twu IlA iiG P Nucea C
- p..
Trojan Prescolt OR Porland General Electric Co.
PWR Turkey PoInt 3 Floda City FL Floria Power & Ught Co.
PWR Turkey Point 4 Florida City FL Florida Power A Ught Co.
PWR Veniont Ywakee Vernon VT Vermont Yankee Nuclear Power Corp.
BWR Vogue 1 Waynesboro GA Georgia Power Co.
PWR Vogue 2 Waynesboro GA Georgia Power Co.
PWR Walerford 3 Taft LA Louisiana Power & Light Co.
PWR lpunng City WNP-1 Richland WA
- /Ikl).
) *IL,__
Richland IN iN WNP-3 Sat-op
[W2i Burlington "KS
'Tennessee Valley Authority ITerinesse Valley Authority Washingt*n Public Power Supply System Washington Public Power Supply System Washington Public Power Supply System Woll Creek Nuclear Operati*g Corp.
I B&W W
W W
GE W
W C-E W
W B&W GE C-E Archltectl Core Power Engineer MWI Stone &
2436 Websler Brown A 3800 Root.
Brown &
3800 Root.............
Ebksm
- 2700, EVQs 2700 Gilbert 2775 Stone &
244 1 Webster Stone&
2441 Webster Bechtel 3293 Bechtel
- 3293, Gilbert 2535 Bechtel 3411.
1
- Bechtel, 2200 6
Bechtel 2200 6
Ebawo.
1593 5
Bechtel 3411 I
Bechtel 3411 1
Ebl
ý s
3390 T
VA 3411 1
WA 3411 1
JEAC 3760 Furns &
3323 1T lee basw 3005 12 echtel 3411 11 T
E B
Net Electric Output MWe 820 1250 1250 839
- 839, 585 781 1032 1032 76 T095 66 l68 04 079 079 075 165 165 g95 242 28 Ullllly Long Idend Lighbng Co.
Ho*,
Ughbng & p*,,-
Co.
0
Weds Bar I VVlHI
- II L
M WNi1P-2 Wo roo ff k
- ng Iiy al MW.
Rating MDC or DER DER MIXO MDC MOc.
MOO MDC MOO MOC MDC MOC MOXO MOO.
MOC.
MOO MOO MIOO SER DER IDER mcn DER ummnsni Wells Bat 2 u B R
E B
SvIl~
SC 5 *..m PWR ginia P C
ower o.
Gi E--
Susquisanna I Q
R cklIWIR BWR T
S......... *'"# "*e" PWR
Table 2-1. General Plant Data - Sorted by Plant Name (Continued)
Reactor Plant City State Utility Reactor NSSS Architect/ Core Pow Type Vendor Engineer mWt YarilRoen Row M A Yankee Atomic Electrc Co.
PWR W
Stone A 600 Webster Zion I Zion IL Co-rnr,-eath Edison Co.
PWR W
Sargent 3250
& Lundy Zion 2 Zion IL Cornmnrmea*ih Edison Co.
PWR W
Sargent 3250
& Lundy otba:
MDC - Maximum Dependable Capacity DER - Design Electric Rating Is
Table 2.4. General Reactor Site Data Plant Name Location Water Source Ult. Heat Sink SSE Tornado Wind Hortz. G's Vert. G's Speed (MPH)
ANO-1 Russeiville, Arkansas Dardanelle Resevoir Same 0.2 0.133 360 ANO-2 Russelville, Arkansas Dardanelle Resovoir Nat. Cooling tower 0.2 0.133 360 Beaver Valley I & 2 25 Mi. NW Pittsburgh, Pa Ohio River Nat. Cooling Towers 0.12 0.08 360 Bellefonte I & 2 7 Mi. ENE Scotsbero, Ala Guntersville Resevoir Nat. Cooling Towers 0.18 0.12 360 Big Rock Point 4 Mi. NE Charlevoix, Mi Lake Michigan Same 0.05 0.05 210 Braldwood I & 2 2 Mi. S Braidwood. Ill.
Kanakee River Braidwood Lake 0.2 0.133 360 Browns Ferry 1, 2, & 3 10 Mi. NW Decatur, Ala.
Tennessee River RiverlMoch. Cooling Towers 0.2 0.133 300 Brunswick I & 2 19 Mi. S Wdmington, NC Atantic Ocean Cape Fear River 0.16 0.107 360 Byron 1 & 2 4 Mi. S Bryon, III Rock River Nat. Cooling Towers 0.2 0.133 360 Callaway 10 Mi. SE Fulton, MO Missouri River Nat. Cooling Tower 0.2 0.133 360 Calvert Cliffs I & 2 40 Mi. S Annapolis, MD Chesapeke Bay Same 0.15 0.1 360 Catawba I & 2 19 Mi. SW Charlotte Lake Wylie Mech. Cooling Towers 0.15 0.1 360 Clinton I Harp Township, IlI.
Salt Creek (N. Fork)
Lake Clinton (Manmade) 0.25 0.25 360 Comanche Peak I & 2 40 Mi. SW Ft. Worth, TX Squaw Creek Resevoir Same 0.12 0.08 360 Cooper 23 Mi. S Nebraska City, NB Missouri River Same 0.2 0.133 360 Crystal River 3 7 Mi. NW Crystal River, Fla Gulf Of Mexico Same 0.1 0.067 360 tJo 0
Table 2.4.
General Reactor Site Data (Continued)
Plant Name Location Water Source Ult. Heat Sink SsE Torndo Wind Horiz. G's Vert.
G's Speed (MPH)
D.C. Cook 1 & 2.
10 Mi. S St.Joseph. Mi Lake Michigan Same 0.2 0.133 360 Davis Besse 21 MI. E Toledo, OH Lake Ede Nat. Cooling Tower 0.15 0.1 360 Diablo Canyon,1 & 2 12 Mi. W San Luis Obispo, CA Pacific Ocean Same 0.75 0.5 200 Dresden 2 & 3 9 Mi. E Morris, Il.
Kanakee River Cooling Lake.
0.2 0.133 200 Duane Arnold 8 Mi. NW Cedar Rapids. 10 Cedar River "-
Mech. Cooling Towers 0.12 0.096 360 Farley V& 2 16 Mi. E Dothan. Ala.
Woodruff Resevolr Mech. Cooling Towers 0.1 0.067.
360 Fermi 2...
30 Mi. SW Detrolt, MI Lake Ede Nat.Cooling Towers 0.15 0.1 360 FItzpatrick.
6 Ml. NE Oswego. NY.
Lake Ontar.o Sae 0.15 0.1 360 Fort Calhoun 1 10 PA. N Omaha; NB Missouri River Same 0.17 0.113 360 Ginna 15 Mi. NE Rodeser, NY Lake Ontario Same 0.2 0.133 132 Grand Gulf I & 2 25 MI. S Vicksburg Mississippi River NaL Cooling Towers 0.15 0.1 360 Haddam Nock 13 Mi. E Mdde1, CT Connecticut River Same 0.15 0.1 360 Hatch I & 2 11 Mi. N Baxley, GA Altahama River, Mach. Cooling Towers 0.15 0.1, 360 Hope Creek I & 2 8 Mi. SW Salem, NJ Delaware River Nat. Cooling TowerS 02 0.2 360 Indian Point 2 & 3 25 Mi. N New York City, NY Hudson River Same 0.15 0.1 300 Kewaunee 20 MP. N Manitowoc. WI Lake Michigan Same 0.12 0.08 360 tj
Table 2-4.
General Reactor Site Data (Continued)
Plant Name Location Water Source Ult. Heat Sink SSE Tornado Wind Horiz. G's Veri.
G'a Speed (MPH)
LaSale, I & 2 12 Mi.. W Morris. IIl.
Illinois River 2058 Acre Cooling Lake 0.2 0.133 360 Limenck' 1' & 2 30 Mi. NW Philadelphia, Pa Schuylill River Nat. Cooling Towers 0.15 0.1 360 Maine Yankee Wicasset, Maine Black River, Montsweag Bay 0.1 0.067 360 McGuire 1 a 2 17 Mi. NW Charlotte, NC Lake Norman Same 0.15 0.1 360 Millstone 1. 2, & 3 5 Mi.SW New London, CT Long Island Sound Niantic Bay 0.17 0.113 300 Monticello' 30 Mi. NW M!Rneoalxoh MN Mississippi River Mech. Cooling Towers 0.12 0.08 360 Nine Mile Point 1 8 Mi. NE Oswego. NY Lake Ontario Same 0.11 0.055 360 Nine Mde Point 2 8 Mi. NE Oswego. NY Lake Ontario Same 0.15 0.1 360 North Anna 1 & 2 40 Mi. KIW Richmond, VA North Anna Resevoir Cooling Pond 0.12 0.08 360 Oconee 1, 2, & 3 30 Mi. W Greenville, SC Lake Keowee Same.
0.1 0.067 360 Oyster Creek t 9 Mi. S Toms River, NJ Atlantic Ocean Barnegat Bay 0.17 0.113 360 Palisades 35 Mi. W Kalamazoo, MI Lake Michigan Mach. Cooling Towers 0.2 0.113 360 Palo Verde 1, 2. & 3 2 Mi. S Wintersberg, AZ Domestic Water Mach. Cooling Towers
,0.27 0.18 300 Peach Bottom 2 & 3 19 Mi. S Lancaster. PA Susquehana River River/Moch. Cooling Towers 0.12 0.08 360 Perry 1 & 2 37 Mi. E Cleveland, OH Lake Ene Nat. Cooling Towers 0.15 0.1 360 Pilgrim 1 35 Mi. SE Boston, MA Cape Cod Bay Same 0.15 0.1 300 t%)
t~3
Table 2.4. General Reactor Site Data (Continued)
Plant Name Location Water Source Ult. Heat Sink SSE Tornado Wind Horz. G's Vert. G'. Speed gMPH)
Point Beach I & 2 15 Mi. N Manitowoc. WI Lake Michigan Same 0.12 0 08 360 Praine Island 1,& 2 40 Mi. SE Minneapolis. MN Mississippi River Mech. Cooling Towers 0.12 0.08 360 Quad Cities lA 2 20 Pi. NE Moline, Ill.
Mississippi River Same 0.24,
0.16 200 Rancho Seco 25 Mi. SE Sacramento, CA Folsom South Canal Nat. Cooling Towers 0.25 0.167 175 River Bend 1 25 MI. N Batton Rouge, LA Mississippi River Mech. Cooling Towers 0.1 0.1 360 Robinson 2 6 Ml. NW Hartsville. SC Lake Robinson Water Discharge Tunnel 0.2 0.133 300 Salem 1 & 2 8 Mi. SW Salem, NJ Delaware River Same 0.2 0.133 360_
San Onof1.
5 MA. S San Clemente, CA Pacific Ocean Same 0.67, 0.144,
- 75.
San Unolre 21 3
5 MP. S San Clemente. CA Pacific Ocean Same 0.67 0.44 5260 Seabrook I Seabrook. NH,.
Atdantic Ocean Same 0.2, 0.133 360 Sequoyah I & 2 18 Mi. NE Chattanooga, TN Tennessee River River/Nat. Cooling Tower 0.18 0.12 360 Shearon Haris 1 20 MP. SW Raleigh. NC Cape Fear River Nat. Cooling Tower 0.15 0.1 360 Shorehar 12 Ml. NW Riverhead, NY Long IslandSound Same -
0.2 0.133.
360 South Texas I & 2 12 Mi. SW Bay City. TX Colorado River,,
7000 Acre Cooling Pond
.0.1 0.067 360 St Lucie I A 2 12 PA. SE FLPlerce, FL NIA N/A-0.1 0.067 360 Summer 26 Mi. NW Columbia, SC Lake Monticello Same 0.15 0.1 360 Surry 1 & 2 8 Mi. S Williamsburg. VA James River Same 0.15 0.1 360 tjj 0
Table 2-4. General Reactor Site Data (Continued)
Plant Name Location Water Source Ult. Heat Sink SSE Tornado Wind Hortz. G's Vert. G's Speed (MPH)
Susquehanna& 2 7M. NE Bawick, PA Susquehanna Rver Nat. Cooling Towers 0.1 0.067 360 TMI-I 10 Mi. SE Harrisburg, PA Susquehanna River Nat. & Mech. Cooling Towers 0.12 0.08 360 Trojan 30 Mi. NW Portland. OR Columbia River Nat. Cooling Tower
'0.25 0.167 200 Turkey Point 3 & 4 25 Mi. S Miami, FL Biscayne Bay Canals 0.15 0.1 225 Vermont Yanke" 5 Mi. S Battleboro, VT Connecticut River RiverlMech. Cooling Towers 0.14 0.093 360 Vogde
& 2 39 Mi. SE Augusta, GA Savannah River Nat. Cooling Towers 0.2 0.133 360 Waterford 3 Tall, LA Mississippi River Same 0.1 0.067 360 Watts Bar I & 2 8 Mi. E Sprng City. TN Chickamunga Lake Nat. Cooling Towers 0.18 0.12 360 WNP-1 Hanford. WA Columbia River Mach. Cooling Towers 0.25 0.167 360 WNP.2 Hanford. WA Columbia River Mech. Cooling Towers 0.25 0.167 360 WNP-3 Satsop, WA N/A Nat. Cooling Tower 0.25
-0.167 360 Wolf Creek 4 Mi. NE Burlington, KA Wolf Creek Cooling Lake 6000 Acre Cooling Lake 0.12 0.08 360 Yankee Rowe 20 Mi. NW Greengield, MA Sherman Pond Same 0.1 0.067 110 Zion I & 2 6 Mi. N Waukegan, I1l.
Lake Michigan Same 0,17 0.113 360 t,,3 I b,)
4*
Table 2-5. Summary or General Licensing Data - Sorted by Plant Name NRC Dockett M Reactor Plant Reactor NSSS NRC Construction Operating Lic. Power Expiration Comm' Date of Notes (so.
0
.+
,Type Region Permit Licence MWt Date-Opt ?
Comm'l Ops 313 ANO-1 PWR,1 B&W IV 12/6/68 5/21/74, 2568 12/6/08 Yes 12/19/74 368 ANO-2
- PWR, C-E IV 12/6172 12/14/78 2815 12/6/12, Yes 4/26/80 334.,
Beaver Valley I
- PWR, W
I 6/26/70 7/2/76 2652 1/29/16 Yes 10/1/76 412.,
Beaver Valley 2 PWR.
W I
5/3/74 8/114/87
-2652 5/27/27 Yes 11/17/87
- 438, Bellefonte 1 PWR O&W II 12-24-74 N/A.
0 N/A Indefinite N/A 439.
Bellefonte 2 PWR B&W II 12/24/74 N/A 0
N/A Indefinite N/A I
III 5/31/60 511/64 240 5/31/00 Yes 3129/63
- 456, Braidwood I PWR W
1II 12/31/75 7/2/87 3411 10/17/26 Yes 7/29/88 457,.
Braldwood2 PWR W
111 12131/75 5/20/88 3411-,
12/18/27 Yes 10/17/88, 259 Browns Ferry I BWR-GE,,
II 5/10/67 12/20/73 3293 5/10/07, yes 8/1/74, 260 Browns Ferry 2 BWR.
GEE 11 5/10/67 8/2/74 3293 5/10/07 YS 3/1/75.
296 Browns Ferry 3 BWR GE II 7/31168 8118/76 3293.
7/21/08 Yes 311177 3252 Brunswick,-'
BWR GE 2/7170 11/12176 2436 2/7110 Yes 3/18/77 324 Brunswick 2 BWR GE II 2/7170, 12/27174 2436 12/6/10 Yes 1113/75 454 Byron 1 PWR W
III 12131/75 2/14/85 3411.
10/31/24
- Yes, 9/16/85
- 455, Byron 2,,
PWR 11W 1
i' 12/31/75-1/30/87 3411 11/6126 Yes 8/21/87 483.
Calaway I PWR VW III 4116176
,10118184 3565 10/18124
- Yes, 12119/84 1ý 317 Calvert Cliffs I PWR C-E I
7/7/69 7/31/74 2700 7/31/14 Yes 5/8/75 318 Calvert Cliffs 2 PWR C-E I
7/7/69 11/30176 2700 8/31/16 Yes 411/77 t,3 t.1 0
STable 2.5. Summary or General Licensing Data - Sorted by Plant Name (Continued)
NRC Dockett 0 Reactor Plant Reactor NSSS NRC Construction Operating Lic. Power Expiration Comm'l Date of Notes (5o-I Type Region Permit Licence MWI Date Opa ?
Commi Ops 413 Catawba I PWR W
II 8/7/75 1/17/85 3411 12/6/24
,Yes 6/29/85 414 Catawba 2 PWR W
II 8/7/75 5/15/86 3411 2/24/26 Yes 8/19/86 461 Clinton I BWR GE III 2/24/76 4/1 7187 2894 9/29/26 Yes 11/24/87 445 Comanche Peak 1 PWR W
IV 12/19/74 N/A 0
N/A Expected '95 N/A 5
446 ComandcePeak2 PWR W
IV, 12119/74 N/A 0
N/A Indefinite N/A 3
298 Cooper BWR GE IV 6/4/68, 1/18/74 2381" 6/4/08 Yes 7/1/74 302 Crystal River 3 PWR B&W III 9125/68 1/28/77 2544 12/3/16 Yes 3113/77 315 D.C. Cook I PWR W
III 3/25169 10/25/74 3250 3125109 Yes 8/28/75 316 D.C.Cook 2 PWR W
III 3/25/69 12/23/77 3411 N/A Yes 7/1/78 346 Davis-Besse PWR B&W III 3/24/71 4/22/77 2772 3/24111 Yes 7/31/78 275 Diablo Canyon I PWR W;
V 4123/68 11/2/84 3338 4/23/08
'Yes 5/7/85 323 DiabloCanyon2 PWR W,
V' 12/9170 8/26/85
, 6/9/70 4
249 Dresdon 3 BWR GE III 10114/66 3/2/71 2527 10/14/06 Yes 11/16/71 331 DuaneArnold BWR GE III 6/22/70 2122/74 1658 6/21110 Yes 2/1/75 348 Fauley I PWR W
II 8/16/72 6125177 2652 8/16112 Yes 12/1/77 364 Farley 2 PWR W
II 8/16/72 3/31/81 2652 8116/12 Yes 7/30/81 341 Fermi 2 BWR GE III 9/26/72 7/15/85 3292 3/20/25 Yes 1/23/88 333 Fitzpatrick BWR GE I
5/20/70 10/17/74 2436 5/20/10 Yes 7/28/75 t0
"JTable2-5.. Summary of General; Licensing:Data Sorted by Plant Name (Continued)
NRC Dockettll Reactor Plant Reactor NSSS NRC Construction Operating Lic. Power Expiration Comm.
Date of Notes (50-
)
Type R0g0on Permit Licence MWt Date Ope ?
Comm'I"Ops 285 Fort Calhoun I PWR C-E IV
-617168 8/9173
, 1500' 6/7/08 8
Yes" 6/20/74 267 FortSt. Vralin HTt GA TIV 9/17168 12121/73 842 91171088 Yes 7/1/79 244 Ginnra
-PWR W
I 4125/66 12/10/84 1520.
4125/06 Yes 7/1/70 416 GrandGulfl BWR GE II 9/4174 11/1/84'
N/A' Indefinite N/A 1I "213 HaddaNeNock.-
PWR W
I 5/26/64 12/27174 1825 5/26/04 Yes 1/1/68 321
- Hatch1Il -
BWR GE II 9/30169, 10113/74 2436 9130/09 Yes 12/31/75 366 Hatch 2 BWR GE II 12127/72 6113/78 2436 12/27/12 Yes 9/5/79 354 Hope Crook I BWR GE I
1114/74,
7/25186, 3293 4/11/26 Yes 12/20/86 247-Indian Point 2 PWR W
I 10/14166,,
9128173 2758 9/28/13 Yes 8/1174 286 Indian Point 3
.PWR W
I 8113/69 4/5/76 3025 8113/09.
