Westinghouse Technology Manual Chapter 17.0 Plant Operations PDF

Westinghouse Technology Manual Chapter 17.0 Plant Operations Westinghouse Technology Manual Plant Operations TABLE OF CONTENTS 17.0 PLANT OPERATIONS.........................

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 Westinghouse Technology Manual Plant Operations 17.0 PLANT OPERATIONS and letdown flow. The steam generators are in the "wet layup" condition (filled to the 100% Learning Objectives: 'level with water) and all secondary systems are secured with the exception of one circulating 1. Arrange the following evolutions in the water pump. The main and feedwater pump proper order for a plant startup from cold turbines are on the turning gear. All pre-startup shutdown: checklists have been completed. a. Start all reactor coolant pumps, 17.2.2 Operations b.ý Place all engineered safety systems in an operable mode, A pressurized water reactor may have a C. Establish no-load Tavg, positive moderator temperature coefficient-at low d. Take the reactor critical, - temperatures due to the soluble poison in the e.° Start a main feedwater pump, moderator. To minimize the-magnitude of the f. Load main generator to the grid, and positive moderator temperature coefficient or g. Place steam generator level control make it negative, the plant is ,brought to near system in automatic. operating temperatures with reactor coolant pump heat before 'the reactor is made critical. To 17.1 Introduction operate the reactor coolant pumps, reactor coolant system pressure must be increased to, approxi This chapter will briefly discuss the ,basic mately 400 psig to satisfy net positive suction procedures for startup, power operation, and head requirements. - (Pressure must be main shutdown of the pressurized water reactor tained below 425 psig while RHR is aligned to described in this manual. The discussion will be the reactor coolant system.)- When operating the general in nature and is designed to show how reactor coolant pumps at low pressures, the the systems previously discussed are utilized, reactor coolant-pump number one seal bypass during'plant operations. -valve must be open to ensure adequate flow to cool and lubricate the pump radial bearing. '17.2. Plant Heatup Pressure is increased by maintaining charging 17.2.1 Initial Conditions flow greater than letdown flow. -When pressure is stable between 400 and 425 psig, the reactor -The nuclear steam supply system (NSSS) is coolant pumps are started to begin'reactor coolant in the "cold shutdown" mode (Tavg = 120*F, "system heatup. Pressurizer heaters are energized pressurizer pressure = 50 - 100 psig, boron 'to begin pressurizer heatup. 'Residual 'heat concentration sufficient to yield 10% shutdown' removal flow is diverted through 'the bypass line margin, pressurizer solid, reactor coolant pumps "to bypass the heat exchanger and allow heatup. off). Decay heat is being removed by the residu al heat removal system (RHR) with letdown from This RHR system alignment is maintained to RHR established for reactor coolant system provide adequate letdown for pressure control cleanup. Pressure in the solid system (Figure" and to remove the excess coolant volume pro 17-1) is being maintained by adjusting charging, duced by expansion due to heatup. During the 17-1 Rev 11198 - USNRC Training Center Technical Training USNRC Technical Center 17-1 Rev T0198  Westinghouse Technology Manual Plant Oneration Westnghuse MaualPlant echolog Operations entire heatup and pressurizer draining process, in service to provide an additional letdown path approximately one-third of the reactor coolant to minimize the time necessary to "draw a bub system volume (30,000 gallons of water) will be ble" in the pressurizer. diverted to the holdup tanks through the chemical and volume control system. The main and auxiliary steam lines are warmed as steam is available during the plant As the reactor coolant system temperature heatup. Main steam isolation valves are opened approaches 200'F, steam