ML19329E140

Chapter 7 of AR Nuclear 1 PSAR, Instrumentation & Control. Includes Revisions 1-18

ML19329E140
Person / Time
Site:	Arkansas Nuclear
Issue date:	11/24/1967
From:	ARKANSAS POWER & LIGHT CO.
To:
References
NUDOCS 8005300720
Download: ML19329E140 (47)

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TABLE OF CONTENTS Section Page 7 INSTRUMENTATION AND CONTROL 7-1 7.1 PROTECTION SYSTEMS 7-1 7.1.1 DESIGN BASES 7-1 7.1.1.1 Vital Functions 7-1

• 7.1.1.2 ' Principles of Design 7-2 7 1.1.3 Functional Requirements 7- 4 3 7 1.1.h Environmental Considerations 7-5 7 1.2 SYSTD1 DESIGN 7- 5 7 1.2.1 System Description - Reactor Protection System 7-5 7.1.2.2 Descriotion - Safeguards Actuation System 7-5b 7.1.2.3 Design Features 7-7 7.1.2.h Su mary of Protective Actions 7-10 7.1.2.5 Relationship to Safety Limits 7-11 7.1.3 SYSTEMS EVALUATION 7-12 3 7.1.3.1 Functional Capability - Reactor Protection System 7-12 7 1.3.2 Functional Capability - Safeguards Actuation System 7-13 71.3.3 Preoperational Tests 7-13 7.1.3.h Component Failure Considerations 7-13 a 3 7.1.3.5 Operational Tests 7-lh 72 REGULATING SYSTEMS 7-14 7.2.1 DESIGN BASES 7-14 7 2.1.1 Compenration Considerations 7-lh

, 7 2.1.2 Safety Considerations 7-15 103 T-1 oi

CONTENTS (Cont'd)

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Section Pm 7 2.1.3 Startup Considerations 7-16 7.2.2 SYSTD4 DESIGN 7-16 7 2.2.1 Description of Reactivity Control 7-16 7.2.2.2 Integrated Control System 7-19 7.2.3 SYSTEM EVALUATION 7-22 7.2.3.1 System Failure Considerations 7-22 7.2.3.2 Interlocking 7-23 7 2.3.3 Emergency Considerations 7-23 7.2.3.h Loss-of-Load Considerations T-23 7.3 INSTRUMENTATION 7-2h 7.3.1 NUCLEAR INSTRUMENTATION 7-2L 7.3.1.1 Design 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 7.3.3.2 System Design 7-29 7.3.3.3 System Evaluation 7-30 7.h OPERATING CONTROL STATIONS T-31 7.h.1 GENERAL LAYOUT 7-31 7.h.2 INFORMATION DISPLAY AND CONTROL FUNCTION 7-31 i 7.h.3 SUWJLRY OF ALARMS 7-31 7.k.h COMMUNICATION 7-32 l

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CONTENTS (Cett'i)

Se?'1 - Face

7.h.3 OCCUPANCY 7-32

7. 4' .6 AUELIARY CONIE0L STAHONS 7-33 7.k.7 SAFETY FEATrJRES 7-33 75 3YSTEMATIC, NONRANDOM, CCle10N 10DE INSTRUMENTKIION FAILURES 7-33 13 1

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10-31-69 Suppler _ent Nc. 13

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LIST OF FIGURES (At rear of Section)

Figure No. Title 7-1 Reactor Protection System Block Diagram 7-2 Nuclear Instrumentation and Protection Systems 7-3 Typical control Circuits for Engineered Safeguards Equipment 7h 3eacter Power Measurement Errors and Control Limits 7-5 Reactor and Steam Temperatures versus Reactor Power 7-6 Reactor Control Diagram - Integrated Control System 7-7 Automatic Control Rod Groups - Typical Worth Curve versus Distance Withdrawn 7-8 Steam Generator and Turbine Control Diagram -

Integrated Control System 7-9 Huclear Instrumentation Flux Ranges 7-10 Nuclear Instrumentation Detector Locations

!7-11 Nonnuclear Instrumentation Schematic 12 Incore Detector Locations 7-13 Typical Arrangement - Incore Instrumentation Channel 7-14 Control Room Layout

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7 INSTRUMENTATION AND CONTROL 71 PROTECTION SYSTDtS The protection systems, which consist of the Reactor Protection System and the Safeguards Actuation System, perform the most important' control and safety func-tions. The protection

• systems extend from the sensing instruments to the final actuating devices, such as trip circuit breakers and pump or valve motor contac-tors.

7 1.1 DESIGN BASES The Reactor Protection System monitors parameters related to safe operation and trips the reactor to protect the reactor core against fuel rod cladding damage caused by departure from nucleate boiling (DNB), and to protect against reactor coolant system damage caused by high system pressure. 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-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:

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

b. Loss of any two reactor coolant pumps. k

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

d. The reactor pressure reaches an established minimum limit.

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

a. The reactor pressure reaches an established maximum limit.

The Safeguards Actuation System automatically performs the following vital fune-tions:

a. Commands operation of injection emergency core coolant.

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

c. Ccmmands closing of the reactor building isolation valves.

The core flooding system is a passive system and does not require Safeguards l Actuation System action. l l

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6-5-68 7_1 Supplement No. 4

7 1.1.1.1 Nonvital Functions ,.

The Reactor Protection System provides an anticipatory reactor trip when the reactor startup rate reaches specified limits.

7 1.1.2 Principles of Design The protection systems are designed to meet the requirements of the IEEE proposed " Standard for Nuclear Power Plant Protection Systems," dated September 13, 1966. Prototype and final equipnent will be subject to qualification tests as required by the subject standard. The tests vill establish the adequacy of equipnent performance in both normal .nd acci-dent environments.

The major design criteria are simmarized 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 ac-tion is required.

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b. No sin 61e component fe.ilure shall initiate unnecessary pro-tection system action, provided implementation does not conflict with the criterion above.

7 1.1.2.2 Redundancy

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All protection system functions shall be implemented by redundant sensors, instrument strings, logie, and action devices that combine to form the pro-tection channels.

7 1.1.2 3 Independence Redundant protection channels and their associated elements vill be elec-trically independent and packaged to provide physical separation.

Separate detectors and instrument strings are not, in general, employed 3 for protection system functions and regulation or control. However, no more than one channel of protection is shared with control at any time so that failures in the sing el channel still have three complete protection channels which fully satisfy the sin 61 e failure criteria of the IEEE standard for reactor protection systems. Sharing instrumentation for pro-tection and control functions is accomplished within the framework of the stated criteria by the employment of isolation emplifiers in each of the multiple outputs of the analmg protection system instrument strings.

The isolation amplifiers are precision operational amplifiers having a closed loop unity gain and a low dynamic output impedance. The effective-ness of the isolation amplifiers has been proven by actual test.

The isolation amplifiers vill block a direct connection across their out-put of 410 vde or peak ac, 300 v rms without perturbing the input signal.

108 7-2 .

5-3-68 Supplement No. 3

This may be stated as a corollary to the design criteria: "a direct 3 short, open circuit, ground fault, faulting to a power source of less than 410 volts, or bridging of any two points at the output terminals of a protection system analog instrument string having multiple outputs shall'not result in a significant disturbance within more than one out-put."

Testing has demonstrated that the protection system design vill meet the

.above criteria.

7 1.1.2.4 Ioss of Power

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

. b. Availability of power to the Safeguards Actuation System shall be continuously indicated. The loss of instrument power, i.e.,

1 vital bus power, to the instrument strings and bistables vill initiate a trip in the affected channels. System actuation re-quires control power from one of the two engineered safeguards d-c power busses so that loss of this power does not actuate the system. The system equipnent is divided between the redun-dant engineered safeguards channels in such a way that the loss of one of the d-c power busses does not inhibit the system's intended safeguards functions.

7 1.1.2 5 Manual Trip Each protection system shall have a manual actuating switch or switches in the control room which shall be independent of the automatic trip instru-mentation. The manual switch and circuitry shall be simple, direct-acting, and electrically connected close to the final actuating device.