Yes 8/30/78 305-Kewaunee -
PWR W
- II1, 8/6168 12/21/73 1650.,
8/6/08 Yes 6/16/74 373 LaSalel BWR GE IIi 9110/73 8/13/82
- 3323 5/17/22 Yes 1/1184 374 LaSafle2 -
9/10173 3/23184 3323 12116123 Yes 10/19/84 352 Limerick I BWR GE.
1 6119/74 818/85 3293 10126/24 Yes 2/1/86 353 Limerick 2 -.-
ý6/1'9174 Yes 1/9/90 309 MaineYankee PWR C-E I
10/21/68, 6/29173 2630' 10/21108
.Yes' 12/28172 369 McGuire 1 PWR W
II 21231/73 7/8/81 3411 6112121 Yes 12111/81 370 McGuire 2 PWR W
II 2/23/73
,5/27/83.
3411,,
3/3123 Yes 3/1/*A to)
Table 2-5. Summary, of General Licensing Data - Sorted by Plant Name (Continued)
NRC Docketi # Reactor Plant Reactor NSSS NRC Construction Operating Lc. Power Expiration Comm'l Date of -
Notes (50.-
)
Type Region Permit Licence MWt Date Ops ?
Comm'l Ops 245 MillstonelI BWR GE I1 5119166 10131/86 2011 5/19/06 Yes 3/1/71 326 Millstone 2 -
PWR C-E I
.12/11170 9/30/75 2700 7/31/15 Yes 12/26/75 423 Millstone 3 PWR W
I 819174 1/31/86 3411 11125/25 Yes 4/23/86 263 Monticello BWR GE-III.
6119167-_.
1/9/81, 1670 9/8/10 Yes 6/30/71 220 Nine Mile Point I -
4/A 2165 12/26/74 1850 4/11/05 Yes 12/1/69 410 Nine Mile Point 2 BWR GE I
6/24/74 7/2/87 3323 10/31/26 Yes 4/5/88 338 North Anna I PWR W
II 2/19/71 4/1/78 2893 4/1/18 Yes 6/6/76 339 North Anna 2 PWR W
II 2/19/71 8/21/80 2893 8121/20 Yes 12/14/80 269 Oconeel PWRI B&W
.11 11/6167 216/73 2568
.2/6/13 Yes 7/15/73
.270 Ooonee2 PWR B&W II 1116/67 10/6/73 2568 10/6/13 Yes 9/9/74 287 O=We3 PWR-B&W II 11/6/67
/719/74 2568 7119/14 Yes 12/16/74 219 Oyster Creek I BWR
.GE I
12/15164 8/1/69 1930 4/9/72 Yes 12/11/69 4
'255 Palisades PWR C-E.
Ill.,
.3/14/67 10/16/72 2530 3/1174 Yes 12131/71 4
528 Palo Verde I PWRJ C-E
.V 5/25/76 6/1/85
'3800 12131/24 Yes 1/28/86 529 Palo Verde 2 PWR C-E V
5/25/76 4/24/86 3800 12/9125 Yes 9119/86.
530 Palo Vrde 3 PWR C-E V
5/25/76 11125187 3800 3/25/27 Yes 1/8/88 277 Peach Bottom 2 BWR GE 1
1/31/68 7/2174 3293 1/31108 Yes 12123/74 278 Poach Bottom 3 BWR GE I
1/31168 7/2/74 3293 1/31108 Yes 12/23/74 440 Perry I BWR GE Ill
,5/3/77, 11/13/86 3579 3/18126 Yes 11118W87 t0J t'..)
00
)
Table 2-5.' Summary of General Licensing Data - Sorted byY Plan t Name (Continued), : ?9 K 4,
NRC Dockett # Reactor Plant Reactor NSSS NRC Construction Operating Lic. Power Expiration Comm'I
- Date of -
Notes (SO-Type Reilon Permit Licence MWt Dale Ops ?
Comm'I Opa 441 Perry, 2 O-WR GE III 5/13/77T N/A 0-%.,
N/A Indefinite N/A I
8126/68 19/15172 1998-,
8/26108 Yes
--12/1/72 266 PointBeach1, PWR W
III 7/19/67 10/5/70, 1518 10/5/10 Yes 12/21/70 301 PolntBeach2 PWR W
III 7125/68 318/73, 1518.
3/8113 Yes 110/1/72 282 Prairie Island I PWR W
III 6125168 4/5/74, 1650 819/13 Yes 12/16/73 306 Prairie Island 2 PWR W
III 6125/68 10/29/74 1650W 10d/29/114 Yes 12/21/74 25 254 Quad cities I BWR GE III 2/15167 12114/72 2511 2/15/07.
Ye 2/18/73 265 QuadCIdes2 BWR GE III 2/15/67 12'14/72 2511 2/15/07, Yes
-3/10/73 312 Rando Seoo PWR B&W V
1011/68
/816/74, 2772.
10/11108 Yes 4/17/75 458 River Bend 1 BWR GE IV 3/25177 11/20/85 2894 8/29125
-Yes 6/16/86 261 Robinson 2 PWR W
II 4/13/87 9/23170
.. 2300.
4/13/07 Yes 3/7/71 272 Salem I PWR W
I 9/125/68.
12/1/76, 3411 9125/08 Yes 6/30/77 311 Salem2 PWR W
I 9125/68_
5/20/81 3411 9/25108 Yes -
10113181 206 SanOnofrel PWR W
V 3/2164 3127/67, 1347..
9/27/2072
-Yes
-1/1/68
-4 361 San Onofre2 PWR C-E V
10/18/73_
97182 3390..
10/18/13 Yes 818/83 362 San Onofre 3 PWR C-E V
10118/73 9116/183 3390.
10118/13 Yes--
4/1/84 443" Seabrookl PWR W
I
--7/7176.
10/17186
. 3411 10/17/26 IMb N/A -
5 327 Sequoyah I1 PWR W
II 5/27170 9117180 3411 5/27110 Yes 7/1181 328 Seqoyah2
- PWR, W
II 5/27/70, 9/15/81 3411 5/27/10 Yes 6/1/82
'-0
Table 2.5. Summary of General Licensing Data - Sorted by Plant Name (Continued)
NR ocke,, 0 Reactor Phant Reactlor NSSS NRC Construction Operating LIc. Pow-or Expiration Comm'l Date of
- Nolo, Aso0-A-}
Type Region Permit Licence MWI
- Dale, Ops ?
Comm'l Ops 400 Shearon Harris I PWR W
I 1/27/78 1/12/87 2775 16/24/26 Yes 5/2/87 322 9hOreham O WI.
GE 1
4/14173_..
7/3/85 121.
4/13/13 M..N N/A 5
4 SothTV T I
o212/22/75.
312288 3800 8/20/27 Yes 1,8/2518.
I IV 1 12/22/*
I1 2/16/8 335
!SL c.ia rIW II 7/!/70 3/1/76 389 IS SI,
- 2.
-UiI!
I"WR PWH C-E W
II II 512/77 3/21173 IL 611(
li/1
- 2700 2700 2775 280 Surry. I PWR W
II 81/25/68
-5/25/72 2441 5/25/12 Y
281 Surry 2 PWR W
II 6/25/68 1/29/73 2441
.1/29113
-Ye 387 SiqA-.
11/2/73 11/12/82 3293 711 722 Y
388 1
_S_.,_- _
2
-W n
d 1211612 1 216/2 8 3/1/16
.4/61/23 I 1I2I/73 6127/84' 3293... 1-3123124 Ye6 289 tThrOO Md, Island I PW I& 1]
as-?:*-4I...
/1 I/o6 ill U114 2AA tThre,u12n I
'G----!
-WI---W rvWM V
2/8/71-11/21/75 250 ITrnlan r,,,,*
PRwe W
II.
4/27167 7/19172, ia lF..4......
4----4-------J-------L I _________
TUIrUky Pint 4 rwM W
II 4/27167, 4/10/73 341 V1 2200
/18108 2/8/11
/27107
/27/07 bs s
s Yes
12/11/87 2/28/73 1593 12/11/07 Yes 424 VogUe I PWR W
II 6/28/74 3/16/87 3411 1/16127 yes 425 Vogle 2 PWR W
II 6/28/74 N/A 0
N/A Expected1 382 Waterflord 3 PWR C-E IV 11/14/74 3116185 3390 12/18/24 Yes7 12/21/76 8/8/83 1/1/84 1 2/22/72 5/1/73 6/8/83 2/1 2/85 5/20/76 "12114/72 9/7/73 11/30/72 611/87 N/A 9/24/85 39 ~
0, 0
ureyont 3 2 1 5
W V
190,,-
wv*
C-IE.
YE
-Yo
- VIP 2/82 m*
1 U IIIII1 1
and ff VWIW VWllff l
2535 -*
5 4
4, 95ý 1
9117 K
eb *' V
Table 2-5. Summary of General Licensing Data - Sorted by Plant Name (Continued)
NRC Dockett
- Reactor Plant Reactor NSSS NRC Construction Operating LIc.
Power Expiration Comm'l Date of Notes (50._._)
Type Region Permit Licence MWt Date Ops 7 Comm'l Ops 390 Watts Bar 1 PWR W
II 1/23/73 N/A 0
N/A Expected '95 N/A 2
391 Watts Bar 2 PWR W
II 1/23/73 N/A 0
N/A Expected '95 NIA 2
460 WNP-t PWR 8&W V
12/24/75 N/A 0
N/A Indelinito N/A 1
3/19/73 4/13/84 3323 12/20/23 Yes 12/13/84 508 WNP-3 PWR C-E V
4/111/78 N/A 0
N/A Indefinite N/A I
482 Wolf Creek PWR W
IV 5/31/77 6/4/85 3411 3/1 1/25 Yes 9/3/85 29 Yankee Rno PWR W
I 11/4/57 12124/63 600 7/9/00 Yes 7/1/61 295 Zion 1 PWR W
III 12/26/68 10119173 3250 12126108 Yes 12131/73 304 Zion 2 PWR W
III 12/26/68 111/14173 3250 1 2/26/08 Yes 9/1 7194 NOTES:
- 1. Construction hafted
- 2. Under active construction
- 3. Construction deferred
- 4. License not expired under 10 CFR 2.109
- 5. Low power icense
- 6. May be doornrrmlioned t,3
- 3.
- PRESSURIZED WATER REACTOR (PWR) SYSTEM OVERVIEW Reactor plant systems may be broadly classified as safety-related or as non safety-related. Light water reactor (LWR) safety-related systems typically are considered to be those-that are required to perform any bf the following safety functions:
Control reactivity I
Provide reactor core cooling and heat removal from the primary system Maintain reactor coolant system integrity Maintain containment integrity Control radioactive releases In order to ensure the performance of these "front-line" safety functions, additional safety related systems are required to perform the following support functions:
Provide adequate motive power (i.e. electric, pneumatic or hydraulic motive power, direct steam turbine or diesel engine drive)
Provide adequate control and instrumentation power (i.e., AC or DC electrical
-control power)
Provide adequate cooling of safety-related equipment (i.e.,:'cooling water, room air cooling).
Provide other support functions needed by front-line or support systems to establish and maintain a safe shutdown condition In theirpresent form, the Nuclear Power Plant System Sourcebook series focuses on front line safety systems and on electric power and cooling water support systems.
-In this section, an overview of PWR systems is provided, focusing on basic system functions and interfaces. In Sections 4 to 6, more detailed comparative information is presented on the different product lines of the three U.S. commercial PWR Nuclear System Supply System (NSSS) vendors: Westinghouse, Combustion Engineering, and Babcock'& Wilcox. Comparative data summaries for PWR systems are found in Sections 4.5, 5.5, and 6.5 (for'individual PWR vendors) and in Section 7 (compilation for all PWRs). The reader should refer to the available Nuclear Power Plant System Sourcebooks identified in Section 1 for summary information on safety systems at specific nuclear power plants.
3.1 Introduction to the Pressurized Water Reactor SThe PW R reactor coolant system (RC S) transports heat generated in a low enrichment, light-water cooled and moderated core to the secondary coolant system via
,external primary coolant loops with steam generators. Control and removal of heat from
-the reactor and conversion of this heat into usable electrical power requires a broad
,, spectrum of operating and auxiliary systems. Additionally, safety systems are required to ensure that postulated accidents at the PWR do not cause undue risk to the health and safety of the public.- The'spectrum of "generic" PWR systems is'listedin Table 3.1-1. As indicated in this table, some systems normally are supplied by the Nuclear System Supply System ;(NSSS) vendor. The remaining systems, or the Balance-of-Plant (BOP), are supplied by the architect-engineer (A-E) who is responsible for the detailed integrated design of the plant.
3.2 PWR Primary System The !PWR NSSS is the primary system, or reactor coolant system (RCS),
which consists of the reactor vessel and two to four external primary coolant loops, each 3/90 3-1 I
one Steam generaitor. and one 6r-two primary coolant pumps. The three U.S.
PWR vendors have produced seven basic plant configurations, as summarized below:
RCS Configuration Westinghouse Combustion Engineering 2-loop 3-loop 4-loop 2-loop 3-loop Number of Plants 6
14 35 14 1
Babcock & Wilcox "lowered" 2-1oop "raised" 2-loop Total PWRs The three models of the Westinghouse NSSS are shown in Figure 3.2-1 (2-loop), Figure 3.2-2 (3-loop), and Figure 3.2-3 (4-loop).r The Combustion Engineering 2-loop NSSS is shown in Figure 3.2-4. The two basic models of the Babcock & Wilcox NSSS are shown in Figure 3.2-5 ("lowered" 2-loop) and Figure 3.2-6 ("raised" 2-loop).
A pressurizer is connected to the "hot leg" of one of the primary coolant loops and serves to control primary system pressure by means of electric heaters (to increase the steam volume in the pressurizer-and raise pressure) and spray (to condense the steam bubble in the pressurizer and lower pressure). RCS coolant inventory is measured by pressurizer water level, which is controlled by the combined letdown to and makeup from the Chemical and Volume Control System (CVCS).
3.3 Reactor Core and Fuel Assemblies The PWR generates heat in a low-enrichment, light-water cooled and moderated core. All PWR fuel assemblies consist of a square array of fuel and burnable poison rods.
The general fuel assembly configurations used by the three PWR vendors are listed below:
Vendor Fuel Assembly Configuration Westinghouse 9 x (6 x 6) 14 x 14 15 x 15 17 x 17 Combustion Engineering 15 x 15 14 x 14 16 x 16 Babcock & Wilcox 15 x 15 17 x 17 plants ApplkaicaL Yankee-Rowe only 2-1oop plants and San Onofre I Some 3-loop and 4-loop plants Most 3-loop and 4 - loop plants, replacing the 15 x 15 fuel elements Palisades only Earlier plants Later plants, replacing 14 x 14 fuel elements in earlier plants All plants except Bellefonte Bellefonte only. Can replace 15 x 15 fuel elements in earlier 3-2 3/90 7
3 I
80
The general trend is toward the denser arrays (i.e., 16 x 16, 17 x 17) which have greater surface area and hence lower linear heat rates and surface heat flux. The result is a greater margin to departure from nucleate boiling (DNB), lower clad temperature and peak centerline (fuel) temperature.
3.4 Reactivity ControlSystems Reactivity control is provided by two independent systems; the control rod system and the Chemical and Volume Control System (CVCS). -The control rod system provides control for short-term reactivity changes (e.g., staitup, shutdown and rapid transients) and is used for rapid shutdown (e.g., reactor trip or scram). All PWRs except SYankee-Rowe (Westinghouse) and Palisades (Combustion Engineering) have multi-finger control rod assemblies that insert into thimbles in the fuel assemblies. The multiple control rod fingers are joined at the top by a "spider" assembly connected to an extension shaft that can be engaged by a control rod drive mechanism (CRDM) in the reactor vessel head.
- Yankee-Rowe and Palisades have cruciform control rods that are inserted between the fuel assemblies. The cruciform control rods also are driven by means of extension shafts that are engaged by CRDMs in the reactor vessel head. "Westinghouse and Combustion
-Engineering PWRs have magnetic jack CRDMs that provide rod motion'in small steps.
Babcock & Wilcox PWRs have roller-nut CRDMs that can provide continuous iather than stepped rod motion. Typically, there are 45 to 83 CRDMs ina PWR.
An automatic reactor trip is initiated by the Reactoi Protecti6n System (RPS) when monitored plant conditions reach specified safety system setpoints.- As indicated in Figure 3.4-1, the RPS causes a reactor trip by opening the circuit breakers supplying power to the rod control system. As a result, the CRDMs are deenergized, allowing the
-control rods to fall into the reactor core.',
The CVCS continuously adjusts boron concentration in the primary coolant to compensati for long-term reactivity changes during normal operation (e.g., fuel bumup, effects of xenon). The CVCS integrates the process of adjusting the primary coolant boron concentration with the RCS coolant inventory control function., The principal CVCS flow
..paths and interfaces are shown in Figure 3.4-1. The CVCS can take the reactor subcritical without use of control iods by significantly increasing the boron concentration in the primary coolant.
"3.5' Heat Transfer Systems for Power-Operation--...
When the reactor is operating at power, the normal heat trahsfer path is by means of three fluid system loops as illustrated in Figure 3.5-L_ The first heat transfer loop is the RCS. This is a closed, single-phase, high-pressure (2200 psig) loop which circulates hot primary coolant from the reactor core, through the steam generators, and returns I"cold" primary coolant to the reactor core via the reactor coolant pumps.. In this heat transfer loop, the reactor core is the heat source and the steam generators are the heat sink..
The 'second heat transfer loop islformed by the Steam and Power Conversion System which is a closed, two-phase, lower pressure secondary coolant loop.' During Spower operation,-this secondary coolant system removes heat from the RCS by boiling water in the steam generatoirs at about 700 to 1000 psig. The main-turbine generators extract power from the steam to generate electricity and exhaust to the main condenser which operates under partial vacuum conditions (20 inches mercury vacuum). Heat is transferred in the main condenser to the tertiary circulating water cooling loop and the condensed steam is returned to the steam generators via the main condensate and feedwater systems which together increase the secondary coolant pressure back up to 1000 psig.
The tertiary coolant loop is'the circulating water system which rejects plant waste heaf to the ultimate heat sink. This isa l6w-pressure,' high flowrate, single-phase coolant loop that may operate on an open cycle, closed cycle, or combined cycle. In an open cycle system, the circulating water pumps draw cooling water from a body of water 3/90 3-3
(i.e. :iti ocean, lake, river) and return all of the heated water back to the body of water. In comparison,: a closed cycle system recirculates the condenser cooling water and utilizes cooling towers or other heat exchangers to reject heat to the atmosphere. Water from a nearby source is needed to provide makeup for evaporation. In a combined cycle cooling water system, part of the plant waste heat is rejected to the atmosphere via cooling towers before the circulating water is returned to the water source:
3.6 Heat Transfer Systems for Shutdown Cooling at High RCS During a normal shutdown, initial shutdown cooling is accomplished by using the main turbine bypass system to direct steam to the main condenser, and the condensate and feedwater systems to' return, the secondary coolant to the steam generators. The circulating water system completes the heat transfer path to the ultimate heat sink. This essentially is the same heat transport path as is used during power operation (see Section 3.5) except that the main turbine is tripped and bypassed and the steam, condensate, and feedwater systems are operating at a greatly reduced flow rate.
When the Steam and Power Conversion System is not available, heat may be removed'from the RCS by the combined operation of the Auxiliary Feedwater (AFW)
System and the secondary steam relief system (SSRS). This heat transfer path involves two cooling loops. In the first loop, heat is transferred from the reactor core to the steam generators by forced circulation, or by natural circulation when the reactor coolant pumps are unavailable. In the secondary cooling loop, the AFW system takes water from a condensate storage tank or other suitable water source.and delivers it to the' steam
-generators where it is boiled and vented to atmosphere via atmospheric dump valves in the SSRS. The atmosphere is the ultimate heat sink in this case. Core heat removal by the AFW system and the SSRS is illustrated in Figure 3.6-1.
