ML19319D684

Chapter 7 to Crystal River 3 & 4 PSAR, Instrumentation & Control. Includes Revisions 1-10

ML19319D684
Person / Time
Site:	Crystal River, 05000303
Issue date:	08/10/1967
From:	FLORIDA POWER CORP.
To:
References
NUDOCS 8003240664
Download: ML19319D684 (59)

Text

{{#Wiki_filter:_ _ - -, _ ___ _ _ _ _ _ _ O O

                                  $/

5 E_ j

                                  ?

O 0118 O sooss:a6fy

O

 \_ ,/                                     TABLE OF CONTENTS Section                                                               Pace 7             INSTPU?'E7TATION AND CONTROL 7-1 7.1                PPOTFOTICU 9Y3Tr'"3     -

7-1 7.1.1 DESIGI BASES 7-1 7.1.1.1 Vital Functicns 7-1 7.1.1.2 Princinles of Desien 7-2 7.1.1 3 Functional Pecuirements 7-3 7.1.1.4 Environmental Considerations T-h 7 1.2 SYSTEM DESIGN 7-4 7.1.2.1 System Descriution - Reactor Protection 67 stem 7-h 7 1.2.2- Descrintion - Safecuards Actuation System 7-6 7.1.2.3 Desien Features 7-7 7 1.2.4 Summarv of Drotective Actions 7-10 7.1.2.5 Felationshin to Rafetv Limits 7-11 7 1.3 SYSTEMS EVALUATION 7-11 7.1.3.1 Funct2cnal Canability Peactor Protsetton System 7-11 7.1.3.2 Functional Canability - Safecuards j Actuation System 7-12a 3 7 1.3.3 Precterational Ter ;-_ 7-13 7 1 3.h Comoonent Failure Considerations 7-13 7.1.3 5 Ocerational Tests 7-14 72 REGULATING SYSTFMS 7-15

7.2.1 DESIGN BASES 7-15 (Deleted) 7

 \J Of19 l                                                       7-1 (Revised 7-15-69) r

CONTENTS (Cent'd) O Section Pm 7 2.2 ROD DRIVE CONTROL SYSTEM 7-15 7 7 2.2.1 Desien Basis 7-15 7.2.2.2 System Desien 7-16 7 2.2 3 System Evaluation 7-19 7.2.3 INTEGRATED CONTROL SYSTEM 7-21 7.2.3.1 Turbine Control 7-22 7.2.3.2 Steam Generator Control T-23 7.2.3.3 Reactor Control 7-2h 7 2.3.h System Failure Considerations 7-2ha h 7.2 3 5 Interlocking 7-2ha 7 2.3.6 Loss of Load Considerations 7-2ha 7.3 INSTRUVENTATION 7-25 O' 7.3.1 NUCLE R INSTRUMENTATION 7-25 7.3 1.1 Desien 7-25 7.3.1.2 Evaluation 7-26 7 3.2 NONNUCLEAR PROCESS INSTRUMENTATION 7-27 7.3.2.1 System Design 7-27 7.3.2.2 System Evaluation 7-28 7.3.3 INCORE MONITORING SYSTEM 7-28 7 3.3.1 Design Basis 7-28 7332 System Desien 7-29 7.3.3.3 System Evaluation 7-30 7.h OPERATING CONTROL STATIONS 7-31 7.h.1 GENERAL LAYOUT 7-31 7.h.2 INFORMATION DISPLAY AND CONTROL FUNCTION 7-31 a i (~~! v 7.h.3

SUMMARY

OF ALARMS 7-31 7.h.h COMMUNICATION 7-h h 7-11 (Revised 7-15-69) l L m

Co:iTE:iTS (Cont'd) Section Page 743 OCCUPANCY 7-32 7.h.6 AUXILIARY CONTROL STATIONS 7-33 747 SAFETY FEARSES 7-33 i O

                                                                                                                    ;

l

                                                                                                                    ;

i 00b21 i O 7-111 1

LIST OF FIGURES L-- ( At rear of Section)

-

Figure No. Title 7-l' Rea:thr Protection System 31ock Diagram

               -T-2a             Nuclear Instrumentation System                                              7
               .7-2b             Reacter Prctectica System 7-2c:           Safeguards Actuation System 7-3             Typica'. Centrol Circuits for Engineered Safeguards
                               . Equipment 7h              Reactor Power Measurement Errors and Control Limits

, 7-5 Reactor and Steam Temperatures versus Reactor Power

               '7-6             Reactor Control Diagram - Integrated Control System i

7-7 Automatic Control Rod Groups - Typical Worth Curve versus Distance Withdrsvn 7-8 Steam Generator and Turbine Control Diagram - l- , s_ Integrated Control System

7-9 Nuclear Instrumentation Flux Ranges 7-10 Nuclear Instrumentation Detector Locations 7 Nonnuclear Instrumentation Schematic 7-12 Incore Detector Locations 7-13 Typical Arrangement - Incore Instru=entation Channel k

2 0122 7-iv (Revised 7-15-69)

    ,       ,         ,.____.n                 ._                               , . - - < -   - - - - - - **   -~' "

7 INETR"ME WAT!ON AND CONTROL t s 71 PROTECTICN SYSI"!MS The protection systems, which concist of the Reactor Protection System and the Safeguards ' Actuation System, perform the most important control and safety lanc - tions. LThe protection systems extend from the sensing instruments to the final actuating devicec, such as trip circuit breakers and pump or valve motor contac-tors. 7 1.1 DECIGN EASEE The Reactor Protection System =onitors p1rameters related to safe operation and trips the reacter to protect the reactor core againut fuel rod cladding damage caused by departure from nucleate boiling (DNB), and to protect against- reactor coolant system damage caused by high r,ystem prescure. The Safeguards Actuation System monitors parameters to detect failure of the reactor coolant system and initiates reactor building isolation and engineered safeguards operation to con-4 tain radioactive fission products in the reactor building. 7 1.1.1 Vital Functions The Reactor Protection System automatically trips the reactor to protect the reactor core under these conditions:

a. When the reactor power, as measured by neutron flux, exceeds the limit set by the reactor coolant flow.

() f_s (The reactor coolant flow is deter-mined by the number of operating reactor coolant pumps. )

b. Loss of all reactor coolant pumps.

c. The -reactor outlet temperature reaches an established maximum limit.

d. The reactor pressure reaches an established minimum limit.

The Reactor Protection Cystem automatically trips the reactor to protect the reactor coolant system under this condition:

a. The reactor pressure reaches an established maximum limit.

The Cafeguards Actuation Eystem automatically performs the folleving vital fune-tions:

a. Commands operatien of injection emergency core coolant.

b. Commands operation of the reactor building emergency cooling system and the:reacter building spray system.

c. Ccamands closing of the reactor building isolation valves.

qQ 0123 7-1

The core flooding system is a passive system and does not require Safeguards Actuation System action. 7..l.1.1.1 Nonvital Functions The Reactor Protection System provides an anticipatory reactor trip when the reactc,r startup rate reaches specified limits. 7.1.1.2 Principles of Design The protection systems are dcsigned to meet the requirements of the IEEE pro-posed " Standard for Nuclear Power Plant Protection Systems," dated September 13, 1966. Prototype and final equipment vill be subject to qualification tests as required by the subject standard. The tests vill establish the adequacy of equipment performance in both normal and accident environments. The major design criteria are summarized in the following paragraphs. 7.1.1.2.1 Single Failure

a. No single component failure shall prevent the protection systems from fulfilling their protective functions when action is required.

b. No single component failure shall initiate unnecessary protection system. action, provided implementation does not conflict with the criterion above.

7.1.1.2.2 Redundancy All protection system functions shall be implemented by redundant sensors, in- O strument strings, logic, and action devices that combine to for= the protection channels. 7.1.1.2.3 Independence Redundant protection channels and their associated elements will be electrically independent and packaged to provide physical separation. Separate detectors and instrument strings are not, in general, employed for pro-tection system functions and regulation or control. Sharing instrumentation for protection and control functions is accomplished within the framework of the stated criteria by the employment of isolation amplifiers in each of the multiple outputs of the analog protection system instrument strings. The isolation amplifiers are precision operational amplifiers having a closed loop unity gain and a low dynamic output impedance. The effectiveness of the 3 isolation amplifiers has been proven by actual test. The isolation amplifiers will block a direct connection across their cutput of L10 vde or peak ac, 300 v rms without perturbing the input signal. This may be stated as a corollary to the design criteria: "a direct short, open circuit, ground fault, faulting to a power source of less than L10 volts, er 7-2 (Revised 3-lh-68) O 012ft

bridging of'any two points at the output ter~inals of a protection system analog 3

              ~ instrument string having multiple outputs shall not result in a significant 4
                ' disturbance within more than one output."

Testing has demonstrated that tne protection system design vill meet the above criteria. ~ 7.1.1.2.h Loss of Power

a. A loss of power in the Reactor Protection System shall cause the af-fected channel to trip.

i 4 i 4 r 1 1 O O o125

i

7-2a (Revised 3-14-68)~

  '~'
          ~,            .,                   _ _ . . .               _.    , , . , . . . . .
                                                                                              . ,.,_-,.,......,,_..,._,,....m._

                            ~

f i As b. Availability of power to the Safeguards Actuation System shall be con-tinuously indicated. The loss of instrument power, i.e., 120v a-c Ecsential Services bua power, to the instrument strings and bistables vill initiate a trip in the affected channels. System actuation re-quires control power frcm one of the two engineered safeguards de power busses so that loss of this power does nct actuate the system. The system equipment is divided between the redundant engineered safe-guards channels in such a way that the loss of one of the de power busses does not inhibit the system's intended safeguards functions. 71.1.25 Manual Trip Each protection sy tem chall have a man'ual actuating cwitch or cwitches in the control room which chall be independent of the automatic trip instrumentation. The manual switch and circuitry chall be cimple, direct-acting, and electrically connected close to the final actuating device. 7.1.1.2.6 Equipment Removal 4 The Reactor Protection System shall initiate a trip of the channel involved when

modules, equipment, or cubausemblies are removed. Safeguards Actuation System i channele chall be designed to provide for servicing a single channel without affecting integrity of the other redundant channels or without compromising the criterion that no single failure shall prevent actuation.

  -%pA    7.1.1.27         Testing Manual-testing facilities shall be built into the protection systems to provide for

a. Preoperational testing to give assurance that the protection systems can fulfill their required functions.

b. On-line testing to prove operability and to demonstrate reliability.

7.1.1 3 Functional Requirements The functional requirements of the protection systems are those specified under vital functions together with interlocking functions. The functional requirements of the Reactor Protection System are to trip the re-actor under the following conditions: 1 a. The reactor power, as measured by neutron flux, reaches an allowable limit set by the number of operating reactor coolart pumps,

b. The losslof all reactor coolant pumps.

c. The reactor outlet temperature reaches a preset maximum limit.

d. The reactor coolant pressure reaches a preset maxi =um limit.  !

(O_) . e. The reactor coolant pressure reaches a preset minimum limit. 0126 7-3 -

f. The reactor startup rate reaches a maximum limit while operating below a preset power level. h Interlocking functions of the Reactor Protection System are to

a. Bypass the startup rate trip when the reactor power reaches a preset value.

b. Inhibit control rod withdrawal on the occurrence of a predetermined startup rate, slower than the rate at which reactor trip is initiated.

The functional requirements of the Safeguards Actuation System are to

a. Start operation of high pressure injection upon detection of a low reactor coolant system pressure.

b. Start operation of low pressure injection upon detection of a very low reactor coolant system pressure.

c. Operate the reactor building isolation valves upon detection of a moderately high reactor building pressure.

d. Start the reactor building emergency cooling units upon detection of a moderately high reactor building pressure.

e. Start the reactor building spray system upon detection of a high re-actor building pressure.

h 7 1.1.4 Environmental Considerations The operating environment for equipment within the reactor building will normally be controlled to less than approximately 110 F. The Reactor Frotection System instrumentation within the reactor building is designed for continuous operation in an environment of 140 F, 60 psig, and 100 per cent relative humidity, and will function with less accuracy at the accident temperature. The environment for the neutron detectors will be limited to 150 F with a rela-tive humidity of less than 90 per cent. The detectors are designed for contin-uous operation in an environment of 175 F, 90 per cent relative humidity, and 150 psig. The Safeguards Actuation System equipment inside the reactor building will be designed to operate under the accident environment of a steam-air mixture. Protective equipment outside of the reactor building, control. room, and relay room is designed for continuous operation in an ambient of 120 F and 90 per cent relative humidity. The control room and relay room ambients will be maintained at the personnel comfort level; however, protective equipment in the control room and relay room will operate within design tolerance up to an ambient tem-perature of 110 F. - 4 O 0127 7k

7 1.2 SYSTEM DESIG4 7.1.2.1 Sy3 ten Descrintion - Peactor Protection System The system as shown in Figures 7-1 and 7-2b consists of four identical 7 protective channels. Each channel controls one RS relay in four identical 2 out of h logics. Each channel forms an AND gate to energize its respective RS relays. Should any cne or more inputs to a given channel initiate a trio the RS relays associated with that channel vill de-energize. Thus , frcm a trip standpoint, each channel is an OR gate. Should any two of the four protection channels de-energize their re-spective RS relays all four of the 2 out of h logics vill opsn, tripping the control rod drive system circuit breakers whien in turn removes power from the drive notors permitting the rods to move into the core shutting devn the reactor. The control rod drive circuit breakers, Figure 3-66, form a logic which is just short of a full 2-out-of h coincidence. The specified breaker com-binations which initiate a reactor trip can best be stated in logic notation as: AB + ADF + BCE + CDEF. This is a 1 out of 2 loeie used twice and is referred to as a 1-out-of-2 X 2 logic. It should be noted that when any 2-out-of-b protective channels trip, all 2-out-of h logics trip, commanding all control rod drive breakers to trio. l' The undervoltage coils of the control rod drive breakers receive their power from the protective channel associated with each breaker. The manual reactor trip switch is interposed in series with each 2-out-of-h () logic and the assigned breakers undervoltage coil. The trip circuits _ and devices are redundant and independent. Each breaker is independent of each other breaker, so e single failure within one trip circuit does not affect any other trip circuit or prevent trip. By this arrangement, each breaker may be tested independently by the manual test switch. One segment of the nanual reactor trip switch is included in each of the circuit breaker trip circuits to implement the " direct action in the final device" criterion. The pover/ flow monitor logic details are shown on Figure 7-2b. There are four identical sets of pover/ flow monitor logic, one associated with each protection channel. Each set of logic received an independent total resctor coolant flow signal (IF), a " number of pump moters in operation" signal, and two isolated reactor power level signals ($). 0128 (D 7-5 (Revised 7-15-69)

                          , y     -                                                     -

O Paragraphs Deleted. 3 Using a flux / flow comnarator, one part of the power / flow monitor continuously compares the ratio of the reactor neutron power to the total reactor coolant flow. Should the reactor power as measured by the linear power range channels exceed 1.075 times the reactor coolant flow, a reactor trip is initiated. All measurements are in terms of per cent full flow or full (rated) power. The flux / flow comparator runs back the over flux trip level in step with a detected decreasing flow thus providing an opportunity for the control system to reduce the reactor power to an acceptable level without a reactor trip. The second element in the power / flow monitor is the pump monitor logic. The pump monitor logic counts the number of pump motors in operation as indicated by the number of closed pump power breakers and initiates a reactor trip if less than three pumps are in operation. 7.1.2.2 Descrirtion - Safecuards Actuation System Figure 7-2c shows the action initiating sensors, bistables, and logic for 3 the Safeguards Actuation System. The major differences between this system and the Reactor Protection System are:

a. Each protective action is initiated by either of two channels with 2-out-of-3 coincidence logic between input signals.

b. Either of the two channels is independently capable of initiating the desired protective action through redundant safeguards equipment.

c. Protective action is initiated by the application of power to the terminating control relays through the coincidence logic.

