ML20212A251

Proposed Tech Specs Deleting All References to Steam Line Low Pressure Safety Injection Function

ML20212A251
Person / Time
Site:	Mcguire, McGuire
Issue date:	10/06/1997
From:	DUKE POWER CO.
To:
Shared Package
ML20212A248	List: ML20212A242 ML20212A251
References
NUDOCS 9710230222
Download: ML20212A251 (57)

Text

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1 .

ATTACHMENTI PROPOSED NEW TECHNICAL SPECIFICATION PAGES ijt.M

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TABLE 3.3-3 .

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE OF CHANNELS TO TRIP OPERABLE MODES . ACTION FUNCTIONAL UNIT

1. Safety Injection, Reactor Trip, Feedwater Isolation, Component Cooling Water, Start Diesel Generators, and Nuclear Service Water 2 1, 2, 3, 4 18

a. Manual Initiation 2 1 Automatic Actuation 2 2 1,2,3,4 14

b. 1 Logic and Actuation l

Relays

c. Containment 3 2 2 1,2,3 15 Pressure-High 4 2 3 1, 2, 3f 19

d. Pressurizer Pressure - Low-Low 3/43-17 Amendment No.

.McGUIRE - UNIT 1

TABLE 3.3-3 (Continued)

TABLE NOTATION

1. ' Trip function may be blocked in this MODE below the P-11 (Pressurizer l Pressure Interlock) Setpoint.

H Trip functjon automatically blocked above P-11 and may be blocked below P-11 when Main Stc&m Isolation on low steam pressure is not blocked. l These values left blank pending NRC approval of three loop operation.

Note 1: Turbine driven auxiliary feedwater pump will not start on a blackout signal coincident with a safety injection signal.

ACTION STATEMENTS ACTION 14 With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least il0T STANDBY within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />; however, one channel may be bypassed.for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLE.

ACTION 15 With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed until perfomance of the next required OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 15a With the number of OPERABLE channels less than the Total Number of Channels, operation may proceed until performance of the next required OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. With more than one channel inoperable, enter Specification 3.8.1.1.

ALTION 15b With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed until performance of the next required OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 16 With the number of OPERABLE channels one less than the Total Number of Channels, o)eration may proceed provided the inoperable channel is placed in t1e bypassed condition and the Minimum Channels OPERABLE requirement is met. One additional channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing per Specification 4.3.2.1.

ACTION 17 With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment purge supply and exhaust valves are maintained closed.

1 McGUIRE - UNIT 1 3/43-24 Amendment No.

_ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . l

TABLE 3.3-4 .

t ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUENTATION TRIP SETPOINTS TRIP SETPOIjLT . ALLOWABLE VALUES' FUNCTIONAL UNIT

1. Safety Injection, Reactor. Trip, Feedwater Isolation, Component Cooling Water, Start Diesel Generators, and -

Nuclear Service Water.

N.A. N.A.

a. . Manual . Initiation Automatic Actuation Logic N.A. N.A.

b.

and Actuation Relays

~

Containment Pressure--High s 1.1 psig s 1.2 psig c.

z 1845 psig 2 1835 psig

d. Pressurizer Pressure--Low-Low i

2. Containment Spray N.A. N.A.

a. Manual Initiation Automatic Actuation Logic - N.A. N.A.

b.

and Actuation Relays Containment Pressure--High-High s 2.9 psig s 3.0 psig c.

Amendment No.

McGUIRE - UNIT 1 3/43-27

TABLE 3.3-5 (Continued ,

ENGINEERED SAFETY FEATURES 'tESPOSSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE LIME IN SECONDS

3. Pressurizer Pressure-Low-low

a. Safetyinjection(ECCS)  ; 27(l)/12(3)

b. Reactor Trip (from SI) s2

c. Feedwater Isolation s 12

d. Containment Isolation-Phase "A"(2) s 18(3)/28(8)

e. Containment Purge and Exhaust isolation s4

f. Auxiliary Feedwater(5) N.A.

g. Nuclear Service Water System s 76(1)/65(3)

h. Component Cooling Water s 76(I)/63(3)

1. Start Diesel Generators s 11

4. Steam Line Pressure-Low Steam Line Isolation s 10 t

4 McGUIRE - UNIT 1 3/43-33 Amendment No.

TABLE 4.3-2 .

ENGINEEREO SAFETY FEATJkES ACTUATION SYSTEM INSTRUENTATION SURVEILLANCE REQUIREENTS TRIP ANALOG ACTUATING MODES CHANNEL DEVICE MASTER SLAVE F0rl WHICH OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE CHANNEL CHANNEL TEST LOGIC TEST TEST TEST IS REQUIRED FUNCTIONAL UNIT CHECK CALIBRATION TEST

1. Safety Injection, Reactor Trip, Feedwater Isolation, Component Cooling Water, Start Diesel Generators, arci Nuclear Service Water N.A. N.A. N.A. R N.A. M.A. N.A. 1, 2, 3, 4

a. Manual Initiation M(I) 1, 2, 3, 4

b. Automatic Actuation N.A. N.A N.A N.A. M(1) Q Logic and Actuation Relays N.A. N.A. 1, 2, 3 N.A. N.A.

c. Containment Pressure- S R Q High N.A. N.A. N.A. 1, 2, 3

d. Pressurizer Pressure- S R Q N.A.

Low-Low i

2. Containment Spray N.A. N.A. R N.A. M.A. M.A. 1, 2, 3, 4

a. Manual Initiation N.A. 1, 2, 3, 4 Automatic, Actuation M.A. N.A. M.A. N.A. M(1) M(1) Q b.

Logic and Actuation Relays N.A. N.A. N.A. 1, 2, 3 R Q N.A.

c. Containment Pmssure-- S High-High 3/43-37 Amendment No.

McGUIRE - UNIT 1

INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued) may be initiated by the Engineered Safety Features Actuation System to mitigate the consequence.S Injection of a automatic pumps start and steam line breakposition, valves cr loss-of-coolant accidents (2) Reactor trip. 3 (1)(Safety feedwater isolation, (4) startup of the emergency diesel generators. (5) )

containment spray pumps start and automatic valves position, (6) containment isolation, (7) steam line isolation, (8) Turbine tri pumps start and automatic valves position, and (10) p. (9) nuclear auxiliary service waterfeedwater pumps start and automatic valves position.

Technical S)ecifications for the Reactor Trip Breakers and the Reactor Trip Bypass Brea(ers are based u)on NRC Generic Letter 85-09 " Technical Specifications for Generic .etter 83-28 Item 4.3," dated May 23, 1985.

The Engineered Safety Features Actuation System interlocks perform the following functions:

P-4 Reactor tripped - Actuates Turbine trip, closes main feedwater valves on T below Setpoint, prevents the opening of the main feedwater valves wf[i'ch were closed by a Safety Injecticn or High Steam Generator Water.

Level signal, allows Safety Injection block so that components can be reset or tripped.

Reactor not tripped - prevents mar.ual block of Safety Injection.

P-11 Defeats the manual block of Safety Injection actuation on low pressurizer pressure and defeats steamline isolation on negative steamline pressure l rate. Defeats the manual block of the motor-driven auxiliary feedwater pumps on trip of main feedwater pumps and low-low steam generator water level.

P-12 On increasing reactor coolant loop temperature P-12 automatically provides an anning signal to the steam dump system. On decreasing reactor coolant loop temperature, P-12 automatically remnves the arming signal from the steam dump system.

P-14 On increasing steam generator level. P-14 automatically trips all feed-water isolation valves and inhibits feedwater control valve modulation.

McGUIRE - UNIT 1 03/43-2

TABLE 3.3-3 ,

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

1. Safety Injection, Reactor Trip, Feedwater Isolation, Component Coding Water, Start Diesel Generators, and Nuclear Service Water

a. Manual Initiation 2 1 2 1,2,3,4 18

b. Automatic Actuation 2 1 2 1,2,3,4 14 Logic and Actuation Relays

c. Containment 3 2 2 1,2,3 15 Pressure-High

d. Pressurizer 4 2 3 1, 2, 3f 19 Pressure - Low-Low McGUIRE - UNIT 2 3/43-17 Amendment No.

TABLE 3.3-3 (Continued)

TABLE NOTATION

1. Trip function may be blocked in this MODE below the P-11 (Pressurizer Pressure Interlock) Setpoint.

ff Trip function automatically blocked above P-11 and may be blocked below P-11 when Main Steam Isolation on low steam pressure is not blocked. l 1

• • These values left blank pending NRC approval of three loop' operation.

Note 1: Turbine driven auxiliary feedwater pump will not start on a blackout signal coincident with a safety injection signal.

ACT!DN STATEMENTS ACTION 14 With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />s:

however, one channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLE.

ACTION 15 With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed until perfomance of the next required OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 15a With the number of OPERABLE channels less than the Total Number of Channels, operation may proceed until perfomance of the next required OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. With more than one channel inoperable, enter Specification 3.8.1.1.

ACTION 15b With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed until perfomance of the next required OPERATIONAL TEST provided-the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 16 With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is 11 aced in the bypassed condition and the Minimum Channels OP F.RABLE requirement is met. One additional channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing per Specification 4.3.2.1.

ACTION 17 With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment purge supply and exhaust valves are maintained closed.

-McGUIRE --UNIT-2 3/4-3-24 Amendment No.

TABLE 3.3-4 .

ENGINEERED SAFETY FEATUR'!S ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

1. Safety Injection, Reactor Trip, Feedwater Isolation, Component Cooling Water, Start Olesel Generators, and Nuclear Service Water.

a. Manual Initiation N.A. N.A.

b. Automatic Actuation Logic N.A. N.A.

and Actuation Relays

c. Containment Pressure--High s 1.1 psig s 1.2 psig

d. Pressurizer Pressure--Low-Low 2 1845 psig 2 1835 psig i

2. Containment Spray

a. Manual Initiation N.A. N.A.

b. Automatic Actuation Logic N.A. N.A.

and Actuation Relays

c. Containment Pressure--High-High s 2.9 psig s 3.0 psig McGUIRE - UNIT 2 3/43-27 Amendment No.

TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIES INITIATING SIGNAL-AND FUNCTION RESPONSE TIME IN SECONDS

3. Pressurizer Pressure-Low-Low

a. SafetyInjection(ECCS) s 27(1)/12(3)

b. Reactor Trip (from SI) s2

c. Feedwater Isolation s 12-

d. - Containment isolation-Fhase "A.(2) s 18(3)/28(4)

e. Containment Purge and Exhaust Isolation s4

f. Auxiliary Feedwater(5) N.A.

g. Nuclear Service Water System s 76(I)/65(3)

h. Component Cooling Water s 76(I)/65(3)

1. Start Diesel Generators s 11

_4._ Steam Line Pressure-Low Steam Line Isolation s 10 l

McGUIRE:- UNIT 2 3/4_3-33 !a udment_No.