3.7 Heat Transfer Systems for Shutdown Cooling at Low RCS The Residual Heat Removal (RHR) System provides for post-shutdown core cooling of the RCS after an initial cooldown and depressurization to about 350*F and 425 psig by the Steam and Power Conversion System or the AFW system and the SSRS. As illustrated in Figure 3.7-1, the RHR system establishes a new closed-loop, low-pressure, single-phase primary heat transfer loop by diverting reactor coolant from an RCS hot leg to the RHR heat exchangers. In most PWRs, the RHR system is a multi-mode system that also performs the low-pressure safety injection (LPSI) function as part of the emergency core cooling system (ECCS).
Heat is transferred from the RHR system to a secondary cooling loop and the reactor coolant is returned to an RCS cold leg. The Component Cooling Water System (CCWS) forms the secondary cooling loop. This is a closed-loop, single-phase, low pressure system that also provides cooling for other safety-related components. Heat is transferred from the CCWS to a tertiary loop that rejects heat to the ultimate heat sink. The tertiary loop is a service water system that may operate on an open, closed, or combined cycle. The service water and the circulating water systems may operate on different cooling jcycles (i.e., a closed cycle service water system and an open cycle circulating water system).
3.8 RCS Overpressure Protection System RCS overpressure protection is provided by power-operated relief valves (PORVs) and/or safety valves mounted on the pressurizer.' The safety valves lift mechanically on high RCS pressure;- A typical pressurizer safety valve is shown in Figure S:3.8-1. The PORVs can be controlled to open at lower pressures, thereby reducing the frequency of challenges to the safety valves. The PORVs may also play a role in feed-and bleed, or bleed-and-feed core cooling (see Section 3.9). The pressurizer safety valves and 3/90 3-4
- the PORVs discharge to a "quench tank" located inside the containment. The quench tank is partially filled with water and is sized to handle modest blowdowns from the RCS.
Rupture disks are generally used to provide overpressure protection for the quench tank.
3.9 Emergency Core Cooling Systems Following a breach in the reactor coolant system pressure boundary, water is "lost from the'RCS at a rate that is determined by several factors,-including break size and locationl The Emergency Core Cooling System (ECCS) is a multi-mode system that injects S makeup -water into the RCS during a loss-of-coolant accident (LOCA) and recirculates Sw
- ater thiough the core-following a LOCA-to provide for long-term post-accident core
-. -cooling. In all PWRs, the ECCS includes pressurized safety injection taniks (SITs) and
"'high-and 10w-pressure safety injection (HPSI and LPSI) pumps. The RCS injection
-Cold legs Cold legs Cold legs (initially)
(initially)
Hot legs
'Hot legs (later)
(later)
'Combustion Engineering -
Cold legs Cold legs GCold legs Babcock & Wilcox Cold legs
'Reactor Vessel Reactor Vessel
'In addition, the ECCS in some Westinghouse plants can be aligned to inject into the upper head of the reactor vessel.
3.9.1 ECCS Injection Phase During the injection phase of operation following a large LOCA, the ECCS operates as an open-loop system and provides -rapid injection of borated water to the RCS to ensure reactor shutdown and adequate core cooling. Following a large LOCA, the RCS is rapidly depressurized, and makeup' is initially provided by 'the safety injection accumulators as RCS pressure drops below accumulator pressurie (i.e., 650 psig). Both ihe high-and low-pressure safety injection pumps are aligned to take a suction on the Refueling Water Storage Tank (RWST) and deliver makeup water to the reactor vessel via
" the RCS Cold legs. Water lost from the RCS during the LOCA is collected in the containment sump. The coolant injection and heat transport paths associated with large LOCA mitigation are shown in Figure 3.9-1.
Following a small LOCA, the RCS may slowly depressurize or remain at or near normal operating pressure. 'RCS pressure behavior will be determined by many factorsi,including the size of the small break and the availability of the steam generators as a heat sink. An RCS heat balancewill be establishid between the heat generated in the reactor core *ahd heat lost-via the small break, the steam generators, and if necessary, the primary power-operated relief valves (PORVs) and safety valves located on the pressurizer.
Maximium RCS pressure is limited by the primary safety valves. In some PWR plants, makeup to~the RCS can be provided by the ECCS high-pressure safety injection pumps at pressures up to ihe primary safety valve setpoint. 1 In these plants, it is a relatively straight forward matter to control RCS coolant inventory following the small LQCA.
In some PWR
-plants, the ECCS high-pressure safety injection pumps have a shutoff head in the range'from'1400 to 1800 psig and, therefore, are not capable of providing makeup at full RCS pressure. In these plants, RCS makeup at high pressure is limited to the capacity (and availability) of the normal charging pumps, therefore it is necessary to depressurize the RCS to enable the high-pressure injection pumps to provide 3/90 3-5
RCS makeup:, RCS depressurization can be accomplished by means of heat transfer to the steam generators using the AFW system and the SSRS as described previously.
Alternatively, it may be possible to reduce RCS pressure by opening the PORVs on the pressurizer (i.e., bleed-and-feed).
3.9.2 ECCS Recirculation Phase After the RWST makeup water supply has been exhausted, the ECCS is placed in the recirculation mode of operation by aligning the suctions of the low-pressure safety injection pumps to the containment sump and isolating the suction path from the RWST. In most PWR plants; the high-pressure safety injection pumps cannot be aligned to take a suction directly from the containment sump. At the time recirculation is initiated, the normally dry containment sump is full of water that has collected from the LOCA and from the operation of the containment spray system. -
I - I Following a large LOCA, the RCS is depressurized to the point that the low pressure safety injection pumps can provide continuous makeup to the RCS and the high pressure pumps may be stopped. If available, heat exchangers in the low-pressure safety injection system may be used during the recirculation phase to transfer heat to the ultimate heat sink via the CCWS and the. service water system. The low-pressure ECCS recirculation loop is comparable to the RHR shutdown cooling loop described in Section 3.7, with the exception that the low-pressure pumps are aligned to take a suction from the containment sump.
During a small LOCA, RCS pressure may remain high, precluding injection by the low-pressure safety injection pumps which typically have a shutoff head on the order of 300 to 400 psig. In this case, the high-pressure recirculation flow, path is established with the low-pressure and high-pressure safety injection pumps operating in tandem. The low pressure pumps take a suction on the containment sump and are aligned to deliver the water to the suction of the high-pressure pumps which then inject water into the RCS via the cold legs (initially) or the hot legs (later). Water returns to the containment sumpthrough the RCS break that caused the LOCA. Heat exchangers in the low-pressure safety injection system may be used during high-pressure recirculation to transfer heat to the ultimate heat sink via the CCWS and the service water system; 3.9.3 High-Pressure, Feed-and-Bleed Cooling Some PWRs have the, capability to use the high-pressure ECCS pumps to implement a post-transient decay heat removal method called feed-and-bleed cooling. In essence, this is little more than-small LOCA mitigation with the pressurizer PORV substituting for a break in the primary system.- If the steam generator is unavailable as a post-transient heat sink, RCS pressure will increase to the point that the pressurizei safety valves and/or the PORVs Will lift. The RCS will remain at high pressure and a heat balance will be established between decay heat generated in the core and heat carried off via the pressurizer safety valves and/or the PORVs: As shown in Figure 3.9-2, feed-and-bleed cooling is implemented by aligning a high-pressure makeup pump to maintain RCS inventory and modulating the PORVs to control RCS cooldown rate. Normally a discharge from the pressurizer safety valves and/or the PORVs is contained in the pressurizer quench tank. This tank is not sized for continuous feed-and-bleed operation, therefore, rupture disks on the tank will burst, venting the tank to the containment. 'The containment cooling "systems are needed to complete the heat transport path to the ultimate heat sink. Normally, RCS coolant inventory is measured by the water level in the pressurizer. During feed-and bleed cooling, pressurizer water level may not be an accurate indication of RCS coolant inventory. Furthermore, repeated cycling of the pressurizer safety valves and/or the PORVs may result in valve failure and an actual LOCA due to a stuck-open valve.
3/90 3-6
S3.10 -
ContainIment an d Containment Auxiliary Systems The containment structure is a physical boundary against the release of fission Sproducts to the environment following a release from the RCS. There are three functionally different types of primary containments used in U.S. PWRs:
"I-Large,;dry (atmospheric)'containment (Large, dry) subatmospheric containmrent Ice condenser (pressure suppression) containment These.pnmary containment designs may b6 constructed of steel or concrete and may be used with or'without a secondary containment. PWR containment types are summarized by functional design in Figures 3.10-1. PWR containments are not inerted.
3.10.1 j Lairge, Dry Containment The large, 'dry, atmospheric containment is the predominant type of PWR containment, being found in 53 of 80 PWR plans. -All Combustion Engineering and
'Babcock '& Wilcox PWRs have large, dry conta'inments.
Example of large, dry containiient configurations are shown in Figure 3.10-2 (Yankee-Rowe steel sphere), 3.10 S
3,(Davis-Besse steel cylinder with concrete shield building), 3.10-4 (Diablo Canyon ieinfoiced concrete cylinder with steel liner).and 3.10-5, (Zion post-tensioned concrete cylinder';rith steel liner). Design pressures for large, dry containments vary considerably, but generally are in the range from 40 to 61 psig.,Abnong plants with large, dry containments, Yankee-Rowe has the lowest containment design pressure at 34 psig.
3.10.2 Subatmospheric Containment pls Subatmosphericcontainments are only found at seven Westinghouse PWR plants, six 3-loop plants, and one 4-loop plant.,All subatmospheric containments are constructed of reinforced concrete with a steel liaer.. An example of the configuration of a suba'tmospheric containment is shown in'Figure 3.10-6 (Millstone 3 4-loop PWR). Design pressures for subatmospheric containments vary considerably, but generally are in the pnge from 45 to 60 psig.
'3.10.3 Ice' Condense'r Containment
" "Ice condenser containments are only found at ten Westinghouse 4-loop plants.
Examples of ice condenser'containments are showing Figure 3.10-7 (Catawba, Sequoyah, and Waits Bar steel cylinder with concrete shield building), and 3.10-8 (typical of D.C.
Cook and McGuire reinforced concrete cylinder with steel liner). An isometric view of the ice condenser is shown in Figure 3.10-9. Due to the pressure suppression effects of the ice condenser, these containments have lower design pressures than" either large, dry containments orsibatmospheric'containment&.'.Typical design pressures for ice condenser containments are in the range from 12 to 30 psig.
3.10.4 Containment Auxiliary Systems,
Regardless of the 'containment type, all -PWR "containment designs have aux'ilir systems to accomplish the functions of contain ment 'iolanon, containment pressure control and hieat removal, containment fissioh product cleanup, and combustible gas control. Systems related to these functions are described below.'
A.
Containment Isolation.
c n
s c
"- During normal operation, PWRcontaimnts typically are closed, or have only a limited amount of "purge" airflow. "Containment cooling during normal operation typically is provided by a recirculating ventilation system, therefore, large diameter ventilation lines penetrating containment can remain isolated.
Following a LOCA, the containment isolation system causes isolation valves 3-7 3/90
and dampers to close in certain lines that penetrate the containment boundary, including any open containment puige lines.
B.
Containment Pressure 'Contro1;'. Heat 'Removal, and Fission Product Cleanup The functions of containment pressure control, heat removal, and fission product cleanup are integrated in the containment spray system in most PWRs.
Containment pressure and fission product concentration in the containment atmosphere are reduced by a containment spray system. The design of this system varies with containment type, but a spray system is found in large, dry containments,'subittmospheric containments, and ice condenser containments.
In most PWRs, the containment spray system initially, injects water from the RWST into the conitainment via spray headers located in the dome of the containment. A chemical additive usually is added to the spray water to enhance its fission product removal capability.- Ii the ice condenser plants, the ice beds perform a pressure suppression' function to limit maximum containment pressure and also provide some "scrubbing" of fission prodkcts in the containment atmosphere.- When the RWST has been emptied in a plant with a
'large; dry containment oi an ice condenser containment, the cofitainment spray pump suction is'aligned to the containment sump and the suction path from the RWST is isolated. In a subatmospheric containment, there typically' are two spray systems; an injection spray system that functions as described above, and a recirculation spray system., When the RWST has been emptied in a plant with a subatmospheric containment, the injection spray system is secured and the recirculation spray system is started.
In large, dry containments, post-LOCA containment heat removal is accomplished by heat exchangers in the containment spriy system or the residual heat removal system and/or containment fan-coolers. The fanr coolers include filter beds for fission product removal. 'A simplified diagram of these containment cooling system heat transport paths is shown in Figuire 3.10-10.
The heat transfer path from the containment spray (or RHR) heat exchangers and the containment fan coolers to the ultimate heat sink is completed by one or two cooling water loops (i.e. the CCWS and/or ihe serice water system).
Plants with subatmospheric coniainments or ice condenser containments typically do not have containment fani coolers for post-LOCA containment heat removal.
C.
Containment Combustible Gas Control PWR containments are noi ine*rted. Post-LOCA combustible gas concentration in the containment can be controlled by hydrogen'recombiners and igniters.
3.11 Component Cooling Systems.
The Component Cooling Water System (CCWS) is a low-temperajure, low pressure, single-phase cooling system that provides cooling f6r a wide range of safety related components.
As illustrated in Figure 3.11-1, the CCWS may pi6vide for component, area, or system cooling in a variety of ways:
A.
Direct component cooling This type of cooling arrangement is illustrated by cooling paths A-A' and applies to cooling'for pump bearings and seals.
3/90 3-8
B.
Area cooling by means of fan cooler units This type of cooling arrangement is illustrated by cooling paths B-B' and
-applies to equipment room coolers that are required to maintain normal and post-accident environmental conditions within limits necessary for long-term operation of components in safety systems.
C.
Fluid system heat removal This type of cooling arrangement.is illustrated by cooling paths C-C' and applies to CCWS cooling for a yariety of systems including the RHR system, containment spray system, and the spent fuel pool cooling system.
D.
Area cooling by means of HVAC Units This type of cooling arrangement is illustrated by cooling paths D-D' and applies to CCWS cooling for normal and emergency heating, ventilating, and air-conditioning (HVAC) systems. In this case, the chiller system forms a closed-loop heat transfer system between the CCWS and the area(s) being cooled.
The CCWS rejects heat to the ultimate heat sink through the cooling loop
- indicated'as the service water system in Figure 3.11-1.- This service water system may operate on an open, closed, or combined cycle as described previously.
-3.12
- Safety System Actuation The role of the Safety System is to actuate components and systems needed to mitigate the consequences of events that challenge limits established,for normal plant operation. The Safety System ^consists of two major subsystems: the Reactor Protection System (RPS) and the Engineered Safety Feature Actuation System (ESFAS). As described previously, the function of the RPS is to initiate a reactor scram when needed.
The ESFAS provides for automatic actuation of a wide variety of components and systems based on the detection of abnormal conditions in the reactor plant. As appropriate, the ESFAS can actuate systems necessary for, RCS coolant inventory control and/or core cooling, containment isolation and cooling, radioactive release control, emergency power, and component cooling.
As illustrated in Figure 3.12-1, the ESFAS includes provisions for manual actuation at the system level (typically from the control room) or at the actuation-train level (typically from the ESFAS output logic cabinets). A manual trip from the control room actuates all components that would be actuated by an automatic ESFAS actuation signal. A manual trip from ESFAS output logic cabinets actuates only the components that are controlled by the respective ESFAS traimn.
The relationship between the ESFAS and other means of actuation is also shown in Figure 3.12-1. Individual remote-manual component controls, which do not use any part of the ESFAS logic, generally are provided in the control room and/or at some other alternate control location. In addition, most motor-driven components can be manually actuated by manipulating their circuit breaker on the respective switchgear panel or motor control center. Other types of power-operated valves often can be controlled locally by manual manipulation of the pilot valves on the pneumatic or hydraulic actuator.
3.13 Onsite Electric Power System The onsite electric power system consists of two parts; the non-Class 1E system which supplies non-safety loads, and the Class 1E system which supplies safety systems.
During normal operation, the entire onsite electric power system is supplied from the output of the main generator and/or the offsite grid. Diesel generators are standby AC power sources for the Class 1E portion of the onsite power system, and batteries are standby DC 3/90 3-9
power sources. During normal operation, the diesel generators are idle, and the batteries are maintained fully charged by battery chargers which also supply the DC power loads.
Large Class IE AC electrical loads (i.e. large pumps and fans) typically are supplied from 6.9 or 4.16 kV switchgear. Smaller Class IE AC loads (i.e. motor-operated valves, small pumps and fans, battery chargers) are supplied from 480 VAC motor control centers. A representative onsite 4.16 kV and 480 VAC power system (Callaway) is shown in Figure 3.13-1.
Most DC loads are supplied from 125 VDC panels, although some plants may have 250 VDC distribution systems to support DC-powered motor-operated valves or other relatively large DC-powered components' Instrumentation power typically is supplied from a 120 VAC system that normally is powered from the 125 VDC system with backup power from the 480 VAC system. A representative 125 VDC and 120 VAC system (Callaway) is shown in Figure 3.13-2.
Loss of the normal (preferred) source of offsite power typically causes an automatic shift to the alternatecsource of offsite power and starts the respective standby diesel generator(s). If both sources of offsite power are unavailable, the non-Class IE and the Class 1E portions of the onsite'electric power system are separated by opening circuit breakers, and the diesel generators are aligned to supply the Class 1E system. The standby diesel generators and batteries can provide adequate power to enable other safety systems to establish and maintain a safe shutdown condition.
The diesel generators are complex systems with integrated diesel and generator control systems 'that interface with a load-sequencing' system that re-energizes selected loads in prescribed sequences when the diesel generator is ready for loading. Diesel generator starting is dependent on a source of DC power (usually the station batteries) for the control systems and generator field flashing. In addition, the following support systems typically are needed for diesel generator operation:
Fuel oil system (including the day tank which is the short-term fuel source)
Fuel oil storage and transfer system (long-term fuel source)
Air start system - ;
Lubricating oil system Jacket cooling,water system Combustion air intake and exhaust system Diesel room cooling system Simplified schematics for'these systems are shown in Figures 3.13-3 and 3.13-4. As shown in Figure 3.13-4, heat from the diesel generator jacket cooling water system and lubricating oil system is transferred to the ultimate heat sink via a service water system. In a few plants, the jacket cooling water system may incorporate a radiator (i.e., a water-to-air heat exchanger) and use the atmosphere as a heat sink for diesel generator operation. A significant amount of heat from diesel generator operation is transferred directly to the atmosphere by the diesel exhaust system and the diesel room ventilation system.
3-10 3/90
Table 3.1-1.
Summary of PWR Systems PWR System NSSS Scope IBOP Scope Reactor X
Reactivity Control System X
Shutdown Cooling System X_
X Reactor' Water Cleanup System X
Containment X
Emergency Core Cooling System X
Habitability Systems X
Containment Spray Systems X
ESF Filter Systems X
Reactor Trip System X
Engineered.Safety Feature Actuation System X
X Safety' Related Display Instrumentation-X X
Non-Safety Control Systems X
On-Site Electric Power System X
Off-Site Electric Power-System X
New Fuel Storage System X
Spent Fuel Storage System X
Spent Fuel Pool Cooling and Cleanup
- System, X
Fuel Handling System X
Service Water System X
Component Cooling Water System X
Ultimate Heat -Sink X
Compressed Air System X
- Process Sampling System X
Chemical and Volume Control System X
Non-safety HVAC System X
Fire Protection System X
Diesel Generator Fuel Oil Storage and Transfer System X
Diesel Generator Cooling Water System X
Diesel Generator Starting-System X
Diesel Generator Lubrication System X
Diesel Generator Combustion Air Intake and Exhaust System X
Main Steam System X
Turbine Generator and Auxiliaries X
Main Feedwater and Condensate System X
Auxiliary Feedwater System X
Radioactive Liquid Waste System X
Radioactive Gaseous Waste System X
Radioactive Solid Waste System X
Radiation Protection System X
I.
