3.0 PROCESS INSTRUMENTATION AND BWR CONTROL

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Process Instrumentation & ControlCotlGeneral Electric Technoloev ManualPrcsIntueato3.0.43.0 PROCESS INSTRUMENTATION ANDCONTROL SYSTEMSThe systems discussed in this chapter are thosethat have to do with process instrumentation orprocess control. The control systems used to alterreactor core reactivity are discussed in Chapter 7.The process instrumentation and control systemsare shown in Figure 3.0-1 and include thefollowing systems: the Reactor VesselInstrumentation System, the Electro HydraulicControl System, and the Feedwater ControlSystem.3.0.1Composite BWR Control SystemsFigure 3.0-2 shows a composite drawing of BWRcontrol systems. In addition to the ElectroHydraulic Control System and Feedwater ControlSystem, which are discussed in this chapter, someChapter 7 reactivity control systems are alsoshown to assist in understanding overall plantresponse. The three Chapter 7 systems included inthe drawing are the Reactor Manual ControlSystem, Recirculation Flow Control System, andReactor Protection System.Reactor Vessel InstrumentationSystem (Section 3.1)The Reactor Vessel Instrumentation Systemprovides information concerning reactor vesselwater level, reactor vessel pressure, reactor vesseltemperature, and core flow rate. This informationis used for control and automatic trip functions.3.0.2Electro Hydraulic Control System(Section 3.2)The Electro Hydraulic Control System maintainsa constant reactor pressure for a given reactorpower level, controls the speed and load on theturbine generator, and provides protection for themain turbine.3.0.3Feedwater Control System(Section 3.3)The Feedwater Control System regulates the flowof feedwater to the reactor vessel in order tomaintain reactor water level. The FeedwaterControl System measures and uses total steamflow, total feedwater flow, and reactor vessel levelsignals to carry out its function.CenterTraining CenterUSNRC Technical Training3.0-13.0-1Rev 0300Rev 0300

:. Steam Flow.S.i.a. SignalFA7I---VeeVesselPressure,"1.-,!MSIV'sStop- ---I.J-ValvesIPT.E. Feed FlowSignalsBypassiH .HValves 1p,.ControlValves-IIVesselTl . .Level Lv""rb"nSignalsGeneratorCondenser &Hotwell--CoreFlowSignalT ineralnzersCondensateBooster PumpCondensatePumpLoPres sndlureLF-ed es ueFeed HeatersHigh Pressure(.Vessel Level :u.TotaSt el veaFlow"Feedwater ControlTotal Steam Flow : ., System (FWCS)Total Feed Flow .JReactor FeedPumpHO00F-igure 3.u-1 Process Instrumentation & Control Systems

Electro HydraulicControl (EHC) System3-2TFA-0]HOMSIV'sISIBypass IHVlaves T---11- -iStop-IValvesH-tr ControlValvesITurGeneratorICoreSCondenser &HotwellRecirculatlonPuIIGeneratorControl RodDrivesFluidw MCBsFeed44 ,IICondensateCeCondensatePumpmReactor ureFeed HeatersIReactorProtection System(RPS)Reactor ManualControl ytSystem (FWC3(20Figure 3.0-2 BWR Control Systems

Reactor Vessel Instrumentation SystemGeneral Electric Technology ManualTable of Contents.13.1.1 System Description .3.1.2 Component Description .3.1.2.1 Reactor Vessel Water Level Instrumentation .3.1.2.2 Reactor Vessel Pressure .3.1.2.3 Core Flow Instrumentation .3.1.2.4 Reactor Vessel Temperature Instrumentation .3.1.3 System Features .3.1.3.1 Bases for Level Setpoints .3.1.3.1.1 High Level Trip .3.1.3.1.2 High Level Alarm .3.1.3.1.3 Normal Operating Level .3.1.3.1.4 Low Level Alarm .3.1.3.1.5 Reactor Level Low .3.1.3.1.6 Reactor Level Low-Low .3.1.3.1.7 Reactor Level Low-LowLow .3.1.3.2 Bases for Reactor Pressure Setpoints .3.1.4 System Interfaces .3.1.5 Summ ary .1112222223333344443.1 REACTOR VESSEL INSTRUMENTATION SYSTEMList of TablesReactor Vessel Level Setpoints and FunctionsReactor Vessel Pressure Setpoints and FunctionsTable 3.1-1 .Table 3.1-2 .List of FiguresReactor Vessel Level Instrumentation RangesCore Flow Summing NetworkFigure 3.1-1 .Figure 3.1-2 .CenterTraining CenterTechnical TrainingUSNRC Technical3.1-i3.1-iRev 0102Rev 0102