TABLE 4.3-2 ,

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ANALOG ACTUATING MODES CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED

1. Safety Injection, Reactor Trip, Feedwater Isolation, Component Cooling Water, Start Diesel Generators, and Nuclear Service Water -

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4

b. Automatic Actuation N.A. N.A N.A N.A. Mf1) M(1) Q 1,2,3,4 Logic and Actuation Relays

c. Containment Pressure- S R Q N.A. N.A. N.A. N.A. 1, 2, 3 High

d. Pressurizer Pressure- S R Q N.A. N.A. N.A. N.A. 1, 2, 3 Low-Low

2. Containment Spray

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4

b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 Logic and Actuation Relays

c. Containment Pressure-- S R Q N.A. N.A. N.A. N.A. 1, 2, 3 High-High l

l McGUIRE - UNIT 2 3/43-37 Amendment No.

, igiIgyMENTATION BASES 3/4.3.1and3/4.3.2REACTORTRIPANDENGINEEREDSAFETYFEATURESACTUATION SYSTEM INSTRUMENTATION (Continued) may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss-of-coolant accidents (1) Safety Injection pumps start and automatic valves position, (2) Reactor trip.

feedwater isolation, containment spray pump (4) startup of the emergency diesel s start and automatic valves position (6) containment generator isolation, (7) steam line isolation, (8) Turbine trip. (9) auxiliary feedwater pumps start and automatic valves position, and (10) nuclear service water pumps start and automatic valves position.

Technical S)ecifications for the Reactor Trip Breakers and the Reactor Trip Bypass Brea(ers are based u)on NRC Generic Letter 85 09 " Technical Specifications for Generic .etter 83-28. Item 4.3." dated May 23, 1985.

The Engineered Safety Features Actuation System interlocks perform the following functions:

P-4 Reactor tripped - Actuates Turbine trip, closes main feedwater valves on T below Setpoint, prevents the opening of the main feedwater valves w[1'ch were closed by a Safety Injection or High Steam Generator Water t.evel signal, allows Safety Injection block so that components can be reset or tripped.

Reactor not tripped - prevents manual block of Safety Injection.

P-11 Defeats the manual-block of Safety Injection actuation on low pressurizer pressure and defeats steamline isolation on negative steamline pressure l rate. Defeats the manual block of the motor-driven auxiliary feedwater pumps on trip of main feedwater pumps and low-low steam generator water level.

P-12 On increasing reactor coolant loop temperature P-12 automatically provides an arming signal to the steam dump system. On decreasing reactor coolant loop temperature, P-12 automatically removes the anning signal from the steam dump system.

P-14 On increasing steam generator level, P-14 automatically trips all feed-water isolation valves and inhibits feedwater control valve modulation.

McGU RE - UNIT 2 B3/43-2

. _ _ - _ _ - - _ _ i

ESFAS Instrumentation 3.3.2 Table 3.3.2 1 (page 1 of 6)

Engineered Safety feature Actuation System Instrtsnentation APPL l CABLE MODES OR OTHER SPEClfitD Rt0VIRED $URVilLLANCE ALLOWABLE TRIP IUNCil0N CONDITIONS CHANNELS CONDitl0NS REQUIREMEN18 VALUE Li1PolNT

1. tafetyinjection

a. Manual Initletion 1,2,3,4 2 B SR 3.3.2.7 kA alA

b. Autonntle 1,2,3,4 2 trains C SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.4 and Actuation SR 3.3.2.6 Relays

c. Contalrvnent 1,2,3 3 0 $R 3.3.2.1 s 1.2 psig s 1.1 psig Pressure

• High SR 3.3.2.5 SR 3.3.2.8 SR 3.3.2.9

d. Pressuriser 1,2,3(4) 4 0 LR 3.3.2.1 a 1835 psig a 1845 pois Pressure a Low Low SR 3.3.2.5 SR 3.3.2.8 SR 3.3.2.9

2. Contaltvnent spray l

a. Manual Initiation 1,2,3,4 2 per s SR 3.3.2.7 NA NA train, 2 trains ,

b. Automatic Actuation 1,2,3,4 2 tralna C st 3.3.2.2 NA W4 L Logic and Actuation SR 3.3.2.4 Relays SR 3.3.2.6

c. Contairvnent Pressure

• High 1,2,3 4 L $R 3.3.2.1 s 3.0 psig a 2.9 psig High $R 3.3.2.5 ER 3.3.2.8 st 3.3.2.9

3. Containment Isolation

a. Phase A lsolation (1) Manual 1,2,3,4 2 8 $R 3.3.2.7 NA NA initiation (2) Automatic 1,2,3,4 2 trains C SR 3.3.2.2 NA NA Actuation $R 3.3.2.4 Logic and SR 3.3.2.6 Actuation Relays (continued)

(a) Above the P 11 (Pressuriter Pressure) Interlock.

McGuire Unit 1 3.3-30 9/3/97

ESFAS Instrumentation

. B 3.3.2 BASES l BACKGROUND Solid State Protection System (continued)

Applicable Safety Analyses, LCO, and Applicability sections of this Bases.

Each SSPS train has a built in testing device that can test the decision logic matrix functions and the actuation devices while the unit is at power. When any one train is taken out of service for testing, the other train is capable of providing unit monitoring and protection until the testing has been completed. The testing device is semiautomatic to minimize testing time.

The actuation of ESF components is accomplished through master and slave relays. The SSPS energizes the master relays appropriate for the condition of the unit. Each master relay then energizes one or more slave relays, which then cause actuation of the end devices. The master and slave relays are routinely tested to ensure operation. The test of the master relays energizes the relay, which then operates the contacts and a) plies a low voltage to the associated slave relays. T1e low voltage is not sufficient to actuate the slave relays but only demonstrates signal path continuity. The SLAVE RELAY TEST actuates the devices if their operation will not interfere with continued unit operation. For the latter case, actual component operation is prevented by the SLAVE RELAY TEST circuit, and slave relay contact operation is verified by a continuity check of the circuit containing the slave relay.

APPLICABLE Each of the analyzed accidents can be detected by one or SAFETY ANALYSES, more ESFAS Fonctions. One of the ESFAS Functions is the LCO, and primary actuation signal for that accident. An ESFAS APPLICABILITY Function may be the primary actuation signal for more than one type of accident. An ESFAS Function may also be a secadary, or backup, actuation signal for one or more other accidents. Functions such as mannal initiation, not l specifically credited in the accident safety analysis, are qualitatively credited in the safety analysis and the NRC staff approved licensing basis for the unit. These Functions may provide protection for conditions that (continued)

McGuire Unit 1 B 3.3-61 9/3/97 i

ESFAS Instrum:ntation

. B 3.3.2 l i

BASES APPLICABLE d. Safety injection-Pressurizer Pres 1ure-low SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY . required to satisfy the requirements with a two-out-of-four logic.

This Function must be OPERABLE in MODES 1, 2, and3(aboveP-11)tomitigatetheconsequences -

of an HELB inside containment. This signal may l be manually blocked by the operator below the l P-11 setpoint. Automatic Si actuation below this  ;

pressure set)oint is then performed by the Containment )ressure-High signal.

This Function is not required to be OPERABLE in-  !

MODE 3 below the P-11 setpoint. Other ESF functions are used ta detect accident conditions and actuate the ESF Jystems 11. this MODE. In MODES 4, 5, and 6, this function is not needed for accident detection and mitigation.

I d

i (cor.tinued) i f McGuire Unit 1 B 3.3-67 9/3/97

.,f.-,----.- , - - - , , . _ .-,y- .. .m.w w w w . . s , , . . .

ESFAS Instrumentation

. B 3.3.2 BASES APPLICABLE 2. Containment Surav SAFETY ANALYSES, LCO, and Containment Spray provides two primary functions:

APPLICABILITY ,

(continued) 1. Lowers containment pressure and temperature after an HELB in containment; and

2. Reduces the amount of radioactive lodine in the containment atmosphere.

These functions are necessary to

. Ensure the pressure boundary integrity of the containment structures and

. Limit the release of radioactive iodine to the environment in the event of a failure of the c"ntainment structure.

(continued)

McGuire Unit 1 B 3.3-68 9/3/97

l ESFAS Instrumentation

, B 3.3.2 l

BASES APPLICABLE a. Enaineered Safety Feature Actuation System SAFETY ANALYSES, interlocks-Reactor Trio. P-4 (continued)

LCO, and APPLICABILITY , could cause an insertion of positive reactivity with a subsequent increase in generated power.

To avoid such a situation, the n3ted Functions have been interlocked with P-4 as part of the design of the unit control and protection system.

None of the noted Functions serves a mitigation function in the unit licensing basis safety analyses. Only the turbine trip Function is explicitly assumed since it is an immediate consequence of the reactor trip Function.

Neither turbine trip, nor any of the other four Functions associated with the reactor trip signal, is required to show that the unit licensing basis safety analysis acceptance criteria are nnt exceeded.

The RTB position switches that provide input to the P-4 interlock only function to energize or de-energize or open or close contacts.

Therefore, this Function has no adjustable trip setpoint with which to associate a Trip Setpoint and Allowable Value.

This Function must be OPERABLE in MODES 1, 2, and 3 when the reactor may be critical or approaching criticality. This Function does not have to be OPERABLE in MODE 4, 5, or 6 because the main turbine, the MFW System, and the Steam Dump System are not in operation,

b. Fnaineered Safety Feature Actuation 'vstem Interlocks-Pressurizer Pressure. P-11 The P-11 interlock permits a nonnal unit cooldown and depressurization without actuation of SI or main steam line isolation. With two-out-of-three pressurizer pressure channels (discussed previously) less than the P-11 setpoint, the operator can manually block the Pressurizer Pressure-Low SI signal and the Steam Line l Pressure-Low steam line isolation signal (previouslydiscussed).

(continued)

McGuire Unit 1 B 3.3-87 9/3/97

ESFAS Instrumentation

, B 3.3.2 BASES APPLICADLE b. Enaineerfd Safety Feature Actuation System SAFETY ANALYSES, interlocks-Pressurizer Pressure. P-11 LCO, and (continued)

APPLICABILITY When the Steam Line Pressure-Low steam line isoir. tion signal is manually blocked, a main stea'n isolation signal on Steam Line Pressure-Negt.tive Rate-High is enabled. This )rovides protection for an SLB by closure of 11e MSIVs.

With two-out-of-three pressurizer pressure channels above the P-11 setpoint, the Pressurizer Pressure-Low Si signal and the Steam Line l Fressure-Low steam line isolation signal are automatically 2nabled. The operator can also enable these trips by use of the respective manual reset buttons. When the Steam Line Pressure-Low steam line isolation signal is enabled, the main steam isolation on Steam Line Pressure-Negative Rate-High is disabled.