.1 339O 3-11
Reactor Figure 3.2-1.
Westinghouse 2-Loop PWR NSSS 3-12 3M90
Figure 3.2-2. Westinghouse 3-Loop PWR NSSS 3-13 3/90 i"
Figure 3.2-3.
Westinghouse 4-Loop PWR NSSS 3-14 3/90
1 plant Steel Sphere
~XWith Concrete IEnc oaure Bldg.d ItReinforced Concrete5 Cylinder With Steel Liner &
Sec. Containment IN1-0 (Verticylnd Post-Tensioned
- Concrete Cylinder With Steel Uner Diagonal Poet-Tensioned Concrete Cylinder With Steel Uiner fi3-D Post-Tensied Concrete Cylinder With Steel Uiner 3-D ýPoet-Tesio.ned Concrete Cylinder, "With Steel Liner &
See Containment SReinforced Concrete Cylinder With Steel Liner &
Sec. Containment 1 plant 7 plants 11 plants I plant 2 plants 1 plant 36 plants 3 plants 6 plants 1 plant 4 plants 6 plants Figure 3.10-1.
Distribution' of PWR Containment Types (7 plants)
(10 plants) 1 3-25 3/90
Figure 3.10-2. - Yankee-Rowe Large, Dry Containment (Steel Sphere) 3-26 3/90
-Steel containmrent-vessel structure Reinforced-concrete secondary containment structure
$1 U) 0 C (DC.*
ow, Cop
-W 0 0
Fw Uo C),I OWJ
-. 3
Figure 3.10-4. Diablo Canyon Large, DryContainment (Reinforced Concrete with Steel Liner) 3-28 3/90
SCONTAINMENT SPRAYS CRAN POLAR CRANE 0
STEAM GENERATORS REACTOR VESSEL INSTRUMENT-TUNNEL CAIT 9 FT BASEMAT Figure 3.10-5. Zion, Large, Dry Containment (Post-Tensioned
'Concrete with Steel Liner),
U 3-29 3190
Containment liner.
(typical)
In-core I n-core Neutron shield-tank coolo In-core instrumentation I Containment i liner (typical) room sump pump shield tank LP. (El. 25 ft 6 in.)
Figure 3.10-6.- Millstone 3 Subatmospheric Containment, (Reinforced Concrete with Steel Liner)
Reactor heater 3-30 3/90
SHIELD BUILDING DOM E*..
~CONTAINMENT
~S PRAY STEL-
-9
.9O CONTAINMENT, VESSEL I-- -- '
~AIR RETUR FAN.
!v DIVIDER
. f
~BARRIER
,,,CRANE
,, WALL
~CONTROL ROD DRIVE
,MISSILE SHIELD C R
÷AV
- SUMP SHIELD BUILDING WALL"
- TOP OF SICE BED ICE CONDENSER VAPOR BARRIER" A6CUMULATOR'
..HYDROGEN IGNITER 1641
..STEAM GENERATOR SVENTILATION FAN AND EQUIPMENT Figure'3.10-7. Sequ'oyah Ice Condenser Containment (Steel Cylinder with Concrete Shield Building)
-- 3-31 3/90
Figure 3.10-8.
Ice Condenser, Containment with Reinforced Concrete Structure-and Steel-Liner 3-32 3/90
-1
AIR HADLIWG UN IT DECK ICE LATTICE FRAME-WALL-WALL PANELS TURN ING VANES -
- CRANE
-AIR DIS-IBUTION DUCTS WALL
-LOWER INLET DOORS
- LOWE R SUPPORT STRUCTURE FLOOR Figure 3.10-9.
General Arrangement of an Ice Condenser 3/90 3-33
°*
WESTINGHOUSE PRESSURIZED WATER REACTORS (PWRs)
In the U.S., Westinghouse has produced 55 PWRs with two, three, and four primary loops. As of March 1990, the numbers of units of each type are as follows:
2-loop 6 units 3-loop 14 units 4-loop 35 units A general orientation to PWR systems is presented in Section 3. Expanding on this introductory material, an overview and brief comparison of major features of Westinghouse PWRs is presented in Section 4.1. The 2-, 3-, and 4-loop Westinghouse plants are described in Sections 4.2 to 4.4, and numerous detailed comparative tables are presented in Section 4.5. In Section 7, the comparative tables for the Westinghouse PWRs are compiled with similar tables'for Combustion Engineering and Babcock & Wilcox PWRs.
.4.1 -
"Westinghouse PWR Overview'
--4.1.1 Primary System In all Westinghouse PWRs, a primary' loop consists of a U-tube steam S generator, a single vertical, centrifugal reactor coolant pump, and connecting loop piping.
The pressurizer is connected to one of the RCS hot legs. The general configuration of the Westinghouse reactor vessel and internals is shown in Figure 4.1-1. The U-tube steam generator is shown in Figure 4.1-2 and the pressurizer is shown in Figure 4.1-3. Typical reactor vessel sizes for the three Westinghouse PWR models are as follows:
2-loop 132 inch i.d.
3-loop 156 to 159 inch i.d.
4-loop 173 inch i.d.
There are four principal models of U-tube steam generators in Westinghouse plants: 27-series, 44-series, 51-series, and Model F. All models have integral moisture separators and steam dryers. Basic design characteristics of Westinghouse steam generators are listed in Table 4.1-1.
The Westinghouse RCS is designed to operate with nearly constant cold leg temperature (Tcold). Hot leg temperature (Thor) and average loop temperature (Tave) increase with power level as shown in Figure 4.1-4.. This is the same RCS temperature control scheffie used in Combustion Engineering PWRs.
4.1.2 Reactor.Core and Fuel Assemblies SWith the exception of Y ankee-R ow e and Shippingport, all W estinghouse commercial PWRs are designed to operate with rod-type slightly-enriched fuel in a 14 x 14 fuel assembly array, a 15 x 15 array, or a 17 x 17 array. Yankee-Rowe has a unique rod type fuel assembly design, and Shippingport had unique plate-type and rod-type fuel assemblies. All Westinghouse PWRs except Yankee-Rowe and Shippingport also use multi-finger control rods that insert into channels in the fuel assemblies. Yankee-Rowe and Shippingport both were designed with cruciform control rods. A comparison of basic Westinghouse core parameters is presented in Section 4.5. A brief description of each rod type fuel assembly design is provided below.
A.
Yankee-Rowe Fuel Assembly (9 x (6 x 6))
The first-generation of Westinghouse PWR fuel assembly, used only in Yankee-Rowe, consisted of nine 6 x 6 subassemblies arranged in a square array (i.e. essentially a 36 x 36 array). The assemblies measure 7.61 inches square.
3/90 4-1
Table 4;1-1.
General Characteri-stics of Westinghouse Steam Generators, Design -Parameters 27 Series 44 Series 51 Series Model F Tube Side (primary)
Design Pressure (lsig) 2,485
.2,485 2,485 2,485 Norm. Operating Press. (psig) 2,000 2,235 2,235 unk.
Design Temp. (°F) unk.
650 650 650 Design Flow Rate Ob/hr) -
2.50E+07 3.40E+07 3.40E+07 3.55E+07 Total Primary Side Vol. (ft3) 553 944 1,080 962 Shell Side (secondary)
Design Pressure (psig) 1,035 1,085 1,085 1,185 Full Power Pressure (psig) 570 755 797 to 960 1,000 Design Temp. (IF) unk.
600 600 600 Full Power Temp. (OF) unk.
514 518 to 540 544 to 559 Steam Flow Rate (lb/br) 1.55E+06' 3.32E+06 3.76E+06 to 3.78E+06
@ Full Power 4.06E+06 Total Secondary Side Vol. (ft3) 2,592 4,580 5,868 3,559 I.
I 4-2 3/90 I
with an overall length of 111.25 inches. As shown in Figure 4.1-5, the cruciform control rods were inserted between the'fuel assemi'blies. Rubbing straps on the outside edges of the fuel assemblies protected the outer fuel rods from wear by the control rods. The fuel rod cladding was thick-wall stainless steel; and spacing between rods was established by ferrules brazed to the fuel rods.
The later-generation Yankee-Rowe fuel continued the geometry of the original fuel assembly, but changed to thinner-wall cold worked stainless steel cladding "or zircaloy. In addition, fuel rod spacing was established by a grid structure.
B.
14 x -14 Fuel Assembly The 14 x 14 fuel assembly was introduced and is still used inseveral early Westinghouse plants -(i.e.,' San Onofre 1, Point Beach) along with a corresponding change in the design of the control rods. The use of "rod cluster "control assemblies" (RCCAs) distributed control rod poison more uniformly by means of multi-finger control rods that can be inserted into full-lenath'thimbles in each fuel assembly. The' 14 x 14 fuel assembly is designed fori use With a 16
","finger" RCCAs. The insertion of.an RCCA into a 14 x 14 fuel assembly is shown in Figure 4.1-6. The thimbles are'structural assemblies that join the end pieces of the fuel assembly. The use of RCCAs instead of cruciform control
- 'rods reduced the occurrence of local hot spots when the rods were-withdrawn.
When first introduced, the 14 x 14 fuel assemblies used stainless steel cladding
-and structural parts, but changed to zircaloy cladding. Gradual replacement of stainless steel by zircaloy in fuel assembly structural parts continued into the
-1970s.
C.
15 x IS Fuel Assembly _
Use of the 15 x 15 fuel assembly was introduced in 1967 with stainless steel cladding-and structural ýparts, but changed to'zircaloy -cladding in 1968.
Gradual replacement of stainless steel by zircaloy in fuel assembly structural parts continued into the 1970s. The 15 x 15 fuel assembly is designed for use with a 20 "finger" rod cluster control assembly (RCCA). Most plants that have used the 15 x 15 fuel assemblies are transitioning to-the 17 x' 17 fuel
-- assemblies.
D.
17 x 17 Fuel-Assembly "
-The Westinghouse 17 x-17 fuel assembly is shown-in Figures '4.1-7 and 4.1-8.
The 17 x 17 fuel assembly was designed to fit in the same geometric and power envelope as the 15 x 15 fuel assembly. The use of the 17 x 17 array resulted in numerous advantages over the 15 x 15 fuel assembly, including:
Core average power density reduced from 7.03 to 5.43 kW/liter (for a 3411 MWt core)
Reduced linear power rating 12% lower surface heat flux, hence increased margin to Departure from Nucleate Boiling (DNB)
About 500OF (278 0C) reduction in peak clad temperature under LOCA conditions.
3/90 4-3
Nominal fuel enrichment is in the range from 2.1 to 3.1 weight percent U-235.
The 17 x 17 fuel assembly is designed for use with a 24 "finger" rod cluster control assembly (RCCA)., The center thimble is reserved for in-core instrumentation., This fuel assembly is mostly zircaloy except for the bottom nozzle which is 304 stainless steel, and various springs and bolts which are Inconel 600 or 718.
4.1.3 Reactivity Control System Core reactivity is controlled by full-length and part-length rod cluster control assemblies (RCCAs) burnable poison rods, and soluble boron in the coolant. The neutron absorber in the control rods typically is Ag-In-Cd (silver-indium-cadmium), while the burnable poison is B4C (boron carbide) that may be in the form of borosilicate glass.
Some Westinghouse plants use B4C for the control rod poison.
As shown in Figure 4.1-9, the RCCAs consist of multiple control rods that are joined at the top by a "spider" assembly and attached to the control rod drive mechanism (CRDM) extension shaft., Magnetic jack-type CRDMs are used to control the full-length and part-length RCCAs. These CRDMs consist of a set of five magnetic coils outside of the CRDM pressure housing, and solenoid-operated plungers and "gripper latches" inside the pressure housing to engage the grooved drive rod extension shafts and hold, insert, or withdraw the control rods. The five sets of CRDM coils are: (a) the stationary gripper coil, (b) the movable gripper coil, (c) the lift coil,* (d) the push-down coil, and (e) the load transfer coil. The action of these coils is programmed so that stationary and movable grippers are alternately engaged with the grooved drive shaft. The stationary gripper holds the drive shaft while the movable gripper is moving to its new, position to raise or lower the control rod through steps of about 3/8 inch. The CRDMs for the full-length rods are designed so that, upon loss of electrical power to the magnetic jack coils, the RCCA is released and falls by gravity into the core. The CRDMs for the part-length RCCAs are designed to hold the rods "as-is" upon loss of power. Details of a magnetic jack CRDM are shown in Figure 4.1-10 and the location of the CRDMS with respect to the reactor vessel and fuel assemblies is shown in Figure 4.1-11. The magnetic jack CRDMs are air cooled.
4.1.4 Containment Westinghouse PWRs have been built with a greater variety of containment designs than either C-E or B&W PWRs, and include the only PWRs with subatmospheric containments and ice condenser containments. The distribution of containment types for Westinghouse plants is shown in Figure 4.1-12.
Examples of containments for Westinghouse PWRs are included in the following sections.,
3/90 4-4
STEAM SWIRL VANE FEEDWATER NOZZLE TUBE WRAPPER.
TUBE BUNDLE SUPPORT TUBE SHEET, Figure 4.1-2.
(Typical
,POSITIVE ENTRAINMENT STEAM DRYERS SECONDARY MAKWAY UPPER SHELL LO ANTIVIBRATION BARS TUBE SUPPORT PLATE VER SHELL WIDER PLATE Westinghouse U-Tube Steam Generator of Series 44 and 51 and Model F) 4-6 3/90
I MODEL D STEAM GENERATOR
Figure 4.1-5. General Arrangement of the Yankee-Rowe 9 X (6 X 6) Fuel Assembly 4-9 3/90
Yankee-Rowe (4-loop)
San Onotre I (3-loop)
Kewaunee (2-loop)
Prairie Island I & 2 (2 -loop)
Comanche Peak I & 2 (4-loop)
Diablo Canyon I
& 2 (4-loop)
Haddam Neck (4.loop)
Indian Point 2 & 3 (4-loop)
Salem 1 & 2 (4-loop)
Shearon Harris 1 (S-loop)
Seabrook 1 (4-loop)
M Post-Tensioned Ginna (24oop)
Concrete Cylinder H.B. Robinson (3-loop)
With.,.
Steel Line Diaonal SPost-Tensloned No Westinghouse Plants
-Concrete Cylinder Nh With Steel Liner
- Concrete Cyl With Steel t Braidwood 1 & 2 (4.loop)
Byron 1" & 2 (4-loop)
Callaway (4-loop)
Farley I
& 2 (S-loop)
Point Beach 1 & 2 (2-loop)
South Taxas 1 A 2 (c-loop)
Summer (3-1oop)
Trojan (4-loop)
Turkey Point 3 & 4 (S-loop)
Vogtie I & 2 (4-loop)
Wolf Creek (4-loop)
Zion I & 2 (4-loop)
No Westinghouse Plants Beaver Valley I & 2 (3-loop)
North Anna I & 2 (3-loop)
Surry 1 & 2 (3-loop)
Millstone 3 (44oop)
I Concrete Cylinder With Stee. nel e-r PRIMARY j
I CONTAINMENT I
j.Steel Cyindertij Sequoyah 1 1 2 (4.loop) 4 Bulldlngret5: Watts Bar I & 2 (4-loop)
I*ShIslduidgi ReiRnfored Concrete5 Catawba 1 & 2 (4-loop)
Cylinder With D.C. Cook 1 & 2 (4-loop)
Steel Liner
! McGuire 1 & 2 (,.loop)
Figure 4.1-12.
Distribution of Containment Types for Westinghouse Reactors 4-16 3/90 CONCRETE CYLINDER WITH MC*ay11rUI=0
4.2 2-Loo Westinghouse PWRs Westinghouse 2-loop plants in the United States include the following:
Ginna
-° Kewaunee Point Beach 1 and 2 Prairie Island 1 and 2 All of these plants had full power operating licenses as of 2/89.
4.2.1 Reactor Core and Fuel Assemblies All 2-loop Westinghouse plants have cores comprised of 121 fuel assemblies that yield power levels in the range from 1518 to 1650 MWt. All 2-loop plants use 14 x 14 fuel assemblies and all have power densities in the range from 87 to 96 kW/liter.
4.2.2 Reactivity Control System Typically there are 29 to 33 rod cluster control assemblies (RCCAs) in 2-loop plants.
4.2.3 Reactor Coolant System The general configuration of a 2-loop Westinghouse reactor coolant system is shown in Figure 4.2-1.
All 2-loop plants have a reactor vessel with a 132 inch inside diameter.
4.2.4 Steam Generators Ginna and Point Beach 1 and 2 use the intermediate-size 44-series steam generators while Kewaunee and Prairie Island I and 2 use the large 51-series steam generators.
4.2.5 Shutdown Cooling Systems Shutdown cooling is accomplished by the Residual Heat Removal (RHR) system which also operates in the Low-Pressure Safety Injection (LPSI) mode as part of the ECCS.
4.2.6 Emergency Core Cooling Systems A representative ECCS for a 2-loop Westinghouse PWR is comprised of the following subsystems:
Three High-Pressure Safety Injection (HPSI) pump trains that inject into the cold legs via a boron injection tank.
Two RHR trains which perform the Low-Pressure Safety Injection (LPSI) function, each with a pump and heat exchanger Three Safety Injection Accumulators, each connected to an RCS cold leg The shutoff head of the HPSI pumps is about 1750 psig, therefore, these "pumps are unable to provide makeup to the RCS at normal operating pressure. In the event of a small LOCA which leaves the RCS at high pressure, it is necessary to first depressurize the RCS before the HPSI pumps can provide makeup. Depressurization can be accomplished by heat transfer from the RCS to the steam generators or by opening the power-operated relief valves on the pressurizer.
The HPSI pumps cannot be aligned to take a suction directly on the containment sump. If needed, the RHR pumps can be aligned in tandem with the HPSI pumps for high-pressure recirculation.
4-17 3/90
The normal charging system which has three low-capacity positive displacement pumps, is not part of the ECCS.
4.2.7 Containment All Westinghouse 2-loop plants have large, dry containments of varioms designs, as summarized below.
Containment Construction Annlicable Plants Steel cylinder with concrete Kewaunee enclosure building Prairie Island 1 and 2 One-dimension (vertical)
Ginna post-tensioned concrete cylinder with a steel liner Three-dimension post-tensioned Point Beach 1 and 2 concrete cylinder with a steel liner The general arrangement of the Ginna containment is shown in Figure 4.2-2 (section views) and Figures 4.2-3 and 4.2-4 (plan views).
3/90 4-18
Steam
- CQ Ja Figure 4.2-I.' Westinghouse 2-Loop NSSS 4-19 3/90
w!
Figure 4.2-2.
Dry Section Views of the Ginna Large, Containment (Sheet 1 of 2) 4-20 3/90
g.A
- .o Figure 4.2-2. Section Views of -the Ginna Large, Dry Containment'(Sheet 2 of 2) 3/90 4-21
S'
'"to a-t*"*t*-:V"
"-0
.. F-~--r--,_
Figure 4.2-3. Plan View of the Ginna Large, Dry Containment Below the Elevation of the Operating Floor 4-22 3ý/YU
Figure 4.2-4.
Plan View of the Ginna Large, Dry Containment, Above the Operating Floor 4-23 3/90
4.3 3-Loon Westinghouse PWRs Westinghouse 3-loop plants in the United States include the following:
Beaver Valley I and 2 Farley I and 2 H.B. Robinson 2 North Anna I and 2 San Onofre 1 Shearon Harris 1 Summer Surry land2 TurkeyPoint 3 and 4 All of these plants had full power operating licenses as of 2/89 except San Onofre 1 which had a provisional operating license.