generator draining is initially as heatup begins. commenced through the, normal blowdown system. If reactor coolant system oxygen con As reactor coolant pressure continues to centration is high, hydrazine is added through the increase, letdown flow will also increase. The chemical and volume control system for oxygen low pressure letdown valve is adjusted (closed) scavenging. Oxygen must-be in specification until the normal letdown pressure (340 psig) is before exceeding 250"F. achieved and then orifice isolation valves are shut as necessary to maintain letdown flow below the After oxygen is within specification, a maximum. hydrogen blanket is established in the volume control tank. This is accomplished by securing Before reactor coolant system temperature the nitrogen regulator, opening the vent from the reaches 350°F, the residual heat removal system volume control tank to the waste gas header, and is isolated from the reactor coolant system and is raising the volume control tank level to force the aligned for at-power operation (emergency core nitrogen to the waste gas system. After the cooling system lineup). All reactor. coolant volume control tank level has raised to approxi system letdown is now through the normal mately 95%, the hydrogen regulator is placed in letdown orifice path to the chemical and volume service and the last of the nitrogen is purged to control system. the waste gas system. Volume control tank level is allowed to return to normal with the hydrogen After the residual heat removal system is regulator maintaining an overpressure of approxi isolated from the reactor coolant system, system mately 15 - 20 psig. pressure is allowed to increase as the pressurizer temperature increases. When pressurizer temperature reaches satura tion temperature for the pressure being main When pressurizer level, as read on the hot tained (450°F for 400 psig), a pressurizer bubble calibrated channels, indicates the no-load pro is established. Reactor coolant system tempera grammed setpoint, charging flow is placed in ture is approximately 250 - 300TF. The bubble is automatic. As system heatup continues, pressur established by maximizing letdown and minimiz izer level will try to increase due to coolant ing charging flow. This will cause the pressuriz ,expansion. Pressurizer level control will com er level to decrease. System pressure will be pensate by reducing charging flow. maintained at 400 psig as the saturated pressuriz er water flashes to steam. Pressure control can When reactor coolant system pressure reach now be accomplished only by heater and spray es 1,000 psig, the emergency core cooling operation. Residual heat removal is maintained system accumulator discharge valves are opened Training Center 17-2 Rev 0198 U/SNRC Technical USNKC Technical Training Center 17-2 Rev 0198 Westinghouse Technology Manual Plant Operations Westinghouse Technology Manual Plant Operations and all emergency core cooling system equipment turning gear, and all electrical power supplied is checked for proper alignment. from off-site.. After reactor coolant pump number one seal The next step in the startup of the plant is to leakoff has increased to at least one gallon per take the reactor critical. minute on all reactor coolant pumps, the number one seal bypass valve is closed. 17.3 Reactor Startup to Minimum Load As pressure increases above P-11, the loyv Reactor startups are normally performed at pressurizer pressure engineered safety feature s no-load temperature where the moderator temper actuation signal is automatically unblocked ature coefficient is at a low or negative value. Pressurizer heaters and spray valves are placed ii automatic control when pressure reaches th( If necessary, the reactor coolant boron normal operating value of 2235 psig. concentration is adjusted-to the required value prior to startup. The required value is calculated When steam pressure is at or above 125 psig;, by performing -a -reactivity balance (estimated e critical condition calculation). For a