Reactor Vessel Instrumentation SystemGeneral Electric Technology ManualRn3.1 REACTOR VESSELINSTRUMENTATION SYSTEMSystem(FWCS) and the Emergency Core CoolingSystems (ECCS).The purpose of the Reactor VesselInstrumentation System is to provide sufficientinformation concerning reactor vessel water level,reactor vessel pressure, reactor vessel temperature,and core flow rate to allow safe plant operation.3.1.2The functional classification of the Reactor VesselInstrumentation System is that of a safety relatedsystem, although some portions are strictly forpower generation.3.1.2.1 Reactor Vessel Water LevelInstrumentation3.1.1The major components of the Reactor VesselInstrumentation System are discussed in theparagraphs which follow.Reactor vessel water level is obtained throughsensors which compare the weight of water in areference column to the height (weight) of thewater in the reactor downcomer annulus.Condensing chambers, external to the reactorvessel, are used to provide a constant referencecolumn of water.System DescriptionReactor vessel instrumentation consists of severalindividual subsystems that monitor reactorparameters such as water level, pressure, flow andtemperature.Reactor vessel water level is measured in thereactor vessel downcomer annulus. Thisparameter is monitored and displayed foroperation on four different ranges.Reactor vessel pressure is measured in the vesselsteam space and displayed to aid the operator insafe plant operation. Both narrow and wide rangepressure indications are provided for normal plantoperation and for full range pressure coverage.The plant power output capability should beproportional to the ability to remove the heatgenerated, so accurate core coolant flow-measurements are required to evaluate corethermal behavior. Since the total flow that passesthrough the core must also pass through the jetpumps, the flow through each jet pump ismeasured and summed to yield total core flow.Reactor vessel instrumentation supplies neededinformation to several systems such as the ReactorProtection System (RPS), Primary ContainmentIsolation System (PCIS), Feedwater ControlCenterUSNRC TechnicalTechnical Training CenterComponent DescriptionDuring normal reactor operation, reactor waterlevel is maintained approximately 17 feet abovethe top of the active fuel (Figure 3.1-1).Maintaining an acceptable water level in thereactor vessel ensures that a sufficient quantity ofreactor coolant is available to dissipate the heatgenerated by the core and the reactor is operatingwithin the initial conditions assumed for thevarious analyzed accidents.The level sensors, most of which indicate locally,are located throughout the reactor building atinstrument racks. From the instrument racks, thelevel is transmitted to nine separate reactor vesselwater level indicators in the control room andvarious trip circuits as shown on Figure 3.1-1 andTable 3.1-1.3.1.2.2 Reactor Vessel PressureReactor Vessel Pressure is sensed in the steamdome area using the same instrument piping thatexists for vessel level instrumentation. The reactorvessel pressure instruments contain numerous3.1-13.1-1Rev 0102Rev 0102