This Function must be OPERABLE in MODES 1, 2, and 3 to allow an orderly cooldown and depressurization of the unit without the actuation of SI or main steam isolation. This Function does not have to be OPERABLE in MODE 4 5, or 6 because system pressure must already be below the P-11 setpoint for the requirements of the heatup and cooldown curves to be met.

c. Enoineered Safety Feature Actuation System Interlocks-T.y-Low Low. P-12 On increasing reactor coolant temperature, the P-12 interlock provides an arming signal to the Steam Dump System. On a decreasing temperature, the P-12 interlock removes the arming signal to the Steam Dum) System to prevent an excessive cooldown of t1e RCS due to a malfunctioning Steam Dump System.

i Since temperatT Ure,sthisused as an meets Function indication of bulk RCS redundancy requirements with one OPERABLE channel in each loop. These channels are used in two-out-of-four logic.

(continued)

McGuire Unit 1 B 3.3-88 9/3/97

ESFAS InstrumentatiCn

. 3.3.2 Table 3.3.21 (page 1 of 6)

Engineered safety feature Actuation systers instrumentation APPLICAsLE MODE $ OR OTHER SPECIFitD REQUIRED $URVilLL ANCE ALLOWAgLE TRIP FUNCil0N CONDlil0NS CHANNELS CONDillONS REQUIREMENis VALUE stIPolNT

1. Safetyinjection

a. Manual Initiation 1,2,3,4 . 2 g SR 3.3.2.7 NA NA

b. Autceatic 1,2,3,4 2 trains C st 3.3.2.2 NA NA Actuation Logic SR 3.3.2.4 and Actuation st 3.3.2.6 C'. lays

c. Contaltnent 1,2,3 3 D SR 3.3.2.1 s 1.2 psig s 1.1 psig Pressure.High st 3.3.2.5 SR 3.3.2.8 SR 3.3.2.9

d. Pressurfter 1,2,3(a) 4 0 SR 3.3.2.1 a 1835 psig a 1845 psig Pressure Low Low sR 3.3.2.$

$R 3.3.2.8 SR 3.3.2.9

2. Contaltvwnt sprey l

a. Manual Initiation 1,2,3,4 2 per g st 3.3.2.7 NA NA train, 2 trains

b. Automatic Actuation 1,2,3,4 2 trains c sk 3.3.2.2 NA NA Logic and Actuation st 3.3.2.4 Relays SR 3.3.2.6

c. Contaltnent Pressure . High 1,2,3 4 E SR 3.3.2.1 s 3.0 psig s 2.9 psig High SR 3.3.2.5 st 3.3.2.8 SR 3.3.2.9

3. Containment Isolation

s. Phase A Isolation (1) Manual 1,2,3,4 2 g $t 3.3.2.7 NA NA Initiation (2) Automatic 1,2,3,4 2 trains C $R 3.3.2.2 NA NA Actuation tR 3.3.2.4 Logic and SR 3.3.2.6 Actuation Relays (continued)

(a) Above the Pall (Pressuriter Pressure) Interlock.

McGuire Unit 2 3.3-30 9/3/97

ESFAS Instrumentation

. B 3.3.2 BASES BACKGROUND Solid State Protection System (continued)

Applicable Safety Analyses, LCO, and Applicability sections

, of this Bases.

Each SSPS train has a built in testing device that can test the decision logic matrix functions and the actuation devices while the unit is ut power. When any one train is taken out of service for testing, the other train is capable of providing unit monitoring and protection until the testing has been completed. The testing device is semiautomatic to minimize testing time.

The actuation of ESF components is accomplished through master and s inve relays. The SSPS energizes the master relays apprspriate for the condition of the unit. Each master r;iay then energizes one or more slave relays, which then cause actuation of the end devices. The master and slave relays are routinely tested to ensure operation. The test of the master relays energizes the relay, which then operates the contacts and aoplies a low voltage to the associated slave relays. The low voltage is not sufficient to actuate the slave relays but only demonstrates signal path continuity. The SLAVE RELAY TEST actuates the devices if their operation will not interfere with continued unit operation. For the latter case, actual component operation is prevented by the SLAVE PILAY TEST circuit, and slave relay contact operatice, is verified by a continuity check of the circuit containing the slave relay.

APPLICABLE Each of the analyzed accidents can be detected by one or SAFETY ANALYSES, more ESFAS Functions. One of the ESFAS Functions is the LCO, and primary actuation signal for that accident. An ESFAS APPLICABILITY Function may be the primary actuation signal for more than one type of accident. An ESFAS Function may also be a secondary, or backup, actuation signal for one or more other accidents. Functions such as manual initiation, not l specifically credited in the accident safety analysis, are qualitatively credited in the safety analysis and the NRC staff approved licensing basis for the unit. These Functions may provide protection for conditions that (continued)

McGuire Unit 2 B 3.3-61 9/3/97

ESFAS Instru entation

. B 3.3.2 BASES APPLICABLE d. Safety in.iection Pressprirer Pressure-Low SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY required to satisfy the requirements with a two-out-of-four logic.

This Function must be OPERABLE in MODES 1, 2, and 3 (above P-11) to mitigate the consequences of an HELB inside containment. This signal may be manually blocked by to: operator below the P-11 setpoint. Automatit ' i actuation below this pressure setpoint is then perfomed by the Containment Pressure-High signal.

This Function is not required to be OPERABLE in MODE 3 below the P-11 setpoint. Other ESF functions are used to detect accident conditions and actuate the ESF systems in this MODE. In MODES 4, 5, and 6, this Function is not needed for accident detection and mitigation.

(continued)

McGuire Unit 2 8 3.3-67 9/3/97

ESFAS Instrumentaticn B 3.3.2

[<

BASES APPLICABLE - 2. Containment Soray SAFETY ANALYSES, .

LCO, and Containment Spray provides two primary functions:

APPLICABILITY ,

(continued) 1. Lowers containment pressure and temperature after an HELB in containment: and 2.. Reduces the amount of radioactive iodine in the containment atmosphere.

These functions are necessary to

• Ensure the pressure boundary integrity of the containment structure and

• Limit the release of radioactive iodine to the environment in the event of a failure of the containment structure.

(continued)

McGuire Unit 2 B 3 . . ' . 'l 9/3/97

. m . . .

,_,,j

ESFAS Instrumentation

, B 3.3.2 BASES l APPLICABLE a. Enaineered Safety Feature Actuation System SAFETY ANALYSES, Interlocks-Reactor Trip. P-4 (continued)

LCO, and APPLICABILITY could cause an insertion of positive reactivity with a subsequent increase in generated power.

To avoid such a situation, the noted Functions have been interlocked with P-4 as part of the design of the unit control and protection system.

None of the noted Functions serves a mitigation fJnction in the unit licensing basis safety.

analyses. Only the turbine trip Function is explicitly assumed since it is an immediate consequence of the roactor trip Function.

Neither turbine trip, nor any of the other four Functions associated with the reactor trip signal, is required to show that the unit licensing basis safety analysis acceptance criteria are not exceeded.

The RTB position switches that provide input to the P-4 interlock only function to energize or de-energize or open or close contacts.

Therefore, this Function has no adjustable trip setpoint with which to associate a Trip Setpoint and Allowable Value.

This Function must be OPERABLE in MODES 1, 2, and 3 when the reactor may be critical or approaching criticality. This Function does not have to be OPERABLE in MODE 4, 5, or 6 because the main turbine, the MFW System, and the Steam Dump System are not in operation,

b. Fnaineered Safety Feature Actuation System Interlocks-Pressurizer Pressure. P-ll The P-11 interlock pemits a normal unit cooldown and depressurization without actuation of SI or main steam line isolation. With two-out-of-three pressurizer pressure channels (discussed previously) less than the P-11 setpoint, the operator can manually block the Pressurizer Pressure-Low SI signal and the Steam Line l Pressure-Low steam line isolation 510n4 (previously discussed).

(continued)

McGuire Unit 2 B 3.3-87 9/3/97 i

ESFAS Instrumentation B 3.3.2 1

BASES APPLICABLE b. Q19lneered Saf reature Actu_a11 ton System SAFETY ANALYSES, 1p;erlocks-Pren rizer Pressure. P-11 LCO, and (continued)

APPLICABILITY When the Steam Line Pressure-Low stes a line isolation signal is manually blocked, a main steam isolation signal on Steam Line Pressure-Negative Rate-High is enabled. This )rovides protection for an SLB by closure of t1e MS!Vs.

With two-out-of-three pressurizer pressure channels above the P-11 setpoint, the Pressurizer Pressure-Low SI signal and the Steam Line l Pressure-Low steam line isolation signal are automatically enabled. The operator can also enable these trips by use of the respective manual reset buttons. When the Steam Line Pressure-Low steam line isolation signal is enabled, the main steam isolation on Steam Line Pressure-Negative Rate-liigh is disabled.

This Function must be OPERABLE in MODES 1, 2, and 3 to allow an orderly cooldown and depressurization of the unit without the actuation of SI or main steam isolation. This Function does not have to be OPERABLE in MODE 4, 5, or 6 because system pressure must already be below the P-11 setpoint for the requirements of the heatup and cooldown curves to be met.

/

/ c. Enaineered Safety Feature Actuation System

) Interlocks-Tyd ow Low. P-12 i

On increasing reactor coolant temperature, the P-12 interlock provides an aming signal to the Steam Dump System. On a decreasing temperature, the P-12 interlock removes the aming signal to the Steam Dum) System to prevent an excessive cooldown of tie RCS due to a malfunctioning Steam Dump System.

Since T i temperaElre,s used this as anmeets Function indication of bulk RCS redundancy requirements with one OPERABLE channel in each loop. These channels are used in two-out-of-four logic.

(continued)

McGuire Unit 2 B 3.3-88 9/3/97

/

'#4.4MhJh.-6AWw h e hm .__ M ed u-.AA h h 9

ATTACHMENT 11 MARKED UP TECHNICAL SPECIFICATION PAGES i

l

~

u v *

\

TABl$3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INST TOTAL NO. MINIMUM FUNCTIONAL UNIT CHANNELS OF CHANNELS CHANNELS TO TRIP APPLICABLE

1. OPERABLE Safety injection, Reactor _

MODES _ ACTION Trip, Feedwater Isolation, Component Cooling Water, Start Diesel Generators, and Nuclear Service Water

a. Manual Initiation 2 1 2 1,2,3,4

b. 18 Automatic Actuation 2 Log 1c and Actuation 1 2

Relays 1, 2, 3, 4 14

c. Containa. 3 2 Pressure-High 2 1. 2, 3 15

d. Pressurizer 4 2 Pressure - Low-Low 3 1, 2, 37 19

-e.

Ste:: Lin: P=:: re La eter L::p:

^per ting-  ?/ te:r 'ine 2/:te:r 'in la ray rterm 2/:tter ?inc 1, 2, ?f 15-h S re: L::p:

Operatfng () (") (") ("] ("}

McGUIRE - UNIT 1 3/4 3-17 l Anendment No. 166

1. F

<' TABLE 3.b3 (Continued)

JADLE NOTATION Pressure Interlock) Setpoint. Trip function mayurizer be blocked in this H

Trip function automatically blocked above P-11 and may be bloc

• • 11 when M,9WW ety S"cctien on low steam pressure is not blocked e elow P-t1 ' Dol.KnOO .