pressurized main and feed pump turbine gland seals ari established, and a condenser vacuum is drawn ' water reactor; a specific critical rod height is Condenser vacuum is established by mechanica 1, chosen and boron concentration is adjusted to a vacuum pumps and/or steam jet air ejectors. value which will produce criticality at the desired rod height. Control rods must always be with As reactor coolant system heatup continues drawn abovethe rod insertion limit prior to the high steam flow engineered safety feature s criticality to ensure adequate "cocked" reactivity actuation signal will be automatically unblocke4I to satisfy shutdown margin'requirements. when Tavg increases above 540"F. The -stean dump system, operating in pressure contro 1 - - -Immediatelyprior to reactor startup, function mode, will 'dump steam to the main condense r - al checks are performed to ensure proper opera when steam pressure reaches a predeterminecI tion of the source and intermediate range nuclear setpoint (normally 1,005 psig which is saturatior I instrumentation channels. A source and interme pressure for the 547°F no-load reactor coolan t diate range channel are recorded and the "source system temperature). The steam dump systen n range high flux at shutdown" alarm is blocked. will dissipate the excess decay and reacto r - _- - 1 1, - 1 1 - coolant pump heat ,and maintain Tavg approxi - The shutdown -rod banks- (if not 'already mately equal to 547"F. The startup feedwate]r withdrawn) are withdrawn in sequence, and system is used to feed the steam generators t(o then, the control banks-are withdrawn in manual maintain level at the no-load value. to achieve criticality. After criticality is achieved, a positive startup rate is established, and power Plant conditions are now as follows: norma level in increased.- When 'power exceeds the operating temperature and pressure, reacto r source range -permissive (P-6)' setpoint, the shutdown, normal condenser vacuum, stean n source range trip is blocked and source range dump to the condenser in the steam pressur(e high voltage deenergized. mode, main and feedwater pump turbines on the Center 17-3 Rev 0198 USNRC Training Center USNRC Technical Training 17-3 : , Rev 0198  Westinghouse Technology Manual Plant Operations Westinghouse Technology Manual Plant Operations Power is then increased to 10-8 amps in the is, chosen, and the turbine load is increased intermediate range where neutron flux is stabi toward 15%. lized (lev iied out) and critical data are taken. After critical data are taken, the reactor power As turbine load is increased, the reactor increase is continued until the "point of adding operator withdraws control rods to maintain Tavg heat" is reached. This is the power level (about Tref. During the load increase, the steam dump 1% power) where the reactor is producing valves will shut as steam pressure- decreases. sensible heat. When the valves are shut, steam dump control is shifted to Tavg control to be ready for a possible The reactor operator hold 1% power while load rejection or reactor trip. Steam generator the turbine-driven main feedwater pump is level continues to be controlled by maanual warmed and placed in service. Feedwater supply operation of the main feedwater regulating is switched from the auxiliary feedwater system bypass valves. 'to the main feedwater pump. When power level exceeds the setpoint of the Reactor power is increased to' about 5% nuclear at-powier permissive (P- 10), tlhe interme power in preparation for rolling the main turbine. diate range rod withdrawal stop and the inrterme Increasing reactor power will cause the steam diate and power range (low setpoint) trips are dump valves to open further to dissipate the manually blocked. excess heat. Steam generator feedwater is controlled manually through the small (4 -6 inch) At or above fifteen percent power, the rod bypass valves to maintain level at the program control system and steam generator level control setpoint. Providing excess reactor power yields