General Electric TechnoloL,v ManualGeeaEeticTcnloaReactor Vessel Instrumentation SystemMaulRatrVseIsrmnainSsepressure transmitters (PT), pressure switches (PS)and pressure indicators (PI). A summary ofreactor pressure trips is shown in Table 3.1-2.temperature monitoring system is designed to maptemperature gradients during startup andshutdown conditions. The data is recorded by amultipoint recorder and a two pen recorder,providing the basis to establish the rate of heatingor cooling performed on the vessel.3.1.2.3 Core Flow InstrumentationTo evaluate reactor core power level and corethermal characteristics, accurate core flowmeasurements are required. Since all core flow,except control rod drive cooling water, must passthrough the jet pumps, the flow through each jetpump is measured and summed to yield total coreflow.3.1.3System FeaturesA short discussion of the system features is givenin the paragraphs which follow.3.1.3.1 Bases for Level SetpointsAll 20 jet pumps have a pressure tap on the pumpthroat which is compared to core inlet plenumarea pressure. The square root of this differentialpressure provides a signal representing jet pumpflow. As indicated on Figure 3.1-2, five jet pumpflow signals are summed and then added toanother five jet pump flow signals to yield loopflow. The two loop flow signals are then summedto yield total core flow.During normal plant operation with bothrecirculation pumps operating, the loop flows aresimply added together in a summation network.However, if one recirculation pump is off and theother is operating, the inactive loop will havereverse flow. The jet pump flow transmitters arenot capable of distinguishing the direction of flowthrough the jet pumps. As a result, a relay logicsystem senses recirculation pump status tosubtract the idle loop flow from the operating loopflow to yield an accurate total core flow outputsignal.Vessel level instrumentation used to initiate safetysystems, cause operational trips, and providecontrol system inputs, are listed in Table 3.1-1.The reactor vessel water level trip setpoints arereferred to as numbered levels. These levels andtheir elevation referenced to instrument zero are:Level 1 (-132.5"), Level 2 (-38"), Level 3( 12.5"), Level 4 ( 33.5"), Level 5(approximately 37"), Level 7 ( 40.5"), and Level8 ( 56.5"). The bases for the various levelsetpoints are discussed in the paragraphs whichfollow.3.1.3.1.13.1.2.4 Reactor Vessel TemperatureInstrumentationThe trip of the main turbine is to protect it againstthe occurrence of gross carryover of moisture andsubsequent damage to the turbine bladeing. Thereactor feed pump turbines are tripped to preventoverfilling the reactor vessel. The reactor coreisolation cooling and the high pressure coolantinjection turbines are tripped, in the event thesesystems have activated, to prevent flooding ofsteam lines.3.1.3.1.2The reactor vessel metal temperature is measuredand monitored to provide temperature datarepresentative of thick, thin, penetration, andtransitional sections of the vessel. TheCenterTraining CenterTechnical TrainingUSNRC Technical3.1-23.1-2Level 8 ( 56")Level 7 Alarm ( 40")While operating at full power, the high levelalarm annunciates at the reactor vessel water levelRev 0102Rev 0102

Reactor Vessel Instrumentation SystemtGeneral Electric Technolo.v ManualRetViabove which moisture carryover in the steam isexpected to increase at a significant rate. Thealarm warns the operator of this undesirablecondition.3.1.3.1.3Normal Operating Level ( 37")Reactor vessel water level can be controlled at anypoint between the high and low level alarms.However, the Feedwater Control System isusually set to maintain vessel level at 37 inches.3.1.3.1.4Level 4 Alarm Trip ( 33")The low water level alarm annunciates at thereactor vessel water level below which steamcarryunder in the water will begin affecting thereactor recirculation flow rate significantly at fullpower because of recirculation pump cavitation. Awater decrease to this point, coincident with areactor feed pump trip, causes the recirculationpumps ( Section 7.2) to runback to apredetermined speed to reduce thermal poweroutput within the capacity of the remainingreactor feed pump(s).3.1.3.1.5Level 3 Trip ( 12")The low level scram function is for protectionagainst high moisture carryover because of steambypassing the dryer under the seal skirt. Thescram occurs while the water level is above thebottom of the dryer seal skirt. The level selectionalso results in a quantity of reserve coolantbetween this level and the top of the active fuel toaccount for evaporation (decay heat boil off)losses, steam void collapse, and other coolantlosses from the reactor vessel following a loss offeedwater flow, without the vessel water leveldecreasing to -132.5", which would initiate theEmergency Core Cooling Systems. This selectedquantity of reserve coolant assumes the ReactorCore Isolation Cooling (RCIC) System isproviding design flow rate. A decrease of reactorCenterTraining CenterUSNRC TechnicalTechnical Trainingvessel inventory to this level also causes actuationof the Primary Containment Isolation System.3.1.3.1.6Level 2 Trip (-38")This setpoint is selected to be low enough so thatthe RCIC and High Pressure Coolant Injection(HPCI) Systems will not be initiated on low levelafter a reactor scram unless feedwater flow hasbeen terminated. The setpoint accounts for theexpected level decrease caused by steam voidcollapse which occurs following any scram. Thesetpoint is selected high enough so that the RCICSystem design flow is sufficient, taking intoaccount system startup time following a loss offeedwater flow, to recover reactor vessel waterlevel and prevent a level decrease to 132.5" witha subsequent initiation of emergency systems. Thevarious system isolations are to prevent or limitthe loss of reactor coolant and the release ofradioactive products to the atmosphere assumingthat the vessel water level decrease was due to aleak from one or more of the effected systems.The recirculation pumps are tripped to insertnegative reactivity using subsequent voidformation, in the unlikely event that the reactordid not scram on a reactor vessel low water levelsignal. This even is referred to as an anticipatedtransient without scram - recirculation pump trip(ATWS-RPT).3.1.3.1.7Level 1 Trip (-132")3.1.33.1-3This level setpoint is selected to be high enoughabove the top of active fuel to initiate the ECCSthus providing adequate time for the-ECCS tofunction in' the event of a Loss of CoolantAccident (LOCA) to provide adequate corecooling and prevent fuel damage.Rev 0102Rev 0102