These values left blank pending NRC approval Note 1:

of three loop ope ration.

signal coincident with a safety injection ackout signal. Turb ACTION STATEMENTS ACTION 14 With Channels theOPERAELE number ofrequirement,OPERABLE be inchannels at least HOT one STANDles hours and in COLD SHUTDOWN within the following n 12 r, 30 hou one channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing per Specification 4.3.2.1, previded thee other OPERABLE.

s chann l i ,

ACTION 15 J With of the number Channels, operation ofmay OPERABLE proceed untilchannels performance umber one less of the n required OPERATIONAL in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. TEST provided the inoperable chan ACTION 15a With the number of OPERABLE Channels, operation may proceed until e Total Numberchannels performance of of the nextless required in the tripped OPERATIONAL TEST provided the inoperable channe condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

channel inoperable, enter Specification 3.8.1.1.With more than one ACTION 15b With the number of OPERABLE n the Total Numberchannels one le of Channels, operation may proceed until performance of the nex required in the tripped OPERATIONAL condition within I hour.TEST provided the inoperable channe ACTION 16 With of the number Channels, o>eration ofmay OPERABLE proceed provided channels one lessc er inoperable the is placedrequirement OPERABLE in tae bypassed is met. condition and the Minimum Channels Specification 4.3.2.1.bypssed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance te ACTION 17 With less than the Minimum Channels OPERABLE requirement

, operation may valvescontinue provided are maintained closed.the containment purge supply and exhaust McGUIRE - UNIT 1 3/4 3-24 Amendment No. 166

_ __ _ _ _ _ - - - ~ ^ '

V U y

TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

1. Safety Infection, Reactor Trip, Feedwater Isolation, Component Cooling Water, Start Diesel Generators, and '

Nuclear Service Water.

a. Manual Initiation N.A. N.A.

b. Automatic Actuation Logic N.A. N.A.

and Actuation Relays

c. Containment Pressure--High s 1.1 psig s 1.2 psig

d. Pressurizer Pressure--l.ow-Low 2 1845 psig 2 1835 psig

. St ;- Lin 'rcssurc - Lcw t 775 psig z.755 p;ig

2. Containment Spray

a. Manual Initiation N.A. N.A.

b. Automatic Actuation Logic N.A. N.A.

and Actuation Relays

c. Containment Pressure--High-High s 2.9 psig i s 3.0 psig McGUIRE - UNIT 1 3/4 3-27 Amendment No. 166

I

) ENGINEERED SAFETY FEATURES RESPONSE TlHES JNITIATING SIGNAL AND FUNCTION RESPONSE TlHE IN SECONDS _

3. Pressurizer Pressure-tow-low

a. Safetyinjection(ECCS) s 27hl/1213) l b. ReactorTrip(fromSI) s2

c. Feedwater Isolation s 12

d. Containment Isolation-Phase "A.12) g 3g(31/28(4) i

e. Containment Purge and Exhaust Isolation s4

f. Auxiliary feedwatert51 N.A.

g. Nuclear Service Water System s 7603/65(3)

{

h. Component Cooling Water s 7603/65(3)

1. Start Diesel Generators s 11

4. Steam line Pressure-Low

] -e -

Sifety !njectier (ECGS}-  ; 1 2'31/22'4L -

b Daar+ar T-ia (frer!!) ;2-

+- Fee &ater !:e h t ha -- -

-le

'c Centainment-4s+ht4en """ ""(2) .. ._ _

_ ; gal /E6GL

=e. Centat m:nt Pur;: :nd Exh:: t- 1:cht4en ;4 Auxili:ry F : h:tu tst , , _ ,,_ _ g,3, l,,  ;.

Nuclear Service M:ter-- -

55'31/70'41

$ Steam Line Isolation s 10 l i -

C:rpen':nt C lin; M:ter  : SEGI /75L41

,j . Start Cie 1 Centratert  ; ::

l l

s ,

McGUIRE - UNIT 1 3/4 3-33 Amendment No. 166

v G .

TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ANALOG ACTUATING MODES CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY FUNCTIONAL UNIT ' CHECK RELAY SURVEILLANCE CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REOUIRED

1. Safety Injection, Reactor Trip, Feedwater Isolation, Component Cooling Water, Start Diesel Generators, and Nuclear Service Water a anual Initiation N.A. N.A. N.A. R N.A. N.A. N.A.

b. tomatic Actuation 1, 2, 3, 4 N.A. N.A K.A N.A.

'gic and Actuation M(1) M(1) Q 1, 2, 3, 4 Relays

c. Containment Pressure- S R N.A.

Q N.A. N.A. N.A. 1, 2, 3 High

d. Pressurizer Pressure- S R N.A.

Q N.A. N.A. N.A. 1, 2, 3 Low-Low

e. Ste?- Line T " ""

Prc:::;rc L=

Q

. .. " f. .

.. ".*. 1, 2, 3

?. Containment Spray

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4

b. Automatic Actuation N.A. N.A. N.A. N.A.

Logic and Actuation M(1) M(1) Q 1, 2, 3, 4 Relays

c. Containment Pressure-- S R Q N.A. N.A.

High-High N.A. N.A. 1,2,3 McGUIRE - UNIT 1 3/4 3-37 Amend:ent No. 166 j

INSTRUMENTATION

) PASES 3/_4 3.1 and 3/4.3.2 REACTOR " RIP AND ENGINEERED SAFETY FEATUR SYSTEMINSTRUNENTATION(Cont'nued) may be itdtiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss-of-coolant accident (1) Safety injection pumps start and automatic valves position, feedwater isolation, containment spray pump (4) startup of elthe emergency generators. (5) dies (2) Reacto isolation, (7) steam line isolationsstartandautomaticvalvesposition,(6) containment (8) Turbine trip, (9) auxiliary feedwater pumpsstartandautomaticvalvespos,ition,and(10)nuclearservicewaterpumps start and automatic valves position.

Technical S)ecifications for the Reactor Trip Breakers and the Reactor Trip Bypass Brea(ers are based u)on NRC Generic Letter 85-09 " Technical Specifications for Generic

.etter 83-28, item 4.3," dated May 23, 1985.

The Engineered following Safety features Actuation System interlocks perfom the functions:

P-4 Reactor tripped - Actuates Turbine trip, closes main feedwater valves on T

below Setpoint, prevents the open<ng of the main feedwater valves wD:h were closed by a Safety injection or High Steam Generator Water Level signal, allows Safety injection block so that components can be reset or tripped.

)

Reactor not tripped - prevents manual block of Safety injection.

P-11 pressure Defeatsandthe manual block of Safety injection actuation on low pressurizer

'~;f stea %ine pres m e and defeats steamline isolation on negative tteamline pressure rate. Defeats the maaual block of the motor-(

driven auxiliary feedwater pumps on trip of main feedwater pumps and low-low steam generator water level.

P-12 On increasing reactor coolant loop temperature, P-12 automatically provides an aming signal to the steam dump system. On decrtasing reactor coolant loop temperature, P-12 automatically removes the anning signal from the steam dump system.

Pd 1 increasing steam generator level, P-14 automatically trips all feed-ster isolation valves and inhibits feedwater control valve modulation.

..[_. m.

< m e* AM .,

y .

~

TABLE 3.3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION HINIMUM TOTAL NO, CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

1. Safety Injection. Reactor Trip, Feedwater Isolation,

- Component Cooling Water, Start Diesel Generators, and Nuclear Service Water

a. Manual Initiation 2 1 2 -1, 2, 3, 4- 18

b. Automatic Actuation '2 Logic and Actuation 1 2 1,2,3,4 14 Relays

c. Containment 3 2 2 1, 2, 3 15 Pressure-High

d. Pressurizer 4 2 3 1, 2, 3f 19 Pressure - Low-Law

c. St::= Line Prc::tre-L =

Ferr Leeps 3/ te:r " e 2/:te:r 'in 2/ t = 'in: 1, 2, 2' 15-Oper: tins #-

=yste= l 44*e-Thr;c L;;p:; - ('*) (**) (**)

Operating (**) (**)

McGUIRE - UNIT 2 3/4 3-17 , Amendmhnt No. 148 g5#Atwauw,w;z .Larerwsm-**wn - -~ ---

~ -

=

IABLE 3.3-3 (Continued)

IABLE NOTATION f

Trip function Pressure may beSetpoint.

Interlock) blocked in this MODE below the P-11 (Pressurizer l

{ H Trip function automatically blocked above P-11 and may be blocked below P-11 when t%pJGahty STC%iRj::hol-RTib tim on lowbteam pressure is not blocked, These values left blank pending NRC approval of three loop operation.

Note 1:

Turbine driven auxiiiary feedwater pump will not start on a blackout signal coincident with a safety injection signal.

4 ACTION STATEMENTS ACTION 14 g With the number of OPERABLE channels one less than the Minimum 4 Channels OPERABLE requirement, be in at least HOT STANDBY within b 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />;

d. however, one channel may be bypassed for up to-4 hours for I channel is OPERABLE. surveillance testing pstr Specificatior. 4.3.2.1, provided th "j

ACTION 15 With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed until performance of the i

} next required OPERATIONAL TEST provided the inoperabic channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. -

.)

S y ACTION Channels, 15a With the number operation may proceed of OPERABLE until performancechannels of the next less than the Tota s

i required OPERATIONAL TEST provided the inoperable channel is 5 placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. With more than i

[ one channel inoperable, enter Specification 3.8.1.1.

(

1 t ft ACTION 15b With the number of OPERABLE channels one less thanJ the Total Number of Channels, operation may proceed until performance of the

$ next required OPERATIONAL TEST provided the ino  :

L placed in the tripped condition within I hour. perable channel is j ACTION 16 .

With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperable '

channel is placed in the bypassed condition and the Minimum j Channels OPERABLE requirement is met. ,

be bypassed4.3.2.1.

for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing perone additional cha Specification ACTION 17-With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment purge supply and exhaust valves are maintained closed.

t McGUIRE - Oh!T 2 3/4 3-24 Amendment No. 148 g

___7 l

-.. = m: =wc . . .

V v v TABLE 3.3-4 I ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMEi!TATION TRIP SETPOINTS

]

l FUNCTIONAL UNIT TRIP SETPOINT I

ALLOWABLE VALUES  !

1.. Safety Injection Reactor Trip.