system are placed in the automatic mode. a constant steam load as the turbine is rolled. As the turbine takes more steam, the steam dump 17.4 Power Operations valves will modulate closed. This makes control of the reactor and steam generator levels much Power level is increased by selecting a easier. A heater drain pump is energized at this desired load and load rate with the turbine time. electrohydraulic control system and allowing the reactor tofollow the turbine load change. As The turbine acceleration rate' is chosen, and turbine load increases, Tavg will tend to decrease. the turbine is accelerated to synchronous speed. The automatic rod control system will sense this With the turbine at synchronous speed, reactor and withdraw control rods to increase reactor power is increased to six percent so that reactor power. power is greater than the initial turbine load. As load'is increased to 30% power, a second The turbine is synchronized with the utilities condensate/booster pump is started, and main electrical grid, and thie generator output breaker is generator hydrogen pressure is increased to its closed: The electrohydraulic control system maximum value (75 psig). automatically assumes five percent of full rated load. After turbine operation and other apolicable As load increases between thirty and fifty instrumentation is checked, a turbine loading rate percent, additional circulating water, feedwater, 17-4 Rev 0198 IUSNRC Training Center Technical Training USNRC Technical Center 17-4 Rev 0198 - Westinghouse Technology Manual SPlant Operations 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 I - Plant shutdown is accomplished by essential ly reversing the steps described in plant startup. Training Center 17-S Rev O198 Technical Training USNRC Technical Center - --17-5 -Rev -0198  Westinghouse Technology Manual Plant Onet:ations W snPlant- n One-ations APPENDIX 17-1 G. Pre-startup checklists completed PLANT STARTUP FROM COLD SHUT II. Instructions DOWN A. Heatup from cold shutdown to hot shutdown I. INITIAL CONDITIONS (Mode 5 to Mode 4) A. Cold shutdown - Mode 5 1. Permission received from operation Keff < 0.99 supervisor for startup 0% rated thermal power Tavg < 200*F 2. Verify shutdown rods withdrawn or verify sufficient shutdown margin avail B. Pressurizer ability 1. Temperature approximately 320"F, with a 3. Verify dr establish RCP seal injection steam bubble established. flow 2. Level approximately 25% with level 4. Begin pressurizer heatup to increase RCS control in automatic. pressure. C. RCS temperature 150 - 160"F CAUTION: Do not exceed a heatup rate of 100lF/hr on the pressurizer, Note: Temperature may be less than lO0*F/hr'_on- the.RCS, or 320"F T 1507F depending on decay heat between pressurizer and spray tem load from the core. perature. Use auxiliary sprays for pressurizer-RCS mixing. D. RCS pressure 100 psig 5. Maintain the RCS temperature < 160°F 1. Charging and RHR letdown established by adjusting flow through the RHR heat exchangers 2. RCS pressure maintained by pressurizer temperature @ 320"F 6. Startup checklist for Technical Specifica tion requirements completed 3. RHR system in operation 7. Begin establishing steam generator water E. Steam generators filled to wet-layup (100% levels to 50% on narrow range indication level indication) (steam generator blowdown system). F. Secondary systems shutdown. Main turbine and main feedwater pump turbines on their 8. Open main steam line isolation valves turning gear 17.6 Center Training Technical , USNRC Technical Training Center USNRC Rev 0198 17-6 Rev 0198 Westinghouse Technology Manual Plant Operations Westinghouse Technology Manual Plant Operations 9. 'If required, commence condensate 2. Complete emergency core cooling system cleanup master checklist 10. Establish condenser vacuum 3. As the RCS pressure increases, maintain letdown flow 120 gpm by increasing the 11. Continue -pressurizer heatup to 430°F setting of the low pressure letdown (RCS pressure 325 psig). Use the low control valve, and by closing the letdown pressure letdown control valve to main orifice isolation valves as necessary. tain letdown flow. RCS pressure control is via heater and spray actuation. 