4.3.1 Reactor Core and Fuel Assemblies The 3-loop Westinghouse plants can be grouped into the following categories based on core thermal output:
1347 MWt 2208 to 2441 MWt 2660 to 2775 MWt San Onofre 1 Robinson,'Surry 1 and 2, and Turkey Point 3 and 4 Beaver Valley 1 and 2, Farley 1 and 2, North Anna 1 and 2, Shearon Harris 1, and Summer All 3-loop cores are comprised 'of 157 tuel' asemblies. San Onofre 1 uses 14 x 14 fuel assemblies and has an average power density of ab6ut 70 kW/liter. The intermediate power level group uses 15 x 15 fuel assemblies and has power densities in the range from 82 to 92 kW/liter. Plants in the high power group use 17 x 17 fuel assemblies to yield power densities of 100 to 108 kW/liter.:
4.3.2 Reactivity Control System The number' of full-length and Westinghouse plants are summarized below:
part-length RCCAs for selected 3-loop lant Robinson Surry Beaver Valley 1 and 2 Fuel Elements 157 157 157 Full-Length RCCA~
45 48 48 Part-Length 8
5 5
A more complete listing is provided in Section 4.5.
4.3.3 Reactor Coolant System The general configuration of a 3-1oop Westinghouse reactor coolant system is shown in Figure 4.3-1. Most 3-loop plants have a reactor vessel with a 156 to 159 inch inside diameter. The only exceptions are the following:
144 inch vessel 172 inch vessel San Onofre 1 Summer and Turkey Point 3 and 4 4-24 3/90
-1
4.3.4 Steam Generators Three vintages of steam generators can be found-among the 3-loop Westinghouse plants, as follows:
27-series San Onofre I
-, 44-series
-Robinson and Turkey Point 3 and 4 51-series Beaver Valley I and 2, Farley I and 2, North Annia 1 and -2, Shearon Harris 1, Summer, and Surry I and 2 A comparison of Westinghouse steam generator design parameters is included in Section 4.1.
4.3.5 Shutdown Cooling Systems For the following 3-loop plants, shutdown cooling is accomplished by the Residual Heat Removal (RHR) system which'also operates in the Low-Pressure Safety Injection (LPSI) mode as part of the ECCS:
Farley 1 and 2 Summer Robinson Turkey Point 3 and 4 Shearon Harris I The San Onofre 1 and the plants with subatmospheric containment (Beaver Valley, North Anna, and Surry) appear to have separate RHR systems.
4.3.6 Emergency Core Cooling Systems In most of the 3-loop Westinghouse plants, the centrifugal charging pumps perform the HPSI function and are-capable of prmviding RCS makeup at the PORV setpoint pressure. These pumps inject into the cold legs via a boron injection tank. Plants in this category are:
Beaver Valley I and 2 Shearon Harris 1 Farley I and 2 Summer North Anna I and 2 Surry 1 and 2
-The remaining 3-loop plants (Robinson, San Onofre 1, and Turkey Point 3 and 4) have separate HPSI pumps and positive displacement charging pumps. The shutoff head of the HPSI pumps is about 1700 psig, therefore, these pumps are unable to provide makeup to the RCS at normal operating pressure. In the event of a small LOCA which leaves the RCS at high pressure, it is necessary to first depressurize the RCS before the HPSI pumps can provide makeup. Depressurization can be accomplished by heat transfer from the RCS to the steam generators or by opening the power-operated relief valves on the pressurizer.
Two different low-pressure ECCS subsystems are found in the 3-loop plants based on containment design. In the plants with a large, dry containment, the LPSI function is performed by the RHR system. In some of the subatmospheric containment
.plants, the low pressure injection/recirculation system is separate from the RHR system and does not include heat exchangers in the flow path to the RCS.
The centrifugal charging pumps, and the separate HPSI pumps, cannot be aligned to take a suction directly on the containment sump. If needed, the low-pressure ECCS subsystem can be aligned in tandem with the high-pressure ECCS subsystem for high-pressure recirculation.
4-25 3/90
4.3.7 Containment Westinghouse 3-loop plants have either large, dry containments or subatmospheric containments of various designs, as described below.
A.
Large, Dry Containment Eight of fourteen 3-loop plants have large, dry containments. The types of construction used in these containments is summarized below.
Containment Construction Aniiffeable Plants Steel sphere with concrete San Onofre 1 enclosure building Reinforced concrete cylinder Shearon Harris 1 with a steel liner One-dimiension (vertical)
Robinson post-tensioned concrete cylinder with a steel liner Three-dimension post-Farley 1 and 2 tensioned concrete cylinder Summer with a steel liner Turkey Point 3 and 4 Examples of large, dry con'tainments for 3-loop Westinghouse plants are shown in Figures 4.3-2 and 4.3-3 (Robinson) and Figures 4.3-4 and 4.3-5 (Summer).
B.
Subatmospheric Containment The following six of the fourteen 3-1o0p plants have Subatmospheric cylindrical containments constructed of reinforced concrete with a steel liner.
Beaver Valley I and 2 North Anna I and 2 Surry land2 Only one otherPWR in the U.S. has a subatmospheric containment: Millstone 3, a 4-loop Westinghouse plant. The North Anna subatmospheric containment is shown in Figures 4.3-6 and 4.3-7.
4-26 3/90
-Figure 4.3-1.
Westinghouse 3-Loop NSSS 4-27 3/90
Figure 4.3-2. Section View of the H.B. Robinson Large, Dry Containment (1-D Post-Tensioned Concrete) 4-28 3/90
-FUEL TRAMOU~ YI'K FUEL KAtIb*G,1 WULDINUG
-1
-I KATw IM3D IL Ik' 0A C.A.TO 0
lvlC
'Iq$$L (1-D Post-T1ensLionr "4O2E o
FLOR 0
I im+
+*
t~~in E L 251s.
30, oo.
AC(1-DO LAtTnsoe pancrete) 4-29 3/90
Figure 4.3-4. Section Views of the Summer Large, Dry Containment (3-D Post-Tensioned Concrete) (Sheet 1 of 2) 4-30 3/90
I ______________________________________
III
'S
- It
ca,4&
- 4.
'I V."...
-- I t1p-,p.
I,,.-:,.
.4 e
- All, AT Figure 4.3-4. Section Views of the Summer Large, Dry Containment (3-D Post-Tensioned -Concrete) (Sheet 2 of 2) 4-31 3/90 E
lIi 21* iN i, li, Ll*
v=i i,* i,. irai*
ti I
-,*lot ft-t gDI*
Figure 4.3-5. Plan View of the Summer Large, Dry Containment (3-D Post-Tensioned Concrete) 4-32 3/90 L-
4-33
=-W vr KI ax P"Traft WMAd CL iff W.-C-C-18 Pftt WNP r-n 2W PrM Do Ta"
ý'Flgure,4.3-6,' Section Views bfthd North Arina Subatmospheric Containmerit"(Sheet I of 2)
LLC L LW-r W-J-1 MA
-Fil IM 4AS~
. P"Of so W=L rMq/
Kvý--O KAN 0.
=-Z
-#Ulm ILCI U T"-W r-ý MM ML CL Mr tL W-4r JW-4 *Ar A
ax O.-C-4.4 We ML.;RAM FF COW" SULD CL ny r-r-9L 39-r CLCI WATS UP-,*
CL W 4kM QCl w.TM UM L =*-ýZWO 3/90
...,..-t F
6 I
.. a1 b
I 2
-I 3
2 S
ýIIl
----I I
Ij q
C
- 1
_'-IIE II/
)JJ]
( I
.1 Al II ii-
11' "A\\-Av bt
-N ii;
NI 15 a.
- lit,
-IN 1i L1 "W,.
I!
Ikv!
-I
(AFRIH Id IF5 Sld FL
d ie
.,q It G
'a
.-;I MT II 7
I!
4 I
J-1 Ilix (S
C;)
CL (a
- 0.
0 ca 0o coCD
. 0 0 tca 4
I I
m pu Figure 4.3-7. Plan View of the North Anna Subatmospheric Containment 4.35 3/90
4-Looo Westinghouse PWRs Westinghouse 4-loop plants in the United States include the following:
Braidwood 1 and 2 Byron 1 and 2 Callaway Catawba I and 2 Comanche Peak I and 2 D.C. Cook I and 2 Diablo Canyon I and 2 Haddam Neck Indian Point 2 and 3 McGuire 1 and 2 Millstone 3 Salem land2 Seabrook South Texas I and 2 Trojan Vogtle 1 and 2 Watts Bar 1 and 2 Yankee-Rowe Zionland2 All of these plants hadfull power operating licenses as of 2/89 except for Comanche Peak 1 and 2, Seabrook, South Texas 2, Vogtle 2, and Watts Bar 1 and 2. The Westinghouse 4 loop NSSS is shown in Figure 4.4-1.
4.4.1 Reactor Core and Fuel Assemblies The 4-loop Westinghouse plants can be grouped into the following categories based on core thermal output:
600 MWt 1825 MWt 2758 to 3817 MWt Yankee-Rowe Haddam Neck All other 4-loop plants The Yankee-Rowe core is -comprised of 76 unique 9 x (6 x 6) fuel assemblies and uses cruciform control rods as shown inFigure 4.4-2. The majority of the other 4-loop plants use 17 x 17 fuel assemblies, with a few still using 15 x 15 fuel assemblies. The Haddam Neck core has 157 fuel assemblies and all others have 193 fuel assemblies. The general arrangement of a 193 fuel assembly core is shown in Figure 4.4-3.
The average power densities of the earlier 4-loop plants (i.e., Yankee-Rowe, Haddam Neck, and Indian Point 2 and 3) are in the range from 82 to 93 kW/liter. All of the later plants using the 17 x 17 fuel elements have power densities in the range from 98 to 109 kW/liter.
4.4.2 Reactivity Control System The number of full-length and part-length RCCAs Westinghouse plants are summarized below:
for selected 4-loop ELant*
Haddam Neck Indian Point 3 McGuire 1 and 2 Seabrook 1 Fuel Elemesnt 157
- 193 193 193 Full-Length RCCAs 45 53 53 58 Part-Length RCCAs 0
0 8
0 A more complete listing is provided in Section 4.5.
Yankee-Rowe is the only Westinghouse plant to use cruciform control rods.
There are 24 cruciform control rods operated by magnetic jack CRDMs plus 8 cruciform fixed "shim" rods. The position of the shim rods can be changed only during refueling.
4-36 4.4 L_
3/90
4.4.3 Reactor Coolant System The general configuration of a 4-loop Westinghouse reactor coolant system is shown in Figure 4.4-1.
Most 4-loop plants have a reactor vessel with a 173 inch inside diameter. The only exceptions are the following:
109 inch vessel Yankee-Rowe 154 inch vessel Haddam Neck 167 inch vessel Comanche Peak 4.4.4 Steam Generators All vintages of steam generators can be found among the 4-1oop Westinghouse plants, as follows:
27-series Yankee-Rowe 44-series Indian Point 2 and 3 S-51-series Most other 4-loop plants Model F Seabrook, Wolf Creek, Callaway (late-model 4-loop plants)
A comparison of Westinghouse steam generator design parameters is included in Section 4.1.
4.4.5 Shutdown Cooling -Systems In most of the 4-loop Westinghouse plants, shutdown cooling is accomplished by the Residual Heat Removal (RHR) system which also operates in the Low-Pressure Safety Injection (LPS1) mode as part of the ECCS.
4.4.6 Emergency Core Cooling Systeims Almost all 4-loop Westinghouse plants have two high-pressure ECCS subsystems; the high-pressure safety injection (HPSI) system and the centrifugal charging pumps which are part of the ECCS: In these plants, the shutoff head of the HPSI pumps typically-is on the order of 1600 to 1700 psig. Each centrifugal charging pump pump can provide approximately 150 gpm makeup to the'RCS at the PORV setpoint pressure.
The only exceptions to this high-pressure injection capability are the Yankee Rowe and Indian Point 2 and 3 plants. These plants have an intermediate-piessure HPSI system'and low-capacity positive displaement charging pumps which" are not part of the ECCS. In'the event of a small LOCA which leaves the 'RCS at high pressure, it is necessary to'first depressurize the RCS before-the HPSI pumps can provide makeup.
Depressurization can be accomplished by heat transfer from the RCS to the steam generators or by opening the power-operated relief valves-on the pressurizer.
4.4.7 Containment Westinghouse 4-loop plants have either -large, 'dry containments, a subatmospheric containment, or ice condenser containments of various designs, as described below. The ice condenser containment is unique to Westinghouse 4-loop PWRs.
A.
Large,` Dry Containment Twenty-four of thirty-five 4-loop plants have large, dry containments. The types of construction used in these containments is summarized below.
4-37 3/90
Containment Construction Bare steel sphere Yankee-Rowe Reinforced concrete cylinder with a steel liner Reinforced concrete cylinder with a steel steel liner and secondary containment Three-dimension post tensioned concrete cylinder with a steel liner Comanche Peak 1 and 2 Diablo Canyon 1 and 2 Haddam Neck Indian Point 2 and 3 Salem I and 2 Seabrook 1 Braidwood I and 2 Byron I and 2 Callaway South Texas 1 and 2 Trojan Vogtle 1 and 2 Wolf Creek Zion 1 and 2 The Yankee-Rowe steel sphere large, dry containment is shown in Figures 4.4 4 and 4.4-5. A reinforced concrete large, dry containment (Diablo Canyon) is shown in Figures 4.4-6 and' 4.4-7. The South Texas three-dimension post tensioned concrete large, dry containment is shown in Figures 4.4-8 and 4.4-9.
B.
Subatmospheric Containment Millstone 3 is the only' 4-loop Westinghouse PWR with a subatmospheric containment. The Millstone 3 containment is'shown in Figure 4.4-10.
Construction is of reinforced concrete cylinder With i steel liner and a secondary containment. All other subatmospheric containments in the U.S are f6und in Westinghouse 3-loop plants.
C.
Ice Condenser (Pressure Suppression) Containment Ice condenser containments are unique to 4-loop Westinghouse PWRs, and ten of the thirty-five 4-loop plants have this type of containments. The types of construction used in these containments is summarized below.
Containment Cofistruction Steel cylinder with concrete shield building Reinfoced concrete cylinder with steel liner Applicable Plants
_Catawba land 2 Sequoyah 1 and 2 Watts Bar 1 and 2 D.C. Cook 1 and 2 McGuire 1 and 2 The Catawba ice condenser containment is shown in Figures 4.4-11 and 4.4 12, and the Watts Bar containment is shown in Figures 4.4-13 to 4.4-15.
Additional details on the configuration of the ice condenser units can be found in Section 3.
3/90 L_
Annlicable Plants 4-38
. Figure 4.4-1.ý Westinghouse 4-Loop NSSS 4-39 3/90
Figure 4.4-2. General Arrangement of the Yankee-Rowe Core 4-40 3/90 I
1800 I00 riDe~r rfnr bri:lf fAfn Ar REGION I ONCE OR TWICE BURNED FUEL REGION 2 ONCE OR TWICE BURNED FUEL SR*EGION 3 FRESH FUEL Figure 4.4-3.'- General'Arrangement ýof a 193 Fuel Assembly Core 4-41 3/90
Figure 4.4-4.-
Section View ofthe Yankee-Rowe Large, Dry-,
Containment (Steel Sphere)
"4-42 3/90
-1 66 Figure 4.4-5.
Plan View of the Yankee-Rowe Large, Dry
-Containment (Steel Sphere) 4-43 3/90
Figure 4.4-6. Section Views of the Diablo Canyon Large, Dry Containment (Reinforced Concrete) 4-44 3/90
I' Figure 4.4-7. Plan View of, the.-Diablo Canyon Large, Dry Cbntainment (Reinforced Concrete)
-'4-45 3/90
L~'
v 1
.wft I,
I mu.V
'I 'I I
N'
(
Ii
-ff Figure 4.4-8.
Section View of South. Texas Large, Dry Containment
- (3-D Post-Tensi0ned Concrete) 4-46 3/90
-t 7
t.
I B
FF q
fO WAN" I
E
i I
Figure 4.4-9.
Plan View of South Texas Large, Dry Containment
'(3.-D Post-Tensionedd Concrete)
"447 3190
l*'f:".*i*
.--- Pessuirzer heater
-.I rmovalspace El. 3 It 8 (n.
- ore l
-- m
-"--Pressurizer tank Figure 4.4-10.
Section View of the Millstone 3 Subatmospheric Containment 4-48+.
- .*+++.
3./90:.*
+
Containment (typical)
I
+ 4-48 3/90
4.5 Westinyhouse PWR Comparative Data This section contains the following tables which present comparative system data for Westinghouse PWRs:
Table 4.5-1 Table 4.5-2 Table 4.5-3 Table 4.5-4 Table 4.5-5 Table 4.5-6 Table 4.5-7 Design Parameters for Representative Westinghouse PWRs Comparison of Westinghouse PWR Vessel and Core Parameters Westinghouse PWR System Comparison - RCS, AFW, Charging and HPSI Comparison of Westinghouse PWR Pressurizer Relief Capacity Comparison of Westinghouse PWR Containments Comparison of Westinghouse PWR Backup Electric Power Systems Comparison of Westinghouse PWR Power Conversion Systems 3/90 4-55
Table 4.5-1.
Design Parameters for Representative Westinghouse PWRs REACTOR PLANT WESTINGHOUSE WESTINGHOUSE WESTINGHOUSE WESTINGHOLSE CHARACTERISTICS 2.LOOP PLANT 3.LOOP PLANT EARLY 4-LOOP PLANT LATE 4-LOOP PLANT Overall GINNA Il.B. ROBINSON TROJAN SOUTiH TEXAS I AND 2 Number of loops 2 without isolation 3 without isolation 4 without isolation 4 without isolation TbmW capacity 1520 MWt 2308 MWt 3411 MWI 3817 MWt Elwiccapacity 470 MWe 665 MWe 1095 MWe I a 1250 MWe Efficiency (net) 30.92%
29.81%
32.10%
32.75%
Coolant pressure in primary circuit at exit fron reactor 2235 psg 2235 psig 2235 psig 2235 psig Coolant ttemperate atinlet 551.9 OF 546 2*F 552.5 "F 560°F Coolant tanpaatum at exit from reactor 634F 642*F 620 °F 628 8°F Coolant flow rate through reacton (total) 66.7136 lb/lw 101.5E6 lb/hr 132.7116 lb/hr 1.39137 lb/l Core Ileight of active con 12 ft.
12 ft.
11.98 ft.
14 ft.
Equivakit dameta" 804 ft.
9.96 ft.
11.06 ft.
11.1 ft.
Number of fuel asamblies 121 157 193 193 Nunber of control rods assemblies 29 53 53 57 Number of fuel elements.n assembly 179 204 264 264 Dianeter of fuel lement 0.3669 in.
0.3659 in.
0.3225 in.
0.374 in.
Area of eat uransfa surface 28,715 sq ft.
42,460 sq.fL.
59,700 sq ft.
69.700 sq.ft.
Mean specific heat flux 150,500 Btu/hi-sq ft.
171,600 Btwfi-sq ft.
Unk.
181,200 Btu/hr-sq ft.
Nunber of fuel rods 21.659 32,028 50,913 50,913 Cor loading 3 region non-undorm 3 region non-uniform 3 region non-unuform 3 region non-uniform Average bunup (firs cycle)
-14.126 MWd/MTN 13.000 MWd/MTh Unk.
15.000 MWd/MTIu Fuel weight (as U02) 120.130 lbs.
175.400 lbs.
222,739 lbs.
259.860 lbs tji C7%
0
Table 4.5-1.
Design 1P6rameters for Representative Westinghouse PWRs (Continued)
REACTOR PLANT WESTINGHOUSE WESTINGHOUSE WESTINGHOUSE WES IGROUSE CHARACTERISTICS 2.LOOPPLANT, 3.LOOP PLANT EARLY 4.LOOP PLANT LATE 4-LOOP PLANT GINNA H.B. ROBINSON TROJAN SOUTH TEXAS 1 AND 2 Reactor Vestsl Ves-PAe-ieght 39.'11 ft.