Feedwater Isolation, Component Cooling Water, Start Diesel Generators, and Nuclear Service Water.. j j

3

a. Manu:? Initiation N.A. N.A. 3

! b. Automatic Actuation Logic N.A. N.A. T

-1 and Actuation Relays j

c. Containment Pressure--High s 1.1 psig s 1.2 psig-

d. Pressurizer Pressure--Low-Low 2: 1845 psig 2: 1835 psig i

e. Ste r Line are::ere Le 2 75 prig 2 755 pri;

{

2. Containment Spray

a. Manual Initiation N.A. N.A.

b. Automatic Actuation Logic .N.A. N.A.

and Actuation Relays

c. Containment Pressure--High-High s 2.9 psig s 3.0 psig McGUIRE - UNIT 2 3/4 3-27 , Amendment No. 148

. _ _ _ _ = _ _ -

_ . _ _ _ - - _ . . w.a v v~ -- . . . >. . ~ ~ ~ - - - -

. , . . . _ . , . - .. , . . . . -ii---'": " ' ~ '

LW2ht45WElid&WGilide

• 1

. IABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TINES i

JNITIATING SIGNAL AND FU.NCTION RESPONSE TIME IN SECONDS

3. Pressurizer Pressure-Low-Low

a. Safety Injection (ECCS) s 2703/12(3)

b. Reactor Trip (from SI) s2

c. Feedwater Isolation s 12 d.

Containment Isolation-Phase "A.(2) s 18133 /28I43

e. Containment Purge and Exhaust Isolation s4

f. Auxiliary Feedwater(s)

N.A.

g. Nuclear Service Water System s 7603/ 6 5(33

h. Component Cooling Water s 7603/65I3)

1. Start Diesel Generators s 11

4. Steam _ine Pressure-Low 2

S:fety Injection (ECCS)  ; 1 2'31/ 2 2'41 ~

b. R ::ttr Trip (frt;;; SI)  ; 2-w-----Feedwater I: let4:n s 12-COntain : t I: lati n-Phas: " Ant 2) g ;graij;3: 4i

e. C :t:f :=t Purge-and Exhaust I:cisticr. ;4

-f. A=ilicry ft:Leter* .

. A.

We'ee- Ser" ice Meter  ; 5;'31 /'c'41 h Steam Line Isolation s 10 i,---CO

penent 00:11:3 S ter  ; 55'31/75'4' j.

St:rt Dic: 1 C :: rater: -

11

r.;

McGUIRE - UNIT 2 3/4 3-33 )

Amendment No. 148 ,

gypy awi.,,-- - - - - --.- -

1

avm.ame - _ _

._7 - --

h v v '

[

2 TABLE 4.3-2

' 1 j

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION M SURVEILLANCE REQUIREMENTS TRIP W ANALOG ACTUATING I MODES CHANNEL DEVICE CHANNEL CHANNEL MASTER SLAVE- FOR WHICH- d OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE -

FUNCTIONAL UNIT CHECK CALIBRATION . TEST TEST LOGIC TEST TEST TEST IS REQUIRED y

1. Safety Injection, Reactor 3 Trip, Feedwater Isolation, p Component Cooling Water, A Start Diesel Generators,

• j and Nuclear Service Water

'4

?

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A.

b. Automatic Actuation N.A. N.A N.A N.A.

N.A. 1, 2, 3, 4 I Logic and Actuation M(1) M(1) Q 1, 2, 3, 4 . ib g

Relays

c. Containment Pressure- S g R Q N.A. N.A. N.A. N.A. 1, 2, 3 High 2

d. Pressurizer Pressure- S R Q N.A. N.A. N.A. N.A. 1, 2, 3 Low-Low

-e. Ste:r Li S O N.A. M.A. M.A. N.A. I, 2, 3

]$g Drecter - Le

2. Containment. Spray y

a. Manual Initiation N.A. N.A. N.A. -R N.A. N.A. N.A. 1, 2, 3, 4

b. Automatic Actuation N.A. N.A. N.A. N.A.

Logic and Actuation M(1) M(1) Q 1,2,3,4  !

D Relays

c. Containment Pressure-- S R N.A. N.A. N.A.

Q N.A. 1, 2, 3 4-High-High 4

6

?8 McGUIRE - UNIT 2  : ?E 3/4 3-37 ,

Amendment No. 148 ;3

~ .n.~ , . a . - - - - -. - +,

'l .'

Q.Q.[QQ& {h[{l l? f Nl$ 0

.- - fi 1

l

.. JNSTRUMENTATION BASES i 3/4.3.1 and 3/4.3.2 REACTOR TRIP AND ENGINEERED SAFETY FEATURES ACTUA SYSTEM INSTRUMENTATION (Continued) may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss-of-coolant accident: 1 i

Injection pumps feedwater isolation,start and automatic valves position, (2) Reactor trip,(1) Safety r

i. containment spray pump (4) startup of- the emergency diesel generators,  ; (5)(3 l isolation, (7) steam line isolations start and automatic valves position (6) containment  !

I (8) Turbine tri

)

pumps start and automatic valves pos,ition, and (10)nuclear p,-(9) service auxiliary feedwater water pumps start and automatic valves position, y I k Technical Bypass BreaS)ecifications for the Reactor Trip Breakers and the Reactor Trip cers are based upon NRC Generic Letter 85-09 " Technical i

i Specifications for Generic Letter 83-28. Item 4.3," dated May _23,1985. *)

The Engineered following Safety Features Actuation System interlocks perform the functions:

P Reactor tripped - Actuates Turbine tri), closes main feedwater valves on T below Setpoint, prevents the open' ng of the main feedwater valves wD:h were closed by a Safety Injection or High Steam Generator Water -

Level signal, allows Safety Injection block so that components can be reset or tripped. -

e,

l Reactor not tripped - prevents manual block of Safety Injection. ~

P-11 Defeats the manual block of Safety Injection actuation on low pressurizer pressure =d is :tedhc r=~e and defeats steamline isolation on 7.

negative steamline pressure rate. Defeats the manual block of the motor- &

driven low steam auxiliary feedwater generator waterpumps level. on trip of main feedwater pumps and low- y b

a P-12 On increasing reactor' coolant loop temperature, P-12 automatically e provides an arming signal to the steam dwnp system. On decreasing reactor coolant loop temperature, P-12 automatically removes the arming signal from the steam dump system.

P-14 On increasing steam generator level P-14 automatically trips 'all feed-water isolation valves and inhibits feedwater control valve modulation.

McGUIRE - UNIT 2 8 3/4 3-2

- , m 2#g ,

ESFAS Instrumentation e

3.3.2 Table 3.3.21 (page 1 of 6)

Ingineered Safety Feature Actuation System Instrumentation APPLICABLE HODES OR OTHER

$PECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNEL 3 CONDITIONS REQUIREMENTS VALUE SETPOINT

1. Safety injection

4. Manual Initiation 1.2.3.4 2 8 SR 3.3.2.7 NA NA

t. Automatic 1.2.3.4 2 trains C SR 3.3.2.2 NA NA Actuation Logic 3R 3.3.2.4 and Actuation $R 3.3.2.6 Relays

c. Containment 1.2.3 3 SR 3.3.2.1 s 1.2 psig Pressure - High D s 1.1 psig SR 3.3.2.5 3R 3.3.2.8 SR 3.3.2.9

d. Pressurizer 1.2.3(8) 4 0 SR 3.3.2.1 a 1835 psig a 1845 psig Pressure - Low Low $R 3.3.2.5 SR 3.3.2.8 SR 3.3.2.9

~  ;

[35f?.1!^; , .

I' '

tn)

!.fil 0

U3 5'2'!'! ' 5 i"i i I i;Is

...m

..;. g:3.:. ..:

2. Containment Spray

a. Manual Initiation 1.2.3.4 2 per B SR 3.3.2.7 NA NA -

train. 2 trains

b. Automatic Actuation 1.2.3.4 2 trains C $R 3.3.2.2 NA RA Logic and Actuation SR 3.3.2.4 Relays $R 3.3.2.6

c. Containment Pressure - High 1.2.3 4 E SR 3.3.2.1 s 3.0 psig s 2.9 psig High SR 3.3.2.5 SR 3.3.2.8 SR 3.3.2.9

3. Containment Isolation

4. Phase A Isolation (1) Kanual 1.2.3.4 2 8 52 3.3.2.7 NA NA initiation (2) Automatic 1.2.3.4 2 trains C SR 3.3.2.2 NA NA Actuation SR 3.3.2.4 -

Logic and SR 3.3.2.6 Actuation Relays (continued)

(a) Above the P-11 (Pressurizer Pressure) interlock.

McGuire Unit 1 3.3-30 5/20/97

..[sc .-.

r ESFAS Instrumentation B 3.3.2 4

4

- BASES-BACKGROUND Solid State-Protection Svstem -(continued)

Applicable Safety Analyses,-LCO, and Applicability sections of this Bases.

Each.SSPS train has a built in testing device that- can test the decision logic matrix functions and the actuation devices while the-unit is at power. When any one train is taken out of service for testing, the other train is capable of- providing unit monitoring and protection until the testing has been congleted. The testing device is semiautomatic to minimize testing time.

The actuation of ESF components is accomplished through master and slave relays. The SSPS' energizes the master relays appropriate for the condition of the unit. Each master relay then energizes one or more slave relays, which then cause' actuation of the end devices. The nester-and >

slave relays are routinely tested to ensure cperation. The test of the master relays energizes the relay, which then operates the contacts and applies a low voltage to the associated slave relays. The low voltags is not sufficient to actuate the slave relays but only demonstrates signal path continuity. The SLAYE RELAY TEST actuates the devices -

if their operation will not interfere with continued unit ,

operation. For the latter case, actual component operation -

is prevented by the SLAVE-RELAY TEST circuit, and slave relay contact operation is verified by a continuity check of the circuit containing the slave relay.

APPLICABLE Each of the analyzed accidents can be detected by one or SAFETY ANALYSES, mont ESFAS Functions. One of the ESFAS Functions is the LCO, and primary actuation signal for that accident. ' An ESFAS APPLICA8ILITY Function may be the primary actuation signal for more than one type of accident. An ESFAS Function may also be a secondary, or backup. actuation signal for one or more other accidents. . . . . ..,..... ,.... ..... ..

- Wr" (L^"d ::d :  : tutti:

h:ht; ci;n:1

t :ti:n f:r ::.11 zi;n:1h:;

f;rf :t:::ce:hr,t :w.2,,t; 1in; beeks (SLM) : t:id: :: t:t- .nt. Functions such as manual initiation, not'specifically credited in the accident safety analysis, are qualitatively credited in the safety analysis and the NRC staff approved licensing basis for the unit.

These Functions may provide protection for conrlitions that (continued)

McGuire Unit 1 B 3.3-61 '

5/20/97 t

a

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

ESFAS Instrumentation

. B 3.3.2 BASES

. APPLICABLE d. Safety Iniection-Pressurizer Pressure-Low SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY required to satisfy the requirements with a two-out-of-four logic.

This Function must be OPERABLE in MODES 1, 2, and 3 (above P-11) to mitigate the consequences of an HELB inside containment. This signal may be manually blocked by the operator below the P-11 setpoint. Automatic SI actuation below this pressure setpoint is then performed by the Containment Pressure-High signal.