4. Prior to reaching 1;000 psig in the RCS, open each of the 'cold leg -accumulator 12. Start the reactor coolant pumps. After isolation valves. -Remove each valve's five minutes running, sample the RCS for power supply. chemistry specifications. Partially open pressurizer sprays for mixing. 5. When RCP no. 1 seal leAkoff is > 1 gpm, or RCS pressure > 1,500 psig, close 13. Stop residual heat removal system pumps RCP seal bypass return valve. Verify no. 1-seal leakoff remains > 1 gpm. 14. Allow RCS temperature to increase to 200"F 6. When RCS pressure reaches 1,970 psig, verify pressurizer low- pressure safety 15. When RCS temperature reaches 200*F,, injection logic auto reset. -determine that primary system water "-- chemistry is within specifications 7. When Tayg exceeds 540°F, verify steam line safety injection logic auto-reset. 16. When condensate chemistry is within specifications as determined by chemical" 8. The steam dump control system is in , lab, 'align condensate and feedwater -. pressure control mode (set at 1,005 psig) Ssystem to normal configuration. :to maintain RCS temperature at 5470F. 17. Verify control rod drive cooling fans on 9. Place RCS pressure control in automatic before RCS temperature reaches 160'F to maintain 2235 psig. 18. Terminate residual heat removal letdown' 10. Establish hot standby conditions of 540 to chemical and volume control prior to 547"F Tavg.' *- " -_'.exceeding 350"F and 425 psig. C. Heatup from-Hot Standby 'to Power B. Heatup from Hot Shutdown to Hot " :Operations. (Mode 3 to Mode 1) Standby-(Mode 4-to-Mode 5) 1. Administrative permission to take the 1. Startup checklist for Technical Specifica reactor critical has beenobtained. tion requirements completed Center 17-7 - Rev U19 Technical Training ý USNRC Technical Training Center ,- 17-7 Rev 0198  -1 Westinghouse Technology Manual Plant Operations 2. Notify system dispatcher of unit startup a. Block source range trip at P-6 and approximate time the generator will be tied on to the system. b. Record critical data at 10-8 amps 3. Notify onsite personnel of reactor startup 8. If the control bank height at criticality is over P/A system. below the minimum insertion limits for the 0 percent power conditions. 4. If shutdown banks have not been with drawn, complete a shutdown margin a. Re-insert all control bank rods to the calculation (assuming SD banks out) and bottom of the core., if desired SD margin will exist, withdraw the shutdown banks to the fully with b. Recalculate the estimated critical drawn position. boron concentration Note: Nuclear instrumentation shall c. Borate to the new estimated critical be monitored very closely in boron concentration anticipation of unplanned reactivity rate of change. d. Withdraw the control bank rods in manual and take the reactor critical 5. Calculate the estimated critical boron concentration for the desired critical 9. Withdraw rods to bring reactor power to control bank rod position (normally 150 approximately 1% on power range steps on Bank D). indicators and select the highest power range channel to be recorded on NR-45. 6. If necessary, conduct a boron concentra tion change to the estimated critical boron 10. Start a main feedwater pump at 1% power concentration. Equalize boron concentra and maintain steam generator levels at 50 tion between the reactor coolant loops and percent narrow range level indication the pressurizer by turning on pressurizer during secondary plant startup by throt backup heaters. tling the feedwater bypass regulating valves and operating the master feedwater Note: Nuclear instrumentation shall pump speed controller and the individual be monitored very closely in steam generator feedwater pump control anticipation of unplanned station in auto. reactivity rate of change. CA UTION: Coordinate all steam Note: Block the source - range high generator steam removal and flux level at. shutdown alarm at significanit feedwater changes with both source range panels. the 'reactor panel operator while rod control is in manual 7. Withdraw the control bank rods in manual and take the reactor critical. 