7 1.1.2.6 Equipment Removal The Reactor Protection System shall initiate a trip of the channel involved when modules, equignent, or subassemblies are removed. Safeguards Actuation System channels shall be designed to provide for servicing a single channel

without affecting integrity of the other redundant channels or without com-promising the criterion that no single failure shall prevent actuation.

.7 1.1.2 7 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 re-liability.

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5-3-68 Supplement No. 3

-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 reactor under the following conditions:

a. The reactor power, as measured by neutron flux, reaches an allowable limit set by the number of operating reactor cool-ant pumps.

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b. Two reactor coolant pumps are lost. (This also covers loss of three pumps and loss of all pumps)

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

d. The reactor coolant pressure reaches a preset maximum limit.

e. The reactor coolant pressure reaches a preset minimum limit.

f. The reactor startup rate reaches a maximum limit while operating below a preset power level.

g. The reactor power as measured by neutron flux reaches an allowable 3 point set by the measured reactor coolant flow.

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 predeter-mined startup rate, slover than the rate at which reactor trip

is initiated.

l l 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 or upon detection of a high 3 reactor building pressure,

b. Start operation of low pressure injection upon detection of a very low reactor coolant system pressure, or upon detection of a high reactor building 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.

I10 6-5-68 7_4 Supplement No.4

e. Start the reactor building spray system,upon detection of a high reactor building pressure.

7 1.1.4 Environmental Considerations The operating environment for equipment within the reactor building vill nor-mally be controlled to less than approximately 110 F. The Reactor Protection System instrumentation within the reactor building is designed for continuous operation in an environment of 140 F, 60 psig, and 100 per cent relative hu-midity, but vill function with less accuracy at the accident temperature.

The environment for the neutron detectors vill be limited to 150 F vith a rel-ative humidity of less than 90 per cent. The detectors are designed for con-tinuous operation in an environment of 175 F, 90 per cent relative humidity, and 150 psig.

The Safeguards Actuation System equipment inside the reactor building vill be designed to operate under the accident environ =ent of a steam-air mixture.

Protective equipment outside of the reactor building, control room, and relay room is designed for continuous operation in an embient 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 vill operate within design tolerance up to an ambient temperature of 110 F.

7 1.2 SYSTEM DESIGN 7 1.2.1 System Description - Reactor Protection System Figure 7-1 is a block diagram of the Reactor Protection System. The system consists of four identical protection channels, each terminating in a non-inverting bistable and reactor trip relay. In the nomal untripped state, each channel functions as an AND gate, passing current to the terminating bistable and holding the reactor trip relay energized only if all channel inputs are in the nomal energized (untripped) state. Should any one or more inputs to a channel become deenergized (tripped), tM terminating bistable in that cbnnnel trips, deenergizing the reactor trip relay. Thus, for trip si 5nn h each chan-nel becomes an OR gate.

Contacts from the four reactor trip relays (RS) are arranged into four identi- 3 cal 2-out-of-4 coincidence networks. Each pair of these coincidence networks controls the power to one of the two identical control rod drive power supplies.

The reactor trip circuits are shown in more detail on Figure 7-2, which is an overall diagram showing the Nuclear Instrumentation System (7-2A), Reactor Protection System (7-2B), and the Safeguards Actuation System (7-20). Figure 2B shows the circuit breakers controlling input power to each control rod drive clutch assembly and the manner in which the reactor trip relays trip these circuit breakers.

7-5 5-3-68 Supplement No. 3 iii

.s R:act r trip is accomplished by interrupting all input pov:r to e:ch drive clutch asst.mbly. Each control rod drive clutch power supply received its input power through_two circuit breakers in series so that opening of either interrupts that source of power. The two control rod drive clutch power ')

supplies operate in parallel so that both must be deenergized for the con-trol rods to trip. Circuit breakers No. 1 and No. 2 control primary power to one clutch assembly power supply, and circuit brekaers No. 3 and No. h control power to the other. Thus, reactor trip is accomplished by tripping one circuit breaker in each of these pairs.

The control rod drive clutch holding coil power supply circuit breakers are equipped with undervoltage coils which must be energized for the circuit breaker to be closed or to remain closed. The holding volta 6e for the under-voltage coil of each circuit breaker is taken from the vital bus.

In each circuit breaker (nos. 1, 2, 3 and h) the undervoltage coils are energized through contacts of trip relays RS1, RS2, RS3, and RSh under nor-mal conditions with all trip relays energized. If trip relays RS1 and RS2, h RS1 and RS4, RS2 and RSh,or RS3 and RS4 become deenergized, each circuit breaker undervoltage coil vill be deenergized, and the circuit breaker will open. Thus any 2-out-of-b trip relays will cause each circuit breaker to open, removing power.

The trip circuits and devices are redundant and independent. Each breaker is independent of each other breaker, so a 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 manual reactor trip switch is included in each .

of the circuit breaker trip circuits to implement the " direct action in the )

final device" criterion.

The power / flow monitor logic details are also shown on Figure 7-2. There sre four identical sets of pover/ flow monitor logic, one associated with 3 each protective channel. Each set of logic receives an independent total reactor coolant flow signal (IF), a " number of pump motors in operation" signal (Pn), and an isolated reactor power level signals (4).

Using a flux / flow comparator, one part of the pover/ 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 h 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 de-tected 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 pover/ 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.

112 6-5-68 T-Sa Supplement No. 4

T.1.2.2 Description - Safety Features Actuation System Figure 7-2C shows the action initiating sensors, bistables, and logic for 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 ini-tiating the desired protective action through redundant Safe-guards equipment

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

There are three independent sensors for each input variable. Each sensor terminates in a bistable device. The outputs of the three bistables as-sociated with each variable are formed into two identical and independent 2-out-of-3 coincident logic networks or channels. Safeguards action is 9

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initiated when.either of the channels associated with a variable becomes energized through the coincident trip action of the associated bistables. )

r The engineered safeguards equipent is divided between redunannt actuation channels as shown in Figure 7.20. The division of equiment between chan-nels is based upon the redundancy of equipent and functions. Where two active safeguards valves are connected in redundant manner, each valve vill be controlled by a separate engineered safeguards channel as shown in Figure 7-20. When active and passive (check valve) safeguards valves are used re-dundantly, the active valve vill be equipped with two OR control elements, each driven by one of the safeguards channels. Redundant safeguards pumps will be controlled in the same manner as rea nnaant active valves. Figure 7-20 shows a typical control scheme for both safeguards valves and pumps.

Separate dual logic channels for reactor building spray pumps and valves are 3 provided. The separate logic permits testing the reactor building spray sys-tem without actually spraying water by starting the pumps with the valves closed and opening the valves with the pumps shut off.

Figure 7-3 shows typical control circuits for equiment serving safeguards functions. Each circuit provides for nomal start-stop control by the plant operator as well as automatic actuation. Normal starting and stopping are initiated by momentary contact pushbuttons or control switches.

1 The control circuit shown for a makeup pump is typical of the controller of 3 a large pump started by switchgear. There are three makeup pumps, two are equipped with single control relays povered from separate safeguards actua-tion channels. The third pump is equipped with two control relays, CR1 and CR2, each of which is povered from separate safeguards actuation channels. )

Energizing the control relays through their associated safeguards actuation channel energizes the pump circuit breaker closing' coil and starts the pump.

The control of the reactor building spray pumps and decay heat removal pumps is by Jans of single control relays in each pump control circuit. Each pump is controlled by separate engineered safeguards channels. Safeguards action is initiated when the pump control relay is energized by its associated en-a gineered safeguards channel.

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 employed as one of two active redundant valves, then it is controlled by a single safeguards actuation channel to CRl. If the valve is employed with a passivt redundant check valve, then the motor operated valve is-controlled by two safeguards actuation channels with CR1 and CR2 connected in an "0R" configuration.

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. 5-3-68 Supplement No. 3

The control relays, when energized by their associated safeguards actuation channels, close the valve through contacts which duplicate the manual CIOSE 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 pemit torque closing.

Air-operated engineered safeguards valves automatically go to their engi-neered safeguards position upon loss of control air. Valves used with active redundant 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 controlled by a single safeguards channel operating in an OR configura-tion. Engineered safeguards action is initiated when power is applied to the electrical actuator.

3 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 mea-sured 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 in-strument strings vould not impair the function of any other instrument or protective channel.