This Function is not required to be OPERABLE in MODE 3 below the P-11 setpoint. Other ESF functions are used to detect accident conditions cnd actuate the ESF systems in this MODE. In MODES 4, 5, and 6 this Function is not needed for accident detecticn and mitigation.

Safety Iniection-Steam Line Pressure-Low m Line Pressure-Low provides protection agai the following accidents: ~

• SLB;

• Feed line k; and

• Inadvertent open o n SG relief or an SG safety valve.

Steam Line Pressur ow provides input to any control functi . Thus, three OPE LE channels on each ste ine are sufficient to sa sfy the protecti requirements with a two-out-of- rce '

logic each steam line, is Function is anticipatory in nature and has a typical lead / lag ratio of 50/5.

(continued)

McGuire Unit 1 B 3.3-67 5/20/97

ESFAS Instrumentation B-3.3.2 BASES-APPLICABLE Safety Iniection-Steam Line Pressure-Low SAFETY ANALYSES, r

LCO, and APPLICABILITY continued)

Stea te Pressure-Low must be OPERABLE in-

/

-MODES 1, and 3 (above P-11) when a se dary side break o tuck open valve could suit in the rapid depres ization of the eam lines.

This signal may be ally b1 ed by the operator below the P-11 nt. Below P-11.-

feed line break is-not- o rn. Inside containment SLB wil e termina by automatic SI actuation via ntainment Press -High, and outside cont nt SLB will be termin d-by the Steam Lin ressure-Negative Rate-High sig for

-steam e_ isolation. This-Function is not re d to- be 0PERABLEL in MODE 4, 5. or 6 cause there is insufficient energy in the secondary side of the unit to cause an accident.

2._ containment sorav Containment Spray provides two primary functions:

1. Lowers containment-pressure and temperature after _

an HELB in containment; and- -

2. Reduces the amount of radioactive iodine in the containment atmosphere.

These functions are necessary to:

• Ensure the pressure boundary integrity of the containment structure; and

' Limit the release of radioactive iodine to the environment in the event of a failure of the ,

containment structure.

(continued)

McGuire Unit 1 B 3.3-68 5/20/97

ESFAS Instrumentation B 3.3.2 BASES APPLICABLE a. Enaineered Safety Feature Actuation System SAFETY ANALYSES, Jnterlocks-Reactor Trin. P-4 LCO, and (continued) s APPLICABILITY could cause an insertion of positive reactivity with a subsequent increase in generated power.

To avoid such a situation, the noted Functions have been interlocked with P-4 as part of the design of the unit control and protection system.

None of the noted Functions serves a mitigation function in the unit licensing basis safety analyses. Only the turbine trip Function is explicitly assumed since it is an inmediate consequence of the reactor trip Function.

Neither turbine trip, nor any of the other four Functions associated with the reactor trip signal, is required to show that the unit licensing basis safety analysis acceptance criteria are not exceeded.

The RTB position switches that provide input to the P-4 interlock only function to energize or de-energite or open or close contacts.

Therefore, this Function has no adjustable trip _

setpoint with which to associate a Trip Setpoint and Allowable Value.

This Function must be OPERABLE in MODES 1, 2, and 3 when the reactor may be critical or approaching criticality. This Function does not have to be OPERABLE in MODE 4, 5 or 6 because

' the main turbine, the MFW System, and the Steam Dump System are not in operation,

b. Enaineered Safety Feature Actuation System Interlocks-Pressurizer Pressure. P-11 The P-11 interlock permits a normal unit cooldown and depressurization without actuation of SI or \

main steam line isolation. With two-out-of-three pressurizer pressure channels (discussed [d previously) less than the P-11 setpoint, the operator can manually block the Pressurizer Pressure-Low and Ste:r Line Prc::Or -L w S!

- ign:10 and the Steam Line Pressure-Low steam

/[t

{ line isolation signal (previously discussed).

(contirJed)

McGuire Unit 1 B 3.3-87 5/20/97

..?- ..

ESFAS Instrumentation

, B 3.3.2 -

BASES-

.. APPLICABLE b. Enoineered Safety Feature Actuation System SAFETY ANALYSES, LCO, and Interlocks-Pressurizer Pressure. P-Q (continued)

APPLICABILITY When the Steam Line Pressure-Low steam line

-isolation signal is manually blocked, a main steam isolation signal on Steam Line Pressure-Negative Rate-High is enabled. This provides protection for an SLB by closure of the MSIVs.

With two-out-of-three pressurizer pressure 5g6g channels above the P-11 setpoint, the Pressuriz 'N Pressure-Low :nd Ot:= Lin: "=:r: L:: !!

> s ta=1e and the Steam Line Pressure-Low steam

. line isolation signal are automatically enabled.

The operator can also enabic these trips by use of the respective manual reset buttons.- When the -

Steam Line Pressure-Low steam line isolation signal is enabled, the main steam isolation on Steam Line Pressure-Negative Rate-High is -

disabled.

l This Function must be OPERABLE in MODES 1,_2 and 3 to allow an orderly cooldcwn and depressurization of tne unit without the actuation of SI or main steam isolation. This Function does not have to be OPERABLE in MODE 4, 5, or 6 because system pressure must already be below the P-11 setpoint for the requirements of the heatup and cooldown curves to be met,

c. Enoineered Safetv Feature Actuation System Interl ocks-T..._-Low Low. P-12 On increasing reactor coolant temperature. the P-12 interlock provides an arming signal to the Steam Dump System. On a decreasing temperature, the P-12 interlock removes the arming signal to the Steam Dump System to prevent an excessive cooldown of. the RCS due to a malfunctioning Steam Dump System.

Since T i temperaNre,sthis used as.an indication Function of bulk RCS meets redundancy requirements with one OPERABLE channel .in each loop. These channels are used in two-out-of-four logic.

(continued)

McGuire Unit 1 B 3.3-88 5/20/97

.m- -. .-

. -- i - _ _ _ _ _ _ _ _ _ _ _

- ESFAS Instrumentation 3.3.2 Table 3.3.21 (page 1 of 6)

Engineered $afety Feature Actuation $ystem Instrumentation APPLICABLE MODES OR OTHER

$PECIFIED REQUIRED $URVEILLANCE ALLOWA8LE FUNCTION TRIP CONDITIONS CHANNELS CONDITIONS REQUIREMENTS YALUE SETPO!NT-

1. - Safety injection

a. Manual Initiation 1,2,3,4 2 8 SR 3.3.2.7 NA NA

b. Automatic 1,2,3,4 2 trains C SR 3.3.2.2 NA NA Actuation Logic $R 3.3.2.4 and Actuation SR 3.3.2.6 Relays

c. Containment 1,2,3 3 SR 3.3.2.1 s 1.2 psig Pressure High D s 1.1 psig SR 3.3.2.5 SR 3.3.2.8 SR. 3.3.2.9

d. Pressurizer 1,2,3(a) 4 0 3R 3.3.2.1 a 1835 psig Pressure . Low Low a 1845 psig SR 3.3.2.5 SR 3.3.2.8 SR 3.3.2.9

~

555i!.!'

5,2. 5.-.

0 a #A bb i'i'i.'b i

.u rn .......

% 3. 0. 0. 0 --

2. Containment $ pray

a. Manual Initiation 1,2,3,4 2 per B SR 3.3.2.7 NA NA

~

train, 2 trains

b. Automatu Actuation 1,2,3,4 2 trains C $R 3.3.2.2 NA NA Lgic and Attuation $R 3.3.2.4 Relays

$R 3.3.2.6

c. Containment Pressure . High 1,2,3 4 E $R 3.3.2.1 s 3.0 psig s 2.9 psig Nigh SR 3.3.2.5 SR 3.3.2.8 SR 3.3.2.9 '

3. Containment Isolation

a. Phase A Isolation (1) Manual 1,2,3,4 2 8 SR 3.3.2.7 NA NA Initiation (2) Automatic 1,2,3,4 2 trains C SR 3.3.2.2 NA NA Actuation SR 3.3.2.4 Logic and SR 3.3.2.6 Actuation Relays (continued)

(a) Above the P.11 (Pressurizer Pressure) interlock.

McGuire Unit 2 3.3-30 5/20/97

• ESFAS Instrumentaticn B 3.3.2 BASES BACKGROUND Solid State Protection System (continued)

Applicable Safety Analyses. LCO, and Applicability sections of this Bases.

Each SSPS train has a built in testing device that can test the decision logic matrix functions and the actuation devices while the unit is at power. When any one train is taken out of service for testing, the other train is capable of providing unit monitoring and protection until the testing has been completed. The testing device is semiautomatic to minimize testing time.

The actuation of ESF components is accomplished through master and slave relays. The SSPS energizes the master relays appropriate for the condition of the unit. Each master relay then energizes one or more slave relays, which then cause actuation of the end devices. The master and slave relays are routinely tested to ensure operation. The test of the master relays energizes the relay, which then operates the contacts and applies a low voltage to the associated slave relays. The low voltage is not sufficient to actuate the slave relays but only demonstrates sional path continuity. The SLAVE RELAY TEST actuates the vices -

if their operation will not interfere with continues . .t operation. For the latter case, actual component op.,ation -

is prevented by the SLAVE RELAY TEST circuit, and slave relay contact operation is verified by a continuity check of the circuit containing the slave relay.

APPLICABLE Each of the analyzed accidents can be detected by one or SAFETY Af;ALYSES, more ESFAS Functions. One of the ESFAS Functions is the LCO, and primary actuation signal for that accident. An ESFAS APPLICABILITY Function may be the primary actuation signal for more than one type of accident. An ESFAS Function may also be a secondary, or backup, actuation signal for one or more other 4"

acc,_i dents.._, .....".+4 a

_ ,4m.... cr <m.

en vic,

. 33"r:=ri: r Pror=re-im,, Lew

,.nnion+ 2 r r u,,n n

~

t__li._

=- p.vooy 5e' EC M. I.I.72 S !2i[..'. #111.

. -----y . 12 2._l.. ~ .12. v_6som_._ s.s.a._.s_

i

' s. v_ l ' ._ l i _ _

visor, JL"3) Outsid: 00ntat = r.t. Functions such as manual initiation, not specifically credited in the accident safety analysis, are qualitatively credited in the safety analysis and the NRC staff approved licensing basis for the unit.-

These Functions may provide protection for condition: that S (continued)

McGuire Unit 2 B 3.3-61 5/20/97

1

,e . 1 ESFAS Instrumentation B 3.3.2 BASES l

APPLICABLE d. Safety Iniection-Pressurizer Pressure-Low SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY required to satisfy the requirements with a two-out-of-four logic.

This function must be OPERABLE in MODES 1, 2, and 3 (above P-11) to mitigate the consequences of an HELB inside containment. This signal may be manually blocked by the operator below the P-11 setpoint. Automatic SI actuation below this pressure setpoint is then performed by the Containment Pressure-High signal.