Reactor Vessel Instrumentation SystemGeneral Electric Technology ManualGeea lcrcTcnIorMaulRatrVseIsrmnainSse3.1.3.2 Bases for Reactor Pressure Setpoints3.1.4A summary of reactor pressure trips is given inTable 3.1-2 and discussed in the paragraphs whichfollow.The interfaces this system has with other plantsystems are listed in Tables 3.1-1 and 3.1-2.System Interfaces3.1.5 SummaryA reactor pressure of 1120 psig trips therecirculation pumps to insert negative reactivityby means of void formation, assuming the reactorfailed to scram on high pressure. This event isreferred to as an anticipated transient withoutscram - recirculation pump trip (ATWS-RPT).The reactor scram setpoint (1043 psig) preventsreactor vessel overpressurization and, inconjunction with safety/relief valve operation,provides sufficient margin to the maximumallowable reactor coolant boundary pressure.Also, when not in the RUN mode, and below thissetpoint, the scrams from Main Steam IsolationValve closure and low condenser vacuum arebypassed to allow operation in Hot Standby atnormal reactor pressure.The high pressure alarm (1025 psig) alerts theoperator to abnormal system pressure.Classification:Safety related systemPurpose:To provide sufficient information concerningreactor vessel water level, reactor vesselpressure, reactor vessel temperature, and coreflow rate, to allow for proper plant operation.Components:Reactor vessel level instrumentation; reactorvessel pressure instrumentation; reactor vesseltemperature instrumentation; and core flowrate.System Interfaces:Reactor Vessel System; Recirculation System;Main Steam System; Reactor Core IsolationCooling System; Feedwater Control System;Recirculation Flow Control System; allEmergency Core Cooling Systems.Water injection by the Core Spray (CS) andResidual Heat Removal (RHR) Systems isdelayed until reactor vessel pressure is reduced to465 and 338 psig to prevent reverse flow andoverpressurization of these emergency corecooling systems.As a part of the RHR initiation logic, therecirculation pump discharge valves close whenpressure decreases to 310 psig to ensure RHRwater enters the reactor vessel on a recirculationsuction line LOCA. Delaying valve closure tothis pressure will ensure the valve will be able toclose since it is designed to close with a maximumdifferential pressure of 200 psi.CenterUSNRC Technical Training Center3.1-43.1-4Rev 0102Rev 0102