41.5 fL 43.83 ft.
43.75 ft ine diineter Ift.
12.96 ft.
13.92 ft.
14.4 ft Number of openings for Inlet md outlet Nozzles 2x2 203 2x4 2x4 Nrnbe oftmits 2
3 4
4 lThwml p0eper unit 650MWt 769.3 MWt Unk.
954 MWt Shell side operating ipause (steSim)
-.9 iu 1005 l ig 1073 psig 1073 psi Tube side openting presur e 2235 psig 2235 pslg 2235 psi5 2235 psi Tube side design flow 33.63 E6 lb/hr 33.93 E6 lb7lw Unk.
Unk.
otleg 9
29in.
29 in.
Cold leg hum dia.
27.5 in.
27.5 in.
27.5 in.
27.5 in.
Be wno -pmsmd Iteam rnermtor 31 In.
31 In.
31in.
32In.
safuty valves First opening presmre 2485 psig 2485 pug 2485 psi5 2485 psi, Capadcty 19,500 Ib/r each do 2 139,300 b/r each of 3 123,100 Ilb/t each of 2 504.953 lb/hr each of 3 117.3 1bf/hMWt each PORV of 2 95.5 lb/lW/MWt each PORVof2 61.6 1bJfr/MWt each PORW c2 210,000 lb/h each PORV of 3 Primay "ele t Pumps Number 2
3
,4 4 4 Pw capacity
- 90.
g."",
88.500 g 88.500- m.
102.500 gpm Coolat temperstuee 55.1 PF 546.5OF
- 552.59- -
650' Presuweflse 252 fL head 261 ft. head 277 ft. head Unk.
Design pressure 24851pig' 2485 psig 2485 psig 2485 psig Design tepersture 650 OF 650 OF 650 OF 650 OF Motor rating (anmeplate) 5,500 hp.
6.000 hp.
6.000 hp.
6,000 hp.
f
'-,4
-4
Table 4.5.2.
Comparison of Westinghouse PWR Vessel and Core Parameters PWR PWR Reactor Plant Core Reactor Core Core Core Average Number Fuel Number of Vendor Type Name Power Vessel Equivalent Active Power Density of Fuel Element Control Rods (MWt) 1.0.
(In) Olm.
(in) Height (in)
(kWlllter)
Elements Geometry (FulIlPart Length)
W 2-loop Ginma 1520 132 96.5 144.
89.00 121 14 x 14 29 F/4 P W
2-loop Point Beach I &2 1518 132 96.5 144 87.00 121 14 x 14 37 (total)
W 2-loop Kewsuee 1650 132 96.5 144 94.90 121 14 x 14 33 (total)
W 2-loop Prairie Island 1 & 2 1650
- 132
-96.5 144 95.90 121 14 x 14 29 F/4 P W
3-loop SanOnore l 1347 144 119.5 144 70.40 157 14 x 14 45 (total)
W 3-loop H.B. Robinson2 2200 156 119.5 144 82.60 157 15 x 15 48 F/S P W
3-loop Swrty I & 2 2441 159 119.5 144 92.00 157 15 x 15 48 F/5 P W
3-loop Turkey Point 3 & 4 2208 172 119.5 144 82.80 157 15 x 15 48 F/S P W
3-loop Beaver Valey 1 2660 157 119.5 143.7 100.00 157 17 x 17 48 F/5 P W,
3-loop Beaver Valley 2 2660 157 119.5 143.7 100.00 157 17 x 17 48 F/S P W
3-loop Farley 1 &2 2652 157 119.5 144 101.10 157 17 x 17 45 (total)
W 3-loop North Anna I &2 2775 157 119.5 144 108.70 157 17 x 17 48 (total)
W 3-loop Shearon Harris 1 2785 157 119.5 144 105.00 157 17 x 17 52 (total)
W 3-loop Summer' 2785 172 119.5 144 104.50 157 17 x 17 48 (total)
W 4-loop YdkRoweR 600 109 75.7 91.9 89.3 to 90.1 76 9 x (6 x 6) 24 (total)
W 4-loop HoddamNeck 1825 154 119.6 121.8 82.00 157 15 x 15 45 (total)
W 4-loop Braidwood I & 2 3411 173 132.7 143.7 104.50 193 17 x 17 53 (total)
W 4-loop Byron I & 2 3411 173 132.7 143.7 104.50 193 17 x 17 53 (total)
W 4-loop Calaway 3425 173 132.7 143.7 109.20 193 17 x17 53 (total)
W 4-loop Catawba I &2 3411 173 132.7 143.7 103.5 to 44.6 193 17 x 17 53 (total) 00 0
I
Table 4'.5-2.
Compiarison of. Westingh`ouse PWR Vessel and Core Parameters'(Conttinued) 4
+
4
+
r-1
I
eacior
.urw Reactor Plant
,,, Name I
PWR Vendor W
'W W
W "W
W W
W W
W W.
W W
"W*
W W
Core Power 3425 3250 3338 2758 3025 3425 3411
- 3339 3411
- 3411, 3817' 3411' 3411'
- 3411, 3411, 3250 Comanche Peak I & 2 D. C. Cook I &2, Diablo Canyon 1 &2 Indian Point 2 Indian Point 3 Millstone 3 Salem I & 2 Seabrook I Sequoyah I & 2 South Texas 1& 2 Trojan,.
Vogue 1'&2-.
Watt Bar 1 &2 Wolf Creek Zion 1 &2' Reactor vessel I.D." lin1 Equivalent Diem.
(In)
- 132.7 132.7 132.7 3/4 132.7
",1'32.7 132.7
.132.7 132.7 132.7 132.7 132.7 132.7
- 132.7 132.7 132.7 132.7 I
-I Active Height (In) 143.7 14:.7 143.7 144 144 143.7 143.7 143.7 143.7 143.7 S16 143.7 143.7 143.7 143.7 144 PWR Type 4.loop 4-loop 4-loop 4-loop 4-loop 4-loop 4-loop 4-1oop0 4-loop 4-loop':
4-loop 4-loop 4-1oop 4-1oop S
I Number I
of Fuel
)
Fuel
- umo*r I Number I
I I
41i 167 173 173 173 "173
'173 173',
173 173 173 173 173 173
,173 173 173
-I W
Core Average.
Power Density (kW/llter)
'103.3 to'104.5 980 to 103.8 102.3 to 104.5 85.00 92.70 103.50 104.50 102.60 104.50 103.50 105 (est.)
105.50 104.50.
103.50 101.90 100.00 of Fuel Elements 193 193 193 193 193 193 193 193' 193 193 193 193 193 193 193 193 Element Geometry 17 x 17 17 x 17 17 x 17 15 x 15
.15 x 15 17 x 17 17 x17 17 x 17 17 x 17 17 x 17 17 x 17 17 x 17
.17 x 17 17 x 17 17 x 17
'15 x 15
- +
'II Number of Control Rod' (Full/Part -Length 53 (total) 53 (total) 53 F/8 P 53 F/8 P 53 F18 P 53 (total) 53 F/8 P (total) 53 (total) 53 F/8 P 57 (total) 53 F/8 P 57 (total) 53 F/8 P 57 (total) 57 (total)
I
Table 4.5-3.
Westinghouse PWR System Comparison
- RCS, AFW, Charging and JIPSI Reactor Coolant System Auxlliary Feodweler System Charging System High Pressure lnjction System NSSS Core'I tiCS MftiCS S/I I AFW Type Capacity I
Type Capacity Capacity i
1 Type Capacity CpacIy Plant Name Vendor MW Loops PORV/SV Model PUMPS Drive glm 0 pala PUmp* Pump gPM @ psrg oIR @ POIl Pumps Pump i
POFy Nate.
Beaver Valley I 2
W 2660 3
313 51 2
M 350 @ 1169 3
Cat 150 @ 2514 150 Same as cenl. charging pumps (a)
I T
7002 1169 Sam acadhrggpm Braldwood I & 2 W
3411 4
2/3 51 1
M 890 @ 1452 2
Cent 150 @ 2526 150 2
Cent 400 t106 0
(a)
' I O
840 @ 1452 1
PD 98 98 Byron I & 2 W
3411 4
213 51 1
M 890 @ 1452 2
Cart 150 @ 2526 150 2
Cent 400 1106 0
(a)
-aI 0
840 @ 1452 1
PO 9g8 98 Callaway tW 3425 4
213
-F
,2 M
6000 1387 2
Coar t50 50 d2.
Cent 42501162 0
(a)
I T
1200 0 1387 I
P0 98 98 Ca1awbeI&2 W
341 4
3/3 51 2
M S0 001392
,2 Cent 15002800 150 2,
Cent 400 1750 0
(a)
,I T
1000 1395 1
P0 98 98 Comanche Peak 1 &2 W
3425 4
V-3 F
-2 M
470 @ 1107 2
Cart unk.
unk.
2 Cent un0 1
T 9000 1107 1
PRI 0
(a)
DC Cook I W
3250 4
.313 51 I
M 450 0 1177 2
Cori 150 @ 2800 150 2
Cent 400 1700 0
(a) (b)
I T
9000 1177 1
po 98 98 D.C. Cook2 W
3250 4
213 5tF 1
M 4500 1177 2
Cart 150 @ 2800 150 2
Cent 400 @ 1700 0
(a) (b)
SI T
900 @ 1177 1
RI 98 g8 Diablo Canyon I & 2 W
3338 4
313 UT 2
M 440 0 1300 2
Card 15002514 10 2
Cent 425 @ 1084 0
I___
I T
880 I unk I
RD 98 @ 2514 198 1
Farley I & 2 W
2652 3
213 61 2
M 350 0 1214 3
Cort 15002800 150 Same s cent charging pumps (a)
I U
70001214 (a)
Ginna W
1520 2
212 44 2
M 2000 1344 3
I:D 0
60 3
Cent 300 @ 1170 0
(a) (c)
SI T
4000 1344 2
M 2000 1080 Hadna Nock W
1825 4
213 27 2
T 4509 1000 2
Cart 360 0 2300 350 2
Cent 9700 1750 0
(a)
RI 30 30 Irdlan P oint*2 W
2758 4
213 44F 2
M 1400 0 1350 3
PID 98 98 3
Cent 4000 1180 0
I T
8000 1350 indian Point 3 W
3025 4
2/3 44F 2
U M 400 0 1350 3
RI) 98 98 3
Cent 40001080 0
(a)
T
-8000 1350 Can 40__0
(
Kewaunee.
W 1650 2
.212 51 2
M 2400 1235 3
IO 50 60 2
Cent 700 @ 1082 0
(a)
I T
2400 1235 1 -
I McGuire I &2 W
3425 4
313 51 2
M 4500 1855 2
Cart 15002514 150 2
Cent 40001106 0
(a)
I I
'T 9000 1730 I1 PO 55 55 Wisdlone 3 W
3411 4
213 51 2
M 575 6 1290 3
Cart 150 0 2800 150 2
Cent 4250 1500 0
(a)
I T
11500 1290 NorthAnna IA2 W
2775 3
213 5IF 2
M 3500 1214 3
Cant 15002500 150 Sam*
cent. charging pumps (a)
I T
700 @ 1214 (a)_
Point Beach I £ 2 W
1518 2,
2/2 44F I
2 200 0 1192 3
PO 60.5 60.5 2
Cent 700 1750 0
(a) (b)
I I
T 4000 1192 1
1 Pratlesl Iand & 2 W
1650 2
212 51 1
M 200 0 1200 3
PD 50.5 50 5 2
Cant 700 1082 0
(a) (d) 51 1
T 2000) 1200 Rl-i-in W
2200 3
213 46F 2
M 300 0 1300 3
PD 77 77 3
Cent 375 0 1750 0
(a)
I T
6OW* @I130 Salem I & 2 W
3338 4
213 51 2
M 4400 1300 2
Cart 1500 2800 ISO 2
Cent unk.
0 (a)
I_ I T
8800 1550 1
PO go 98 I
I SanOnolre I W
1347 3
212 27 1 I M
2350 1035 2
Cent 002400 0
Same ecent. charging pumps (a)
I I I T
300 @ I110 Seabrook W
3411 4
213 F
I M
710 a 1322 2
Cent 1500 2800 150 2
Cent 425 1750 0
(a)
II I
T 7100e 1322 1
RI 8
98 I
I 00 0
Table 4.5-3. Westinghouse PWR System Comparison - RCS, AFW, Charging and HPSI (Continued)
._--
r'
-It Pressure Ineit onn Svlter NS Core R
AFW Type Capacity 9
Type Capacity Capacity Type Capacity Capacity Plant Name Vendor MWt tI.
PORpVISV Modal PUMpS Drive OPM 0 psa9 psPump gpm @psio gpm! 2PORV Pumps Pump I gpm @ polg J ppm @ PORV Notes Sequoyah 1&2 W'
3411 4
213 51
,2 M
440 0 1257 2
Caor 150 @ 2514 150 2
Cent 425 @ 1084 0
31 T
8800 1257 1
PO 55 a
55anhrgn Shearon Harris I
W 2785 3
313 UT 2
M 400 @ 1265 3
Cent 150 @2514 150 Same as cent. charging pumps 1
,T 900
- 1265, SouthTeas &2, W
3817 4
213 F,
3ý M
540 1435 2
Caor 160Q2513 10 3
Cent,,8000@1235 0
11 T
5400 1435 1
PD 135 35 Summer
,W 2785 3
unk.
UT 2
M 4000 1211 3
Carn 150 @ u 150 Same as cant. charging pumps I
T 10100 1211 Surry I & 2 W
2441 3
213 51F 2
M 3500 1183 3
Cort 150 0 2485 150 Same as cent. charging pumpS (a)
I I
T 700M 1183
__I Trolan W
3411
.4 213 S1,
- 1.
T 960 0 1474 2
Carl 150@2800 150 2
Cent 425 @ 1700 0
(a)
I 1
0 9600 1474 1
PD gg8 98I Turkey Point 3 &4 W
2208 3
213 44F 2
T 600 1203 3
FO 77 77 2
Cent 300 @ 1750 0
(a)
Vogolo I &2 W
3411 4
212 51 2
M 6300 1517 2
Cars 15002514 150 2
Cent 425@ 1162 0
(a) 1 T
11170 1517 1
POD 8 0 unk.
g8 I
Waits Bar I & 2 W
3411 4
,213 51 2i M
470 1600 2
Carl 15002514 150 2
Cent unk.
0 (a)
'I T
940 0 1600 1
IP 9803200 98 1
Woo Creek W
3411 4
213 F
2 M
800 0 1387 2
Cent 15002514 150 2
Cent 4250 1161 0
(a)
,1 T
12000 1387 1
IPO 9802514 g8 Yankeenow" W
600 4
-112 27, 1
T 8001200 3
Carl 33 33 3
Cent 187 Q 650 0
(a) (s)
ZionI &2 W
3250 4
213 51 2
M 4500 1343 2
Cora 15002800 150 2
Cent 400 0 1084 0
(a)
I T
9000 1343 1
1P3 98 go J.,
I-A 61 00 I Iwll.* w.......
... IF--.......
F......
AuxiiaIry tFoowater S)ystem Chanrging ayeom
Table 4.5-3.
Westinghouse PWR System Comparison - RCS, AFW, Charging and HPSI (Continued)
General Note:
All pump capacities are stated on a per-pump basis, AFW pump capacity is stated in terms of rated capacity. Charging and high pressure injection pump capacity is stated in terms of rated capacity and approximate capacity when RCS pressure is at the PORV setpoint (i.e. for "feed and bleed" operation).
Codes used in this table include:
Type drive:
Type pump:
RCS PORV/SV:
S/G model:
0,'
M = electric motor T = steam turbine PD = positive displacement Cent = centrifugal number of RCS power-operated valves (first number) and safety valves (second number).
UT = U-tube (see note (a))
OT = once-through Notes:
(a) The Westinghouse small inventory steam generators (series 27 and 44) require twice the feedwater flo-w to prevent dryout as compared to the later versions (series 51 and F). The exception is Yankee-Rowe (ref. NUREG/CR-3713, Section 3).
(b). At Point Beach, and D.C. Cook, the motor-driven AFW pump in each unit can feed steam generators in both units.
(c) Ginna has a main AFW system with two motor-driven and one turbine-driven pump as well as a standby AFW system which has tow motor-driven pumps located in a separate area.
I (d) At Prairie Island, the motor driven AFW pump at each unit normally supplies the opposite unit.
(e) At Yankee Rowe; charging and SIS provide backup foriAFW.
0
Table 4.5-4.
Comparison or Westinghouse PWR Pressurizer Relief Capacity NSSS
- RCS Manufacturer Capacity Lowest
- RCS Capacity Lowest Plant Name Vendor PORV's (Iblhr/MWt) Setpolnt (psig) SV's (Klb/hr) SetpoInt (psig)
Beaver Valley 1 & 2 W,
3 Masoneilan 79.9 2335 3
345 2485 38-20771 Braldwood I & 2 W
2 Unk.
210" 2335 3
420 2485 Byron 1 & 2 W
2, Unk.
210*
2335 3
420 2485 Callaway W
2' Unk.
210*
2335 3
420 2485 Catawba I & 2 W
3 Unk.
210" 2485 3
- 420, Unk.
ComanchePeakl&2 W
21,1, Unk.
210*
2185 3
420; 2485 D.C. Cook 1 W
3 Masoneilan 64.6 2335 3
129.2 2485 D.C. Cook2 W
2,
'38-20721 61.8 2335 3
123.5 2485 Diablo Canyon 1 & 2 W,
3 Unk.
Unk.
Unk.
3 Unk.
Unk.
Farley 1 & 2 W
2 Copes-Vulcan 79.2 2335 3
130.1 2485 D-100-160 Ginna W
2 Copes-Vulcan 117.8 2335 2
189.5 2485 1 D-100-160 I_
I II I
Haddam Neock W
2 Copes-Vulcan 115.1 2270 3
160.7 2485 D-100-160 0_
Indian Point 2 W
2, Copes-Vulcan 78.7 2335 3
147.9 2485 I
D-100-160 Indian Point 3 W'
2, Copes-Vulcan 78,.7, 2335 3,
147.9 2485
_1D-I00-160 1
,,_1 Kewaunee W
2 Copes-Vulcan 106 2335 2
209.1 2485 D.100-160 0*'
Table 4.5-4.
Comparison of Westinghouse PWR Pressurizer Relief Capacity (Continued)
- RCS Manufacturer Capacity Lowest
- RCS Capacity Lowest Plant Name Vendor PORV's (Ib/hr/MWt) Setpolnt (psig) SV's (KIb/hr) Setpolnt (pslg)
McGuire 1 & 2 W
3 Unk.
210*
2335 3
420 2485 Millstone 3 W
2 Unk.
210' 2335 3
420 2485 North Anna 1 & 2 W
2 Masonelian 76 2335 3
137 2485 38-20721 Point Beach 1 & 2 W
2 Copes-Vulcan 117.9 2335 2
189.7 2485 D-100.160 Prairie Island 1 & 2 W
2 Copes-Vulcan 5
2335 2
209.1 2485 D-100-160 Robinson W
2 Copes-Vulcan 95.5 2335 3
130.9 2485 D-100-160 Salem I & 2 W
2 Copes-Vulcan 63 2350 3
125.8 2485 D-100-160 San Onofre I W
2 ACF Industries 80 2190 2
178.2 2500 70-18-9 DRTX_
Seabrook W
2 Unk.
Unk.
Unk.
3 Unk.
Unk.
Sequoyah I & 2 W
N/A Unk.
Unk.
Unk.
Unk.
Unk.
Unk.
Shearon Harris 1
,W 3
Unk.
210*
2335 3
,380 2485 South Texas 1 &2 W
2 Unk.
210*
2485 3
505 2485 Summer W
Unk.