This Function is not required to be OPERABLE in MODE 3 below the P-11 setpoint. Other ESF functions are used to detect accident conditions and actuate the ESF systems in this MODE. In MODES 4, 5, and 6, this Function is not needed for accident detection and mitigation. 3

c. %fet Inicett:n Ster Line "c :ure-Leer eam Line Pressure-Low provider protecti -

aga t the following accidents:

. SLB, Feed line eak; and

=

Inaavertent ope of an SG relief or an SG safety va .

Steam Line P sure-Low provide o input to any control f tions. Thus, three 0 BLE channels on eac team line are sufficient to isfy the prot ive requirements with a two-out-o hree 1 e on each steam line.

This Function is anticipatory in nature and has a typic:1 10:d/1:g r:ti: Of50/5.

(continued)

McGuire Unit 2 B 3.3-67 5/20/97

ESFAS Instrumentation B 3.3.t BASES APPLICABLE se My Iniection-steam Line Pressure-Low

. SAFETY ANALYSES, (continued)

LCO, and- -

APPLICABILITY Stea ine Pressure-Low must be OPERABLE in MODES ,

. and 3 (above P-11) when a seconda side break- stuck open valve could resu n the rapid- dep urization of the ste ines.

This signal way b - nually bloc y the operator below the P- setpo . Below P-11, feed line break is not ocern. Inside containment SLB will te ated by automatic SI-actuation via tainrent sure-High, and outside cont nt SLB will be t inated by the Steam L1 ressure-Negative - Rate-H1 signal - for stea ne isolation. - This Function is t tred to be OPERABLE in MODE 4, 5,_ or 6 because there is insuffic 2nt energy in the secondary side of the unit to cause an accident.

2. containment Sorav Containment Spray provides two primary functions:

1. Lowers containment pressure and temperature after _

an HELB in containment; and

2. Reduces the amount of radioactive iodine in the containment atmosphere.

These functions are necessary to:

Ensure the pressure boundary integrity of the containment structure; and Limit the release of radioactive iodine-to the environment in-the event of a failure of the containment- structure.

1 (continued)

McGuire Unit 2 B 3.3-68 5/20/97 ,

t

f ESFAS Instrumentaticn B 3.3.2 BASES-APPLICABLE a. Enaineered Safety Feature Actuation System SAFETY ANALYSES, Interlocks-Reactor Trio. P-4 (continued)

LCO, and APPLICABILITY could cause an insertion of positive reactivity with a subsequent increase in generated power.

To avoid such a situation, the noted Functions have been interlocked with P-4 as part of the design of the unit control and protection system.

None of the noted Functions serves a mitigation '

function in the unit licensing basis safety analyses. Only the turbine trip Function is explicitly assumed since it is an inmediate consequence of the reactor trip Function.

Neither turbine trip, nor any of the other four Functions associated with the- reactor trip i signal, is required to show that the unit -

licensing basis safety analysis acceptance criteria are not exceeded.

The RTB position switches that provide input to t.he P-4 interlock only function to energize or de-energize or open or close contacts.

Therefore, this Function has no adjustable trip _

setpoint with which to associate a Trip Setpoint and Allowable Value.

This Function :mr t be OPERAELE in MODES 1, 2, and 3 when the reactor may be critical: or -

approaching criticality. This Function does not have to be OPERABLE in MODE 4, 5, or 6 because the main turbine, the MFW System, and the Steam Dump System are not in operation.

b. Encineered Safety Feature Actuation System Interlocks-Pressurizer Pressure. P-11 The P-11 interlock permits a nonnal unit cooldown and depressurization without actuation of SI or main steam line isolation. With two-out-of-three pressurizer pressure channels (discussed previously) less than the P-11 setpoint, the operator Pressure-Low can manually

nd block

c, Lin: the Pr:;;r Pressurizer".5I313Df

-La 0:

- Staam and the Steam Line Pressure-Low steam line isolation signal (previously discussed).

(continued)

McGuire Unit 2 B 3.3-87 S/20/97

ESFAS Instrumentation B 3.3.2 BASES APPLICABLF. b. Enoineered_ Safety Featur_e_ Actuation System SAFETY ANALYSES, LCO, and Interlocks-Pressurizer Pressure. P-11 (continued)

APPLICABILITY When the Steam Line Pressure-Low steam line isolation signal is manually blocked, a main steam isolation signal on Steam Line Pressure-Negative Rate-High is enabled. This provides protection for an SLB by closure of the MSIVs.

With two-out-of-three pressurizer pressure channels above the P-11 setpoint, the Pressurizer / Gd si h Pressure-Low erd Steem Line Pre 55ere-Lew- SI /

ign:6and the Steam Line Pressure-Low steam line isolation signal are automatically enabled.

The operator can also enable these trips by use of the respective manual reset buttons. When the Steam Line Pressure-Low steam line isolation signal is enabled, the main steam isolation on Steaa Line Pressure-Negative Rate-High is disabled.

This Function must be OPERABLE in MODES 1, 2, and 3 to allow an orderly cooldown and depressurization of the unit without the -

actuation of SI or main steam isolation. This Function does not have to be OPERABLE in MODE 4, 5, or 6 because system pressure must already be below the P-11 setpoint for the requirements of the heatup and cooldown curves to be met.

c. Enaineered Safety Feature Actuation System Interlocks-T yg-Low Low. P-12 On increasing reactor coolant temperature, the P-12 interlock provides an anning signal to the Steam Dump System. On a decreasing temperature, the P-12 interlock removes the arming signal to the Steam Dump System to prevent an excessive cooldown of the RCS due to a malfunctioning Steam Dump System.

i Since temperat T,ydre,s thisused as an Function indication meets of bulk RCS redundancy requirements with one OPERABLE channel in each loop. These channels are used in two-out-of-four logic.

(continued)

McGuire Unit 2 B 3.3-88 5/20/97 1

3

Attachment til Technical Justification Rerr. oval of Steam Line Low Pressure Safety injection A. Summary of Safety injection Function The safety injection function is one of the engineered safety features (ESF) of McGuire Nuclear station. The occurrence of a limiting fault, such as a loss of coolant accident or a steam break, requires a reactor trip plus actuation of one or more of the Engineerod Safety Features in order to prevent or mitigate damage to the core and Reactor Coolant System components, and insure Containment integrity, in order to accomplish these design objectives the Engineered Safety Features system has proper and timely initiating signals which are supplied by the sensors, transmitters and logic components making up the various instrumentation channels of the Engineered Safety Features Actuation System.

Safety injection involves the portion of the ESF related to the automatic actuation logic sssociated with pressurizer pressure low-low (1845 psig), containment pressure high ( 1.1 psig) and steam line pressure low (775 psig). Safety injection also includes a manual actuation function. The actuation logic causes a reactor trip, feedwater isolation, component cooling water safety re-alignment, diesel generator start, nuclear service water safety realignment and safety ro-alignment of the emergency cooling water systems (ECCS). Finally, the safety injection function is fulfilled by ECCS injection of borated FWST water into the reactor coolant system by the high head pumps (NV), the intermediate head pumps (NI) and low head pumps (ND).

The safe;y injection accident mitigation capability with respect to emergency core cooling is described in the IJFSAR 6.3.3.2. The UFSAR description covers large break LOCA small break LOCA, inadvertent steam line opening and steam line rupture. The proposed technical specification will remove the actuation of safety injection for plant conditions resulting in steam line pressure below 775 psig. All other aspects of safety system actuation logic, system actuation and ECCS components alignment and operation are unaffected by this technical specification change.

B. Affect of Change on Design Basis Accidents The low steam line pressure safety inject lon signal is relevant to DBAs that involve a depressurization of the secondary side of the plant. Each DBA was first evaluated to determine if the r3moval of the iow steam pressure cafe'y injection would affect the accident consequences.

The accidents which could be affected by the removal of tne low steam liae pressure safety injection were analyzed and results were compared to the acceptance criteria for those accidents.

Table 1 provides a summary of all of the Chapter 15 accidents with regard to safety injection actuation.

Accidents which cause a secondary side depressurization were identified to determine a need for a specific evaluation. This review ider.tified the steam line break and the feedwater line break as

, requiring analysis. The analysis determined the acceptance criteria of these accidents were still satisfied when evaluated without credit for low steam line pressure safety injection. The analysis was true for both the old and replacement steam generators.

• The steam line break accident was analyzed to demonstrate short term cooling capability. A spectrum of break sizes were evaluated to determine the limiting break size. For smaller breaks (including the limiting break size), the safety injection actuation on low pressurizer pressure occurs prior to low steam line pressure safety injection. However, for larger steam line breaks the setpoint for low steam line pressure safety injection is reached prior to low pressurizer pressure safety injection. The larger spectrum of breaks were analyzed without credit for the low steam line pressure safety injection. The results of this analysie found that there would be a slight increase l

1

~ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

y ,

l in time required for safety injection to actuate. The low pressurizer safety injection would actuate in these accidents due to the cooldown and depressurization of the reactor coolant system in response to the secondary side energy removal. The Departure from Nucleate. Boiling Ratios (DNBRs) were analyzed with this time delay in safety injection. The DNBRs for these cases were

- found to be less limiting than those calculated for the limiting break size. Therefore, the removal of steam line low pressure safety injection does not adversely affect the DNBR, fuel failure or dose consequences of the main steam line break accident. Other acceptance criteria would not be expected to be affected by the small change in timing of the safety injection signal.

In addition, to the Chapter 15 accident analysis, the Chapter 6 containment response to mass and energy releases was evaluated without credit for steam line low pressure safety injection.

The evaluation demonstrated that for steam line breaks inside of cantainment, the high containment pressure safety injection set point is reached prior to the pressure associated with steam line low pressure safety injection.- Therefore the existing containment response evaluation is not adversely affected by the removal of the low steam pressure safety injection. This also assures that the existing environmental qualification envelope for McGuire is not affected by this change. For steam line breaks outside of containment the maximum required breaksize is 1.0 ft*,

which results in transients with safety injection caused by low pressurizer pressure prior to low steam line pressure safety injection.

The feedwater line break accidents were analyzed to demonstrate long term core cooling capability. During a feedwater line break, the secondary system will depressurize if the break occurs between the main feedwater check valve and the steam generator. However, breaks are only required to occur at the terminal ends of feedwater piping (i.e. at the feedwater pump or at the steam generator). For a feodwater line break at the main feedwater pump, the main feed check valve will prevent depressurization of the steam generator. For a feedwater line break at the steam generator, a safety injection on high containment pressure will occur prior to safety injection on steam pressure. Therefore, the elimination of the steam line low pressure safety injection does not adversely impact the feedwater line break accident.

In summary, a review was cenducted of all design basis accidents to identify those which result in a low steam pressure safety injection. Affected accidents were evaluated to verify that the accident analysis were within acceptance criteria. This review revealed that accident analysis results were within current analysis acceptance criteria.