(nPr alElectric Technolo ManualRecoVselntrmtainSsmReactor Vessel Instrumentation SystemTABLE 3.1-1 REACTOR VESSEL LEVELSETPOINTS AND FUNCTIONSFUNCTIONSLEVEL 56"High Level Trips Main Turbine and Reactor Feed Pump Turbines. ClosesRCIC and HPCI Steam Supply Shutoff Valves. 40"High Level Alarm 37"Normal Operating Level 33"Low Level Alarm, RFC Runback 12"Reactor Scram, Primary Containment Isolation, Start Standby Gas Treatment System, ADS Permissive-38"Initiate HPCI, RCIC Systems, Recirculation Pump Trip (ATWS-RPT)-132"Initiate Core Spray, Residual Heat Removal, ADS and Diesel Generators,and Main Steam Line IsolationUSNRC Technical TrainingTraining CenterCenter3.1-53.1-5zevuiuRe y U01,2

General Electric Technology ManualGeeaEeticTcnloReactor Vessel Instrumentation SystemMaulRatrVseUSNRC Technical Training CenterIsrmnainSse3.1-6Rev 0102

Reactor Vessel Instrumentation SystemGeneral Electric Technolo.v ManualeTABLE 3.1-2 REACTOR VESSEL PRESSURESETPOINTS AND FUNCTIONSFunctionsPressure Setpoint1120 psigTrip Recirculation Pump1043 psigReactor Scram1025 psigHigh Pressure Alarm465 psig & 338 psigCore Spray, RHR Initiation & Valve Interlocks310 psigRecirculation Pump Discharge Valve Closureon LPCI Initiation Signal125 psigRHR Isolation (Shutdown Cooling Mode)CenterTraining CenterUSNRC TechnicalTechnical Training3.1-73.1-7RevU1UZRev 0102

General Electric Technoloev ManualGeneral.Ele.ReactorCenterTechnical Training CenterUSNRC TechnicalVessel Instrumentation S stem3.1-83.1-8Rev 0102Rev 0102

Head Flange 723"E"200-psetRange-1G);uCmMain Steam Nozzle 640"I'l.o.RLevel #8 573"veIl 554"NormalSLevel#4529"30 (0)Instrument "0" 517"(D:;.CoreSpray Nozzle 484"Feedwater Nozzle 483"CD479"Level !#2 #249(Typ 3)0FuelZone.75Range125-10'Level #1 384"Top of Active Fuel 358"2/3 Core Covered 303"CD0Bottom of Active Fuel 208"M-eewter ozzle, 483"CD390160550Level #3r-"(Narrowa g-a WidegJet Pump Instrument Tap 132"SCL Nozzel and Instrument Tap 94"Inside Bottom Head "0"-SO(Typ 2)mmfn

JET PUMPFLOWS23 ---45 -- ,-,-.LOOP AFLOWLOOP BFLOWFlFlJET PUMPFLOWS12E-I113ap"lB171878101415"3-20CLOSED IF ONE PUMP ISRUNNING WHILE THEOTHER IS TRIPPED;OTHERWISE OPENCLOSED IF BOTH PUMPSARE RUNNING OR IF BOTHPUMPS ARE TRIPPED;OTHERWISE OPENTOTAL COREFLOW RECORDER00Figure 3.1-2 CORE FLOW SUMMING NETWORK

EHC Control SystemGeneral Electric Technology ManualTable of Contents3.2 ELECTRO HYDRAULIC CONTROL SYSTEM .3.2.1 System Description .3.2.2 Component Description .3.2.2.1 Pressure Control Unit .3.2.2.2 Load Control Unit .3.2.2.3 Speed Control Unit .3.2.2.4 Valve Control Unit .3.2.3 System Features .3.2.3.1 Normal Operation .3.2.3.2 Plant Shutdown and Cooldown .3.2.4 System Interrelations .3.2.4.1 M ain Steam System .3.2.4.2 Recirculation Flow Control System .3.2.5 Summ ary .1111.2223334444List of FiguresPressure Control BandElectro Hydraulic Control System LogicFigure 3.2-1 .Figure 3.2-2 .CenterTraining CenterUSNRC Technical Training3.2-i3.2-iRev 0102Rev 0102