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 identi-cal instrument' strings. The total removal of any one instrument string vill not prevent the system from performing its intended itnetions.

71.232 Independence The redundancy, as described above, is extended to provide independence in the Reactor Protection System. Each instrument string feeding into one pro-tective channel is operationally and electrically independent of every other instrument string. Each protective channel is likewise functionally and electrically independent of every other chnnnel.

Only in the coincidence output are the channels brought into any kind of com-mon relationship. Independence is preserved in the coincidence circuits through insulation resistance and physical separation of the coincidence net-works and their switching elements.

The Safeguards Actuation System instrumentation and control have electrically and physically independent instrument strings. The output 'of each bistable is ic 7-7 5

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_ ** electrically independent of every other bistable. Independence is preserved in the coincidence networks through insulation resistance and physical separa- s tion of the switching elements.

7 1.2 3 3 loss of Power --

The Reactor Protection System initiates trip actica upon loss of power. All

bistables operate in a normally energiced state and go to a deenergized state to initiate action. Ioss of power thus automati.taly focees the bistables into the tripped state. Figure 7-2B shows the system in a deenergiced state.

The Safeguards Actuation System instrumentation ctrings terminate in bistable trip elements similar to those in the Reactor Protection System. Loss of in-strument power up to and including the bictables force the bistables into the tripped state initiating safeguards action. The loric networks and the equip-ment control elements are povered from the Engineered Safeguards d-c power buces 1 and 2. Electrical safe 6uards equipment is powered from one of the Engineered Safeguards a-c power busses. Ioss of engineered safeguards power to the logie networks or to the safeguards equignent does not initiate safe 5uards action as described in 71.1.2.4.

7 1.2 3 4 Manual System Trip

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The manual actuating devices in the protection cyctems are independent of the automatic trip circuitry and are not subject to failures that make the auto-matic circuitry inoperable. The manual trip devices arc independent control switches for each power controller. The independent control evitchec, however, are all actuated through a mechanical linkage to a co==on manual trip switch or pushbutton.

7 1.2 3 5 Equipment Removal The removal of modules or subassemblies from vital sections of the Reactor Pro-tection System vill initiate the trip normally associated with that portion of the system. The removal criterion is implemented in two ways: (1) advantage is taken of the inherent characteristics of a normally energiced system, and (2) interlocks are provided.

An inherent characteristic is illustrated by considering the power supply for one of the reactor protection channels. Removal of this power supply automati-cally results in trip action by virtue of the resulting loss of power. No inter-lock is required in such cases. Other instances require a system of interlocks built into the equipment to insure trip action upon 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 6 Testing The protection systems will meet the testing criterion and its objectives. The test circuits vill take advantage of the systems' redundance, independence, and coincidence features which make it possible to initiate trip cignals manually in any part of one protective channel without affecting the other channels.

This test feature vill allow the operator to interrogate the systems from the input of any bistable up to the final actuating device at any time during reactor operation without disconnecting permanently installed equipment.

llh 78 REVISED, 2-8-68

The test of a bistable- consists of inserting an analog input and varying the input until the bistable trip point is reached. The value of the inserted test signal represents the true value of the bistable trip point. Thus the test verifies:not ly set.

only that the bistable functions but that the trip point is correct-Prestartup 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.

As shown in Figure 7-2B, the power breakers in the reactor trip circuit may also be manually tested during operation. The only limitation is that not more than one power supply may be interrupted at a time without causing a reactor trip.

T.1.2.3.7 Physical Isolation The physical arrangement of all elements associated with the protection systems will reduce the probability of a single physical event i= pairing the vital func-tions of the system. For example, pressure measurements of reactor coolant pressure vill be divided. among four redundant pressure taps so as to reduce the probability of collective damage to all sensors by a single accident.

System equipment will be distributed between instrument cabinets so as to re-duce the probability of. damage to the total system by some single event.

Wiring between vital elements of the system outside of equipment housing vill be routed and protected within the unit so as to maintain redundancy of the systems with respect to physical hazards.

7.1.2.3.8 Primary Power Source The primary source of control power for the Reactor Protection System is the vital busses described in 8.2.2 7 The source of power for the measuring ele-ments in the Safeguards Actuation System is also from the vital busses. Com-mand circuits from the Safeguards Actuation System coincidence logic that ex-tend to Engineered Safeguards Equipment controllers are povered frem the Engi-neered Safeguards d-c busses. Engineered Safeguards equipment such as pump and motor cperators and their starting contactors are povered 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, cuch as redundant equipment, redundant channels, and coincidence arrange-ments permitting in-service testing, have been employed to implement reliability of protective action. The best grades of commercially available components vill be used in fabrication. A system fault analysis will be made considering the modes of-failure and determining their effect on the system vital functions.

Acceptance testing and periodic testing vill be designed to insure the quality and reliability of the completed systems.

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7 1.2 3 10 Instrumentation for Emergency Core Cooling Initiation The instrumentation system makes use of both physical and electrical isola- -

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tion. The high pressure and lov pressure injection systems are activated by both low reactor coolant and reactor building pressure signals originat-ing 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.

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 secon-dary shield. The transmitters are physically separated from each other and located outside the secondary shield inside the reactor building. The trans-mitters ' electrical outputs leave the reactor building through separate pene-trations.

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 devices. The bistable trip devices of a given logic function are physically separated by cabinet barriers. Each pressure transmitter and its associated bistable 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 safeguards 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 control power.

The operation of the engineered safeguards channels and the trip relays form-ing the system logic is described in 71.2.2.

The high order of system redundancy assures compliance with the single failure criteria of 7 1.1.2.1.

7 1.2.4 Su=ary of Protective Actions The abnormal conditions that initiate a reactor trip are as follows:

Steady State Trip Value or Trip Variable No. of Sensors Nor=al Range Condition for Trip Neutron Flux 4 0-100%

107 (rated5%)of power full NeutronFlux/ -4 Flux 3 to 4 pumps (1) Ratio of reac- 3 Reactor Coolant 16 Reactor Cool- tor power to total Flow ant Pump Mon- reactor coolant itors flow exceeds 1.07 2 Flow Tubes (2) Ioss of two or more reactor cool-ant pump motors. j

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ii8 1 5-3-68 Supplement N6. 3

Steady State Trip Value or Trip Variable No. of Sensors Normal Range Condition for Trip Startup Rate 2 0-2 Decades / min 5 Decades / min Reactor Coolant 4 2,120-2,250 2,350 psig

, Pressure psig 2,050 psis Reactor Outlet 4 520-603 F 610 F Temperature Actions initiated by the Safeguards Actuation System are as follows:

Steady State Action Trip Condition Normal Value Trip Point High Pressure Icv Reactor 2,120 - 2,250 psig 1,800 psig Injection Coolant Pressure or Ifigh Reactor Atmospheric 10 psig 3 Building Pres-sure lov Pressure Very lov Reactor 2,120 - 2,250 psig 200 psig Injection Pressure or High Reactor Atmospheric 10 psig 3 Building Pres-sure Start Reactor High Reactor Atmospheric 4 psig Building Emer- Building Pres-gency. Cooling sure Unit and Reactor Building Isola-tion

- Reactor Building High Reactor Atmospheric 10 psig Spray Building Pres-sure 7 1.2 5 Relationship to Safety Limits Trip setpoints tabulated in 7 1.2.4 are consistent with the safety limits that have been established from the analysis described in Section 14. The set point for each input, which must initiate'a trip of the Reactor Protec-tion 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 a trip of the Safeguards Actuation System are established at levels that vill insure that correctira action is in progress in sufficient time to prevent an unsafe condition. T%ctors such as the rate at which the sensed variable can change, ~ instrumentation and W A 4 -

7-11 a l19 l i

5-3-68 Supplement No. 3

. calibration inaccuracies, bistable trip times, circuit breaker trip times, control rod travel times, valve. times, and pump starting times have been _.)

. considered in establishing the margin between the trip set points and the ~

safety limits that have been derived.

The flux trip set point of 107 5 per cent is based upon the tolerances and error bands shown in Figure 7-4. The incident flux error is the sum of the errors at the. output of the measuring channel resulting from rod motion, e.nd instrument drift during the interval between heat balance checks of nuclear instrumentation calibration.