Unk.
Unk.
Unk.
Unk.
Unk.
Unk.
Surry 1 & 2 W
2 Copes-Vulcan 86 2335 3
120.5 2360 IA58RGP Trojan W
2 Copes-Vulcan 61.6 2350 3
123.1 2485 I D-100-160
Table 4.5-4.
Comparison of Westinghouse PWR Pressurizer Relief Capacity (Continued)
NSSS.RCS Manufacturer Capacity
- Lowest,
- RCS Capacity Lowest Plant Name Vendor PORV's (IblhrlMWt) Setiolnt' (psig) SV's (Klb/hr) Setpolnt (psig)
Turkey Point 3 & 4 W
2 Copes.Vulacn 95.1 2335,_
3 132.8 2485 5-131642 Vogtlel&2" W
2 Unk.
210' 2235 2-420.
2485 Watts Bar 1 & 2 Wý 2
Unk.
Unk.
Unk.
3 Unk.
Unk.
Wolf Creek W
'2 N/A 210" 2235 3
420 2485 YankeeRo W-
- 1.
Dresser 118 2400 2
153 2485 31533 VX Zion 1 & 2 W
-,2...
Copes-Vulcan 764.6 2335
. 3 129.2 2485 0
D-100-160-,
- Klb/hr rating LA 4-4 I
Table 4.5-5.
Comparison of Westinghouse PWR Containments NSSS Arch./
Prim.
Concrete Internal Containment Design Design Enclosure Plant Name Vendor Engineer Cont.
Construction Construction Diameter Free Volume Pressure Leok Rats Building?
TYPO Subtype (feet)
(ft3)
(psal)
% vollday Beaver Valley 1 & 2 W
Stone A Sub Concrete Cylinder wl Reinforced 126 1.80E+06 54 0.1 No Webster Alm Steel Liner Braidwood I & 2 W
Sargent Dry Concrete Cylinder w/
3-D 140 2.90E+06 61 0.1 Nb
& Lundy Steel Liner Prestressed Byron 1 & 2 W
Sargent Dry Concrete Cylinder wI 3-D 140 2.90E+06 61 0.1 M
& Lundy Steel Liner Prestressed Calaway W
Betchel Dry Concrete Cylinder w/
3-1) 140 2.50E+06 60 0.1 Nb Steel Liner Prestressed Catawba I & 2 W
Duke ice Steel Cylnder Reinforced 115 1.22E+06 30 0.2 Yes Power Cond.
Comanche Peak 1 & 2 W
Gibbs &
Dry Concrete Cylinder w/
Reinforced 135 2.98E+06 50 0.1 tb Hill Steel Liner D.C. Cook I & 2 W
AEP Ice Concrete Cylinder wl Reinforced 115 unk.
12 0.25 tN Cord.
Steel Uner Diablo Canyon I & 2 W
Pac. Gas Dry Concrete Cylinder wI
, Reinlorced 140 2.63E+06 47 0.1 Nb
& Elect Steel Liner Farley I & 2 W
Betchel Dry Concrete Cylinder w/
3-.0
,130 2.03E+06 54 0.3 ND Steel Urnor Prestressed Gkina W
Gilbert Dry Concrete Cylinder wl
, 1-D Vert.
105 9.97E+05 60 0.1 Nb Steel Uner Prestressed Ha&InamNedo W
Slone&
Dry Concrete Cylinder wl Reinforced 138 1.71E+06 40 0.1 Nb Webster Steel Uner.. -
Indian Point 2 W
LE*
Dry Concrete Cylinder wl Reinforced 135 2.61E+06 47 0.1 Nb S.....
Steel Uner Indian Point 3 W
L*
Dry Concrete Cylinder wl Reinforced 135 2.61E+06 47 0.1 lb S..
SteelUnr Line_
Kewaunee W
Pioneer Dry Steel Cylnder 108 unk:
46 0.5 Yes McGuire
& 2 W
Duke Ice Concrete Cylinder wl
-'Reinforced 115 unk.
28
- 0.2 N:
Power Cortdi
' Steel Uner Millstone 3 W
Stone &
Sub Concrete Cylinder wl Reinforced 140 1.03E+07 45 0.9 Yes I
I Webster Atm Steel Liner North Anna I & 2 W
Store &
Sub Concrete Cylinder w/
Reinforced 126 unk.
45 0.1 IN Webster Atm.
Steel Uner Point Beach 1 & 2 W
Betchel Dry Concrete Cylinder wl 3-D 105 unk.
60 0.4 tb Steel Unor Prestressed a%
Table 4.5-5.
Comparison of Westinghouse PWR Containments (Continued)
NSS8 Arch./
Prim.
Concrete Internal Containment Design Design Enclosure Plant Nime'M Vendor-. Engineer Cont.
Conetructlon, Construction Diameter Free Volume Pressure Leak Rate Building?
112Type Subtype (feet)
(ft3)
(plal)
% vollde, Prairie isJand I & 2 W.
Pioneer.
Dry.
.Steel Cylinder.-
105 unk.
- 41.
0.5 Yes Robinson 2 W W, Dry Concrete Cylinder wl t-D Vert.
130 2.1OE+06 42 0.1 N
Steel Uner Prestressed Salem 1 & 2 W
PSE&G, Dry Concrete Cylinder wil Reinforced 140 2.62E+06 47 0.1 Nb
_Steel Uner San Ofre, I W
Belchel Dry Steel Sphere 140 1.44E+06 47 0.5 Yes Seabrook I,,,,
W UL3 Dry Concrete Cylinder wl Reinforced 140 2.70E+06 65 0.5 Yes Steel Uner..
Sequoy.h1 & 2 W.
WA Ice Steel Cyinder 106 unk.
10.8 0.5 Yes Shearon Harris 1, W
Dry Concrete Cylinder w/
Reinforced 130 2.50E+06 45 0.3 NM S tee l U n e r SouthTem A
&2 W
Brown Dry Concrete Cylinder w/
3-D 150 3.30E+06 56 0.3 W
" I.I..
I I
Steel Liner
'Prestressed Summer W
Gilbert Dry Concrete Cylinder wl 3-D 126 unk.
55 0.2 INb "Steel Uner Prestressed Surry 1 & 2 W
Stone &
Sub Concrete Cylinder wl Reinforced 126 1.80E+06 60 0.1 Nb Webster Atm.
Steel Uner Trojan W
Betchel Dry Concrete Cylinder wl 3-D 124 2.00E+06 60 0.2 N,
"__ ___
_Steel Uner Prestressed Turkey Point 3 & 4 W
Betchel Dry Concrete Cylinder wl 3-D 116 1.55E+06 59 0.25 NM Steel Uner Prestressed Vogle I a 2 W
BeIcel Dry Concrete Cylinder wl 3-D 140 2.70E+06 52 0.1 N)
- I. -...
Steel U ner Prestressed Wat Bar 1 & 2---- -.
W TWA-Ice Steel Cylinder 115 unk.
0 5 Yes Cond Wolf Creek W
Betchel Dry Concrete Cylinder wl 3-D 135-2.50E+06 60 0.1 NW
____Steel Uner Prestressed I
YankeeRowe W
Stone&-
Dry BareSteelSphere,
.125 1.02E+06, 34-3 Nb W e b s t e r I
Zion 1 & 2 W
Sargent Dry Concrete Cylinder wl 3-D 141 2.86E+06 47 0.1 NW
_a Lundy'l Steel Urier Prestressed it 0%
.1
Table 4.5-6.
Comparison of Westinghouse PWR Backup Electric Power Systems Shared Dedicated Continuous
- of NSSS Diesels Diesels Rating Diesel Batteries Reactor Plant Vendor per Plant per Unit (k W)
Manufacturer per Plant Voltage Notes Beaver Valley 1 & 2 W
- None 2
2600 Gen. Motors 5 "
125
- Unit 1 4 e".Unit 2 (2 are diesel batts.)
Braldwood 1 & 2 -
-W -
Now 2
5500--
unk.
2 125 Byron I &2 W,
Non 2""
5500 unk.
2-125 Callaway W
None-,
2 6201 unk.
2 125 Catawba 1& 2 W
1 2
7000 unk.,
7" 125
- Diesel batt. -1 DG and 3 2'
125V DC batts for sale shtdwn Comanche Peak 1 & 2 W
Namn 2
-7000 unk.
2" 125 D.C.Cookl &2 W
None 2
3500" Worthington 4
250
'Unit I
__3600" Worthington
- Unit 2 Diablo Canyon I & 2 W
1 2
2600 A1oo 6
125 Farley 1 & 2 W
3 2600 Fairbanks-Morse 7
1 25 2
4600 Fairbanks-Morse Gira W
Noe 2
1950 AMoo 2
1 20
'480 VAC diesel generator HaddamNeck W
Now 2.-
2850 unk.
2 12 5, Indian Point 2 W
None 3
1750 Mee 4
1 25, IndIan Point 3.
W Nora 3
1 175.0 AIO 3
125 Kewaune W
Non 2
2850 Gen. Motors 2
125 McGulre I & 2 W
None 2
4000.
unk.
4 125 Millstone 3 W
Nora 2
4986 Fairbanks-Morse 4
1 25 North Anna I & 2 W
None 2
2750 Fairbanks-Morse 8
125 Point Beach I & 2 W,
2 2850 Gen. Motors 2
125 00 W
Table 4.5-6.
Comparison of Westinghouse PWR Backup Electric Power Systems (Continued),
Shared Dedicated Continuous of NSSS Diesels Diesels Rating 1Di'esel Batteries Reactor Plant Vendor per Plant, per Unit (kW)
Manufacturer per Plant Voltage Notes Prairle Island 1 & 2 W
2
-2850 Gen. Motors 4
125 (model 99.-20)
Robinson 2 W
Nona 2
2450 Falrbanks.Morse 2
,1 25
- Dedicated shutdown diesel 1 "
2500 Falrbanks-Morse generator Salem I & 2 W....
3 2600 Aoo 6
125 1
- 40000" 2
250 Gas turbine San Onofre 1 W
None 2
unk.
unk.
2 125 Seabrook 1 W
None 2
6083 Falrbanks-Morse 4
125 Sequoyah 1 & 2..
W Now 2
-3600 BruceGM unk.
unk.
Shearon Hardr 1
W None 2
6500 unk.
2 1 25 South Texas 1 & 2.
W Noam 3
5935 Cooper Energy unk.
unk.,
I -
____1_1 Services Summer.
W None 2
. 4250 Fairbanks-Morse unk.
unk.
Surry I & 2 W
1 1
2850 Gen.Motors 4
125 Trojan W
None
- 2.
4418 Gen. Motors 2
125 1
250 Turkey Point 3 & 4 W
2.._
2500 Schoonmaker GM 4
125 Vogtle I & 2 W
None 2"
7000 unk.
4 125 Watts Bar 1 & 2 W
None 2
4750 FaIrbanks-Morse unk.
unk.
Wolf Creek-W
- Now, 2
,6201 unk.
4
-125ý Yankee Rowe W
None 3.*
400 Gen. Motors 3
125 "480 VAC diesel generator,
Zion 1 & 2 W
1 2
4000 Cooper-Bessemer 5
125
.4*
%0
'0
Table 4.5-7.
Comparison of Westinghouse PWR Power Conversion Systems NSSS Arohlto Turbine Gen. Turbine Bypass Condenser Normal.
8 Main FW FW Pump Shutoll Head Capacity Plant Name Vendor Engineer Conatructor Cap.
(MW@)
Capability
(%)
Cooling Type Hoaa Sink Pumpa Drive Type (palg Lapml Beaver Vabley I & 2 W
Slone&
D-"-esne 833 85 Closed Iop N.L CoolingTower.
2 Ic
- unk, uni Webster Ligh Co Braidwood I & 2 W
Sargent &
Commonwealth 1120 40 Once Though Braidwood Lake 2 (50%)
turbine unk.
unk.
' Lun Edison 1 (50%
AC unk unk Byron I & 2 W
Sargent &
aommo iAth 1120 40 CIasedLoop NaL. CoolingTowers 2 (50%)
turbine unk unk.
Ltundy I.
Edison 1 (50%1 AC unk-unk Callsway W
Bechtel Daniel 1120 40 Cb0adLoop NaL. CoolingTower 2 (67%)
turbine unk.
unk.
Calawba 12 W
DukePower Duke Power 1129 100 CbtmedLoop Mech. CoolingTowers 2 (50%)
AC -
unk 18.400 Comanche PeakI&,2 W
Gibbs & Hi Brown & Root 10se unk.
Once Thrug
.Squw Cro Resivoi 2 (50%)
turbine 986 19,800 D.C.Cook I&2 W
AEP AS EP 1060 85 OrceTýoi Lake ichigan 2 (70%)
turbine 1.138 16.750 DiabloCanyon I &2 W
PacalcGasA PacdlicGas&
1087 40 OnceThrough PacdcOcean 2
turbine unk.
unk.
Electric Electric Farley I & 2 W
SCSdfeil*I Daniel 823 40 CimodLoop Mech. Cooling Towers 2
turbine 1.474 15.000
@5250 RPM Gum W
Gilbert Bechtel 470 40 Oncm ouo Lake Onlmo 2 (50%)
AC 1.166 14.000
@853 iega HaddmNeck W
Ste one Store&
62 40 Theriough Connecuikt River 2 (50%)
AC 1,100 9.600
__________________Webster Webster' Indian Poirn 2 W
- UEAC, Wedl 849 40 OnceThrough Hudaon River 2
turbine unk.
1.530
___1____70 p
sg i
IndMn Point 3 W
UE&AC Vhd.o 065 45 Once Through Hudson River 2
turbine unk.
1,530
""______@970 vsaa Keniuies W
FEI FEI.
503 40 OnceTmrough Lake Michigan 2 (60%)
AC 2.276 10,000 McGuve
& 2 W
DukePower Duke Power 1129 100 OnceThough LakeNorman 2
turbine unk.
18,000 Milatone 3 W
Stone&
Storme6 1142 40 OnceThrough Niantic Bay 2(50%)
turbine 1.235 19.650 Webster Webster 1 (50%)
AC unk unk NorthAnnal&2 W
Stone&
Stoema&
916 40 OnceThrough Cooling Lake 3 (50%)
AC 986 16.250 Webster Webster PoinBeach I &2 W
Bechtel..
Bechlel
-4865 40 OnceThough LakeMichigan 2 (50%)
AC 1.062 760
_....._0941 Dsa Prairie Island 1 A 2 W
FEI Northern 503 10 CoedLoop MochCooling Towers 2 (65%)
AG 1.050 unk States Power Robinson 2 W
Ebwoo EIb 665 40 Onh Through Lake Robinson 2
AC 1.040 12.690 Salem I & 2 W
Public Service UE&C 1106 40 OnceThough Delaware River 2 (50%)
turbine unk.
16.613 Electirc& Gas
@884 vsig SanOnolre I W
Bechtel Bechtel 436 10 Once lTrugh PacifcOcen 2 (50%)
AC 1.165 14.000 1__
1 1
1 0@853 rain Seabrook I W
UE&C UE&C 1150 50 OnceThough Alaric Ocen 2 (50%)
turbine unk.
17.200 to 1019 pesg Sequoyah I & 2 W
"NA "NA 1148 unk.
CombinedCycle Nat.CoolongTower unk.
unk.
unk.
unk I& Tennessee River
-P
Table 4.5.7.
Comparison of Westinghouse PWR Power Conversion Systems (Continued)
'-3 NSSS Arohiteclt Turbine Gen.
Turbine Bypass Condenser Normal R Main FW FW Pump Shutofl Head Capacity Plant Hame Vendor Engineer Constructor Cop. (MWe)
Capability 1%) Cooling Type Heat Sink Pump*
Drive Type (palo) t9Pmj Shearon Hams 1 & 2 W
F Daniel 860 81 CbsedLoop Nat Cooling Tower 2 (50%)
AC 1.517 15.115
@ 1031 osil South Texas I A 2 W
Bechtel uam 1250 unk.
Closed Loop 7000 Acre Cooling Pond unk.
unk.
unk.
unk.
Summer W
Daniel Gilbert 885 unk Once Through James River unk.
unk unk.
unk Surry 1 & 2 W
Stone&
SionenA 781 40 OnceThvough James River 2 (50%)
AC unk.
13,800 Webster Webster..
Trojan W
Bechtel Bechtel 1095 40 CksadLoop Nat. Cooling Tower 2 (70%)
turbine 1,409 19,800 rt 3...4..@
876 1st0 Turkey Point 3 4
W Bechtel Bechtel 68e 40 OnceThrough Mech. CoolingTowers 2 (80%)
AC 1.149 tJ.000
- vogli, I a1
(@ 81,5 Mign Vogle 1 £2 W
Bechtel Georgia 1079 40 CledLoop NaL Cooling Towers 2
turbine unk.
24.400 Power Co.
@ 1300 DS3I Walls Bar
£
& 2 W
TA TVA 1177 40 Cbuedlnop Nat. Cooling Towers 2
turbine unk.
23.600
@ 819 psig WotI Creek W
- Bechtel, Daniel 1128 unk.
Once Tough 6000 Acre Cooling Pond 2(67%)
turbine unk.
unk.
S&L Yar*,.e*,P W
Sione 1 Slonea&
167 unk.
OnceThough Sherman Pond 3
AG unk.
2.160 Zion___I__A_2______
Webtler Webster 1
1 1 _1 0945 patg IVn
! 8 2 Stonel&
m a.th 1040 40 OnceThough Lake Michigan 2 (50%)
turbine 1.600 15.800 Webster Edion1
@I 160 paig 1 (50%)
AC unk.
unkI CD 0
Westinghouse Technology Manual Chapter 19.0 Combustion Engineering Plant Description
Westinghouse Technology Manual Combustion Engineering Plant Description TABLE OF CONTENTS 19.0 COMBUSTION ENGINEERING PLANT DESCRIPTION...
19.1 Introduction..................................................
19-1 19.2 M echanical Systems..
19-1 19.2.1 Reactor Coolant System.....................................
19-1 19.2.2 Steam Generator..........................................
19-1 19.2.3 Emergency Core Cooling Systems.............................
19-1 19.2.4 Control Element Assembly and Drive Mechanism..................
19-2 19.3 Plant Protection and Monitoring Systems.
19-2 19.3.1 Reactor Protection System (RPS).
19-2 19.3.2 Core Protection Calculators (CPC)............................
19-2 19.3.3 Core Operating Limits Supervisory System (COLSS)................
19-2 19.4 Summ ary.............................................
LIST OF FIGURES 19-1 Reactor Coolant System - Elevation View...........................
19-2 Reactor Coolant System - Plan View..............................
19-3 Steam Generator Secondary Side................................
19-4 Emergency Core Cooling Systems...............................
19-5 Full Length CEA....................................
19-6 Control Element Drive Mechanism...............................
19-7 Simplified Reactor Protection System.............................
19-8 Coincidence Logic Matrix AB...................................
19-9 CPC Software Block Diagram..................................
19-10 Core Operating Limits Supervisory System Block diagram...............
19-3 19-5 19-7 19-9 19-11 19-13 19-15 19-17 19-19 19-21 19-23 USNRC Technical Training Center 19-i Rev 0198 Westinghouse Technology Manual Combustion Engineering Plant Description 19-1 USNRC Technical Training Center 19-i Rev 0198
Westinghouse Technology Manual Combustion Engineering Plant Description 19.0 COMBUSTION ENGINEERING PLANT DESCRIPTION 19.1 Introduction This chapter provides a basic introduction to the Combustion Engineering (CE) technology by discussing the major differences between a
-Westinghous'e design and a CE design.,The first part of the discussion -will be about the mechani cal systems, specifically the reactor, coolant system,-the'steam generatorthe emergency core cooling systems, the-control element assembly, and the control element drive mechanism. The second part will discuss plant protection and monitoring systems.,
19.2 Mechanical Systems 19.2.1 Reactor Coolant System The reactor coolant system consists of two'.
'heat transport loops, each of which has two reactor coolant pumps'and one steam generator.