EHC Control SystemECCnrlSseGepneral Elec tric Technolov ManualSystem requires reactor power to be changed first,followed by a change in turbine generator output.An increase in reactor power causes an increase inboth reactor vessel and turbine throttle pressure.The pressure increase is due to increased "heatgeneration by the reactor core producing moresteam without a subsequent increase in steam flowrate. The throttle pressure increase is sensed bythe pressure control system and the pressurecontrol system signals the turbine control valvesand/or bypass valves to open wider,accommodating the increased steam production.This increase in turbine steam flow compensatesfor the reactor vessel pressure rise.3.2 ELECTRO HYDRAULIC CONTROLSYSTEMThe purposes of the Electro Hydraulic Control(EHC) System are to:1. Provide normal reactor pressure control bycontrolling the steam flow consistent withreactor power.2. Provide the ability to conduct a plantcooldown.The functional classification of the EHC Systemis that of a power generation system.3.2.1Reducing reactor power decreases reactor vesselpressure and turbine throttle pressure. The controlsystem responds to the decrease in throttlepressure by throttling the turbine control valvesand/or bypass valves in the closed direction,decreasing turbine steam flow. Reducing steamflow stops the steam pressure decrease and lowersgenerator output. Using this control system, theturbine follows or is slaved to the reactor.System DescriptionPressure changes in a direct cycle boiling waterreactor can have a pronounced effect on reactorpower. If pressure is increased in a BWR duringpower operation, steam voids, which havesignificant reactivity effects on the core duringpower operation, collapse, increasing coremoderator content. This increases neutronmoderation resulting in more thermal neutronsbeing available, more fissions occuring and,increasing reactor power. As reactor powerincreases, pressure tends to increase even further,and a snowball effect is produced.3.2.2Only the major EHC logic sections are discussedin the paragraphs which follow.3.2.2.1 Pressure Control UnitIf reactor vessel pressure decreases, some of themoderator flashes to steam because the reactorvessel is in a saturated state. This flashingincreases the void content in the reactor coreresulting in more neutron leakage, fewer fissions,and a reduction in reactor power. The powerreduction tends to decrease reactor pressure evenfurther.Because of the effects mentioned above, apressure control system, the Electro-HydraulicControl System (EHC), was developed. The EHCCenterTechnical Training CenterUSNRC TechnicalComponent DescriptionThe pressure control unit consists of two pressureregulators and pressure-percent steam flowconverters (Figure 3.2-2).The pressure regulators are the proportional typewhich require a 30 psi difference between turbineinlet pressure and the pressure setpoint (pressureerror) to open the control valves to the 100 percentposition. Therefore, the pressure at the turbineinlet varies 30 psi from 0 percent power to fullpower, or 3.33% flow/psi.3.2-13.2-1KY UIUhRev 0102