713 SETDE EVAMATION 713.1 Functional Capability - Reactor Protection System The Reactor Protection System has been designed to limit the reactor )mer 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. MwNr trip times of the protection channels are listed below,

a. Temperature - 5 see

b. Pressure - 0 5 see

c. Flux - 0 3 see

d. Pump Monitor - 1.0 see Since all uncertainties are considered as cumulative in deriving these times, the actual times may be only one-half as long in most cases. Even these maximum times, when added to control rod drop times, provide conservative protective action.

The Reactor Protection System vill limit the power that might result from an unexpected reactivity change. Any change of this nature vill be detected and arrested by high reactor coolant temperature, high reactor coolant pres-sure, or high neutron flux protective action.

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 power range channels. A startup rate trip from the intermediate-range chan-nels is incorporated in the Reactor Protection System.

A rod withdrawal accident at power vill immediately recult in a high neutron flux trip.

Reduced reactor coolant flow results in a reduced allowable reactor power.

The reactor coolant pump monitor operates to set the appropriate reactor power limit by adjusting the power level trip point. A total loss of flow results in a direct reactor trip, independent of reactor power level.

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5-3-68

'. Supplement No. 3

Twomajormeasurementsfeedthepover/flowmonitar: (a) reactor coolant 3 flow, and (b) neutron power level. The flow tubes which provide the re-actor coolant flow measurement vill exhibit no change during the reactor

. life. A periodic calibration of the flow transmitters vill be made. The neutron power level signal vill be recalibrated by comparison with a rou-tine heat balance. The power range channels use detectors arranged to ef-fectively average the measurement over the length of the core as described in 7 3 1.1.2. Therefore, their output is expected to be within 4 per cent of the calibrated value during nornal regulating rod group position changes and the need for additional calibration thereby eliminated.

The loss of reactor coolant vill result in a reduction of reactor coolant pressure. The low pressure trip serves to trip the reactor for such an occurrence.

A significant turbine-side steam line rupture is reflected in a drop of re-actor coolant pressure. The low reactor pressure trip shuts down the plant for such an occurrence.

7132 Functional Capability - Safeguards Actuation System The Safeguards Actuation System is a graded protection system. The progres-sive actions of the injection systems as initiated by the Safeguards Actua-tion System provide sufficient reactor coolant under all conditions while minimizing the possibility of setting the entire system in operation inad-vertently.

The key variable associated with the loss of reactor coolant is reactor pres-sure. In a loss-of-reactor-coolant accident, the reactor pressure vill 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. High rdactor building pressure at 10 psig is used to 3 provide diversification in actuation of both high pressure injection and low pressure injection.

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, and high and low presaure injection 3 systems.

7133 Preoperational Tests Valid testing of analog sensing elements associated with the protection sys-tems vill be accomplished through the actual manipulation of the measured variable and comparison of the results against a standard.

Routine preoperational tests vill be perfomed by the substitution of a cali-brating signal for the sensor. Simulated neutron signals may be substituted in each of the source, intermediate, and power range channels to check the operation of each channel. Shulated pressure, temperature, and level signals t

l2l 7-13 i -

, 5-3-68 Supplement No. 3

q 1 d' I t

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 T' against ' standards during shutdowns for refueling, or whenever the true status of any measured variable cannot be assessed because of lack of a5reement among the redunannt measurements.

The final defense against sensor failure during operation vill be the plant operator. The redundancy of measurements provides more than adequate oppor-tunity for comparative readings. In addition, the redundancy or the systems reduces the consequences of a single sensor failure.

7134 Component Failure Considerations The effects of failure can be understood through Figure 7-23. In the Reactor 3 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. With a " tripped," open circuit fault in one channel, the system would be able to tolerate a minimum of two '.'cannot trip," short circuit failures within the same measured variable before complete safety protection of the variable vere 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-tection 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 second breaker in the affected primary power supply.

The Safeguards Actuation System is a 2-out-of-3 input type of system. It can tolerate one fault of the "cannot trip" variety in each of the coincidence

networks. For this tfpe of fault, all remaining inputs must function correctly.

A " tripped". input fault allows any one of the two remaining inputs to initiate action.

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. 5-3-68 Supplement No. 3

Primary power input to both protection systems has been arranged to minimize the pcssibility of loss of power.to either protection system. Each channel of the protection system vill be supplied from one of the four- vital busses de-scri'oed 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 equipment.

7.1.3.5 Operational Tests The protection systems are designed and have the facilities for routine manual operational testing.

Most inputs to the protection systems originate frc= an analog =easure=ent 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 vill uncover mea-surement faults. These elements plus the bistables and relays of the protec-tion systems require a periodic dyramic 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 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.

7.2 REGULATING SYSTEMS 7 2.1 DESIGN BASES 7.2.1.1 Ccmpensation Censiderations Reactor regulation is based upon the use of movable control rod assemblies" and chemical neutron absorber (boric acid) dissolved in the reactor coolant.

Relatively fast reactivity effects including Doppler, xenon, and moderator tem-perature are controlled by the control rods , which are capable of rapid compen-s ation . Relatively slow reactivity effects, such as fuel burnup, fission pro-

-duct buildup, samsrium buildup, and hot-to-cold moderater deficit , are control-led by soluble bcron.

It is possible to change che reactor coolant system boric acid concentration to

" follow" xenon transients over approximately 70 per cent'of each core cycle with-out control rod operation. However, to reduce vaste handling requirements re-sulting from chemical shim operation, control rods are used throughout core life for xenon transient associated with normal power changes. Chemical shim is used in conjunction with control rods to compensate for equilibrium xenon conditions.

CControl rod, rod, and control rod assembly (CRA) are used interchangeably in this section and elsewhere in this report.

L 7- '

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At the beginning of first core life when the moderator temperature reactivity coefficient may be cero or slightly positive, the control rod drive response is faster than necessary to maintain the power error within the allowed deadband.

Analog computer analysis shows that the only change in control response when a positive moderator coefficient of reactivity exists is an increased frequency of control rod motion.

The reactor controls are designed to maintain a constant average reactor coolant temperature over the load range from 15 to 100 per cent of rated power. The steam system _ operates on constant pressure at all loads. The average reactor coolant temperature decreases over the range from 15 per cent load to zero load.

Figure 7-5 shows the reactor coolant and steam temperatures over the entire load range.

Input signals to the reactor controls include reactor coolant average tempera-ture, megawatt demand, and reactor power as indicated by out-of-core neutron detectors. The soluble boron dilution is initiated manuslly and terminated automatically or manually. Manual rod control is used below 15 per cent of rated power. Automatic or manual rod control may be used above 15 per cent of rated power.

Increasing power transients between 20 and 90 per cent power are limited to ramp changes of 10%/ min and step in:reases of 10 per cent. Power increases from 15-20 per cent and above 90 per cent are limited to 5%/ min. Decreasing power transients between 100 and 15 per cent power are limited to ramp changes of 10%/ min 'and step decreases of 10 per cent. The turbine bypass system per-mits a load drop of h0 per cent, or a turbine trip from h0 per cent load without safety valve operation. The turbine bypass system and safety valves permit a 100 per cent load drop without turbine trip to satisfy " blackout" requirements as described in 14.1.2.8.2.

7.2.1.2 Safety Considerations 7.2.1.2.1 Shutdown Margin The control rods are provided in sufficient number to allow a hot shutdown that is greater than 1 per cent suberitical with rod assembly of greatest worth fully withdrawn and a typical level of soluble boron (Figure 3-1).

7.2.1.2.2 Reactivity Rate Limits

- The maximum average rate of change of reactivity that can be inserted by any group of rods does not exceed 5.8 x 10-5 (Ak/k)/sec. (The accidental withdrawal of the rod group of greatest worth is discussed in 14.1.2.2 and lk.l.2.3. )

The maximen norma more 'thma 3 x 10 g(Ak/k)/sec.

rate of pure Reactivity water addition controldoes maynot change reactivity be exchanged between worth rods and soluble boron consistent with the design bases listed above.

7.2.1.2.3 Power Peaking Limits The nominal reactivity available to a power regulating control rod group is

'" limited so that established radial and axial flux-peaking limits are not ex-ceeded with the rod group in any position at power leve'Is up to 100 per cent

-power.