There are three inderendent sensors for each input variable. Each sensor ter-minates in a birtable device. Tha outputs of the three bistables associated with each variable are formed into two identical and indenendsnt 2-out-of-3 coincident logic networks or chanrels. Safeguards action is initiated when either of the channele associated with a variable becomes energized through 0129 O 7-6 (Revised 7-15-69)

the coincident trip action of the associated bistables. The engineered safe- , e- guards equipment is divided between redundant actuation channels as shown in (,g/. Figure-7-2C. The division of equipment between channels is based upon the re-dundancy of equipment and' functions. Where two active safeguards valves are connected in redundant manner, each valve vill be controlled by a separate en-gineered safeguards channel as shown in Figure 7-2C. When active and passive (check . valve) safeguards valves are us'ed redundantly, the active valve vill be equipped with ~two OR control elements, each driven by one of the safeguards channels. Redundant safeguards pumps vill be controlled in the same manner as redundant active valves. Figure 7-2C shows a typical control scheme for both safeguards, valves and pumps. Figure 7-3 shows typical control circuits for equipment serving safeguards functions. Each circuit provides for normal start-stop control by the Plant operator as well as automatic actuation. Normal starting and stopping are initiated by raomentary contact pushbuttons or control switches. The control circuit shown for a decay heat removal pump is typical of the con- 5 troller of a large pump started by switchgear. There are two decay heat re-moval pumps; they are equipped with single control relays powered from separate safeguards actuation channels. 4 E O f 4 1 1 f L. 7-6a (Revised k-8-68)

                                                                ,     ,        ,-,. ,m y,-y-,,m-,,w-- --y:- , -s-- m-   r

fS channels. Energizing the control relays through their associated safeguards (_) actuation channel, energizes the pump circuit breaker closing coil and starts the pump. The control circuit for a reactor building isolation valve is typical of a motor-operated valve which is required to close as its engineered safeguards action. If the valve is e= ployed as one of two active redundant valves, then it is controlled by a single safeguards actuatio'n channel to CRl. If the valve

       'is employed with a passive redundant check valve, then the motor operated valve is controlled by two safeguards actuation channels with CR1 and CR2 connected in an "0R" configuration.

The control relays, when energized by their associated safeguards actuation channels, close the valve through contacts which duplicate the manual CLOSE pushbutton and at the same time override any existing signal calling for the valve to open. A valve limit switch opens just before the valve seats to per-mit torque closing. Air-operated engineered safeguards valves automatically go to their engineered safeguards position upon loss of control air. Valves used with active redun-dant valves are equipped with a single electrical actuator for control by a single engineered safeguards channel as shown in Figure 7-2C. Valves used with redundant passive valves are equipped with two electrical actuators, each con-trolled by a single safeguards channel operating in an OR configuration. Engi-neered safeguards action is initiated when power is applied to the electrical actuator. O V The control of the reactor building spray pumps is by means of single control i relays in each pump controller. Each pump is controlled by separate engineered safeguards channels. Safeguards action is initiated when the pump control re-lay is energized by its associated engineered safeguards channel. 7 1.2 3 Design Features 7 1.2 3 1 Redundancy The Reactor Protection System is redundant for all vital inputs and functions. Redundancy begins with the sensor. Each power range input variable is measured four times by four independent and identical instrument strings. Only one of the four is associated with any one protective channel. The total and complete removal of one protective channel and its associated vital instrument strings would not impair the function of any other instrument or protective channe] . There are two source range channels and two intermediate range channels, each with its own independent sensor. The Safeguards Actuation System is also redundant for all vital inputs and functions. Each input variable is measured by three independent and identical instrument strings. The total removal of any one instrument string will not prevent' the sytten from performing its intended functions. . fs . 7-7 t

7 1.2 3 2 Independence The redundancy, as described above, is extended to provide independence in the Reactor Protection System. Each instrument string feeding into one protective lh channel is operationally and electrically independent of every other instru-ment string. Each protective channel is likewise functionally and electrically independent of every other channel. Only in the coincidence output are the channels brought into any kind of common relationship. Independence is preserved in the coincidence circuits through insulation resistance and physical separation of the coincidence networks and their switching elements. The Safeguards Actuation System instrumentation and control have electrically and physically independent instrument strings. The output of each bistable is electrically independent of every other bistable. Independence is preserved in the coincidence networks through insulation resistance and physical separation of the switching elements. 7 1.2 3 3 Loss of Power The Reactor Protection System initiates trip action upon loss of power. All bistables operate in a normally energized state and go to a deenergized state to initiate action. Loss of power thus automatically forces the bistables into the tripped state. Figure 7-2B shows the system in a deenergized state. The Safeguards Actuation System instrumentation strings terminate in bistable trip elements similar to those in the Reactor Protection System. Loss of in-strument power up to and including the bistables forces the bistables into the tripped state initiating safeguards action. The logic networks and the equip-ment control elements are powered from the Engineered Safeguards D-C Power Bus 1 and 2. Electrical safeguards equipment is powered from one of the Engineered Safeguards A-C Power busses. Loss of engineered safeguards power to the logic networks or to the safeguards equipment does not initiate safeguards action as described in 7 1.1.2.4. 7 1.2.3.L Manual System Trip The manual actuating devices in the protection systems are independent of the automatic trip circuitry and are not subject to failures that make the auto-matic circuitry incperable. The manual trip devices are independent control swi' ches for each IcVer centrcller. The independent centrol switches, hcw- 1 ev er, are avitch or all actuated througn a mechanical linkage to a cot =cn manual trip pushbutton. 7.1.d.2.) squipment Eemoval The removal of modules or subassemblies from vital sections of the Reactor Pro-tection the System will initiate the trip normally associated with that portion of system. The removal criterion is implemented in two ways: (1) advantage is taken of the (2) interlocks are inherent characteristics of a normally energized system, and provided. An innerent characteristic is illustrated by considering the power supply for one of the reactor protective channels. Removal of this power supply g T- S (Revised 1-15-60) 0132

automatically results in trin action by virtue of the resulting loss of pover. No' interlock is required in such cases. Other instances require a system of ( interlocks built into the equipment to insure trip action uron removal of a portion of the equipment. The. Safeguards Actuation System provides for servicing without affecting the integrity of the redundant channels.

                                                                                        ?
   -7.1.2.3.0       Testine The protection systems vill meet the testine criterien and its objectives.

The test circuits vill take advantage of the systems' redundance, independence, and coincidence festures whi'c h make it nossible to initiate trip signals man-unlly in any part of one nrctective channel vithout affecting the other chan-nels. This test feature vill allow the operator to interrogate the systems from the ' input of any bistable un to the final actuating device at any time during re-actor operation without disconnecting permanently installed eo.uipment. The test of a bistable consists of inserting an analog input and varying the innut until the bistable trip point is reached. The value of the inserted' test signal represents the true value of the bistable trip roint. Thus the test verifies not only that the bistable functions but that the trip point is correctly set. o restartup testing vill follow the same procedure as the on-line testing except that calibration of the analog instrument strings may be checked with less re-straint than during reactor operation. O(/ As shown in Figure 7-2b and 3-66, the power breakers in the reactor trip circuit [7 may also be manually tested during oteration. The only limitation is that not more than one power supply may be interrupted at a time without causing a re-actor trip. 7.1.2.3.7 "hysical Isolation The physical arrangement of all elements associated with the protection systems will. reduce the probability of a single physical event impairing the vital func-tiens of the system. For example, pressure measurements of reactor coolant pressure vill be divided between four redundant pressure taps so as to reduce the probability of collective danage to all sensors by a sinFle accident. Systeat equipment vill be distributed between instrument cabinets so as to re-duce the probability of damage to the total system by some single event.

    ' firing between vital elements of the system outside of eculpment housing vill be routed and protected v1 thin the unit so as to maintain the true redundancy of the systems.with respect to physical hazards.

7.1.2 3.8 Prinarv onver source The primary source of control pcver for the Reactor "rotection System is the 120v a-c Essential Services busses described in 8.2.2.7 The. source of pcVer p- for the measuring-Q) y 7-0 (Revised 7-15-69)

clcments in the Safeguards Actuation System is also from the 120v a-c Essential Services busses. Command circuits from the Safeguards Actuation System coinci-g d;nce logic that extend to Engineered Safeguards Equipment controllers are pov-ered from the Engineered Safeguards d-c busses. Engineered Safeguards equip-m;nt, such as pump and motor operators and their starting contactors, are pov-cred from the Engineered Safeguards a-c busses. 7.1. 2. 3. 9 Reliability Design criteria for the Reactor Protection System and the Safeguards Actuation System have been formulated to produce reliable systems. System design prac-tices, such as redundant equipment, redundant channels, and coincidence arrange-ments permitting in-service testing, have been employed to implement the reli-ability of protective action. Tne best grades of commercially available com-ponents will be used in fabrication. A system fault analysis vill be made con-sidering the modes of failure and determining their effect on the system's vital functions. Acceptance testing and periodic testing vill be designed to insure the quality and reliability of the completed systems. 7.1.2.3.10 Instrumentatien for Emergency Core Cooling Initiation The instrumentation system =akes use of both physical and electrical isolation. The high pressure and low pressure injection systems are activated by both lov 3 reactor coolant and reactor building pressure signals originating from three pressure transmitters measuring the reactor coolant system pressure, as shown in Figure 7-11, and three pressure transmitters measuring the reactor building pressure. g Two reactor coolant pressure transmitters are connected to one reactor pipe; the third transmitter is connected to the other reactor outlet pipe. Each transmitter has a separate tap on the reactor coolant piping inside the secondary shield. The transmitters are physically separated from each other and located outside the secondary shield inside the reactor building. Tne transmitters' electrical outputs leave the reactor building through separate penetrations. The three reactor building pressure transmitters are connected to the reactor building through independent taps. The transmitters are physically separated from each other and are located outside the reactor building. The output of each transmitter provides isolated signals to its associated bistable trip de-vices. The bistatle trip devices of a given logic function are physically sep-arated by cabinet barriers. Each pressure transmitter and its associated bi-stable trips are powered from separate battery-backed vital bus power sources, the same power sources which power the reactor protection channels. Two, iso-lated 125 volt d-c engineered sa'feguards control power sources are used for the power to the engineered safeguards channels and logic, as shown in Figure 7-2. Each major function is, therefore, activated from two independent sources of centrol power.

 'The ope % tion of the engineered safeguards channels and the trip relays forming the system legic is described in 7.1.2.2.