17-8 Rev 0198 USNRC Training Center Technical Training USNRC Technical Center 17-8 Rev 0198  Plant Operations Westinq house Technolopv Manual PatOeain 11.Turbine has been on turning gear at least 20. After rod control is placed in automatic, one hour check steam pressure less than steam dump set point and steam dump valves 12. Increase reactor power by manual adjust full closed, then transfer steam dump to ment of the control bank until the steam Tavg mode. dump is bypassing steam flow equivalent to 8 percent nuclear power. 21. Above 15 percent power, transfer steam generator feedwater regulating valve 13.Verify the unit auxiliary and startup control to auto when level is at setpoint transformer cooling systems are aligned and steam flow equals feed flow. for automatic operation. 22. Continue turbine load increase to 100% 14. Start the turbine, bring it up to speed, and connect the generator to the grid. Trans a. Start secondary system components fer station power from the startup trans as required during power escalation. former to the unit auxiliary transformer. Additional components would include items such as condensate pumps, 15. Increase generator load at the desired rate, heater drain pumps, feedwater while maintaining Tavg by manual rod pumps, and condenser circulating control. water pumps. 16. Transfer feedwater flow from bypass b. Maintain rate of load increase within valves to the main feed regulating valves. plant design limits. These limits Maintain programmed level during this would include the loading limits process. imposed upon the main turbine and the limits imposed by boron dilution 17. When reactor power increases above 10 rates. 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. 17-9 Key Rtev UIY 0198 USNRC Center Training Center USNRC Technical Training 17-9  Clo Cf -J CD -L C, 0 a. 0u  (0.601, 228) 228 200 :1.0,184) - I CL a 150 BANKDC B I/A IJ. C- zR 0 0~ 0 zCC 100 (0.0, 98) "0BANKD 0 0 / - - - - - - - - - - - - -- - - - - - - - - - - - - - - - - - - - -- - - - - - - 50 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -- - - - - - -- - -- - - - - - - - - - - - (.148,0).2.4.6.8 1.0 FRACTION OF RATED THERMAL POWER Figure 17-2 Control Bank Insertion Limits, 4-Loop Operations 17-13  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 V 8/90  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 viii 8/90  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 X 8/90  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 xix 8/90  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 tOj 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 CO. Power 3411 1129 MUG Clinton 1 Clinton IL Illinois Power Co. BWR GE Sargent 2894 930 DER _A Lundy _ : Comancho Peak I Glen Rose TX Texas Utilities Electric Co. PWR W Gibbs & 3425 1150 D.II CorminchiePeai 2 Peak__ Rose 2_ Glen GlenRose TX TX ~ Utilitis Electric Texus Uhtiltes Electric Co. Co PWR W Gibbs Gibbs 6 _ 3425 3425 1150Hill 1 150 tIP ____________ _________ ___________________________ I-ill_____ _____  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 Nebraska Public Power District BWR GE Burns & 2381 764 Moc Cooper Brownvllle NE I Pow_ ___ Red Level FL Flonda Power Corp.. PWR B&W Gilbert 2544 821 MIX Crystal River 3 Bridgman MI IndianafMtchigan Power Co. PWR W AEP 3250 1020 MDC DC.Cook.I Orldgman B.r- MI IndianatMichigan Power Co. PWR W AEP 3411 1080 MDC OC.Cook2 OH Toledo Edion Co. PWR B&W Bechtel 2772 860 MDC Davis-Besse Oak Harbor Avila Beech CA Pacifc Gas A Electnc Co. PWR W Pacific Gas & 3338 1073 MDC Diablo Canyon I Electric Avila Beach CA Padfic Gas A Electric Co. PWR W Pacific Gas & 3411 1087 MDC Dablo Canyon 2 Electric Morris IL Commorwealti Edson Co. BWR GE Sargent 2527 772 MoC Dresden 2 S................. & Lundy , Morris IL Commovwealth Edson Co. BWR GE Sargent 2527 773 MoC Dresden 3 II....... & Lundy Palo IA Iowa Electic Light & Power Co. BWR GE Bechtel - 1658 515 MDC Duane Arnold Ls).Dothan AL Alsbama Power Co. PWR W Bechtel 2652 813 MDC Farey 1 Dothan AL Alabama Power Co. PWR W Bechtel 2652 823 MD.; Farley 2 Newport Mill DetoltEdisonCo. 