General Electric Technoloev ManualGeeaEeticTcnlogEHC Control SystemMaulECCnrlSseThis effect is shown in Figure 3.2-1. Also shownin this figure is a curve for reactor vessel pressure.The curve is not linear primarily because ofpressure drops across the flow restrictors,MSIV's, and steam line piping which areproportional to the flow squared. The relationshipbetween pressure error and steam flow wasdetermined by experimentation and given a rapidresponse which is relatively stable.The two pressure regulators compare the turbineinlet or throttle pressure with the pressuresetpoint, normally set at 920 psi, and generate asteam flow demand based on the error. Thecontrolling regulator is selected by applying 10psi bias to one of the regulators. The regulatorcontaining the 10 psi is called the backupregulator and will take over if the primaryregulator output fails downscale (decreasingvalve position demand).At 100% power conditions, pressure inputs toboth regulators would consist of the 920 psigpressure setpoint and a throttle pressure of 950psig. The primary pressure regulator output wouldbe 30 psi error; with the backup regulator havingan output of 20 psi. The 30 psi signal is convertedto 100% steam flow demand in the pressure-steamflow converter which is fed to a high value gate(HVG). Both regulator outputs are transmitted tothe HVG, which allows the highest input signal topass. The highest signal is then fed to a demandcircuit and the pressure/ load low value gate(LVG).3.2.2.2 Load Control UnitThe major part of the load control unit is the loadset motor. The load set motor is used to set thedesired maximum load. Once the load set motorhas been adjusted up or down, the load set valuewill remain constant. The operator controls theposition of the load set motor by using the loadCenterTraining CenterUSNRC TechnicalTechnical Trainingselector increase or decrease pushbuttons. Theoutput of the load set motor is summed with thespeed control unit output and transmitted to thepressure/load LVG. The load set motor is usuallyadjusted to yield a 100% output signal.3.2.2.3 Speed Control UnitThe speed control unit consist of two separatespeed/acceleration controllers. Eachspeed/acceleration controller receives a speedsignal from a turbine shaft speed pickup unit andcompares it to an operator selected speed toproduce a speed error signal. The shaft speedsignal is differentiated to produce an accelerationsignal which is compared to an operator selectedacceleration reference signal. The resultingacceleration error signal is then integrated toproduce an equivalent speed error signal. Thelowest value of the two speed errors and the twointegrated acceleration errors is selected andsummed with the load set signal. The speed setpoint and acceleration set point are selected by theoperator using pushbuttons.3.2.2.4 Valve Control UnitThe valve control unit establishes the steam flowdemand signals to the control valves and/or thebypass valves. Within the valve control unit is thepressure/load low value gate (LVG) whichreceives the output from the pressure control unit,the load limit and the combined output from thespeed control unit and the load control unit.The load limit value establishes the maximumamount of rated reactor steam flow which isallowed to go to the turbine and is normally set at100%. The value of the load limit is determinedby a manually adjusted potentiometer. Themaximum combined flow limits the total steamflow that can be passed by the control and bypassvalves. The maximum combined flow signal is3.2-23.2-2Rev 0102Rev 0102

EHC Control SystemECCnrlSseGeneral Electric Technolopv Manualoperator starts reducing recirculation flow toreduce core flow. As core flow is decreased, moreboiling occurs in the core which causes reactorpower and the steam generation rate to decrease.With the control valves still passing 100% steamflow and the reactor producing less that 100%steam flow, reactor pressure decreases. Thedecrease in reactor pressure causes a decrease inturbine inlet pressure. As turbine inlet pressuredecreases the pressure error signal decreases,causing the control valves to begin closing.Finally at 944 psig turbine inlet pressure, thepressure error is reduced to 24 psig which calls for80% steam flow (24 psig x 3.33/1 80%).normally set at 125 % by the control room operatorusing a manual potentiometer.The output of the pressure/load LVG is the controlvalve demand. This LVG is the device whichmakes the turbine a slave to the reactor. Thebypass valve demand is established by subtractingthe pressure/load LVG output from the pressurecontrol unit output. Any difference passes throughthe circuit as a bypass valve demand.3.2.3System FeaturesA short discussion of the system features is givenin the paragraphs which follow.3.2.3.2 Plant Shutdown and Cooldown3.2.3.1 Normal OperationDuring a plant shutdown, reactor power isdecreased to a point that steam production iswithin the capacity of the bypass valves to pass,usually about 10% generator load, and then theturbine is tripped by the operator. The turbinestop, control, and combined intermediate valvesall trip closed. The bypass valves open inresponse to the signal generated from the bypasscontrol summer. With a " 10" signal beingsummed with the "0" signal from the control valvedemand, a " 10" steam flow signal is generatedand transmitted to the bypass valves. The bypassvalves then control reactor pressure during thepower decrease.The following parameter and controller setpointvalues are listed for a reactor power output of100%.3283 MWtReactor power880 MWeGenerator power1010 psigReactor pressure950 psigpressureTurbine inlet920 psigPressure setpoint30 psigPressure error115%Max. Combine flow limit100%Load limitrpm1800speedTurbine965 MWeLoad selector(110% of rated)The turbine stop valves are open fully, and thecontrol valves are positioned to pass 100% turbinesteam

3.0 PROCESS INSTRUMENTATION AND CONTROL SYSTEMS The systems discussed in this chapter are those that have to do with process instrumentation or process control. The control systems used to alter reactor core reactivity are discussed in Chapter 7. The process instrumentation and control

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