The high order of system redu.2dancy assures conpliance with the single failure criteria of 7.1.1.2.1. }g 7-10 (Revised 3-1-68)

    ~
 /h

( ) 7 1.2.h Su==ary of Protective Actions s_- The abnormal conditions that initiate a reactor trip are as follows: Steady State Trip Value or Trie Variable No. of Sensors Normal Range Condition for Trio Neutron Flux 4 0-100% 107 5% of full (rated) pcVer Neutron Flux / h Flux 3 to k pumps (1) Ratio of reac- 3 Reactor Coolant 16 Reactor Coolant ter power to total F] ov Pump Monitors reactor coolant 2 Flow Tubes flov exceeds 1.075 (2) More than one reactor coolant pump motor is lost. Startup Rate 2 0-2 Decades / min 5 Decades / min Reactor Coolant 4 2,120-2,250 2,350 psig Pressure psig 2,050 psig Reactor Outlet b 520-603 F 610 F l)

  ',_/

Temperature 1 fx 0 0135 l

l l 7-10a (Revised 3-1-63) l t

{} ' Actions initiated by the Safeguards Actuation System are as follows: Steady State Action Trin' Condition Normal-Value Trip Point High Pressure - . Low Reactor 2,120 - 2,250 psig 1,800 psig Injection Ccolant Pressure High Reactor Atmospheric 10 psig 3 Building Pres-sure Low Pressure Very Low Reactor 2,120 - 2,250 psig 200 psig Injection Pressure High Reactor Atmospheric 10 --ig 3 Building Pres-4 sure Start Reactor High Reactor Atmospheric h psig Building Emer- Building Pres-gency Cooling sure Unit and Reactor Building Iso-f lation D

   - \~ I      Reactor. Building       High Reactor             Atmospheric                  10 psig Spray                   Building Pres-sure i

7.1.2 5 Relationship to Safety Limits , Trip setpoints tabulated in 71.2.h are consistent with the safety limits that have been established from the analyses described in Section 1h. The set 4 point for each input, which must initiate a trip of the Reactor Protection System, has been established at a level that will insure that control rods are inserted-in sufficient time to protect the reactor core.~ Likewise, the set points for parameters initiating c trip of the Safeguards Actuation System are established at levels that will insure that corrective action is in prog-ress in sufficient time to prevent an unsafe . condition. Factors such as.the rate at' which_ the -sensed- variable can change, instrumentation and calibratien - ' inaccuracies,: bistable trip times, circuit breaker trip times, control rod travel times, valve times, and pump starting times have been considered in es-tablishing the margin between the trip set points and the safety limits that have been derived. i The flux' trip set point of 107.5 per cent is based upon the tolerances and er-ror bands shown in Figure 7 L. The ' incident flux error is the sum of the er-rors at. the output of the measuring channel resulting from rod motion, and in-strument drift during the interval lbetween heat balance checks of nuclear in-strumentation calibration.' o136 7-11 (Revised 3-1-68) f

7.1.3 SYSTEMS EVALUATICH 7.1.3.1 _ Functional Capability - Beactor Protection System O The Reactor Protection System has been designed to limit the reactor power to a level within the design capability of the reactor core. In all accident evaluations the time response of the sensors and the protective channels are considered. Maximum trip times of the protection channels are listed below.

a. Temperature - 5 sec.

b. Pressure - 0 5 sec.

c. Flux - 0.3 sec.

d. Pump monitor - 1.0 sec.

Since all uncertainties are considered as cumulative in deriving these times, the ac'ual times may be only one-half as long in most cases. Even these max-imum tive times, when added to control rod drop times, provide conservative protec-action. The Reactor Protection System will limit the power that might result from an unexpected arrested reactivity by high change. reactor coolantAny change of this nature vill be detected and or high neutron flux protective action. temperature, high reactor coolant pressure, An uncontrolled rod withdrawal from startup will be detected by the abnormally fast startup rate in the intermediate channels and high neutron flux in the h power range channels. nels is incorporated in Athe startup ratet trip from the intermediate-rangc chan-Reactor rotection System. A rod withdrawal accident at power will immediately result in a high neutron flux trip. Reduced reactor coolant flow results in a reauced allowable reactor power. The power / flow monitor operates to set the appropriate reactor power limit by adjusting the power level trip point. A total loss of flow results in a di-rect reactor trip, independent of reactor power level. Tw; =Mor measurements feea the rcwer/ flow moniter: (a) reactor ecolant flew, anc (b) neutran power level. 3

                                     ~he ficw tubes which provide the reacter coolant flow measurement will exhibit no change during the reactor life. A periodic calibration of ine flow transmitters will le r.ade. Tne neutrcn pcwer level signal will be reenlibrated by comparisen with a rcutine heat balan e. The pcwer range channela use aetectcrn arranged ic s f recti'.ely average the mea-surement over tne lencth of the ccre as descriled 1: 7.,3.1.1.2. Therefere, their   cutput ic expectt: to : ' witnc L per cert of t5e callt rst ed value aur-ing ncrmal regulat;nc rc.i grcup pczition chsnge and the need for additicnal calibration thereoy eliminated.

A loss of reacter coolant will recult in a reduction cf reacter ecclant pres-sure. The low pressure trip servec to trip the reacter for such an occurrence. O 7-12 (se.~ ired 3-2-65) 0137

< -'y A significant turbine-side steam line rupture is reflected in a drop of re-q_j actor coolant pressure. The low reactor pressure trip shuts down the Station - for such an occurrence. 7.1.3.2 Functional Carability - Safeguards Actuation System The Safeguards Actuation System is a graded protection system. The prcgres-sive actions of the injecticn systems as initiated by the Safeguards Actuation System provide sufficient reactor coolant under all conditions while minimi:- ing the possibility of setting the entire system in operation inadvertently. The key variable associated with the loss of reactor coolant is reactor pres-sure. In a loss-of-reactor-coolant accident, the reactor pressure will fall, starting high pressure injection at 1,800 psig. If high pressure injection does not arrest the pressure drop, then low pressure injection starts upon a signal of 200 psig. The injection systems are initiated by a detected reactor building high preFsure trip of 10 psig acting in parallel with the reactor coolant low pressure trips. rx i l \_ '

,a                                                                          n,
   )                                                                        U l .qJUg 7-12a (Revised 3-1-68)

                                   '   a 1

O , i f I I

                                                        ]

l i l l

 \

O . 1 i 1 i l l

                                                        \

i O: ,

                                           "" -s

G V The key variable in the detection of an accident that could endanger reactor building integrity is reactor building pressure. A reactor building pressure of 4 psig initiates operation of the reactor building emergency cooling unit and isolation of the building while a higher pressure of 10 psig initiates operation of the reactor building sprays. 7133 Preoperational Tests Valid testing of analog sensing elements associated with the protection sys-tems will be accomplished through the actual manipulation of the measured variable and comparison of the results against a standard. Routine preoperational tests will be perfor=ed by the substiu. tion of a cali-brating signal for the sencor. Simulated neutron signals may be substituted in each of the source, intermediate, and power range channels to check the operation of each channel. Simulated pressure, temperature, and level signals may be used in a similar fashion. This type of testing is valid for all ele-ments of the system except the sensors. The sensors should be calibrated against standards during shutdowns for refueling, or whenever the true status of any measured variable cannot be assessed because of lack of agreement among the redundant measurements. The final defense against sensor failure during operation will be the Plant operator. The redundancy of measurements provides more than adequate oppor-tunity for comparative readings. In addition, the redundancy of the systems reduces the consequences of a single sensor failure. 7 1 3.4 Component Failure Considerations The effects of failure can be understood through Figure 7-2B. In the Reactor Protection System, the failure of any single input in the " tripped" direction places the system in a 1-out-of-3 mode of operation for all variables. Fail-ure of any single input in the "cannot trip" direction places the system in a 2-out-of-3 mode of operation for the variable involved, but leaves all other variables in the normal 2-out-of-4 coincidence mode. If the fault were c the

     " tripped", open circuit mode, then the system would be able to tolerate a minimum of two "cannot trip", short circuit failures within t'a same measured variable before complete safety protection of the variable were lost. With one " tripped". open circuit fault, a second identical fault within the same variable would trip the reactor.

A similar fault relationship exists between channels as a result of the 2-out-of-4 coincidence output. One " trip" faulted channel places the system in a 1-out-of-3 or single-channel mode. A "cannot trip" faulted channel places the system in a 2-out-of-3 mode. At the final device, a " trip" faulted power breaker does not affect the pro-tective channel mode of operation, reactor trip being dependent upon one of two-breakers in the unaffected primary power supply to the control rod drives. A breaker faulted in the "cannot trip" mode leaves the system dependent upon the seccnd breaker in the affected primary power supply. A U The ' Safeguards Actuation System is a 2-out-of-3 input type of system. It can toleiate one fault of the _"cannot trip" variety in each of the cois.cidence 0139 7-13

networks. For this type of fault, all remaining inputs must function correctly. A " tripped" input fault allows any one of the two remaining inputs to initiate h action. Primary power input to both protection systems has been arranged t .> minimize the possibility of loss of power to either protection system. Each cahnnel of the protection syst2m will be supplied from one of the four 120v a-c Essential Services busses described in 8.2.2.7. The operator can initiate a reactor trip independent of the automatic protection action. The engineered safeguards have been connected to multiple busses to minimize total loss of safeguard capability. The individual parts of the Safeguards Actuation System can be placed in operation through manual operator controls independent of the automatic protection equipa nt. 7135 Operational Tests The protection systems are designed and have the facilities for roi. tine manual operational testing. Most inputs to the protection systems originate from an analog ceasurement of a particular variable. Every input of this type is equipped with a continuous readout device. A routine check by the operator of each reading as compared to the other redundant readings available for each variable will uncover mea-surement faults. These elements plus the bistables and relays of the protec-tion systems require a periodic dynamic test. Each system provides for routine testing. Each bistable may be manually tripped, and the results of that trip traced through the system logic and visually indicated to the operator. The h trip point setting of each bistable may be verified by the application of an analog signal proportional to the measured variable, and that signal may be varied until the bistable element trips. 0140 0 7-lk

72 FEGULATTNG SYSTEM 3

  ~
 / ~s 7.2.1         DESIGN BASES C

Reactor output is regulated by the use of novable centrol rod assemblies and 17 soluble boron dissolved in the coolent. Control of relatively fast reactivity effects includinz Deppler, xenon, and moderator terperature effects , is ecccm-plished by the control rods. The control resconse speed is designed to over-come these reactivity effects. Relatively slow reactivity effects , such as fuel burnup, fission product buildup, samarium buildup, a.nd hot-to-cold mcder-ator reactivity deficit, are controlled by soluble beren. Control rods are normally used for centrol of xenon transiants associated with normal reactor power changes. Chemical shin shall be used th con.1 unction with centrol rods to compensate for equilibrium xenon cenditions. Deactivity cen-trol may be exchanced between reds and saluble boren consistent with limita-ti:ns on power peaking. Peactor regulation is a composite functio *. of the Integrated Control System and Rod Drive Control System. Design deta for these subsystems are riven in the following sections. 7.2.2 ROD DRIVE CONTP.0L SYSTO The rod drive control system (RDC) includes drive controls, power suoplies, position indicators , onerating canels and indicators , safety devices , and en-closures. (~') 7.2.2.1 Desien 9 asis L.J' The rod drive control system design bases are catagorized into safety consider-ations , reactivity rate limits, startup considerations and operational consider-ations. 7.2.2.1.1 Safety Considerations

a. The control t od assemblies (CRA) shall be inserted into th e core upon receipt of protective system trip signals. Trip command ha' priority over all other commands.

b. No single failure shall inhibit the protective action of the "od drive control system.

7 2.2.1.2 Reactivity Rate Limits The speed of the mechanism and group red worth provide the ree.ctivity change rates required. For design purposes the maximum rate of change of {esetivity that can be inserted by any group of rods has been set at 1.1 x 10- AK /K/s. The drive controls , i.e. , the drive mechanims and rods codbination, have an inherent sp:ed-limiting feature. p NY 7-15 (Revised 7-15-69) 0141

7.2.2.1.3 Startup considerations 7 The rod drive control systen design baseJ for startup are as follows:

a. Peactor regulation during startup shall be a manual overation.

b. Control rod "out" motion sball be inhibi's ~ vhen a high startup rate (short period) ir the source range or intermediate range is detected.

7.2.2.1.h Operational Considerations For operation of the reactor, functional criteria related to the rod drive control system are:

a. CRA Dositionine The rod drive control system provides for controlled withdrawal, controlled insertion and holding of the control rod assemblies (CRAs), to establish and maintain the power level reouired for a given reactor coolant boron concentration.

b. Position Indication Continuous rod position indication, as well as full-in and full-out position indication, shall be provided for each control rod drive.

c. System Monitoring The rod u. 've cent. ol system design includes provisions for routinely monitoriryr conditions that are important to safety and reliability.

7.2.2.2 System Desien The rod drive control system provides for withdrawal and insertion of +he control rod assenblies to maintain the desired reactor output. This is achieved either throuch automatic control by the Integrated Control Syster discussed in Section 7.2.3, or through manual control by the operator. A

• noted previouslv, this control ecmpensates for =hort term reactivity '

changes. It is achieved through the positioning in the core of sixty-one , control rod assenblias and eight axial power-shaping rod assedblies. The sixty-one rods are grouped for control and safety purposes into seven groups. Four groups function as safety rods , and three groups serve as regulating rods. An eighth group serves to regulate axial power peaking due to xenon poisoning. Seven of the eight groues may be assigned from four to twelve control rod assehblies. Eight rod assemblico are used in group eight. Control rods are arranged into groups at the control rod drive control cystem patch panel. Typically, twenty-eight rods , including the axial power shpaing rods , are assigned to the regulating groups , and forty-one rods are assigned to the safety rod groups. A typical red grouping arrangement is shown below: 7-16 (Revised T-15-69)

            .-                                   .~  _
                                                            ~       . . ~         -   .- _- _-                       --       -            - . __ -

1 a

                        . Safety Pods                  Pegulatine Reds-             Axini Power Shaning Pods

() Group 1 - 8 Group 5 - 12 Group 6 - 8 Group 2 - 12 Group 6 - 4 Group 3 - 9 "roup 7 - h Group h - 12 During startup the safety rod groups are withdrawn first, enabling withdrawal of the regulating control groups. The sequence allows operation of only one

                .egulating rod group at a time except where reactivity insertion rates are lov (first and last 25% of stroke), at which time two adjacent groups are operated simultaneously in overlapped fashion. These insertion rates are.shown in Fig-ure 7-7.

i As fuel is used, dilutien of soluble boren in the reactor coolant is necessary. When Group 6 is more than 955 withdrawn, interlocks permit dilution. ' The reac-tor controls insert Group 6 to concensate for the reduction in boron concentra-tion by dilution. The dilution is automatically terminated by a pre-set' volume measuring device. Interlocks are also provided on Group 6 rod position to terminate dilution at a pre-set insert limit. 7 2.2.2.' System Equipment l The rod drive control system consists of three basic components: (1) control l rod drive motor power supplies; (2) system logic; (3) trip breakers. The power supplies consist of four group power supplies , an auxiliary power supply, and two holding power supplies. The group power supplies are of a redundant six-phase half-wave rectifier design. In each half of a group power supply, recti-fication and switching of power is accomplished through the use of Silicon Con-1 trolled Rectifiers (SCRs). This switching sequencially energizes first two, then three, then two of the six CRA motor stator vindings in stepping motor fashion, to produce a rotating magnetic field for the control rod assembly motor to position the CRA. Switching is achieved by gating the six SCRs on for the , period each vinding must be energized. As each of the six vindings utilize SCRs to supply power, six gating signals are required. I Gating ' signals for the grcup power supplies are generated by a motor driven programmer, consisting of a 60-cycle, reversible synchronous motor driving a multichannel photo-optic encoder. The coded light beam excites photo-detectors, generating signals which are amplified to form the Silicon Controlled Rectifier gating signals. The programmer is redundant (except for motor and gears), thus . providing' separate but . synchronized gating signals to the dual power supply units. Command signals to pos'. tion the control rod drive are introduced at the j programmer motor. i i Identical power supplies are used for the regulating (control) groups and for - the auxiliary power supply. . Each half of each group power supoly is capable of driving up to 12 drive mechanisms--the maximum number that may be in any one group. The power supplies have dual power inputs, each half fed from separate power sources and each half being capable of carrying the full load. O 7-17 (Revised 7-15-69) 0143

        ,,                 ,, , . , - .        . _ .                        . , .                . , . . , , . .        _ , ,   ,_...._,, ,.        ...m.