-l 24 _7-15

7 2.1.2.h Power Level Limits

_T The _ reactor automatic controls incorporate a high limit and a lov limit of power -

level demand to the reactor. Limits are imposed on reactor megawatt demand by lack of feedwater flow capability and reactor coolant system flow capability.

7.2.1.3 Startup Considerations Over the life of the nuclear unit, startup vill occur at various temperature levels and after varying periods of downtime. Examples of regulating system design requirements as related to startup are

a. Control rod and/or control rod group " withdraw inhibit" on high start-up rate (short period) in the source range and intermediate range.

b. Reactor trip on high startup rate in the intermediate range.

c. Startup control mode. This mode prevents automatic rod withdrawal below 15 per cent power.

d. In startup control mode, the controls are arranged so that the steam system follows reactor power rather than turbine system power demand.

The controls vill limit steam dump to the condenser when condenser vacuum is inadequate.

e. Sufficient control rod worth is provided to override peak xenon and return to power following a hot shutdown or hot standby. During cold shutdown it vill be necessary to increase boron concentration to main-tain shutdown margin. Following a cold shutdown, boron concentration changes will be made during startup. A number of rod assemblies (or groups), sufficient to provide 1 per cent shutdown margin during start-up, are required to be withdrawn before a dilution cycle.

f. . Minimum pressurizer water level conditions must be met before and dur-ing startup.

7 2.2 SYSTEM DESIGN 7 2.2.1 Description of Reactivity Control 7.2.2.1.1 General Description The reactor controls move control rods to regulate the power output of the re-actor and maintain constant reactor coolant average temperature above 15 per cent rated power. ; As shown in Figure 7-6, the megawatt demand signal is added to the reactor coolant average te=perature error to form a reactor power level demand signal. The reactor power level demand signal is compared to the aver-age reactor power level measured by the power range detectors in the nuclear instrumentation.. When the resulting reactor power level error signal exceeds the deadband, the output signal is a control rod drive " withdraw" or " insert" command to the controlling rod group. For reactivity control limits see 3.1.2.2.

L2.5 T e

7 2.2.1.2 . Reactivity. Control' Reactivity control is maintained by movable control rod assemblies and by soluble boron dissolved in the reactor coolant. -The-moderator temperature coefficient (cold to hot critical), as well as long-tarm reactivity changes caused by fuel burnup and fission product poisoning, are controlled by adjusting soluble boron concentration. . Short-term reactivity changes caused by power change, xenon poi-soning, and moderator temperature change from 0 to 15 per cent power are con-trolled by control rods.

First-cycle values for the reactivity components and control distribution are listed in Tables 3-h and 3-5

' Twenty-one of the 69 control rod assemblies are assigned to automatic control of reactor power' level during the first core cycle. Thereafter, 25 rod assemblies are used. These control rod assemblies are arranged in four sy= metrical groups which operate in sequence. The position of one automatic group is used as an index to soluble boron dilution. Soluble boron adjustment is initiated manually and terminated automatically. The position of this group acts as a "per=issive" to ' restrict the start of dilution to a " safe" rod position pattern. The posi-tion of the same group terminates dilution automatically.

During reactor startup, control rods are withdrawn in a predetermined sequence in symmetrical scoups of four or more rod assemblies. The group size is preset, and individual control rod assembly assignments to a group are made at a con-trol rod grouping panel. However, the operator can select any individual con-trol rod assembly and any rod group for motion as required.

A typical control rod group withdrawal _ scheme is as follows:

First Cycle Equilibrium Cycle Group 1 16 CRA's 12 CRA's Group 2 12 CRA's 12 CRA's Group 3 12 CRA's 12 CRA's Group h 8 CRA's 8 CRA's Group 5 h CRA's '

8 CRA's Group 6 8 CRA's- ,

Regulating , 9 CRA's Group 7 5 CRA's Groups 4 CRA's Group 8 h CRA's , , h CRA's An automatic sequence-logic unit is used for reactor control with four regulat-ing rod groups in the power range. This unit allows operation of no more than one control rod group simultaneously except over the last 25 per cent travel of one group ' and the first 25 per cent travel of the next group when overlap-ping motion of two groups is permitted. This tends to linearize the reactivity insertion from group to group as shown in Figure T-7

(' .

As fuel burnup progresses, dilution of the solubl'e, boron is controlled'as fol-

"' ' lows:

1 - :

'. s i26

r

When the partially withdrawn active control rod group reaches the fully with-drawn point, interlock circuitry permits setting up a flow path from a demin- T/

eralized water tank, in lieu'of the normal flow path of borated makeup, to the reactor coolant system. Demineralized water is fed to the reactor coolant sys-tem, and borated reactor coolant is - removed.

She reactor' controls insert the active regulating rod group to compensate for the reduction in boron concentration. When the control rod group has been in-serted to the 75 per cent withdrawn position, the dilution flow is automatically blocked. The dilution cycle is also terminated automatically by a preset timing device, which is independent of rod position. Normally, a dilution cycle is re-quired every several days.

7 2.2.1.3 Reactivity Worth The maximum.vorth of any group of the four automatic control groups is approxi-mately 1.2% Ak/k. At design speed, a group requires approximately 6 minutes to travel full stroke. This rate of control rod group travel results in a re-activity rate of 5.8 x 10-5 (Ak/k)/sec.

The maximum rate of reactivity addition with the soluble boron system, i.e. , in-jecting unbarated water from the makeup system at 70 gpm maximum, is 3.0 x 10-6 (Ak/k)/sec.

Table 3-5 shows a shutdown reactivity analysis. The rod worth provided gives a shutdown margin of 5.1% Ak/k or more under normal conditions, and a margin in excess of 1% Ak/k with the CRA of greatest worth stuck in the withdrawn position.

Under conditions where cooldown to reactor building ambient conditions is re-quired, concentrated soluble boron vill be added to the reactor coolant to pro-duce a shutdown margin of at least 1% Ak/k. The reactivity changes from hot zero pover to a cold condition, and the corresponding increases in boric acid concentration, are listed in Table 3-6. .

7 2.2.1.h Reactor Control The reactor control is made up of analog computing equipment with inputs of mega-watt demand, core average power, and reactor coolant average te=perature. The output of the centroller is an error signal that causes the control rod drive to be positioned until the error signal is within a deadband. A block diagram of the reactor control is shown in Figure 7-6.

First, reactor power level demand (Nd) is co=puted as a function of the megawatt demand (MW d ) and the reactor coolant system average temperature deviation (AT) from the set point, according to the following equation:

Nd

• E lmwd +K2 (ET + [ETdt) ,

. 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 i is introduced as a part of the demand signal after proportional plus reset (in-tegral) action is applied. For.the te=perature deviation, K2 is the adjustable gain and T is the adjustable integration factor.

19, i

i f. / 7-18 t-

The reactor power level demand (Iid ) is then compared with the average re-actor Pcuer level signal (ii i), which is derived from the nuclear instrumen- ,

tation. The resultant error signal (lid - N1) is the reactor power level error signal (E p).

When the reactor power level error si 6nal (E )p exceeds the deadband settings, the control rod drive receives a cor=and that withdraws or inserts rods de-pending upon the polarity of the power error signal.

The following additional features are provided with the reactor power con-troller:

a. An adjustable low limit on the megawatt demand signal ( Wd ) to cut out the automatic reactor control action.

b. A high limit on reactor power level decand (Nd )*

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

Separate from, but related to the auto =atic reactor control system is the 3 reactor coolant flow signal. The reactor coolant flow is measured and limits the unit negawatt demand and respective steam generator feedvater demand to within the capability of the available reactor coolant flow.

In addition to measuring the reactor coolant flow, digital logie units continually ecnitor the number of energized reactor coolant pumps and es-tablish a =aximum limit on the megawatt demand.