8WR GE Detroit 3292 1093 MD, Fermi 2 Edison Fitzpatrick Scribe NY New York Power Authority ,WR GE Stone & 2436 778 MIX SI Webster _____ Fort Calhoun NE Ornahe Public Power District PWR C-E Gibbs & 1500 478 MDC Fort Calhoun I. Hill *__ _ GA & Lundy Sargent 842 ___ ___ 330 MDC.___ _ Fort SL Vran Plattevllle _.. Services Company Public CO._______ of Colorado.HRiM................... U) 1520 470 MMC Geria, - Ontario NY Rochester Gas & Electric Corp. PWR W Glilbert- Port Gibson MS System Energy Resources. Inc. BWR GE.- Bechtel 3833 1142 MDC Grand Gulfl Port Gibson MS System Energy Resources, In.. BWR GE Bechtel 3833 1250 DER Gand Gulf 2 Ibddent Nack CT Conneticut Yankee Atomic Power Co. PWR W Stone A 1825 569 MDC HaddamNed Webster _______ _____ Baxley GA Georgia Power Co. BWR GE SCSI 2436 756 MDC Hiatch I *++ + ". +< + Bechtel Baxley GA Georgia Power Co. BWR GE SCS / 2436 768 MDX Hatch 2 Bechtel Salem NJ Public Services Electrc & Gas Co. BWR GE Bechtel 3293 1067 MDC. Hope Creek I Indian Point 2 Indian Point NY Como riodaEdionCo. PWR W UE&C 2758 849 Mt'  Table 2-1. General Plant Data - Sorted by Plant Name (Continued) Reacto, Plant Icily S81819 jUstitty Reactor USSS Core Powe r Net Electric, Indian Point -3 Archlteg Indian Point N Teo PWR Vendor Engineer MWI MDC or DER W 965 w1 LavalleI PWR "Commonwealt Edson Co. W Pioneer MDC 503 LaSalle 2, BWYR GE Sargent FILR 3323 & Lund 1036 MOC Limerick I Pladelphia Power a Uoht Co. 1WR Sargent Pottalow GE 33293 A L und Limerick 2 DER Pottstown P--adelphia Power £ Ught Co. DWR, GE Bechtel 3293 Maine Yie JBechtel 13293 - MieYankeeo Atoiuc Power Co. GE PWR C-E 810 McGuire I Stonel A Webster MOC Duke Power Co. 2630 NG! Cornelius 11290 MOC McGuire 2 PWR Duke Power Duke Oe. Ca 3411 PWR MDr, t4j Waterlord_ Northeast Utilities Duke Power 1129 Millstone Cca 34-11 BWR ;-E Millstone 2 Northeast utilities EMoaw 27011 Waterford MDC CT,I 863 Monicllton3 "PWR---- 2700ý Aonticello IN_ INolther States Power Co. Stone & 71142 NeMonilelloin BWR 3117 Webster GE Bechtel 5386 Nine Milo Point1 1675 0 PWR GE ;cribs Vlra P ower Co. Niagael Mohawk NrtiAnenie io~ wY 1850ý3 GE_ Stone _& tineral NoruthArms 2 M PWR Webster 33233 SMovie -&, tinstal Virginia Power Co. W 915 Dortw"I1 a Webster 289 3 MIDC Duke Power Co. PWR Stone £, W Webster Oconee S 2893 _ PWR Duke/ 08W 8 46ý Oconee2 3 Duke Power Co. Bechtel 2568 Duke/ B8W 846:::: OyswCreek Si 2566 MDC "" uk Poe Co. Be8;tel OystereCseo1 NJ GUtlNuclear Corp. F PWRR--W - ut Haven MI ConSumners Pwer Co. PaloVades SI PWR C-E Fninterburg AZ Arizona Public Service Co. Palo Verde 1 W PWR C. -E Palo~2 Ved Wintersburg ArLzona Public Service Co. PWR C-E 3800 1221  4-' Table 2-1. General Plant Data - Sorted by Plant Name (Continued) 4' 1 City State Utility Reactor NSSS Architect/ Core Power Not Electrical MW. Raling Reactor Plant MUC or DER Type Vendor Engineer MWI Output MWe PWR C.E Bechtel 3800 1221. MDC Palo Verde 3 Wintersburg AZ Arizona Public Service Co. BWR GE Bechtel 3293 1051 M Peach Bottom 2 Peach Bottom PA Philadelphia Power & Light Co. BWR GE Bechtel 3293 1035 MDC Peach Bottom 3 Peach Bottom PA Philadelphia Power A Light Co. The Cleveland Electric Illuminating Co. BWR GE Gilbert 3579 1205 MDC Perry 1 North Perry OH BWR GE Gilbert 3579 1205 DEIn Perry 2. North Perrya OH 1 Cleveland Electric Iluminating Co. BWR GE Bechtel 1998 570 MDC Pilgrim I Plymouth MA Boston Edson Co. Electric Power Co. PWR W Bechtel 1518 485 MDC Poit Beech I TroCreeks WI eWicornsin Pont Beech 2 Two Creeks WI, Wiaconsin Electric Power Co. PWR W Bechtel 1510 485 MDC PWR W Pioneer 1650 503 MLD Prairie Island I Red Wing MN Northern States Power Co. NorthernStates PowelCo. PWR W Pioneer 1650 503 MDC in' Prairie Island 2 RedWing MN M BWR -. GE