I Unlike the control groun never supplies, the holding never supply is used to maintain the safety rods fully vithdravn; consec.uently, switching is not re-quired, f. cix-chace d-c rover sunnly is used for this purpose. Two holding rcver curplies nre nr-vided. Fach is rated to furnish power to one vinding lll rf LF nechanicnn. Tna auxilisr" rewor sunnly 1 used to positien the safety rod groups and to prcvide sincle rnd control. The safety rod groups are maneuvered with the aux-ilinry tower surrly, and then, .rhen fully positioned, are transferred to the holding Lucses described above. After positioning the safety rods, the auxiliary rover surply is available to the regulating groups, through transfer relays, tc cerve ~ither as a single red controller, should repositioning of a single red b: n" cec sary, or, as a snare group controller, should one of the group control power supplies require maintenance. The system logic encompasses those functionn which connand control rod notion in the manual or automatic modes of cr^ ration, includine CDD sequencine, safety and protection features, and the n1nunl trip function, t'ajor components of the logic system are the Operator's Control Danel, C A rosition indication panels, automatic senuencer, and relay 1 cie. Switches are erovided at the crerators control panel for selection of the de-cired rod control node. Contrcl nodes are: (1) Automatic mode--where rod mo-tion !s ccmmanded bv the7 nterrated Control Ayster; and (2) "anual mode--where rod notien ja cenmanded by the operator. "anual control permits operation of _

 +   ningle rnd or a rrcun of rods. Alarm     lamps en the PDC panel alert the oper-ator to the cystems status at all times.

The secuence section of the loric system utilizes rod rosition signals to gen-erate centrol interlocks which regulate rod group withdrawal and insertion. The sequencer operates in both automatic and manual modes of reactor control, nnd controls the reculating grours only. Analog position signals are gu. erated tv the reed witch matrix en the CPA , and an average group rosition is gener-ated by an avt-~Hng network. This averere sincal serves as inrut te electronic cet reint trir unitc which are activated at approximately 25 and at 75 per cent of creup rod withdrnwal. Tvo bistnble units are provided for each regulating group. Outputs of these histables actuate " enable" relays which permit the rod groups to be commanded in automatic or manual mode. In additicn to the sequencer, relay logic monitors are provided in the " enable" circuits which prohibit out of secuence conditiens. The selection of manual c- t",1 rr'-  : 'uance byrac~ n'de functions rerrit intentional out-of-m<nuence ccmdi*ir r. Thic c'nlitien is indicated to the overater. inrute t, t": v'te- loric fron the r eactcr Drotection System and the Inte-rrated C'ntrol 'vstem rrovide interlock control ever rod motion. These inter-loc: n cause cri notion co .nsnd linas and control mode selection to be inhibited. In tho - drive m'ntrol system, two methods of rosition indication are tro-tided; an at ocluta resiticn indicator and a relative rosition indicator. The absolute position transducer censists of a series of magnetically operated reed cwit cher ncunted in a tube parallel to the notor tube extension. Each switch is hermetically sealed. Switch 'centacts c1cse when a pernanent magnet mounted en the urrer end cf the lead screv extensien ecces in close proximity. As the O 7-1; (Revised 7-15-69) 0144

lead screv (and the control red assembly) moves, the switches cperate sequen- 7 7 tially, prod' icing an analez voltsge proportional to resitien. The accuracy of (l the analog signal is spproxinately 11.1 per cent of full scale (139 1.,. ) and the readout has approximately +2.1 per cent of full scale accuracy. Other reed switches included in the sane tube with the position indicator ratrix provide full-in and full-out limit indications. The relative position transducer is a snall rulse-sterping motor, driven from the power supply for the rod drive notor. This small rotor is courled to s cotentiereter with an outpur sienal accuracy of 1,n.7 rer cent of full scale positien, producing a resdout with an accuracy o# 1,1.7 rer cent of full scale. Ecd drive control systen tric breakers sre provided to inte,rrunt rcver to the control rod assenbly noters. When power is removed, the ro?]er nuts disengage from the lead screv and a gravity free-fall trin of the CPA occurs. Pvn series trip methods are trovided for renoval of never to the CEA motors. First, a trip is initiated when Peactor Protection System logic interrunts never to the undervoltage (UV) coil of the main AC feeder breaker' . Secondiv, a trip is initiated when the Silicon Centrol Sectifier rating power and the DC holding pcuer is interrupted. As parallel tower feeds are provided on both holding and Fating power, interruption of both feeds is required for trip action in either method of trip. Trip circuitrv is shown in Figure 3-66. AC power feed breakers are of the three-pole, stored-energy tvre, and are equipped with instsntaneous underveltare trip coils. Fach AC feed breaker is housed in a separate metal-clad enclosure. The secondary trip breakers are also of the stored-energy type with two parallel-connected instantaneous undersolt-age trip coils consisteing of two 2-pole breakers nechanically ganged to inter-I') \# rupt DC busses. All breaker ~ are motor-driven-reset to provide remote re?et capability. Each undervoltage trio coil is operated from the Peactor Protec-tien System. 7.2.2 3 systen Evaluation 7.2.2.3.1 Safety Consideratiens A reactor trip occurs whenever power has been removed from the rod drive motors. The design provides two stored energy breakers which do not require cover to interrupt the electrical feeds to rod drive centrol power sunnlies, and a sec-ond set of circuit-interructing devices in series en the outnut of the power supplies. All devices have interrupting capacity of sufficient rating to open under any group load conf *.guration. Peactor trip is further assured by pro-viding series trip devices, split buses and provisions for neriodic testing. Trip redundancy is provided by eeries breakers while availability and test-ability are provided through dual power sources. Redundant power supplies permit testing of the trip action of each power-interruptina device without loss of plant availability. Reactivity shutdown margin urovided by the safety rods is sssured by diversifi-cation of their power buses. This feature, as shown in Figure 7-1, utilizes four separate ; buses , each having a seuarate trip device, to never the safety rods. A failure in one bus does not reflect into the ether huses, therefore, a single failure in the distribution system for the safety rods does not Dre-vent a plant shutdown. O) L 7-19 (Revised 7-15-69) f 4 fh

I In summatt, series redundant trir devices having adequate rating, testability and a " split bus" arrange ~ent innure safety of reactor trip circuits. 7.2.2.3.? Peactivit" Late Lirits The decired rate of ch9nce o" C"A renetivity insertion and uniform reactivity dintribution over the enre are trovided for by the control rod drive and power sur41v decirn, and the selection of rods in a groun. The motor, lead screw and rover cuptly designs are fixed to nrovide a uniform rate of sneed of 30 in./nin. The reactivity chenca is then controlled by the rod group size. To incure flexibility in thin area, a natch panel has been included in the power nupply to enable the interchange of rod vorta between rod grouns. Any rod may be patched into any rroup with the exception of Groun 8. Uniforn and synmetrical crour insertion rate is provided for by synchronous withdrawal of all rods in that groun. Such synchronous withdrawal is achieved by the desicn of the rouer supplv. A grour never supply onerates synchronously hv having its load (h to 12 CRA motor vindines) connected in parallel on the output of the SCR's. As the proernmmer gates on the SCP's, all rods in a group havn the same motor vindine energized simultaneously, nroducinc synchronous motion o# the entire groun.

 '/nnitors are nrovided to sense asymnetric rod patterns. These nonitors alarm the condit. ion to the onerator, connuter, and the IC9      Depending upon the pcuer cetting, action is initiated by the ICS to insert rods and reduce power.

7.P.P.3.3 Startup Considerations The rod drive controle receive interlock signals from the ICS and nuclear in- h ntrumentat i on ('IT ) . These innuts are used to inhibit automatic node selection holow 15" rated power rond to inhibit out notien for high startun rates , re-ntectivelv. In addition to the startup considerations, dilutien controls, to termit re-moval of reactor shutdown concentrativ..s of boron in the reactor coolant, are nrovided. This control hypasses the norna3 reactor coolant dilution controls, described in Section 7.2.2.2, providing all safety rods are withdrawn from the core and the operator initiates a continuous feed and bleed cycle. An addi-tional interlock on rod Group 5 inhibits the use of this circuit when rod nrour 4 in nore than W vithdrawn.

7. '. ' . 3. h Operat!cnal Considerations The control rod assenbly positioning system provides the ability te nove any rod to any noaition required consistent with reactor safety. As noted in Fec-t ien 7.2.2.3.0 a uniforn npeed is provided by the drive system. A fixed rod rasition when notion ic not required is obtained by the power sur;1y ability to enerrine two 9djacent vindings of the CFA motor stator. This static ener-gizing of the vindings maintain a latebed stater and fixed rod position.

Donitien Indication As previously described, tvo separate rosition indication signals are provided. The absolute positien sensing systen produces signals proportional to CRA po-l sition from the reed switch matrix located en each CRA nechanism. The relative ll 7-20 (Revised 7-15-69) 0146 l l

position indicat in system nroduces a siinal proportional to the number of CFA 7 7-s motor power pulses from a stepping motor and precision totentiemeter for each ( ,) CRA nechanism. Position indicating readout devices mounted on the operator's censole consist-of 69 single rod position meters and h control group average position meters. The operation of a selector switch permits either relative or absolute position information to be displayed on the single rod meters. The control-group-avdrree meters display the arithmetic average of the relative positien signals of all C2A's in a group. A selector switch on the enerator's console permits the grout neters to display either the positions of all safety rod groups (Groups 1 h) or the nesitions of all regulating rod croups (Groups 5-7) and the axial power shaping red groups (Group 8). Indicator lights are provided on the single-rod meter panel to indicate when each rod is; (1) fully inserted, (2) fully withdrawn, (3) under control and (h) whether a fault is present. Indicators on the operator's console show full insertion, full withdraval, under-control and fault indication for each of the eight control rod grouns. Failures which could result in unclanned control rod withdrawal are continu-ously monitored by fault detection circuits. Iten failures are' detected, in-dicator lights and alarms alert the operator. Fault indicator lights remain . on until the fault condition is cleared by the operator. A list of indicated faults is shown below: p/ g (1) Asymmetric rod patterns. (2) !Aotor rotation faults. (3) Sequence faults. (h) Red position sensor faults. (5) Trip faults. (6) Safety rods not withdrawn. (7) Programmer lamp faults. 7.2.3 INTEGRATED CONTROL SYcTE!! The Integrated Control 3ystem maintains constant average reactor coolant tem-perature and constant steam pressure in the nuclear unit during steady state and transient operation between 15 and 100 per cent rated power. Figures 7-6 and 7-8 show the overall system. The systen is based on the Integrated Boiler-Turbine concept videly used in focsil-fuel-fired utility plants. It combines the stability of a turbine-following systen vith the fest response of a boiler-following system. Optimum overall unit nerformance is maintained by limiting steam pressure variations; by limitinF the unbalance that can exist amont the steam generator, turbine, and the reactor; and by limiting the total unit load demand upon loss of capabi'ity of the steam generator feed ,systen, the reactor, or the turbine generator. r"N U 7-21 (Revised 7-15-69) 014:7

Figure 7-6 shows the reactor control rcrtien of the Interrated Control Eystem described in 7.2.3.3. Figure 7-8 shcus the steam cenerator and turhine cen-trol portion of the Intecrated Centrol Syster. This control receive. inruts g of neaavatt denand, svster frecuency, e,d etean *ressure, and su*nlies autout

?!cnals to the turbine byrars valve, + ebine steed chancer, and st eam renero-or feedvater flov centrols vith chan inr oneratine conditiens.

The turbine and stean generator are canable of automatic contrn1 from zero no"er to rated rover with ortional ranual control, mhe reactor controls are desirned for nanual operation belov 15 per cent rated nover and for automatic or manual operation above 15 per cent rated no"er. The turbine is operated as a turbine-following unit vith the turbine contrn1 valve pressure set point variea in proportion to reravatt error. The steam generater is operated as a boiler-following sytten in which the feedwater flov demand to the stean generator is a sunration of the necavntt denand 3rd th" stean pressure error. The Integrated Control System obtairs a load demand siens1 from the system dis-patch center or fron the operator. /. frequency loop is ndded to natch the speed droor of the turbine steed controls. The losd demand le restrained by a maxinun load liniter, a ninimun load liniter, a rate lir. iter, and a runback limiter. In nornal operation the necavatt denand ('md) linits would be set as follovs:

                         "ax!mun load limit       1005 Mininun loed linit       15

Pate limit 10Z/nin The runbacks act to runback and/or linit the load demand on any of the fo] low- O ing conditions:

u. o ne or nore reactor coolant nunce nre inor.erntiva.

h. Total feedvater finv lacs total feedveter demand b ' rore th m 5 ror cent.

c. Assymetric rod withdraval patterns exist.

d. The generator senarates from the 506 kv hus.