-The reactor coolant exits the reactor vessel and is transported through hot leg (Th)-piping to the steam generators. The reactor coolant leaves the steam generator through two cold legs (Te), each "containing a reactor coolant pump. In each loop, "the coolant isfetumed to the reactor vessel.'
19.2.2
' Steam Generator The CE steam generators are vertical, invert
, ed, U-tube, tube - and shell heat exchangers similar to the Westinghouse design. Each of the two steam :generators in a CE plant are much larger than those in a four loop Westinghouse plant with the same rated electrical output. - Each CE steam generator has 8,400 tubes,providing 86,000 square feet of heat transfer area. Figure 19-3 shows the design features of a CE steam
".generator.
"19.2-3 Emergency Core Cooling Systems The emergency c6re cooling systems (Figure 19-4) consist of the'high head injection system (HPSI), the low head injection system (LPSI),
-and the safety injection tanks (SITs).,
The high head injection system consists of
- two trains. Borated water is taken from the refueling water storage tank during the injection phase or from the containment sump during the irecirculation phase and pumped to the cold legs through motor operated valves.
The HPSI pumps have a discharge pressure of 1600 psig.
Three non-safety related positive displacement
,.pumps in the chemical -and volume control system provide normal makeup_ to the RCS.
Figure 19-1 shows an elevation view of the These pumps charge water from boric acid reactor coolant systenm.' Figire' 19-2 -shows a makeup tanks into the RCS during an accident, plan view of the system. The hot leg piping is but since they are non-safety related, this flow is 42" in" diameter, an'd the cold leg piping is 30".
not taken credit for in the FSAR accident analy The reactor 'co6lant system is designed to 2500" sis.'
psia, with normal operating pressure around'"'-
2250 psiaý. Tavg at 100% pow~er is 583"F.
The low pressure injection system, or shut down cooling system, consists of two trains.
- Water is taken from the refueling water storage S*
,tank during the injection phase.
The LPSI
.pumps have a discharge pressure of,150 psig.
19-1
-,.Rev-0198 SWestinghouse Technology Manual Combustion Engineering Plant Description
-- USNRC Technical Training Center
WestinEhouse TechnoloEv Manual Combustion Engineering Plant Description The LPSI pumps are capable of taking a suction from the recirculation sump, but the HPSI system is designed to perform the recirculation function. When the LPSI system is aligned for shutdown cooling, the LPSI pump takes a suction on the RCS hot leg and discharges the water through the shutdown cooling heat exchangers to the RCS cold. legs. Note that the shutdown cooling heat exchangers are normally aligned in the containment spray flowpath.
There are four safety injection tanks, one on each cold leg. The SITs are filled with borated water and pressurized with nitrogen. The normal pressure in the tanks is approximately 600 psig.
19.2.4 Control Element Assembly and Drive Mechanism A CE control element assembly (CEA) has a spider and hub design with five fingers which are nearly one inch in diameter and consist of boron carbide pellets. A CEA is shown in Figure 19-5.
The control element drive mechanism is a mag netic jack design (Figure 19-6), except five coils are used instead of three. A control element drive mechanism control system (CEDMCS) is used to automatically or manually move the CEAs.
-19.3 Plant Protection and Monitoring Systems 19.3.1 Reactor Protection System (RPS)
A simplified CE RPS is shown in Figure 19
- 7. First of all, CE uses separate instruments for protection and control. If one of the protection channel parameters exceeds'its trip value, the associated bistable will trip. : This will deenergize the 'trip relay in that 'channel. The six logic matrices consist of a series-parallel contact network (Figure 19-8) and are used to determine whether the two out of four coincidence trip logic has been satisfied.
When a logic matrix determines that the trip coincidence is satisfied, the associated logic matrix relays deenergize, opening the associated trip path contacts. When these contacts open, all circuit breaker control relays deenergize and all reactor trip circuit breakers open. Eight reactor trip circuit breakers are in the circuit between the motor generator sets and the CEDM coils. One pair of breakers on each side must open for the CEAs to trip into the core.
The engineered safety features actuation system operates very similar to the RPS de scribed above.
19.3.2 Core Protection Calculators (CPC)
Core protection calculators (Figure 19-9) have been added to the newer CE plants to generate reactor trip signals based upon local power density and DNBR, which prevents these limits from being exceeded during anticipated operational occurrences. The CPC is, a digital computer that continuously calculates a conserva tive value of plant local power density and DNBR using safety channel inputs from RCS flow, RCS pressure, RCS temperatures, reactor power, and flux distribution.
19.3.3 Core Operating Limits Supervi sory System (COLSS)
The core operating limits supervisory system (Figure 19-10) is a plant computer program which provides comprehensive and continuously updated information. The program consists of on-line power distribution, DNBR correlation, USNRC Technical Training Center 19-2 Key UIY5 S....
v 0198 Combustion Eneineering Plant Description Westinehouse Technology Manual USNRC Technical Training Center 19-2
Westin2house TechnoloEy Manual Combustion Engineering Plant Description calorimetric power, and maximum linear power generation rate calculations. When the COLSS is operable, the plant Technical Specifications allows the plant to operated closer to the kw/ft and DNBR limits.
19.4 Summary This chapter discussed the major differences between a Westinghouse design plant and a Combustion Engineering design plant. The CE plant has two reactor coolant loops, each of which has two reactor coolant pumps and one steam generator.
The emergency core cooling systems in a CE plant consist of a high pressure injection system which is also used for the recirculation phase, a low pressure injection system which is also used for shutdown cooling, and four safety injection tanks.
The CE reactor protection system uses a two out of four coincidence logic for reactor trips and engineered safety features actuations. The core protection calculator and the core operating limits supervisory system allows the plant to operate closer to the kw/ft and DNBR safety limits.
USNRC Technical Training Center 19-3 Rev UiY
19-3 Rev 0198 Combustion Engineering Plant Description Westinghouse Technology Manual USNRC Technical Training Center
FEEDWATER CD PRESSURIZER INLET (D
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0COOLANT COOLANT
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198-19 STEAM OUTLET 26 STEAMj E
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I Figure 19-4 Emergency Core Cooling Systems 19-11 198-19 Iz z
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0 0
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SPIDER WITH NUMERICAL IDENTIFICATION SPRING SPRING RETAINER HOLDDOWN SPRING PLENUM STAINLESS STEEL SPACER B4 C PELLETS SPACER SILVER-INDIUM-CADMIUM SPACER Figure 19-5 Full Length CEA 19-13 198-19 HHZ>
H--Z 134" B4C
REED SWITCH ACTUATING MAGNET Figure 19-6 Control Element Drive Mechanism 19-15 0396
INPUTS FROMNSSS 1 Z U4 0 0 1 MEASUREMENT CHANNELS IIII BISTABLES A CHANN LOGIC MATRICES A-B LOGIC RELAYS
ý A1, ý AB2 AE3 A13 TO 120 VAC VITAL INSTRUMENT BUS NO 1 TO 120 VAC VITAL 480 VAC -30 INSTRUMENT BUS NO 3 RhIIN I**
TO 120 VAC VITAL INSTRUMENT BUS NO 4 TRIP PATHS TRIP CIRCUIT BREAKER CONTROL RELAYS INPUTS FROM NSSS MEASUREMENT CHANNELS POWER LEVEL RATE OF CHANGE OF POWER REACTOR COOLANT FLOW STEAM GENERATOR WATER LEVELS STEAM GENERATOR PRESSURES PRESSURIZER PRESSURE THERMAL MARGIN LOSS OF LOAD CONTAINMENT PRESSURE MA AXIAL POWER DISTRIBUTION TRI NUAL PNO I
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Westinghouse Technology Manual Chapter 20.0 Babcock and Wilcox Plant Description
Westinghouse Technology Manual Babcock & Wilcox Plant Description TABLE OF CONTENTS 20.0 BABCOCK & WILCOX PLANT DESCRIPTION........................
20.1 Introduction................................................
20.2 Mechanical Systems...................
20.2.1 Reactor Coolant System............
20.2.2 Once Through Steam Generator......
20.2.3 Emergency Core Cooling Systems.....
20.2.4 Control Rod Drive Mechanism.......
20.3 Control Systems......................
20.3.1 Integrated Control System..........
20.3.2 Reactor Protection System..........
20.4 Summary 20-4 LIST OF FIGURES 20-1 20-2 20-3 20-4 20-5 20-6 20-7 20-8 20-9 20-10 Babcock & Wilcox Reactor Coolant System.............................
Cutaway View of B&W Once Through Steam Generator.....................
Once Through Steam Generator Heat Transfer Regime......................
High Pressure Emergency Core Cooling System.........................
Decay Heat Removal System......................................
Core Flood Tanks.............................................
Control Rod Drive Mechanism.....................................
Integrated Control System (Simplified)...............................
Integrated Control System (Detailed).................................
Reactor Protection System Channel Logic..............................
20-5 20-7 20-9 20-11 20-13 20-15 20-17 20-19 20-21 20-23 USNRC Technical Training Center 20-i Rev 019l
20-1 20-1 20-1 20-1 20-1 20-2 20-2 20-3 20-3 20-3 Westinghouse Technology Manual Babcock & Wilcox Plant Description
°...........°....
°
°
°...
.°.
°......
20-i Rev 0198 USNRC Technical Training Center
Westinphouse Technology - Manual Bbok&Wlo ln ecito 20.0 BABCOCK & WILCOX PLANT DESCRIPTION 20.1 Introduction This chapter provides a basic introduction to the Babcock & Wilcox technology by discussing "the major differences between a Westinghouse sdeign and a B&W design: The first part of the discussion will be about the mechanical systems, specifically the reactor coolant system, the steam generator, the emergency core cooling systems, and the control rod drive mechanism.
The second part will discuss the control systems, specifically the integrated control system and the reactor protection system.
20.2 Mechanical Systems 20.2.2 Once Through Steam Generator The purpose of the steam generator is to take the heat from the primary coolant flowing inside of the tubes and make steam using the secondary water flowing around the tubes. This purpose is accomplished in the once through steam genera tors (OTSG), which is a slightly different design than the Westinghouse U-tube design.
Instead of having U shaped tubes, the OTSG (Figure 20-2) uses a straight tube design. There are approximately 16,000 tubes in the OTSG.
The OTSG is a counterflow heat exchanger.
That is, the primary coolant enters the tubes at the top of the OTSG and flows straight through the tubes to the bottom of the OTSG. The feedwater enters the OTSG at the bottom and flows to the top of the tube bundle. At the primary outlet, the 20.2.1 Reactor Coolant System flow splits into two paths, each going to a reactor S2,-:
-,coolant pump.
The reactor coolant system consists of two heat :transport loops, each of which has two The operation of the OTSG is also slightly reactor coolant pumps and one steam generator.
-?different from a U-tube,steam generator design.
- The reactor coolant is transported through hot leg The steam generated in a U-tube steam generator S(Th) piping connecting the reactor vessel to the is saturated steam. Also,, the amount of heat steam generators. The heat generated in the core transfer area is constant with power. ;In an
- inside-the reactor-vessel is transferred to the OTSG, the steam at the outlet of the OTSG has a secondary system in the' steam generators. -The -minimum of 50*F superheat, and the heat transfer coolant leaves the steam generator through two area varies with power.
cold leg (Tc) connections; each containing a reactor coolant pump. In each loop, the coolant At the bottom of the OTSG (Figure 20-3), the
'is returned to the reactor vessel.
a
., feedwater is heated to approximately a saturated a
a condition in the subcooled region. The water Figure 20-1 shows the major comp6nefits of, -.begins to boil in the nucleate boiling region, and' a raised loop design of 2the reactor coolant sys-:-,
at-the outlet of this region is about 95% steam.
tein. The hot leg piping is 38" in' diameter, and,
- -The steam then enters the film boiling region,
-the cold leg piping is 32"'. The reactor coolant ;.-where it is heated into saturated steam. Finally, system is designed to 2500 psig, with normal the steam enters the superheat region and receives operating pressure around 2195 psig. Tavg at enough heat to provide the minimum of 50°F 100% power is 601*F.
superheat.,
-fl-lao
- Rev ULYO USNRC Technical Training Center
"%-.1 20-1
-Babcock & Wilcox Plant Description Re
,2LY
Westinghouse Technology Manual Babcock & Wilcox Plant Description The sizes of these heat transfer regions change with power. As power increases, the feedwater flow increases. The subcooled region will increase in size. The nucleate boiling region will also increase greatly in size. The size of the film boiling region is approximately a constant over power.
The increase-in size of the subcooled region and the nucleate boiling region results in a decrease in the size of the superheat region. However, the steam at the outlet of the OTSG still has a minimum of 50°F superheat.
These changes in the amount of heat transfer area allows the operator to actually control primary temperature with feedwater if the control rods are not available.
20.2.3 Emergency Core Cooling Systems The emergency core cooling systems consist of the high head injection system, the low head injection system, and the core flood system.
The high head injection system (Figure 20-4) consists of two trains. Water is taken from the borated 'water storage tank and pumped to the cold legs through motor operated valves. The valves can be throttled to control high pressure injecti6n flow. The' pumps in the high head system are used as the makeup pumps during normal operation.
The low pressure injection system (Figure 20-5), or decay heat removal system, consists of two trains. Water can be taken from the borated iWater storage tank during the injection phase or fibr6i the recirculati6n sump during the recircula tion phase. The pumps discharge to the core flood nozzles on the reactor vessel.
The core flood system (Figure 20-6) consists of two tanks. The core flood tanks are filled with borated water and pressurized with nitrogen.
The normal pressure in the tanks is a!pDroximate ly 600 psig. The tanks discharge into the core flood nozzles on the reactor vessel.
20.2.4 Control Rod Drive Mechanism The control rod drive mechanism for a B&W plant is also slightly different from a Westing house drive mechanism.
Instead of using a stepping motor, the mechanism uses a leadscrew and roller nut assembly.
The major parts of the B&W drive mecha nism are shown in Figure 20-7. A synchronous reluctance motor is used to provide the driving force for the control rod drive mechanism. The motor stator is located outside of the motor tube and the rotor on the inside of the motor tube.
When energized, the upper part of the segmented arms of the rotor are pulled out, which pivots the roller nuts on the opposite end of the arms into the lea'dscrew. For every rotation of the roller nuts around the leadscrew, the leadscrew will move 0.750 inches. To prevent the leadscrew from rotating during rod motion, there is a torque taker on the top of the, leadscrew. The torque taker transmits the torque to the torque tube and prevents rotation of the leadscrew. The torque taker also has a permanent magnet on it to close reed switches for rod position indication.
The control rod drive mechanisms are de signed to drop the rods upon a loss of power.
With no power to the drive mechanism, the segmented arms will pivot to the inward position due to springs. This causes the roller nuts to disengage the leadscrew, and the rod will fall.
USNRC Technical Training Center 20.2 Rev U19
Rev 0198 Westinghouse Technology Manuial Babcock & Wilcox Plant Description USNRC Technical Training Center 20-2
Westinghouse Technology Manual Babcock & Wilcox Plant Description 20.3 Control Systems 20.3.1 Integrated Control System The B&W plants use an integrated control system (ICS) to simultaneously control the main turbine, main feedwater flow control valves, main feedwater pumps, and the control rods.
The ICS is shown in simplified form in Figure 20-8 and more detailed in Figure 20-9. The basic function of the system is to match generated megawatts to desired megawatts.
There are four major subassemblies in the ICS. These are:
"* Unit load demand,
"* Integrated master,
"* Feedwater demand, and
"* Reactor demand.
The unit load demand subassembly acts as the setpoint generator for the ICS. The operator can input the desired load and the desired rate of load change into this subassembly, and these, signals are transmitted to the remainder of the ICS.
There are several functions of the integrated master subassembly. First, this subassembly controls the load of the turbine generator by positioning the turbine control valves. Another function is to feed the demand signal to the feedwater and reactor demand subassemblies.
To do this, the integrated master modifies the signal being sent. This subassembly also con trols the position of the steam dump valves. The final purpose is to maintain a constant load on the turbine, even when plant conditions are chang ing. For example, if circulating water tempera ture is higher than normal, the vacuum in the main condenser will be lower (higher absolute pressure). The Output of the turbine will be less due to the loss of efficiency.'. The number of megawatts generated will be less than the desired megawatts.
The error signal will cause an increase in the output of the feedwater and reactor demand subassemblies. The integrated master performs its functions by controlling at a c6ristant steam pressure. If pressure goes up, the turbine valves will open to lower pressure (pick up more load), and vice versa.
The feedwater demand signal originates in the unit load demand and is modified by thle integrat ed master. There is a separate control -for each OTSG. The demand signal controls the position of the startup feedwater regulating valve and the Smain feedwater regulating valve,- which are Soperated in sequence. -That -is, the startup feedwater regulating valve opens first and then the main valve. To maintain the proper differen tial pressure across the feedwater regulating
'valves, the feedwater demand subassembly also controls the speed'of the main feed pumps.
The reactor demand subassembly controls the position of the,control rods for the purpose of controlling reactor coolant system temperature.
The demand signal again' comes from the unit
,load demand and is modified by the integrated
- master.,
20.3.2 -Reactor Protection System
-.The reactor protection system for a B&W plant is significantly different from that of a Westinghouse plant.
The reactor protection system is shown in Figure 20-10. If one of the monitored parame ters exceeds its trip value, the associated contact in that channel will open. This will deenergize the trip relay in that channel, which tells the other USNRC Technical Training Center 1 20-3 Rev 0198 I I Westinghouse ° Technologly-Manual Babcock & Wilcox Plant Description
WetngoseTchoog.
ana.abok..ilo.Pat ecI-t.o.
channels that one channel has seen a trip condi tion. If a second channel receives a trip signal (from the same or a different parameter), the reactor will trip.
- Therefore, the reactor protection system for a B&W unit is two out of four reactor protection system channels, and not based upon a certain coincidence of only one parameter.
20.4 Summary This chapter discussed the major differences between a Westinghouse design plant and a Babcock and Wilcox design plant. The B&W plant has two reactor coolant loops, each of which has two reactor coolant pumps and one steam generator. B&W plants use once through steam generators.
The emergency core cooling systems in a B&W plant consist of a high pressure injection system (which is also used for normal makeup to the reactor coolant system), a low pressure injection system (which is also used for decay heat removal), and a core flood system.
An integrated control system is used to control the main turbine, main feedwater flow control valves, main feedwater pumps, and the control rods in the B&W design plant. The reactor protection system coincidence for a B&W unit is two out of four channels of any combina tion of monitored parameters exceeding their setpoints.
USNRC Technical Training Center 20-4 Rev, 0198 Westinghouse Technology Manual Babcock & Wilcox Plant Descrintion
Figure 20-1 Reactor Coolant System Supports and Restraints 20-5
0596 INLET HANDHOLE.
INSTRUMENT TAP AUXILIARY INLET SECTION MANWAYS TUBE BUNDLE SHROUD OUTLETS (2)
INLETS (2)
LOWER TAP 2 REACTOR COOLANT OUTLETS SUPPORT Figure 20-2 Integral Economizer Once-Through Steam Generator 20-7 (5)
STEAM OUTLET L__
FEEDWATER INLET T
SUPERHEAT REGION 100% I 0
100%
100%
w I-LL 0a
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0 PERCENT FULL LOAD Figure 20-3 Heat Transfer Areas 20-9 0596 1
100%
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CONTROL KREACTOR POWER
,ROD DRIVE MOTORS ONCE THROUGH VALVE FEEDWATER STEAM GENERATOR PUMP Figure 20-8 Simplified Integrated Control System 20-19 0796-12
SP-----TRACK S
+S RUNBACKS HIGH LOAD LIMIT LOW LOAD LIMIT UNIT LOAD DEMAND SUBASSEMBLY INTEGRATED MASTER SUBASSEMBLY f
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