7 2.2.2 Integrated Control System The Integrated Control System =aintains constant average reactor coolant temperature and constant steam pressure in the nuclear unit during steady state and transient operation between 15 and 100 per cent rated power. Fig-ures 7-6 and 7-8 show the overall system. The system is based on the In-te6 rated Boiler-Turbine concept videly used in fossil-fuel-fired utility plants. It combines the stability of a turbine-following system with the fast response of a boiler-following system. Optimum overall unit performance is maintained by limiting steam pressure variations; by limiting the unbal-ance that can exist amon5 the steam generator, turbine, and the reactor; and by limiting the total unit load demand upon loss of capability of the steen generator feed system, the reactor, or the turbine generator.

Figure 7-6 shows the reactor control portion of the Integrated Control System described in 7 2.2.1.4. Figure 7-8 shows the steam generator and turbine control portion of the Integrated Control System. This control receives in-puts of megawatt demand, system frequency, and steam pressure, and supplies output signals to the turbine bypass valve, turbine speed changer, and steam generator feedvater flow controls with chan61ng operating conditions.

The turbine and steam generator are capable of auta=atic control from zero power to rated power with optional manual control. The reactor controls are designed for manual operation below 15 per cent rated power and for auto-matic or manual operation above 15 per cent rated power.

7-19 lU 5-3-68 g Supplement No. 3

The turbine is. operated as a turbine-followin5 unit with the turbine con- 'J trol valve.~ pressure set point varied.in proportion to megawatt error. The -

steam generator is operated as a boiler-following system in which.the feed-water flow l

l n

Y, l 7-19a i

l i29 5-3-68 p

Supple:nent No. 3

d';=and to the steam generator is a smtion of the megawatt demand and the steam pressure arror.

Tho Integrated Control System obtains a load demand signal from the system dis-patch center or from the operator. A frequency loop ic added to match the speed droop of the turbine speed controls. The load demand is restrained by a maxi-mum load limiter, a minimum lond limiter, a rate limit u , and a runback limiter.

In normal operation the limits would be set as folleva:

Megawatt Demand (MR )

d Maximum load limit 100%

Minimum ' load limit 15%

Rate limit 10%

The runbacks act to runback and/or limit the load demand on any of the following conditions:

a. One or more reactor coolant pumps are inoperative.

b. Total feedvater flow lags total feedvater de=and by more than 5 per cent.

c. Thefourshim/safetyrodgroupsarenotfullywithdrawn.

d. Assynetric rod withdrawal patterns exist.

e. The generator separates from the 500 kv bus.

The output of the limiters is a megawatt demand signal which is applied to tha turbine control, steem generator control, and reactor control in parallel.

Th3 reactor control responds to the megawatt demand signal as described in 7 2.2.1.4.

7 2.2.2.1 Turbine Control i

Tha megawatt demand is compared with the generator megawatt output, and the re-sulting megawatt error signal is used to chan6e the steam pressure set point.

Tha turbine valves then change position to control steam pressure. As the megawatt error reduces to zero, the steam pressure set point is returned to the steady state value. By limiting the effect of megavatt error on the steam prassure set point, the system can be ad, justed to permit controlled variations in steam pressure to achieve any desired rate of turbine response to megawatt demand.

7 2.2.2.2 Steam Generator Control Ctntrol of the' steam generator is based on matching feedvater flow to megawatt demand with bias provided by the error between steam pressure set point and stram pressure. The pressure error increases the feedvater flow demand if the pr;ssure is lov. It decreases the feedvater flow demand if the pressure is high.

l The basic control actions for parallel steam generator operation are: l 1

a. Megawatt demand converted to feedvater demand.

I30-7-20 REVISED, 2-8-68 )

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

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

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

e.. Feedvater demand compared to feedvater flow 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 water level. The conversion to level control is automatic and is introduced into the feedvater control train through as auc-tioneer. At low loads below 15 per cent, the turbine bypass valves vill oper-ate to limit steam pressure rise.

The steam generator control also provides ratio, limit, and runback actions as shown in' Figure T-8, which include

a. Steam Generator Load Ratio 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 lohding or unloading rates to those-that are compatible with the turbine and/or the steam gen-erator.

c. Water Level Limits 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 low load control.

d. Reactor Coolant Pump Limiters These limiters restrict feedvater demand to match reactor coolant pumping capability. For example, if one reactor coolant pump 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.

c. Reactor Outlet and Feedvater Low Temperature Limits These limiters reduce feedvater demand when the reactor outlet tem-perature or the feedvater temperature is lov.

i r '

M '

)

t +

.. 7 l l

l

t

f. Feedvater Pump ' Capability 9

A feedwater pump capability runback signal limits the megawatt demand signal whenever total feedwater flow lags total feedvater demand by 5 per cent.

' 7 2.3 -SYSTEM EVALUATION 7.2.3.1 System Failure Considerations Redundant sensors are available to the Integrated Control System. The operator

~

can select any of the redundant sensors from the control room.

Manual reactivity control is available at all power levels.

Loss of electrical power to the automatic controller reverts reactor control to

! the manual mode.

i I

\

p ..

i l

I I

1

[

l32 1'

, 7-22

7.2.3.2' 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 predetermined insertion and withdrawal pat-tern of the four regulating rod groups.

Control circuitry allows manually selected operation of any single control rod assembly or control rod group throughout the power-range.

An interlock will prevent actuation of both withdrawal and insertion of control rods simultaneously 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 manual mode.

The automatic' sequence logic limits regulating rod motion to one group out of four at one time except at the upper and lower 25 per cent of stroke where operation of two groups is permitted to linearize reactivity versus strcke.

Maximum and minimum limits on the reactor power level demand signal (Nd ) Pre-vent the reactor controls from initiating undesired power excursions.

Maximum and minimum levels on the megawatt demand signal (!N d ) prevent the re-actor controls from initiating' undesired power excursions.

T.2.3.3 Emergency considerations Loss of-power to the control rod drive magnetic clutch initiates a reactor trip.

When emergency conditions arise that exceed the capability of the control sys-tem, the operator can revert to the manual control mode.

7 2.3.h Loss-of-Load 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 h0 per cent load re-

. duction or a turbine trip from h0 per cent load without safety valve action.

The^ controls will limit steam du=p to the condenser when condenser vacuum is inadequate, in which case the safety valves may operate. The combined actions of the control system, the turbine bypass valve, and the safety valves permit a 100 per cent lead 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 without turbine. trip. (The " blackout" provisions are discussed in 14.1.2.8.2.)

TDiefeaturesthatpermitcontinuedoperationunder'loadrejectionconditions include

(

133 4 -23 s4

t

a. Integrated Control System 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 changes, turbine control is through the speed changer to maintain constant steam pressure.

s

.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-portant in providing load matching.

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

Consider, for example, a sudden load rejection greater than 10 per cent. When the turbine generator starts accelerating, the governor valves and the intercept valves begin to close to maintain set fre-quency. At the same time the megawatt demand signal is reduced, which reduces the governor speed changer setting, feedwater flow de-mand, and reactor power level demand. As the governor valves close, the steam pressure rises and acts through the control system to rein-force the feedvater flow demand reduction already initiated by the reduced megawatt demand signal. In addition, when the load rejection is of sufficient magnitude, the turbine bypass valves open to reject j-excess steam to the condenser, and the safety valves open to exhaust steam to the atmosphere. The rise in steam pressure and the redue-tion in feedwater flow cause the average reactor coolant temperature to rise which reinforces the reactor power level demand reduction, already established by reduced megavatt demand, to restore reactor coolant temperature to set value.

As the turbine generator returns to set frequency, the turbine con-trols revert to steam pressure control rather than frequency control.

This feature holds-steam pressure within relatively narrow limits and prevents further large steam pressure changes which could impose additional load changes of opposite sign on the reactor coolant sys-tem. As a result, the reactor, the reactor coolant system, and the steam system run bach rapidly and smoothly to the new load level.

T.3 INSTRUMENTATION 7 3.1 NUCLEAR INSTRUMENTATION The nuclear instrumentation system is shown in Figure 7-2A. Emphasis in the design is placed upon accuracy, stability, and reliability. Instruments are redundant at every level. The design criteria stated in 71.1.2 have been ap-plied to the design of this instrumentation.