The outnut of the liriters is a nefavett denand sienal which  :- nW i< te < turbine control, steam renerator control, and reactor control in ra~dl.2.

• The reactor control responds to the nera.rstt demand sien9 m @rcr! H ir 7.2.3.3.

7.2.3.1 Turbine control The teaavatt denand is comrared with the .?eneracor necau9tt outr,ut, :nd the resulting megavatt error sirnal is used to change the stear. nrecsure set noint. The turbine valves then chanre ronttion to control stear rressure. ts tho m23avatt error reduces to zero, the stean pressure set noint 13 returned to the steady state value. By liniting the effect of neravatt erro- on the stean pressure set point, the systen can be ad,iusted to territ controlled variations in steam tressure to acnieve any desired rate of turbine resronse ts nerawatt denand. l 7-22 (Revised 7-15-69) 0148

7 2 3.2 Stean Generator control (' ) , Control of the stean generator is based on matching feedvater flow to megawatt l7 demand.vith bias provided by the error between steam pressure set point and steam pressure. The tressure error increases the feedvater flev demand if the pressure is lov. It decreases the feedvater flow demand if the tressure is hi gh. The basic control actions for parallel steam generator operation are

a. Megawatt denand converted to feedvater demand.

b. Steam pressure compared to set tressure, and the pressure error con-verted to feedvater demand.

c. Total feedvater demand comruted from sum of a and b.

2

d. Total feedvater flow demand split into feedvater flow demand for each steam generator.

                -e. Feedvater demand cemtared to feedvater flov for each steam generator.

The resulting error signals position the feedvater flow controls to match feedvater flow to feedvater demand for each steam generator. For operation below 15 per cent load, the steam generator control acts to main-tain a preset minimum downcomer vater level. The conversion to level control is automatic and is introduced into the feedvater control train through an auc-tioneer. At low loads below 15 ter cent, the turbine bypass valves vill oper-() ate to limit steam pressure rise. The steam genericer control also provides ratio, limit, and runback actions as shown in Figure 7-8, which include

a. Steam Generator Load Patio Control Under normal conditions the steam generators will each produce one-half of the total load. Steam generator load ratio control-is pro-vided to balance reactor inlet coolant temperatures during operation with more reactor coolant pumps in one loop than in the other.

b. Rate Limits Rate limiters are manually set to restrict loating or unloading rates 4 ~to those that are compatible with the turbine and/or the stemn gen-erator,

c. Water Level Linits A maximum vater level limit prevents gross overpumping of feedvater and insures superheated steam under _ all operating conditions.

A minimum water level limit is provided for llow load control.

d. Reactor Coolant Pune Limiters

 . m

_) These limiters restrict feedvater demand to match reactor coolant pumping capability. For example, if one reactor coolant pu=p is not

           .        - operating, the maximum feedvater demand to the steam generator in the loop with the inoperative pump is limited to approximately one-half normal.

7-23 (Revised 7-15 49) 0 T4'?

7

e. Reactor Outlet and Feedvater Lov Temnerature Limits These limiters reduce feedvater demand when the reactor outlet ten-perature or the feedvater temperature is lov.

f. _Feedvater Purm Canability A feedvater pumn canability runback signal limits the megawatt de-mand signal whenever total feedvater flov lags tctal feedvater de-mand by 5 per cent.

7 2.3 3 Reactor control The reactor control is nade up of analog computing equipment with inputs of megawatt demand, core power, and reactor coolant average temperature. The output of the centroller is an error signal that causes the control rod drive to be positioned unti? the error signal is within a deadband. A block diagram of the reactor contrcl is shown in Figure 7 6. First, reactor power level demand (Nd) is computed as a function of the mega-vatt demand (Wd) and the reactor coolant system averaFe temperature deviation (hT) from the set point, according to the following equation: 1 tid -EWd+K2 1 @ + 7 ) H dt) Megawatt demand is introduced as a part of the demand signal through a propor-tional unit having an adjustable gain factor (K1 ). The temperature deviation is introduced as a part of the demand signal after proportional vlus reset (integral) action is arplied. For the temperature deviation, K2 is the adjust-able gain and T is the adjustable integration factor. The reactor power level demand (N d ) is then compared with the actual reactor power leval signal (Ni ), which is derived from the nuclear instrumentation. Tne resultant error signal (?i d - Ni ) is the reactor power level error signal (Fp ). When the rcactor power level error signal (E ) exceeds the deadband setttingc, the control rod drive receives a command thak withdraws or inserts rods de-pending upon the polarity of the power error signal. The folleving additional features are provided with the reector power control-1er:

a. /in adjustable low limit on the megawatt denand sicnal (07 d ) to cut out the automatic reactor control action.

b. A high limit on reactor power level demand (fid)*

c. An adjustable low limit on reactor power level demand (Nd )-

O 7-21s (Fevised 7-15-69) 0150

2 ( Separate frem, but related to, the automatic reactor centrol system is the re-actor coolant flow signs 1 syeten. Pcver to each reactor coolant pump motor is monitored as an indication of' reactor coolant flov. Logie units continuously corpare the nusber of-energized pumps to the messured reactor power to sense If the flew is lov,

                                       ~

that the flow is adeauste for the operating pcVer level. the reactor pcVer level derand is reduced by the Integrated Control System.

         -7.2.3.4        Svsten Failure censiderations Pedundant sensors are available to the Integrated Control Prsten. The opera-

tor can select any of the redundant sensers from the centrol room.

I Manual . reactivity control is available at all power levels. i uoss of- electrical pover to the sutecatic controller reverts reactor control

. to the manual mode.

T.2.3.5 Interlocking Control rod withdrawal is prevented on the occurrence of a positive short pe-riod below 10 per cent power.

The automatic sequence logic sets a predeternined insertion and withdrawal pat-tern of the four regulating rod groups.

Control circuitry allows nanually selected operation of any single control rod ( ) .. or control rod group throughout the power range. An interlock will prevent actuation of both withdrawal ud insertion of con-trol rods simultsneously with the insertion signal overriding the withdrawal. Control rod drive switching circuits allow withdrawal of no more than a single control rod group in the nanual mode. The automatic sequence logic limits regulating rod motion to one group out of four at 'one time except at the upper and lover 25 per cent of stroke where operation of two groups is permitted to linearize reactivity versus stroke. Maximum and minimum limits on the reactor never level demand signal (Nd ) Pre-vent the' reactor controls from initiating undesired power excursions. Maximum and minimum levels on the megawatt denand signal (W4 d ) prevent -the re-actor controls from initiating undesired power excursions. T.2.3.6 Loss-of-Lcad considerations The nuclear. unit'is designed to accept 10 per cent' step load rejection without

safety valve action or turbine bypass valve action. The combined actions of the control' system and the turbine bypass valve permit a ho per cent load re-duction or a turbine trip from h0 per cent load without safety valve acticn. The coc trols vill limit stean dump to the. condenser when condenser vacuum is O 7-24a (Revised 7-15-69) 0151

inadequate, in which case the safety valves may operate. The combined acticns of the control system, the turbine bypass valve, and the safe 'y valves permit I lll a 100 per cent loed rejection without turbine trip. This permits the unit to ride through a " blackout" condition, i.e. , sudden rejection of electrical load down to auxiliary load withcut turbine trip. (The " blackout" provisions are discussed in 1L.1.2.8.2.) The features that certit continued operation under load rejection coaditions include: ,

a. Intecrated Control Fvstem During normal operation the Integrated Control System (see Figure 7-8), controls the unit load in response to load demand from the sys-tem dispatch center or from the operator. During normal load changes and small frequency chanFes , turbine centrol is through the speed changer to naintain constant steam pressure.

During large load and frequency upsets, the turbine governor takes control to regulate frequency. For these upset conditions , frequency error at the input to the integrated control system becomes more in-Dortant in trovidine load matching.

b. 100 Per Cent Seller Canacity in the Steam System This provision acts to reduce the effect of large load drops on the reactor system.

g Consider, for example, a sudden load rejection greater than 10 per cent. When the turbine generator starts accelerating, the governor valves and the intercept vs1ves begin to close to maintain set fre-quency. At the same time the megavatt demand signal is reduced , which reduces the covernor speed changer setting, feedvater flow de-

           -*nd, and reacter never level demand. As the governor valves close, tqe steam pressure rises and acts throuch the control system to rein-torce the feedvater flow demand reduction already initiated by the reduced meFavatt demand signal. In addition, when the load rejection is of sufficient macnitude, the turbine bypass valves onen to re.iect excess steen to the condensyr, anma +h. e safety valvec open to exhsust iteam to the atrosphere.    .ne rise .n otet     -    - --

g, _ ion in feedvater flev cause the average reactor coolant temperature o rise which reinferees the reacter power level demand reduction, already estohlished by reduced megawatt demand, to restore reactor ecom "+ t'.cperature to set value. As the turbine cenerator returns to set : equency, the turbine con-trols revert to steam pressure control rat.'er than frequency centrol. This feature holds steam pressure within re.atively narrev limits and prevents further large steam pressure chtnges which could impose additional load changes of opposite sign en the reactor coolant sys-tem. As a result, the reactor, the reactor coc'. ant system, and the steam system run beck rapidly and smoothly to the new load level. 0\ . 7-2hb (Revised 7-15-69) s

73 I::STR321;TATIC:: 731  :.JcLEAa I::sTac2:."rATIc::

          .We nuclear instrumentation system is shcun in Figurc 7-2A. Emphasic in the
          ' design is placed upon accurtcy, stability, and reliability.            Inctru=ents are redundant at every level. 2 c design criteria stated in 7 1.1.2 have been applied to the design of this instru=entation.

7 3 1.1 Desicn The nuclcar instrumentation has eight channels of neutrun infor=ation divided into three rances of sensitivity: source range, inumediate range, and power range. The three ranges cc=bine to give a continucus =casurement of reactor 4 power fro source icvel to approximately 125 per cent of rated power or ten decades of infor=ation. A minimum of one decade of overlapping infomation is provided between successive higher rances of instru=cntation. The relationship between instrument ranges is shown in Figure 7-9 4

;

The sourec range instrumentation has J*6 redundant count rate channels origi-nating in two high sensitivity proportionaA .:.aters. These channels are used over a counting range of 1 to 102 counts /sec as displayed on the operator's centrol concolc in tems of log counting rate. The channels clso measure the rate of change of the neutron level as displayed for the operator in te ms of startup rate from -1 to +10 decades / min. !io protective functions are associated with the source range because of inherent instru=entation limitations encountered in this range. However, one interlock is provided, i.e., a control rod withdraw hold and alarm on high startup rate in either channel. The intemediate range instru=entation has two log !! channels originating in two 4 identical electrically gn==c-compensated ion chambers. Each channel provides seven decades of flux level infomation in terms of log 1on cha=ber current and ] startup rate. The ion cha=ber output range is from 10-11 to10-ka= peres. The

!          startup rato range is from -1 to +10 decades per minute.           Protective action on
;          high startup rate is provided by these channels. A high startup rate on either channel causes a reactor trip.         Prior to a reactor trip, high startup rate in either channel vill initiate a control rod withdraw hold interlock and alarm.

The power range channels have four linear level channels originating in 12 un-

compensated ion chambers. The channel output is directly proportional to re-actor power and covers the range fr a O to 125 per cent of rated power. The system is a precision analog system w.1ch employs a digital technique to pro-

' vide highly accurate signals for instru=ent calibration and reactor trip set point calibration. The gain of each channel is adjustable, providing a means i for calibrating the output against a reactor heat balance. Protective action i on high flux level consists of reactor trip initiation by the power range chan-nels at preset flux levels. Additional features pertinent to the nuclear instru=entation system are as fol-lows: l a. Independent power supplies are included in each channel. Primary ! O power originates from the 120v a-c Essential Services busses described t u U 0153 7-25 _. ,, _ . ._- ~ , _ - . -- _

in 8.2.2.7 Where applicable, isolation transformers are provided to insure a stable, high-quality power supply,

b. The proportional counters uced in the cource range are decigned to be secured when the flux level is greater thtn their useful operating rance. This is necessary to obtain prolonged operating life.

c. The inter =ediate range channels are supplied with an adjustable source of gc=a-co:pensating voltage.

7 3 1.1.1 Test and caibration Test and calibration facilities are built into the cystem. The test facilities vill meet the requirements outlined in the discuccion of protection syste=c testing. Facilities for calibration of the various channel amplifiers and measuring equipment will also be a part of the syste=, 7 3 1.1.2 Power Range Detectors Twelve uneo pensated ionization chc=bers are used in the power range channela. Three chambers are accociated with each channel, i.e., one near the botto= of the core, a ::ccond at the midplanc, and a third toward the top of the core. The outputs of the three chcmbers are combined in their respective linear am-plifiers. A means is provided for reading the individual chc=ber outputs as a manual calibration and test function during normal operation. h 7 3 1.1 3 Detector Locationc The phycical locations of the neutron detectors are chown in Figure 7-10. Tnc power range detectorc a'e located in four primary positions, 90 degrees apart around the reactor core. The two cource range proportional counters are located on opposite sides of the core adjacent to two of the power range detectorc. The two intermediate range compencated ion chc=bers are also located on oppocite sidea of the core, but rotated ?O degrees from the ccurce range detectors. 7 3 1.2 Evaluation The nuclear instrumentation vill monitor the reactor over the 10 decade range from source to 125 per cent of rated power. The full power neutron flux level at the power range detectors will be approximately 109 nv. Die detectors en-ployed will provide a linear response up to approximately 4 x 1010 nv before thev are saturated. The intermediate range channels overlap the source rance and the power range channels in an adequate =ancer, providir6 the continuity of inforration needed during startup. O 7-20 0154

The axial and radial flux distribution within the reactor core vill be measured (,) by the incore' neutron detectors (7 3 3). The out-of-core detectors are pri=crily

        .for reactor safety, control, and operation inforcation.