Sargent 2511 769. MDC QuadCites 1 Cord", ''., , IL I Edisonflowa-Ilkinols Gas a Electric Commonwealth - & Lundy.. OuddCities2 Cordova IL Commonwealth Edlsonilowa-llhnols Gas A Electric BWR GE Sargent 4 2511 769 MIX,, PlWR &W ,Bechtel 2772 873 MDC RanchoSew Clay Station - CA Sacramento Municipal Utility District RWR GE Stonm & 2894 936 MDC River Bend 1 St. Franciaville LA Gulf States Utlittes Co. -. I.Webster - _______ ______ -I _- PWR W E -"m. 2300 665 MDO Robinson 2 Hartsville SC Carolina Power a Light Co. PWR W Gas& 3411 Pacific 1106 MDC SalemI Salem N , Public Services Electric &GasCo. S, , ,-; , ,,. - , Electric ____.......... & Gas Co. Publc Services Electric PWR W Pacific Gas & 3411 1106 MDC Salem 2 Salem NJ SElectric - EdisovnSan Diego Gas A Electric PWR W Bechtel., 1347, 436 MDC SenOnofre I SanClemenle CA ,Southern Calilornis EdoSan Diego Gas A&Elecic Southem Calfontoiad PWR C-E Bechtel 3390 1070 LAjC SanOnofre2 SanClemente CA CA Edison/San Diego Gas A Electric PWR Southern Californma C-E Bechtel 3390 1080 MDC SanOnofre3 SanClemerfe Newl-HampshweYareUnt PWR W LUE&C, 3411 1150 MDC SeabrookI Seabrok' NH PWR W TVA 3411 1148 MDC S.. Sequoyhl I Soddy-Daisy TN Tennessee Valley Authority PWR W TVA 3411 1148 Sequoyah 2 Soddy-Daisy TN Tennessee Valley Authority PWR W Ebasco 2775 860 1MDC" Sheaon Harris I New Hill NC Carolina Power & Light Co  Table 2-1. General Plant Data - Sorted by.Plant Name (Continued) Reactor Plant City Stale utility Reactor INSSS Archltectl Vendor Core Power Net Electric al MW. Rating Type Engineer MWI Output MWe Brookhaven NY 8Wfl MDC or DER Long Island Lighting Co. Stone & 2436 820 DER South Texas I )alhacios TX- Websler Houston LUghting & Power CO. PKR W Brown A 3800 1250 South Texas 2 Ullllly Ho , Long Ughbng Idend Co. Co. & p ,,- Lighbng Root. P a-_clo- TX ,Houslon Lighting & Power Co. W PWfl Brown & 3800 1250 MIXO PWR Root............. SL Luci I Hlulchin on Island FL Florida Power & Ught Co. C-E Ebksm 2700, 839 St. Lucle 2 Hutchinson Istand FL Florida Power & Light Co. PWR C-E 839, ummnsni '.. EVQs 2700 Parr PWR SvIl~ SC -_o, Caroýia Electric & Gas Co. MDC W Gilbert 2775 585 5Surry I..m. "..id i..L Nk 1 MOc. I A virginia Power Co. PWR Stone & 2441 Surry 2 TGravel Nock ,7,,,,, p.,... -.... Webster ginia Power Co. W MOO I'WH Stone& 2441 781 Q r -- M Susquisanna andI I orwlck. , -. ,. UIE-- Webster 1032 MDC B- Pennsytvania Power & Light Co. Bechtel 3293 R kmmmu~f i. --... i I I cklIWIR P-ensnlvania Power A Light Co. BWR Gi MOO 0 Bechtel 3293, 1032 1hrte Mile Island I I Lonidonderr, Twu IlA iiGN Pucea..C T S......... '"# " e" ' p.. PWR B&W Gilbert 2535 76 MOC Trojan Prescolt OR Porland General Electric Co. PWR MDC W Bechtel 3411. 1T095 Turkey PoInt 3 Floda City FL Floria Power & Ught Co. PWR MOC Bechtel, 2200 6 Turkey Point 4 Florida City W MOXO FL Florida Power A Ught Co. PWR 66 W Bechtel 2200 6 |68 Veniont Ywakee Vernon VT Vermont Yankee Nuclear Power Corp. BWR MOO. GE Ebawo. 1593 5 04 Vogue 1 Waynesboro GA Georgia Power Co. PWR MOC. Bechtel 3411 I 079 Vogue 2 Waynesboro W MOO GA Georgia Power Co. PWR Bechtel 3411 1 079 Walerford 3 Taft W MOO LA Louisiana Power & Light Co. PWR Ebls ý 3390 T 075 C-E MIOO Weds Bar I :ng Iiy IN 'Tennessee Valley Authority PWR W TVA 3411 1 165 SER Wells Bat M 2 ' VVlHI II L lpunng City iN ITerinesse Valley Authority PWR W WA 3411 1 165 DER WNP-1 Richland WA Washingt n Public Power Supply System PWR B&W JEAC 3760 IDER /Ikl). ) IL,__... WNi1P-2 Richland Washington Public Power Supply System PWR BWR GE Furns & 3323 1T g95 WNP-3 Sat-op [W2i Washington Public Power Supply System PWR lee mcn C-E Ebasw 3005 12242 DER Wo ff rook Burlington "KS Woll Creek Nuclear Operati g Corp. P#R B E R uB B echtel 3411 11 28 I I  Table 2-1. General Plant Data - Sorted by Plant Name (Continued) Reactor NSSS Architect/ Core Pow Reactor Plant City State Utility mWt Type Vendor Engineer PWR W Stone A 600 YarilRoen Row MA Yankee Atomic Electrc Co. Webster PWR W Sargent 3250 Zion I Zion IL Co-rnr,-eath Edison Co. & Lundy PWR W Sargent 3250 Zion