7-2h bk

~7 3;1.1 Design The nuclear instrumentation-has eight channels of neutron information divided into three ranges of sensitivity: source range, intermediate range, and power

-range. ..The three ranges combine to give a continuous measurement of reactor power from source-level.to.approximately 125 per centoof rated power or ten decades of-information. A minimum of one decade of overlapping information is provided between successive higher ranges of instrumentation. The relationship between instrument ranges is shown in Figure 7-9

~The source range instrumentation has two redundant count rate channels origi-nating in two high sensitivity proportional counters. These channels are used over a counting range of 1 to 105 counts /see as displayed on'the operator's contro1' console in terms of log counting rate. The channels also measure the rate of change of the neutron level as displayed for the operator in terms of startup rate from -1 to +10 decades / min. No protective functions are associ-ated with the source-range because of inherent instrumentation limitations en-countered in this range. However, one interlock is provided, i.e., a control rod withdraw hold and alans on high startup rate in either channel.

The intermediate range instrumentation has two log N channels originating in

.two identical electrically gamma-compensated ion chambers. Each channel pro-vides seven- decades of flux level information in terms of log ion chamber cur '

rent and startup rate. The~ ion chamber output range.is from 10-11 to 10-k am-peres. The startup rate range is from -1 to +10 decades per minute. Protec-

-tive action on high startup rate is provided by these channels. A high start- l up 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 inter-lock and alars.

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.from 0 to 125 per cent of rated power. The system is a precision analog system which employs a digital technique to pro--

vide highly accurate signals for instrument calibration and reactor trip set point calibration. The gain of.each channel is adjustable, providing a means for calibrating the output against a reactor heat balance. Protective action on high flux level consists of reactor trip initiation by the power rangechan-nels at preset-flux levels.-

Additional feat'ures pertinent to the nuclear instrumentation system are as fol-lows:

a. Independent power supplies are included in each channel. Primary power originates from the vital busses described in 8.2.2 7. Where applicable, isolation transformers are provided to insure a stable, high quality power supply.

bL The proportional counters used in the source range are designed to be' secured when the flux level is greater than their'useful operating range. This:is necessary to obtain prolonged operating life.

c. The. intermediate range channels are supplied with an adjustable source

'y- ;of gamma-compensating . voltage.

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T.3.1.1.1 Test and Calibration Test .and calibration facilities are built into the system. The test facilities will meet the requirements outlined in the discussion of protection systems; testing.

Facilities for calibration ~ of the various channel. amplifiers and measuring.

equipment will also be a part of the system.

7.3.1.1.2 Power Range Detectors Twelve uncompensated ionization chambers are used in the power range channels.

Three _ chambers are associated with each channel, i.e. , one near the bottom of the core, a second at the midplane, and a third- toward the top of the core.

The outputs of the-three chambers are combined in their respective linear am-plifiers. A means is provided for reading the individual chamber outputs as a manual calibration and test function during normal operation.

T.3.1.1.3 Detector Locations The physical locations of the neutron detectors are shown in Figure 7-10. The power range detectors are located in four primary positions, 90 degrees apart around the reactor core.

Tne two source range porportional counters are located on opposite sides of the cora adjacent to two of the power range detectors.

The two intermediate range compensated ion chambers are also located on opposite sides. of the core, but rotated 90 degrees from the source range detectors.

7 3.1.2 Evaluation The nuclear instrumentation wil1~ 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 vill be approximately 109 nv. The detectors em-ployed vill provide a linear response up to approximately b x 1010 nv before they are saturated.

The intermediate range channels overlap the source range and the power range channelsLin an adequate manner, providing the continuity of information needed during startup.

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-

.marily for reactor safety, control, and operation information.

7 3.1.2.l' ' Loss of Power-The nuclear instrumentation draws its primary power from redundant battery-backed vital busses described in 8.2.2.7 bb 7-26 -

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7 3 1.2.2 Reliability and Component Failure The requirements established for the reactor protection system apply to the nuclear instrumentation. All channel functions are independent of every other channel, and where signals are used for safety and control, electri-cal isolation is employed to meet the criteria of 71.1.2.

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 will be held to +l/2 ,

per cent at rated power over a 30-day period. Routine tests and recalibra-tion vill insure that this degree cf deviation is not exceeded. Bistable trip set points of the power range channels vill also be held to an accuracy of +1/2 per cent of rated power. The accuracy and stability of the equip-ment vill be verified by vendor tests.

732 NONNUCIEAR PROCESS INSTRUENTATION 732.1 System Design The nonnuclear instrumentation measures temperatures, pressures, flows, and levels in the reactor coolant system, steam system, and reactor auxiliary systems. Process variables required on a continuous basis for the startup, operation, and shutdown of the nuclear unit are indicated, recorded, and controlled from the control room. The quantity and types of process instru-mentation provided vill insure safe and orderly operation of all systems and processes over the full operating range of the plant. The amounts and types of various instruments and controllers shown are intended to be typical ex-amples of those that vill be included in the various systems when final design 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 follows:

a. In general, resistance elements are used for temperature measure-ments. Fast-response resistance elements monitor the reactor out-let temperature. The outputs of these fast-response elements supply signals to the protective system.

b. Pressures are measured in the reactor coolant system, the steam system, and the reactor auxiliary systems. Pressure si Enals for high and low reactor coolant pressures and high reactor building pressure are provided to the protection systems.

c.- The number of reactor coolant pump motors in operation is monitored.

3 In addition, there is a direct measurement of reactor coolant flow.

This information is fed to the reactor controls and reactor protection system.

l3[ 7-27 6- 5-3-68 Supplement No. 3

d. Flow in the steam system is obtained through the use of calibrated feedwater flow nozzles. Flow information is utilized for control and 3 protective functions in the steam system. Steam generator level mea- '

surements are-provided for control and alarm functions.

e. Pressurizer level is measured by differential pressure transmitters calibrated to operating temperature and pressure. The pressurizer level is a function of the reactor coolant system makeup and letdown flow rate. The letdown flow rate is remote manually controlled to the required flow. Pressurizer level signals are processed in a level controller whose_ output positions _the makeup control valve in the makeup line to maintain a constant level.

f. Reactor coolant system pressure is maintained by a control system that energizes' pressurizer electrical heaters in banks at 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 System Evaluation R:dundant instrumentation has been provided for all inputs to the protection systems and vital control circuits.

Where vide process variable ranges are required and precise control is involved both-vide-range and narrow-range instrumentation are provided.

Where possible, all instrumentation components are selected from standard com-mercially available products with proven operating reliability. I All electrical and electronic instrumentation required for safe and reliable operation vill be supplied from redundant vital a-c instrumentation busses.

T.3.3 INCORE MONITORING SYSTEM 7 3.3.1 Design Basis The incore monitoring syctem provides neutron flux detectors to monitor core performance. No protective action or direct control functions are performed 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 disturbances during power operating modes. Data obtained vill provide measured power dis-l tribution information and fuel burnup data to assist in fuel management deci-sions.

7-28g

ri - 7332 System Design 7 3 3 2.1 System Description The 'incore monitoring system consists of assemblies of self-powered neutron detectors located at 52 preselected radial positions within the core. The 3 incore monitoring locations are shown on Figure 7-12. In this arrangement, an incore detector assembly, consisting of six local flux detectors and one background detector,is installed in the instrumentation tube of each of 52 fuel assemblies . (Figure 3-62) . The local detectors are positioned at six different axial elevations to provide the axial flux gradient. The outputs of the local flux detectors are referenced to the background detector output so that the differential signal is a true measure of neutron flux. As shown in Figure 7-12, seventeen detector assemblies are located to act as symmetry monitors. The remaining 35 detector assemblies, plus five of the 17 symmetry monitors, provide monitoring of every type of fuel assembly in the core when quarter core symmetry exists. Readout for the incore detectors is performed by the ec=puter system rather 3 than by individual indicators. This system sounds alarms if local flux conditions exceed predetermined values. When the reactor is depressurized, the incore detector assemblies can be in-serted or withdrawn through guide tubes which originate at a shielded area in the reactor building as shown in Figure 7-13 These guide tubes, after completing 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 assemblies. The instrumentation tube then serves as the guide for the incore detector assembly. The incore detector assemblies are fully with-drawn only for. replacement. During refueling operations, the incore detector assemblies are withdrawn approximately 13 feet to allow free transfer of the fuel assemblies. After the fuel assemblies are placed in their new locations, the incore detector assemblies are returned to their fully insarted 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 de-tectors which will produce a hi t;h relative accuracy between individual de-tectors. The detector signals must be compensated for burnup of the neutron sensitive material. The data handling system integrates each detector out-put 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 of rated power. The data handling system computes an average power value for the entire core, normalized to the reactor heat balance. This average power value is compared to each neutron detector signal to provide the core power distribution pattern. 7-29

                                                        .-  i 5-3-68 Supplement No. 3

'7.3.3.3' System Evnluation 7.3.3.3.1 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 NRX and NRU reactors, and have been operated at fluxes beyond those expected in normal pressurized water riactor service. 7.3.~3.3.2 B&W Experience Self-powered,inco$eneutrondetectorshavebeenassembledandirradiatedinThe Bibcock & Wilcox Company Development Program that began in 196h. Results from this program have produced confidence that self-powered detectors used in an in-core' instrument system for pressurized water reactors vill perform as well as, if not better than, any system of incore-instrumentation currently in use. The B&W Development Program includes these tests:

a. Parametric studies of the self-powered detector.