7 3 1.2.1 Loss of Power The nuclear instrumentation draws its primary power from redundant battery-backed 12CV a-c Essential Services busses described in 8.2.2.7 7 3 1.2.2 Reliability and Component Failure The requirements established for the reactor protection system apply to the nu-clear instrumentation. All channel functions are independent of every other channel, and where signals are used for safety and control, clectrical isolation is employed to meet the criteria of 7 1.1.2. 7 3 1.2 3 Protection Requirements The relation of the power range channels to the Reactor Protection System has been described in 7.1. To maintain the desired accuracy in trip action, the total error from drift in the power range channels vill be held to 1/2 per cent at rated power over a 30 day period. Routine tests and recalibration vill insure that this degree of deviation is not exceeded. Bistable trip set points of the power range channels will also be held to an accuracy of 1/2 per cent of rated power. The accuracy and stability of the equipment vill be verified by vendor tests. 732 NommCLEAR PROCESS INSTRUPEUTATION 7 3 2.1 System Design The nonnuclear instrumentation =casures temperatures, pressures, flows, and levels in the reactor coolant system, steam system, and reactor auxiliary sys-tc=s. Process variables required on a continuous basis for the startup, op-eration, and shutdown of the nuclear unit are indicated, recorded, and con-trolled from the control room. The quantity and types of process instrumenta-tion provided will insure safe and orderly oneration of all systems and pro-cesses over the full operating range of the Plant. The amounts and types of various instruments and controllers shown are intended to be typical examples of those that vill be included in the various systems when final desi6 n details have been completed. The nonnuclear process instrumentation for the reactor coolant is shown in Figure 7-11 and on the reactor auxiliary system drawings in Sections 5, 6, 9, and 11. Process variables are monitored as shown on the , nonnuclear instrumentation and reactor auxiliary system drawings and are as l follows:-

a. In concral, reaistance elements are used for temperature measurements.

Fast-response resistance elements monitor the reactor outlet tempera-ture. The outputs of these fast-response elements supply signals to the protective system.

b. Pressures are measured in the reactor coolant system, the steam sys-

_ tem, and the reactor auxiliary systems. Pressure signals for high

   \s l                                                   7-27                     0155

and low reactor coolant pressures and high reactor building pressure are provided to the protection systems. g

c. Reactor coolant pump :.otor operation is monitored as an indication of reactor coolant flev. ?nis infomation is fed to the reactor con-trols and reactor protection system. In addition, reactor coolant flov signals are obtained and indicated by continucus measurement of the pressure drops across the reactor coolant side of each steam gen-erator.

d. Flow in the steam system is obtained through the use of calibrated feedvater flow nozzles. Flov information is utilized for control and protective functions in the steam system. Stec= generator level cen-sure=ents are provided for control and alarm functions.

e. pressurizer level is measured by differential pressure transmitters calibrated to operating te=perature and pressure. The pressurizer level is a function of the reactor coolant system makeup and letdown flow rate. The letdova flow rate is re=ote manually controlled to the required flow. pressurizer level signals are processed in a level controller whose output positions the makeup control valve in the

          =akeup line to maintain a constant level.

f. Reactor coolant syste= pressure is maintained by a control system that energizes pressurizer electrical heaters in banks et preset pressure values below 2,175 psig or actuates spray control valves if the pres-sure increases to 2,230 psig.

7 3 2.2 Syste= Evaluation O Redundant instrumentation has been provided for all inputs to the protection systems and vital control circuits. '4here vide process variable ranges are required and precise control is involved, both vide-rance and narrow-rande instrumentation are provided. '4here possible, all instrumentation components are selected fro: standard com-mercially available products with proven operating reliability. All electrical and electronic instru=entation required for scfe and reliable operation vill be supplied from redundant 120v a-c Essential Eervices busses. 733 IN CRE MONITORING SYSTEM 7331 Design Basis The incore nenitoring system provides neutron flux detectors to conitor core 1 performance. No protective action or direct control functions are perfor=ed by this system. All high pressure system connections are terminated within the reactor building. Incore, self-powered neutron detectors measure the neutron flux in the core to provide a history of power distributions and dis-turbances during power operating modes. Data obtained vill provide measured power distribution information and fuel burnup data to assist in f uel manage-ment decisions. 7-28 (Revised 1-15-68) 0156

7332 system Design 7 3 3 2.1 system Description t The incore =cnitoring system consists of asse=blies of self-powered neutron 1 detectors and calibration tubes located at 52 preselected radial positions within the core. The incere scnitcring locations are shown en Figure 7-12. In this arrangenent, an inecre detector assembly, ccnaisting cf seven local flux detectors, one background detector, and a calibratien tube, is in-stalled in the instrumentation tute of each of 52 fuel assemblies (Figure 3-62). The 1ccal detecters are positioned at seven different axial eleva-tions to provide the cxial flux gradient. The outputs of tne local flux detectors are referenced to the background detector cutput so that the dif-ferential signal is a true =easure of neutron flux. As shown in Figure 7-12, seventeen detector asse=blies are located to act as sy==etry monitors. The remaining 35 detector asse=blies, plus five of the 17 sy==etry monitors, provide =onitoring cf every type of fuel assembly in the core when quarter core sy==etry exists. Readout for the inccre detectors is performed by the unit co=puter system rather than by individual indicators. This system sounds alarms if local flux condi-tions exceed predetennined values. When the reactor is depressurized, the incore detector assemblies can be in-(' serted or withdrawn through guide tuben which originate at a shielded area in the reactor building as shown in Figure 7-13 These guide tubes, after com-pleting two 90 degree turns, enter the bottom head of the reactor vessel where internal guides extend up to the instrumentation tubes of 52 selected fuel as-se=blies. The instrumentation tube then serves as the guide for the incore detector assembly. The incure detector assemblies are fully withdrawn only for replace =ent. During refueling operations, the incore detector assemblies are withdrawn approxi=ately 13 feet to allow free transfer of the fuel assemblies. After the fuel asse=blies are placed in their new locations, the incore detec-tor assemblies are returned to their fully inserted positions in the core, and the high pressure seals are secured. 7 3 3 2.2 calibration Techniques The nature of the detectors permits the manufacture of nearly identical detec-tors which will produce a high relative accuracy between individual detectors. The detector signals must be compensated for burnup of the neutron sensitive material. The data handling system integrates each detector cutput current and generates a burnup correction factor to be applied to each detector signal before printing out the corrected signal in terms of per cent cf full power. The data handling system cc=putes an averc6e Power value for the entire core, nomalized to the reactor heat balance. This average power value is compared to each neutron detector signal to provide the core power distribution pattern. O 7-29 (Revised 1-15-68)

                                                       /

7333 System Evaluation 73331 operating Experience The AECL has been operating incore, self-powered neutron detectors at Chalk River since 1962. They have been successfully applied to both the ERX and NRU reactors and have been operated at fluxes beyond those expected in normal pres-surized water reactor service. 73332 B&W Experience Self-powered, incore neutron dtectors have been assembled and irradiated in The Babcock & Wilcox Company Develop =ent Program that began in 1964. Results from this program have produced confidence that self-powered detectors used in an incore instrument system for pressurized water reactors vill perform as well, if not better, than any system of incore instrumentation currently in use. The B&W Develop =ent Program includes these tests:

a. Parametric studies of the self-powered detector.

b. Detector ability to withstand PWR environment.

c. Mdtiple detector assembly irradiation tests.

d. Back6 round effects.

e. Readout syste= tests.

f. Mechanical withdrawal-insertion tests,

g. Mechanical high pressure seal tests.

h. Relationship of flux measurement to power distribution experiments.

Preliminary conclusions drawn from the results of the test programs at the B&W Lynchburg Pool Reactor, the B&W Test Reactor, and the Big Rock Point Nuclear Power Plant are as follows:

a. The detector sensitivity, resistivity, and temperature effects are satisfactory for use.

b. A multiple detector assembly can provide axial flux data in a single channel and can withstand reactor environ =ent. An asse=bly of six local flux detectors,. three background detectors, and two themo-couples has been successfully operating in the Big Rock Point Reac-tor since May 1966.

c. Data collection systems are successful as read-out systems for incore monitors.

d. Background effects will not prevent satisfactory operation in a PWR environment.

O 7-30 0158

Irradiation of detector assemblics and evaluation of perfo=cnce data are con. (] tinuing to provide detailed design infomation for the incore instru=entation V system. 7.4 OPERATING CONTROL STATIO!!S Following proven power station design philosophy, all control stations, switches, contrcliers, cnd indicators necessary to start up, operate, and shut down ecch nuclear unit vill be locatei in one control room. Control functions necessary to =aintain safe conditions after a loss-of-coolant cecident vill be initiated from the centrally located control room. Controls for certain cuxiliary sys-te=s may be located at remote control stations when the system controlled does not involve power generation centrol or emergency functions. 7.h.1 GE !ERAL IAY0 LIP The control room vill be designed so that one man can supervise operation of the Plant during normal steady-state conditions. During other than normal op-erating conditions, other operators will be available to assist the control op-erator. The control room vill be arranged to include operating benchboard cu-bicles to house frequently used and emergency indicators and controllers at close proximity and visibility to the operator. Vertical panel sections of the cubicles vill house less frequently used controllers and informational displays. ' 7.h.2 INFORMATION DISPIAY AND COITfROL FlWfION The necessary information for routine monitoring of the nuclear units and the Plant vill be displayed on the control room benchboard cubicles in the ime-dinte vicinity of the operator. Information display and control equipment frequently employed on a routine basis, or protective equipment quickly needed in case of an emergency, vill be counted on the benchboards. Recorders and radiation monitoring equipment vill be mounted on the vertical panel sections of the cubicles. Infrequently used equipment, such as indicators and controllers used primarily during startup or shutdown, vill be mounted on side panel sections of the cubicles. A computer for each unit vill be available in the control room for alarm moni-tering, performance monitoring, and data logging. On-demand printout is avail-able to the operator at his discretion in addition to the computer periodic log-ging of the unit variables. 7k3 stM'ARY OF AIARMS l Visible and audible alarm units vill be incorporated into the control room to warn the operator if unsafe conditions are approached by any system. Audible reactor building evacuation alarms are to be initiated from the radiation mon-itoring system or manually by the operetor. Audible alarms will be sounded l1 in appropriate areas throughout the Plant if high radiation conditions are present. ' O Oi59 7-31 (Revised 1-15-68) t k

7.4.4 COMMUNICATION Independent Plant telephone and paging syste=s vill be furnished to provide the control roc = operator with conctant " hands free" cc==unication with all areas of the Plant. Acoustical phones vill be supplied in areas where the background l1 noise level is high. Co==unication outside the Plant vill be through the full period leased lines of the Florida Telephone Co=pany. 7.4 5 OCCUPAUCY Safe occupancy of the control room during abnormal conditions vill be provided for the design of the control room. Adequate shielding vill be used to main-tain tolerable radiation levels in the control room for maximu:: hypothetical accident conditions. The control room ventilation system vill be provided with radiation detectors and appropriate alarms. Provisions vill be made for the control room air to be recirculated through HEPA and charcoal filters. Emer-gency lighting vill be provided. The potential magnitude of a fire in the control room will be limited by the following factors: a. Materials used in the control room construction vill be nonfla=mable.

b. Control cables and switchboard viring vill be constructed of materials that have passed the fla=e test as described in Insulated Power Cable Engineers Association Publication S-61-402 and National Electrical Manufacturers Association Publication WC 5-1961.

c. Purniture used in the control roo= will be of metal construction.

d. Combustible supplies such as logs, records, procedures, manuals, etc.,

will be li=1ted to the a=ounts required for Plant operation.

e. All areas of the control roo= will be readily accessible for fire extinguithing.

f. Adequate fire extinguishers vill be provided.

g. The control roo= vill be occupied at all times by a qualified person who has been trained in fire extinguishing techniques.

The only fic==cble =aterials inside the control roc = vill be:

a. Paper in the form cf logs, records, procedures, manuals, diagra=s, etc.

b. The coaxial cables required for nuclear instru=entation.

c. S=all arounts of ec=bustible =aterials used in the =anufacture of various electronic equiptent. The above list indicates that the fla==able materials vill be distributed to the extent that a fire vould be unlikely to spread. Therefore, a fire, if 9 7-32 (Revised 1-15-68). 0160

           . started, vould be of such a small magnitudo that it could be extinguir.:hed by the operator using a hand. fire extinguisher. The resulting smoke and vapors vould be removed by the ventilation system.

,, Essential auxiliary equipment will be controlled by either stored energy, closin6-type, air circuit breakers which vill be accessibic and can be manually closed in the event d-c control power is lost, or by a-c motor starters which

           .have individual control transfor= crc.

7.k.6 AUXILIARY C0!! TROL STATICI!S Auxiliary control stations will be provided where their use simplifies control of auxiliary systems equipment such as vaste evaporator, cample valve selectors, chemical addition, etc. The control functions initiated from -local control sta-tiens vill not dir'.ctly involve either the engineered safeguards equipment or the reactor control system. Sufficient indicators and alams will be provided so that the central control room operator is made aware of abnomal conditions involvin6 auxiliary systems equipment controlled by local stations. 7.4.7 SAFETY FEATURES The primary objectives in the control room layout are to provide the necessary controls to start, operate, and shut down the nuclear units with sufficieat in-formation display and alarm monitoring to insure safe and reliable operation under normal and accident conditions. Special emphasis vill be given to main-taining control integrity during accident conditions. The layout of the engi-O neered safeguards section of the control board will be designed to minimize the time required for the operator to evaluate the system performance under accident conditions. Any deviations from predetermined conditions vill be alarmed so that the operator may take corrective action using the controls provided on the control panel. k

• m.'"

0l61 7-33

CHAIREL l CHAICEL CHA!CEL CHAICEL 1 l 2 3 4 I HIGil NEUTRON FLUX _ g HICH REACTOR i OUTLET TEMP. ' l OLAS P SSUR3 "0R" l "0R" "0R" "0R" GATE l GATE GATE GATE LOW REACTOR FOR FOR FOR FOR COOLANT PRESSERE TRIP l TRIP TRIP TRTP HIGH REACTOR START- l UP RATE (below 10 per  ; cent rated power) LOSS OF REACTOR COOLANT PUMPS l I INPtTFS TYPICAL OF ALL FOUR CHANNELS l l I I{ BISTABIE BISTABLE BISTABLE O l l BISTABLE I A B C D ___________J c/4 e/4 COINCIDENCE CODICIDENCE P y ROD DRIVE ROD DRIVE POWER SOURCE NO. 1 POWER SOURCE NO. 2 BREAKERS BREAKERS REACTOR PROTECTION SYSTEM BLOCK DIAGRAM CRYSTAL RIVER UNITS 5 & 4 0162  % =- eicuRe 7-i

1 \

      \

k 4. C. et f.3 W W W O@ s

                                                      '"*4';P x :... - ::
                                                 ....c Pt.t.

P y y , , SUPPL T L 3n C. .st ,, .$ s s

                  - - . . . . . , < .                 ..P,,,,,
                                                                           .- -       h...          .,,            h
                                                                                                                   .                    ...Oe
                                                                                                                                       - . ~ .            -        -
                                                 .... c1 P..

Sb8Pa t

                                                                                      /- /.. A /-                                               so
                                                                                                                                               ..f t es, II.
                                                                                                                                                                /
         ::'t <                                                            nP, ia i Q                             @Q e.<
                                                 . . . , (u__

j h,,.,...,q,-o'a

                                                                                                                    ;.......

P.st .

                                                                                      /                                                Oe SOPPL ,                                                                             .. ,5 61, f.

y

                   . .,( .wi g s . ,3 . 16t C..a.,18
                                                  .. .. g se t 7

P..

                                                                                      /_ . . ...

St.P Pt f f

                                                                                                                                         ,o L
                                                                       ~               ~7'                                   s, PD.F. ut,t.$

I SUPPLT

                                                                    ~
                                                   ..c -                                   ......

t

                                ....           ...      W       ,,,

I , Spar L T

                                                                                                                             % '. ..      '....o "d                                  I                                     O,             $0 FD.E. .st,f.S b

I SUP't? I

                                                                     ~
                                                   ...<            ' ~-                 .....
                                                                                                                                                                   -l
  ~

O/ /O P0.f. l I SU'PL , I t 0163

1 av9 8 < te . O ,9 O x r;, 4 y - LJ ' o ter t .e it. Oa ,

 ~                  J-        r;,

y 4 7 i' ' L_J >3

       ,,,.g
   ,.W H f 60. I'l' 8.eiui.

8 1 gy. , 3ga r,, y , 4 gialf t, .ast W) .f u f 0

   ,rDilitiDe i'li t8                gy, , ,gm O

_ < ,..... u.a

          .ue.u :

H _ - < ieingite, co.i.a s s,.. r-1 Q s .e n o.o.,:a i n , i . .. .c, = i.e...... .....$. 9 ......

                                          >, '-er I- I L_I

. m n , . .i 1

   ...u,.....,2...i  ..u r e.

9 g ...n ,, 1 ;f. GJ L_ ' > #0el . l I nt t f. Et att O. E 40ft ti tu Belf t. g 3 .v. alt l 1 9 g ~~, HUCLEAR INSTRUMENTATION SYSTEM - . . . , , ...u,- O >' 1 CRYSTAL RIVER UNIT 3

    .,u.....1.                                               ; Co...  .. .
                                           ....s.

____m e,.....u.." 3_ FIGURE 7 2a

                                                                           }hk          AMEND. 7 (715-69)

( CCometa f *aestt ( 1, 1, BC 0 L.e 4.* Petiswas ,  ;, gy g MIS #60s g

         ,, g (3F484199 BC Ft.e St at, g, 94't        l, t                            -8 g                      t e.P .

89ule tIW(1 seme B19 00%  : *, DC ICM ** Stest se eaTE ..p

                                                                                                              --                          ..I FtBI 988            Pe*                  'e 19 8 684708                              I'p;                                             a-._                          . _._
                                     .C.u   , .C f 0e 1908                e  p es II leafvet          I      e                     .: 3 P0pte-FLOm .6 07 F             ; e                          ;g
                                                         #C Ft .e          3,,yrga
                                                                       . . .        ,        n-              :: .                          ,,

1> SC

                                                       . .e                                                 ..,

a s. a.-',- Pye 2;p

                      .FL.G.B.Pu.er

e. . .,  ;,_

e. .;- ,,,.

                                                                                          , ,e ,-    t_.w                                  ,,>,
                                                                                                      -r                '

I 1... 1 ., e

                                                                                                  ',              I                        "         
                                                        =                . t-{.3                  i               j ..,

J ..,

                                                      . .e                                        i

s. 1 (se eB4794

• I T '" T ,, e ,.

                                                        =                                         i                     1 i                   g,.                                             t - - - _ 7 _ __ _ _. _ _

i ... i .mi t

                                                                                                          ..... , .          , t,,s
                                                                                                                            =' 4~-

3:% -- Z, t e :- . p O.. . .,,,- ', , m.c o i SC Flee L 00, . fles-,0el s g g ., CDsPestige Convent ses gesel 1 et . C I f Sit Flevet 3 e. 500

                                                       ' L 8'
                               '                     LQ3P 3 D-aC py,y                 ,gg flat, Pye.e c9amt           can,act      ._ A

v. .am e son on 4.

f . 0165

e.. .. . u s.

   -c                       .
      ],                   1.                                                          e
                                                                                                                                          .  .c.o.c = .
                                                                                                                '.- ,                 >     . , . is n., em ,s..

.>  :: a s s

    ~'                       "
     .,                   = .                                                            el g                      EH,3t,o..,..._,
                                                                                                                                             , , ,=_ss ,=
                          ~~ .
                                                         .tssa.

si

                                                            =s,=

_.,v .t

                                                                   . vime
                                                        .5 ,4mt...,se.
                                                                      .o...t.cu

a. a q e a. .

j

     -.                   :: .                                                                                                    s
     ..'                 ~

e - h > %*, .','0 "'*"

                                     =,

k,,at!

                                      ,m 8                     %                 >
                                                                                    ,,,,,,.u.                                       ,
                                                                                    .......c.

7 l

                                                                                    ..u .xo            n-,

IT '"" TI "'

                                                                                    ' " ' ~ ~ " ~
                                                                                                                                         ==

n. ."

1, 1 l g. g ' s I .., I ."o i i

 .u       m 1

I

1. l l

. ._. _ ._ r - - - - --'I . n ai= 3-

         ""'                                                                      . . . u .i. a
                                                                                                   <                      a mu u.. ..itw.m .                                         >....,,..,.,,

l 7... o um

                                                                                                              -ei... .    .

s i Uc '.h', :: h-4 '

                                                                                                                  ,: * .t L .,
                                                                                                   *' 'l

• i p ,.'9815'.=,a., ,..i, , a

                                                                                                                                  ,0. .

8 LII ,0. cp.,, **,60. Sif , .G ( r.3

                                                                                                        @        .= -i.         n. ou           c.it REACTP8) PROTECTION SYSTEM CRYST.'4. RIVER UNIT 3 0l66                                  hi =               ._. riouRe 7-26 AMEND. 7 7
                                                                                                                                                                           'g

               -,..- .~~ . -        .-             .,                     . - - - _ - . -                                 - -       -__                -                --                                 - _ _ _ _

J-2- 1 3 7 r i - i O l i e .  ; 4 1  ? 4 i fis ' ~5 l smamaos acTuna systu  ! {l li I ! l-+ i _- ii il i  :  :-'

- [ ,

j [  ; O--* - t+ .-C *.

                                          ;                                                                   8       I i

I O._. i ==

o ..l.

i-e l-

,                                                                                          s                          [-

i ). r.r-.4$, [

                                                                                                                                                                ~-

4

r. 4 ,. i

                                                                                                                                                                                          ~

6Y. i. D 1

                                                                                                                                                                                =

j o .. _ O

                                                                                          -w

c. 4. A. [i

o._.o _.- ! O, ") I . TE2"

.

(~C a. o. f o .. 1 i 9 ... .

                                                                                          "~

a. o_f,, . .

L _. i

                                                                                             ;.      y.           ..                                          o ~~~-

v-4 m. t., o o e :::=.= ~ O-_ c.

                                                                                             +4t G-..

c ,. a. l

                                                                                                                                                              * = = = -

i

                                                                                            -o &, j .,

. t. t. {%

o .-

                                                                                                     =        ~
                                                                                                     ;        ;       -

O - Q t" g ".p. 1( k .. r nc I .1 i 1 SAFEGUARDS ACTUATION SYST2M ! CRYSTAL RIVER UNIT 3 O- ~ ~

L FIGURE 7-2c 0167 . 7 <7 i ,>

g TvPsCAL ; E ACTOR ButLOtMG . SOL ATIOM VALVC COMTROL CsRCusT FOX t u (CtRCu tT BRE ME,

                                                                                                                                        ^
                               \
                               ;                                      ll                                                             :    %

VALVE p_

                               -N  -
                                                                      ,,C
                                                                                              MOTO R
                                                                                                                                     ' ~
                               -N-                                    ,,C                                                          _r_

d

                                                                      ..O                                                                          (
                                                                      ,,o i,

m (+) pm 2 r Aou * - ~ ~ *1 Et C RI .a.CR 2 SAIEGuAQDS CRt AFEGuARDS 3 Tyg ACTUATIO N ALTUATiOu , ;__.J SAr( trrBTEM. SN ET _ _ _ _: AcTu tr OPEM A. g si2 SY*8 C. S II II 125 VOLT D 1 ii Q C.LOSED CLOSE c,SS U N

• STAI'0" " IE"Y

                                                                                                                                 /

N gg7 l STOP 3g7g .j _18,.=

          -d_            : :                              - m -                                    C   ,

l $ CC

                   ~

ggj SL% 7 LS C.A l 3kS 17TS OPEM1 4r ; ,= ; ees (~ ~ UCN

                                                               ;4 Ii
                                                                O
                                                                         -ll      -
                                                                                      =n iI i ' 5 LS h              N Y

COMTACTS EwowM su O SWITCHES AMD C%)MTACTS SHOWM WITH \/ALVE IM FULL OPEM POSITIOM I' " D' # C - MAtu COMTAC. TOR g CLOSiMG ' # O- MAtN COUTACTOR OPEMsMG - T BREER CLOS CR 3-COMTPOLRELAY TC-CtRCutT BREARER TRsP qq qgy7pg( pggy TS-TouQuC SMtTCH 50-tNSTAMTAutous CNER' L% - LtMsT SWITCH :1-TiVE OVERCURREMT i

                                                                                                                $$-CIRCutT BREARER COL 15?.-CtRCU1T BRE A%ER AOV' 18G- AustLI ARY TRtPPtMG '

MOTt: CRE ,%wCWu OOTTE OutY WHEM QEDUuCANT REQ L)) REO LIMIT SWITCH CONTACT DEVE@kAEMT

                     %Lv0        iNTERMLDIATE      VALVE         COMTACT RILL OPtwWVE POSsTm FtAL CLOSED           FUNCTIO M I      CLOSED          Optu          OPEN         SPARE E      CLOSED          OPEM         OPEN      OPEN sMD LT 3       OPEN        ClostO         CLO SCO      OPEu uuiT 4       OPEN        CLOSED         CLOSED           % PARE 5       OPEN           OPEN        CLOSE.O    TORQUESW E5YPA%

r. OPEM O PE M CLOSED CLOSED NO LIGHT 7 CLOSED CLOSED oPtu '040ut %w BYPASS

( , S CLOSED Clotto OPEM GO-tu C\RCutT f g OPEuMG TORQUE SWITCH OPEM5 OM MECHAutCAL OvCRLOAD IM OPEuiMG DIRECTION CLOSim TORQVL SWtTCw OPEMS OM MLCHAuiCAL O OVERLOAO nM CLostMG DiREtinou.

1 iCAy M2 AT f:CMOVAL PUM P CONT AOL CIRCunT 70R REACTOR BustotNG spRAv puup j l WOTOR mRytR CournOL) a cogynot) CT %A 4l M - : CT PUMP ^ PVVP m _ _ ee

           "' L      ,            MOTO R                                         

VOTOR LT ^ ,, -f  ; ; ..

~
     "O      50   50                                               g Tinr

si st 1

L hO.L. STOP ^[

                                      ~
                                                   "=           ==                        "
                             .L
                             --Si e                                                                          _

NARM EROM'

                                     --50                   Ml
**M       Ctb   --g                  ~~

CRn S AF EGUAA0% ACTUATION Rt

                       ..gSg                                            SYGTEM O

e. - ,

GUARDS

"'N      ICR h)        TC Y             ..e Oggg 4i 3  g y

EEutMTED POSLTLOM CONTACTL SHOWM tu OEtutRGtztD PostTIOu CRl-CouTROL RELAY uG CON M- M aim COMTACTOR

 ' 0)L

UXRENT RELAY RELAY iTROL %WLTCH LOARY SWITCH LEL AY Osb05EO t

CONTRotig TYPICAL CONTROL CIRCUlTS FOR ENGINEERED SAFEGUARDS EQUIPMENT CRYSTAL RIVER UNITS 3 & 4 0169 E= - emuRe 7 3

l 108

                                                                                               - n h Design s,           -1W. 5                                                                      -

Overpower

                - 107          N      "    *    #           #

d - 3 _ W = Megavatt De=acd

                                                                                               - U2
                                                                                               - 111 108                                                        - 110                 l 107 5                                                   1 l

1W - 109 i

        ,   - 102 3               -
                                                                                              - 108 Limit

_ ic7 1

            - 100                 -
                                                                                              - IC6 1M. 5 I -                         -

1W - 105 e j _ _ 102 Nd -

                                                                                              - 10h
       "                     "            Limit    ..