      .b. Detector. ability to withstand PWR environment.

c. Multiple detector assembly irradiation tests.

d. Background effects.

e. Readout system 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 te=perature effects are satisfactory for use.

b. A multiple detector assembly can provide axial flux data in a single channel and can withstand reactor environment. An assembly of six local flux detectors, three background detectors, and two thermo-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 vill not prevent satisfactory operation in a PWR environment.

7-30 140 o

Irradiation of detector asse=blies and evaluation of perfomance data are con-tinuing to provide detailed design information for the incore instrumentation system. 74 OPERATHIC COITIROL STATIONS Following proven power station design philosophy, all control stations, switches, controllers, and indicators necessary to start up, operate, and shut down the nuclear unit vill be located in the contro.'. room. Control functions necessary to maintain safe conditions after a loss-of-coolant accident vill be initiated from the centrally located control room. Controls for certain auxiliary systems may be located at remote control stations when the system controlled does not involve power generation control or emergency functions. Since there vill be no flam=able materials in the control room except for charts, records, and some of the electronics, a fire resulting in loss of redundant con-trol systems is not considered possible. Shieldin5 around the control room is provided to insure operator safety during any K4A. Iio incident or combination vill result in loss of control room functions. 7 4.1 GEIIERAL IAYOUT The control room will be des 1 6ned so that one man can supervice operation during nomal steady state conditions. During other than normal operating conditions, other operators will be available to assist the control operator. Figure 7-14 shows the control room layout for the unit. The control board is divided into relative areas to show the location of control stations and information display pertaining to various subsystems. 7 4.2 DEORMATI0IT DISPIAY AIID COITIROL FUIICTICII The necessary infomation for routine monitoring of the nuclear unit and the plant vill be displayed on the control room benchboard cubicles in the immedi-ate vicinity of the operator. Infor=ation display and control equipment fre-quently employed on a routine basis, or protective equipment quickly needed in case of an emergency, will be mounted on the benchboards. Recorders and radia-tion monitoring equiIment 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, will be mounted on side panel sections of the cubicles. A plant computer vill be available in the control room for alarm monitoring, perfomnce monitoring, and data logging. On-demand printout is available to the operator at his discretion in addition to the computer periodic logging of plant variables. 743 SUWARY OF AIARIE Visible and sudible alarm units vill be incorporated into the control roo= to warn the operator if unsafe conditions are approached by any system. Audible reactor building evacuation alarms are to be initiated from the radiation non-itoring system or manually by the operator. Audible alarms will be sounded in appropriate areas throughout the plant if high radiation conditions are present. l4} 7-31 REVISED, 2-8-68

                                - -         .. -.          - - - - .                   - .                  - ~ _.-                            - .                  .
,       T.4.4             C000RJNICATION PlantLtelephone_and' paging systems will be provided with redundant power supplies                                                                            y
      ..to provide the control room operator with constant communication with all areas                                                                               si of the plant. ' Acoustic booths will be supplied in areas where the background noise level is high. : Communication outside the plant will be through the Bell System and the Arkansas Power & Light Company microwave system.

7. k . 5 ' ' OCCUPANCY. l

      < Safe occupancy of. the control room during abnormal conditions will be provided for in the design of the' auxiliary building. Adequate shielding will be used j        to maintain tolerable radiation levels in the control room for maximum hypo-l        thetical accident conditions. The control room ventilation system will'be pro ,
,        vided with radiation detectors and appropriate alarms. Provisions will be made for the control room air to be recirculated through absolute and charcoal fil-i        ters.       Energency lighting vill be provided.

t -

      ~ The potential magnitude of a fire in the control room will be limited by the

_following factors:

a. The control room consturction will be of noncombustible materials.

;                   b. Control cables and switchboard wiring will be constructed such that 1                        they have_ passed the flame test as described in Insulated Power Cable

Engineers' Association Publication S-61-402 and National Electrical Manufacturers Association Publication WC 5-1961.

;                   c. Furniture used in the control room will be of metal construction.

1

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

will be limited to the enounts required for plant operations.

'                   e. All areas of the control room will be readily accessible for fire ex-i                        tinguishing.-
;                   f.. ' Adequate fire extinguishers will be provided.

I g.- The control room will be occupied at all times by a qualified person

,                        who has been trained in fire extinguishing techniques.

The only flammable materials inside the control room will be

a. Paper in the form of logs, records, procedures, manuals, diagrams, etc.

b. .The coaxial cables required for nuclear instrumentation.

c. ~Small amounts of combustible materials used in the manufacture of various. electronic equipment.

The above list ' indicates that the flammable materials will be distributed to

       .the extent that a fire would be_unlikely to spread. Therefore, a fire, if 7-32
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started, would be of such a small magnitude that it could be extinguished by the cperator using a hand fire extinguisher. The resulting smoke and vapors wculd be rencved by the ventilation system. Essential auxilinry equipment will be controlled by stored energy, closing-type, air-circuit breakers which will be accessible and can be manually closed in the event d-c control'pcwer is lost. 7.4.6 AUXILIARY CONTROL STATIONS Auxiliary control stations will be provided where their use simplifies con-trol of auxiliary systems equipment such as sample valve selectors, chemical addition, etc. 7.L.7 SAFETY FFATURES Tne primary cbjectives in the control recm laycut are to prcvide the necessary centrcls to start, cperate, and shut dcun the nuclear unit with sufficient in-formation display and alarm menitoring to insure safe and reliable operation under normal and accident ecnditions. Special emphasis will be given to main-taining control integrity during accident conditions. The laycut of the en-gineered safeguards section of the centrcl board will be designed to minimize the time required fcr the operator to evaluate the system performance under accident ccaditions. Any deviations from predetermined conditicns will be alarmed se that the operater may take corrective action using the controls pro-vided on the control panel. 75 SYSTEMATIC, NONRANDOM, COMMON f0DE INSTRUMENTATION FAILURES Arkansas Pcwer & Light Ccmpany is following the prcgress of Babcock & Wilcox - in this area. A Babccck & Wilcox Tcpical Report will be submitted on this 13 subject in early 197 C We will inccrporate this repcrt er applicable sec-tiens therect in the FSAR. 7 10-31-69 /^Q Supplement No. 13 143 e  ;

                                         ~

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- a z indicated Reactor Power with - 2% Heat Balance Error a a

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5  ? o 2 N 3 i' g Indicated Reactor Power with + 2% Heat Balance Error g

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   -                                         Temperature
   ,_h 560                                                                                    i Reactor inlet Temperature 1

5# - - _ _ __ __ __ _ _ . _ _. __ _ _ _ . _ T at 925 psia sat 520 0 20 40 60 80 100 RatedReactorPower(2.452MWt),f. RE4CTOR AND STEAM TEMPERATURES VERSUS REACTOR POWER Figure 7-5 b l

                                                                                       ~ ~ ~

I 4 7 --== R00 10 OTHER CONTROL ORIVE GROUPS

                                ,-                                                      CONTROL R00 AUTOMATIC                         ~

ORIVE GRour SEQUE NCE F- LOGIC l ._ in, T"I I l STEAM REACTOR GENERATOR --

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