3 - 1C3 id g.- 108 g - 100 g

                                                                                              - 102 g                     g                     g                    g - 1Crl. 5       =

g- j j g - 1W 101 Y- Y - Y - 102 N 5 -

                                                                                              - 100 W E                    $                     E             Lknit   N g-                    g    -

g - y- - M L t

       .                     .                    +                     +

g - g - g - 100 - 98 gg 5 9 9 V

                                                                      ,u   -
                                                                                              - n f - 90                    -
                                                                           - 102 N d      n   - 96 Ia -                  $

e a

                                                                      $ga _

23

                                                                                  ""i t j - 95 N-                    N   -                 E     -             E a - 100           5 - 9h 3                                                                                   8 g-3 g   -

1 g - 33 g a - 93 w w w wW-3- 3 90 3 _ 3+ - - 92

                                                                                              - 91 Reactor Nip at 107 5% + 0 5% Indicated Power l

l l REACTOR POWER MEASUREMENT ERRORS & CONTROL LIMITS CRYSTAL RIVER UNITS 3 & 4 l j7} E FIGURE 7-4

O 62o Reactor outlet Temperature 600 1 g Steam 4 ( A~~.f ** N

                                                                      \
   - 580 h                             I Reactor Average j                                  hmperature                                   \

e 560 N Reactor Inlet Temperature O suo - - _ . T at 925 osia sat 520 0 20 40 60 80 100 Rated Reactor Power (2,452 MWt),% l REACTOR & STEAM TEMPERATURES p VERSUS REACTOR POWER k-

                                 .              CRYSTAL RIVER UNITS 3 & 4                                l hE-                              FIGURE 7-5

                                                  - TO OTHER,CONTI ROD DRIVE GR3;

( AUTOMATIC

                   - - - - -=            SEQUENCE
 '                                         LOGIC l

l l I l l STEAM l GENERATOR FEEDWATER - l I l 'c,a 'c2 .

                ;

I I l r I l l . d DEADBAND +

                                                                             - RE l                                                                             -      ST
                                                                             -- CC Tc . RE Th     RE T    A'i di . DC Ep . RE i . R' d

N; . Rt "e 0172 f (

,0L PS f L CONTROL ROD DRIVE GROUP STEAM Th p in, STEAM REACTOR GENERATOR

                                                             - FEEDWATER
                                       #                        T c, ,

i lb LOOP NO. I Th i N; I - AVERAGE = Tc,, I 1b N AVERAGE E aT-. f.- d- EMAND AND _ - COMPARE COMPUTER COMPARE LOOP NO. 2 Th 2 l AVERAGE - Tc SET 2a

                                                                                                                   *2b LEGEND ACTOR COOLANT SYSTEM EAM SYSTEM NTROL SYSTEM ACTOR COOLANT SYSTEM COLD LEG TEMPERATURE 2CTOR COOLANT SYSTEM HOT LEG TEMPERATURE ERAGI REACTOR COOLANT SYSTEM TEMPERATURE SIATION OF AVERAGE TEMPERATURE FROM SETPOINT                                                   REACTOR CONTROL DIAGRAM 2CTOR POCER LEVEL ERROR, N         -N INTEGRATED CONTROL SYSTEM 3 VELOCITY DEMAND CRYSTAL RIVER UNITS 3 & 4 3ADATT DEMAND XTOR POWER LEVEL                                                                                   Em           FIGUR E 7-6 ACTOR P00ER LEVEL DEMAND AMEND. 3 (3-i-68) 0173 i

O O O 4.0 3.5 - y Max. Differential j [ 4 3.0 Worth = 1. I x 10-4 (ak/k/ sec) e ' [ J bu 2.5 ,- / f

                                        $                                                    )
                                                                                               /

3 2.0 u  ! o ' l. 5 '

           --                            .                                       /

N n

                                        $  l.0
                                                                          -l
                                                                            )                           i
                                                                  /

O. 5 ,

                                                              /        /
                                                                     /

a$ 0 # l 8 ' ' ' ' ' Yo 0 20 40 60 80 100 0 20 40 60 80 100 OEh Group 5 Position Group 7 Position

        ,1          a n .- =N h l1l N I$n                                                    0     20      40     60     80           100

, , hOhz Group 6 Position 93e %ag:- c; m2 m cg<m 8

4. m Distance Wi thdrawn, f.

3 m g>mg 7 " 5No ag 5%.

g,. m m-s , _ - '* . . -se-<. I D $0C ll R( QL E hC T-] (- i M Ai f C0eTROLL E R ] - e MAR IMUM LOAD L SMIT l 4 t Mitl mom LOAD LIMIT ~j h R ATE LSMIT l h

                                                                                          $ NIM /$4F ETY 800$ 50T OUT Runsacu LIMITER AS$fMETRIC 000$                1
                                                                                                                                            < m MEEasATS ses OVTPWT a       MEG &o&TT i

I E RROR PREG $y g < ERROR I l Suu l N PRE $$URE $ET Poluf ( ADJUSTED) l CouTROLLER l l CouTROLLER l 1 1 T.R. . .E T R. , .E STPAS$ $PE E0 V AL VE CORTROLS RE ACTOR C00LAsf c P- 1 RE ACTOR COOL ANT RE ACTOR C00Lauf PUMP CAPA81LITV - PUMP L IMITER RusteCE $1Ga&L REACTOR b C00 Leaf PUMP 2 RE ACTOR C00Lauf I TEMP LIMITER gg L0s FEEDwATER 7 TEMP LIMITER ft

                                                                                                                         $YE AM GENERATOR                g LEVEL LIMITER                    1 i

l CO.TRauR l l CO.TRaut l .,, FEEDWATER FL0m

                                                                                             $f ARTUP                         CouT ROLS FEED VALVE                          LOOP I
     -n 1                                                    0175                                           -A      IR --

y .

r, N i l f #1Wlatt 51i f. 0i!T LOSS 08 L0a0 L t Et#0 f i t 0 P Wa** C AP ACI T T SDC . SYSTid CISPaTCH Cimita ng,,,g Ruh9&CE $1GhaL & Olf f tRENCE P 3 - STian PRt 35ual I Pg5ET . 5f t aw 5ET Pets 5 tat

                              ,g ,'                                                                                          e, . FEt0mafft PLou
        *5                    C0kTROL H . wamual SET 70ini N/a . Manual /aut0 Mafic L . 5f tau GistRaTOR LEVEL C0pp aR AT Os                                                           T, . RE ACTOR OUTLif YturtRafvat T . #taCTC# InLti Tturtaafget c

it ~' I f . Ef tDs4TER YturinaTURE Tht uJ"8tR Sut5CaiPT5 af fit 10 LOOP 1440 LOOP 2 fitD' af '# Otua 0 CattutaTOR

              < - TOT al fitDeattR Olwan0 ST[aM 6tilf aTOs tea 0 Cafl0 Con 1ROL i

RtACTOR COOL as T Puur 3 RiaCTOR C00t ati RiacT0e C00L AET PUwP L IMITE R ' JMP CaPaRILITT Rt ACTOR duntaCE 5, Gnat

   ~
                    - C00L ani                     f PUuP 4 RiaCTOR C00LaWT cf     FM        'g2           I e2          TEMP Liptfit
            ' A                                                        -

Lou fitCaaf tR I g2 T E MP LlulitR 5f tau GlutR ATOR k

• 2 atWEL LIMITER

                                     'f2         l CouTROL                    l ConTROLLit l j

81 E Deaf t R c[.'TEt3 sf a"v, STEAM GENERATOR AND TURBINE CONTROL DI AGR AM

                                                                                 'Sto 'at" t00P 2                                                        INTEGRATED CONTROL SYSTEM l                             h                             CRYST AL RIVER UNITS 3 &                                        )

i __ rito.ute rLO. L1 =.. FIGURE 7 8 J

t. . _ _ _

7 O 10 10 i 1

                                              -- 10'                   - 125 1E --                                                     "
                                                                   .- -lM
                                                       ~

10 8 -

                                                  - 10 '          3 m __ io E   _   -10 10       --

1 *- - 1 5 i

                                                    .LO*I               I
    >  10        --

f

   ".                                         -   8
   !!      5 1o       _-

C c y ,8 - -10'9 5 10 -- j g _ tos - .,10-10 h 105 __ y - 10 _ 10'U 3 10 -- 8

                    #-          103 10       --

O q o. V 3-- 102 1 -- g g

          -1              -- 10 10       --

3 e

         -2

__1 k 10 -- 3 Iy A 8 8 o y y

NUCLEAR INSTRUMENTATION FLUX RANGES CRYSTAL RIVER UNITS 3 & 4 l 0177 i =- FIGURE 7 9 l __ _ _ .. - - -

{} PC CIC UCIC UCIC

                             .' 4        4.
               'o'-l  ,  ,

4.>- c, . . :

                                                        .<t I

I I a,', i i

                                                        , l, >    ,

x-

                                                      ,g.k-
              ' ,v . ' '                                *
                            ,' e' ', . 4
                              '. -[ '; . -

UCIC UCIC O cic ec LEGEND PC - PROPORTIONAL COUNTER - SOURCE RANGE DETECTOR CIC - COMPENSATED ION CHAMBER - INTERMEDIATE RANGE DETECTOR UCIC - UNCOMPENSATED ION CHAMBER - POWER RANGF DETECTOR l l NUCLEAR INSTRUMENTATION DETECTOR LOCATIONS

 'd                                               CRYSTAL RIVER UNITS 3 & 4 l

0178 N hE- FIGURE 7- 10

s____., c_4 Ct;JH ' LNL  :* ( e , ____A___, r ( $ l

'-~ ~u. AT IA

_ X"^l j j ua i , t Al l Qa' ent ssu pirc a l ' go e ( Uta i V - _ _ __,_ - - - _ a _ l9 W Aa' l i v v. l,f A* v'a U 'a A ____l i i l

:2____

i

                                                    .1 m              ,       v                N                                  i

(~ '$ i O , i __ pop a~, a i

                                                                      ;

( _ U to A _. .______) - LO l 1 M, 1 ll j l et..cios i r_-____-_, , j cyg i m.ao. , l  : , i p. l oncr = w_i i 3 i,r _ _ +~ma c. ] j -- L _ _sv____a mu i F^;r.;-;r.r, _7_ V

                   , = Tc                           _._ _ s_ _ ,

i s N, __ ,,___ ,,____,

                                                                                                                                   ;

4,

          !        l-- --- J d b d h..h.M.h                                                         l g;g'
                                                                                                                                                            ~--                (

y.y.y. y y.y.yo, j

            ;

1 'l 1 1 .1 , , l

a ,, n. , ,... ta j

v-- N, l 'e V4 (' i ea' ~ v, i lA Al (  : VV m A,- V ~ ',l l lQo Q ,l u I Ani r An, v A v = - ,l p - > _2 _ _ _ _ m _ -

             ,                  l                        ,      Am               \                                                    l                                  c --

Fv vr-----, i

                                                                                                                                                                    - 1 ";;
                                                                                                                                                                  ,' -- 1    s' I                  i
                                                                                                                                                                         ' ~ ~

l - - - - -

                                                        --- N: N,o,S.L                      ,___--_------- ---_           -.
                                                                                                                                    .,. _ _ _ - - _]

l i n i i___-____a s . i i 1

                                                                                                                                          ~ __ f oven voueToa l i

l utaa o. l ,A, l DIl cMac oaa.T.es' l l I. WMP I '

                                                                                                                                          ,___] pump powiroa l i

D  ;

                                                                                    , -- + V OT1' L c Tc
                                                                                                                                                     ,A, I               I I                                                                   ,     MAKLUA                                                          y, l                                                                   g       WATER                                                        g a_ _gwg I

l i

                %---_       --- _---                    _---_-_-__)

r

                                                                          ~

0179 r e

O . e eO @ g retai core monitors

               @                     @           l     based on 1/4 core e          @   S O,        e              Sv==etrv-90          e         el          @-

e @ ie O_ e e @i e e O symmetry Monitor $ SO # e. O e e C e _! SO 9 0 'O e1 9 O' O I g Combination total core O eeO el ano sy...try monitors, e e Ii d l I i o a s b AXIAL FLbX

SHAPE

                                   \     ACTIVE CORE E     j                                LENGTH E     E e

4 h h BACKGRCVND DETECTORS l LOCAL FLUX DFTECTORS INCORE DETECTOR LOCATIONS CRYSTAL RIVER UNITS 3 & 4 E. FIGURE 7-12 AMEND.1 (1 15 68) w-

( l l

                                                                                                       ~~
                                                                                                    ,1 OETAIL-A                ,   \                                           l C

• INSTPuvENT i 9. 1

                                                                                                   ;           ;
                                                            *,       CAlleRATION                        -

l AREA k /

                                                  ,                         . .          ,     ,                 .        .;
                                                                                    ,                                                 CAllBRAtl0N TUBE CONNECTOR q,                    _-     _-_--_I                                                              -

h Ni _j j _ ll _ H j a . ELECTRICAL JJ jp -.

                                                                                                                           , ,g'        CCNNECTOR 4

OUTER . INCORE FLUX TUBE C WONITER ASSEwSLY

                                                              ,                                                      ,,U         ! y 2500 PSI SEAL Y Y L                          ,

INNER TUBE

 %)                         -   -

e a .

                                                          <           .   (     .*-   ,

a *, , . * /

                        .               ,                                     i        ,                                              OUTER T'J8E
                                                                              ,                                        DETAll-A m
                                                                            /

I l l TYPICAL ARRANGEMENT 4 INCORE INSTRUMENTATION CHANNEL CRYSTAL RIVER UNITS 3 & 4 0182 1/)s_. eiouRe 7.i3 . .}}