C. Affect of Change on GDC,10 CFR 50.46 and 10 CFR 50 Appendix K General Design Criterion (17,21,22,35), the associated sections of the UFSAR, and CFR 50.46 and 10 CFR 50 Appendix K were reviewed to ensure continued compliance with relevant regulatory requirements. A brief summary of the impact of this change on these regulatory requirements is summarized in this section.

The ECCS will continue to have sufficient capacity and capability to assure that specified acceptable fuel design limits and the design condhions of the reactor coolara pressure boundary are not exceeded and that the core is cooled during anticipated operational occurrences and accident cond;tions. The review of DBAs and analysis of affected accidents with respect to ECCS performance and core cooling demor strate this capability.

Attemate sources of electric power for ECCS and the ability to withstand a single failure are unaff 9cted by the removal of the low steam pressure safety injection. The ability of the ECCS to cool the core in the event of a failure of any single active component needed for safety injection following an accident is unaffected bf the removal of steam line low pressure safety injection. The remaining safety injection signals have redundant trains for actuation. The elimination of the low steam pressure safety injection signal does not affect any design aspects of the manual, high

containment pressure or low pressurizer pressure safety injection. Safety injection and ECCS single failure tolerance and power requirements are unaffected by this change.

The combined reactivity control system capabikty associated with ECCS is unaffected by the change. The primary mode of actuation for the ECCS remains automatic for all DBAs. The actuation is initiated by signals of suitable diversity and redundancy in that safety injection can be initiated by high containment pressure or low pressurizer pressure. Manual actuation remains '

completely unaffected by tb9 removal of the low steam line pressure safety injection signal. There is no effect on the available source of water from the FWST ar d diversion of flow from ECCS is unaffected by the removal of the low pressure steam line safety injection. The ECCS injection lines provide the same isolation provisions at the interface with the reactor coolant system. There is no change in the number or type of valves used to form the interface between low pressure portions of the ECCS and the reactor coolar,t system providing the same assurance that the ECCS will not be subjected to a pressure greater than its design pressure. There is no affect on the automatic or remote manual valve controls and the same interlocks are provided, with the exception of those required for steam line low pressure.

10 CFR Part 50, Sec. 50.46, and Appendix K to 10 CFR Part 50 were reviewed as they relate to the ECCS being designed so that its cooling performance is in accordance with an acceptable evaluation modol. In particular the acceptance criteria with regard to peak cladding temperature, maximum calculated cladding oxidation, maximum hydrogen generation, coolable core geometry and long term cooling. The accident analysis and evaluation described above demonstrates the ability of the ECCS to meet these regulatory requirements without credit for steam line pressure safety injection.

McGuire Nuclear Station will remain in compliance with General Design Criteria and 10 CFR 50 following the requested removal of the low steam line pressure safety injection actuation logic, s

D. Benefits The removal of a safety injection function reduces the likelihood of spurious safety injections. A spurious safety injection has several consequences with regard to plant operation. Safety injections constitute a challenge to plant safety systems and can cause plant operating transients.

in addition, a spurious safety injection will result in thermal cycles and present reactor operators with challenging operating conditions. An analysis of safety injections at Westinghouse plants indicate that approximately a 15% reduction in spurious safety injections will result from the elimination of secondary side safety injection. These benefits should be directly applicable to McGuire Nuclear Station.

E. UFSAR Revisions Required The low steam line pressure safety injection is discussed in the following sections of the McC uire UFSAR:

3.1,6.3.2.2 2,7.3.2.4.2. Table 7 7,15.1.4.1,15.1.5.1,15.1.5.2 and 15.2.8.1 These sections will be revised to reflect the removal of this safety injection signal following approval of the amendment by the Nuclear Regulatory Commission.

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[ -TABLE 1

SUMMARY

OF SI ACTUATION FOR ALL LICENSING BASIS TRANSIENTS

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UFSAR SI Actuating Signal l- Transient Section Actuation (ifapplicable)

Increase in Feedwater Flow 15.1.2 No Increase in Steam Flow 15.1.3- No Steam Line Break 15.1.5 Yes Low pressurizer pressure Turbine Trip i5.2.3 - No Loss of AC Power 15.2.6 No Feedwater Line Break 15.2.8- Yes High containment pressure Partial Loss of Forced Reactor Coolant 15.3.1- No Flow L Complete Loss of Forced Reactor 15.3.2 No Coolant Flow Locked Rotor - 15.3.3 No ,

Uncontrolled Bank Withdrawal from 15.4.1 No y Suberitical Uncontrolled Bank Withdrawal at Power 15.4.2 No Dropped Rod 15.4.3 a No Single Uncontrolled Rod Withdrawal 15.4.3 d No Startup of an Inactive Reactor Coolant 15.4.4 No Pump at an Incorrect Temperature

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Boron Dilution 15.4.6 Yes Boration provided by manual -

operator action Inadvertent Loading a .d Operation of a _15.4.7 No Fuel Assembly in an Impropor Position Rod Ejection -15.4.8 No -

Inadvertent ECCS Operation 15.5.1 Yes Initiating event is inadvertent SI actuation -

Inadvertent Opening of a Pressurizer 15.6.1 Yes Low pressurizer pressure Safety or Relief Valve Instrument Line Break 15.6.2 No Steam Generator Tube Rupture 15.6.3 Yes None for overfill and short-term core cooling evaluations; Manual operator action for dose evaluation Loss of Coolant Accident - Peak Clad 15.6.5' Yes Low pressurizer presure or high Temperature containment pressure -

Loss of Coolant Accident - Mass and 6.2.1.3 Yes Low pressurizer pressure or high Energy Release containment pressure Steam Line Break Inside Containment - 6.2.1.4 Yes High containment pressure Mass and Energy Release

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Steam Line Break Outside Containment - 3.11- Yes Low pressurizer pressure Mass and Energy Release

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Attachment IV No Significant Haza.'ds Evaluation as per 10 CFR 50.92 and Environmental Impact Will removal of the low steam pressure safety injection,

1. Involve a significant increase in the probability or consequences of an accident previously evaluated?

Answer Probability Accident initiators can affect the probability of a previously evaluated accident. The addition of a new device or piece of equipment to the plant may introduce a new accident initiator. No new equipment is added to the plant as a result of this change. The proposed removal of the low steam line steam pressure will involve removing the steam line pressure safety injection function.

This results in a reduction in the likelihood of spurious safety injections. Spurious safety injections can result in inadvertent ECCS actuations. Inadvertent ECCS Actuation is a UFSAR accident (UFSAR 15.5.1). Therefore, this change vill result in a reduction in the probability of an accident previously evaluatec' Routine plant operating practices and conditions will not be altered by the removal of the safety injection function. Therefore, there is no operating practice or condition change that could increase the probat:ility of occurrence of a previously evaluated accident.

There is no significant increase in the probability of an accident previously evaluated.

Consequences Accide(.ts previously evaluated that could be adversely affected are the steam line break and the feedwater line break. These accidents will result in secondary side depressurization with pressure reaching the current actuation setpoint. The review of these accidents found that the consequences of the previous accident analysis acceptance criteria remain satisfied. The specifics of the accident analysis is discussed below.

The steam line break accident was analyzed to demonstrate short term cooling capability. A spectrum of break sizes were evaluated to determine the limiting break size. For smaller breaks (including the limiting break size), the safety injection actuation on low pressurizer pressure

, occurs prior to low steam line pressure safety injection. However, for larger steam line breaks the setpoint for low steam line pressure safety injection is reached prior to low pressurizer pressure safety injection. The larger spectrum of breaks were analyzed without credit for the low steam line pressure safety injection. The results of this analysis found that there would be a slight increase in time required for safety injection to actuate. The low pressurizer safety injection would actuate in these accidents due to the cooldown and depressurization of the reactor coolant system in response to the secondary side energy removal. The Depar*ure from Nucleate Boiling Ratios

. (DNBRs) were analyzed with this time delay in safety injection. The DNBRs for these cases were found to be less limiting than those calculated for the limiting break size. Therefore, the retaoval of stearr line low pressure safety injection does not adversely affect the DNBR, fuel failure or dose consequences of the main steam line break accident. Other acceptance criteria would not be expected to be affected by the small change in timing of the safety injection signal.

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C in addition, to the Chapter 15 accident analysis, the Chapter 6 containment response to mass and energy releases was evaluated without credit for steam line low pressure safety injection.

The evaluation demonstrated that for steam line breaks inside of cuntainment, the high containment pressure safety bjection set point is reached prior to the pressure associated with steam line low pressure safety injection. Therefore the existing containment response evaluation is not adversely affected by the removal of se low steam pressure safety injoction. This also assures that the existing environmental qualification envelope for McGuire is not affected by this change. For sieam line breaks outside of containment the maximum required breaksize is 1.0 ft',

which results in transients with safety injection caused by low pressurizer pressure prior to low steam line pressure safety injection.

The feedwater line break accidents were analyzed to demonstrate long term core cooling capability. During a feedwater line break, the secondary system will depressurize if the break occurs between the mah feedwater check valve and the steam generator. However, breaks are only required to occur at the terminal ends of feedwater piping (i.e. at the feedwater pump or at the steam generator). For a feedwater line break at the main feedwater pump, the main feed check valve will prevent depressurization of the steam generator. For a feedwater line break at the steam generator, a safety injection on high containment pressure will occur prior to safety injection on steam pressure. Therefore, the elimination of the steam line low pressure safety injection does not adversely impact the feedwater I;ne break accident, in summary, a review was conducted of all design basis accidents to identify those which result in a low steam pressure safety injection. These accidents were then evaluated to verify that the accident sinalysis were within acceptance criteria, ihis review revealed that all accident analysis results were within current analysis acceptance criteria.

Therefore, there is no significant increase in the consequences of a previously evaluated accident.

Conclusion Elimination of the low steam line pressure safety injection results in no significant increase in the probability or consequences of an accident previously evaluated.

(OR)

2. Create the possibility of a new or different kind of accident frcm any accident previously evaluated; Answer There is no introduction of new equipment or operating practices that could result in a new operating condition. The plant will continue to operate in the same method with the same complement of equipmerst with the exception of the actuation logic associated with the steam line low pressure safety injection. Therefore, there is no new operating condition that would be expected to generate a new sequence of events which could generate a new or different accident.

There is no new equipment that could interact with other plant structures, systems or components.

The low pressure safety injection equipment is the only plant equipment affected by this change.

There are no new equipment failure modes which might result in a new or different accident.

Affected accidents were evaluated to validate that the accident sequence would not deviate in a fashion which would create a new or different accident. The analysis of the feedwater line break and steam line break did not reveal any new or different type of accident.

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Removal of the low steam line pressure safety Iniection will not create the possibility of a new or different kind of accident from any accident previously evaluated; (OR) .

3. Involve a significant reduction in the margin of safety? '

Answer The margin of safety relevant to this change is represented by the margin of physical protection.

provided by fuel cladding and the reactor containment. Effects of this change on the safety analysis was described under question 1 above. The results of the analysis demonstrate that DNBR, fuel clad integrity and containment response were not significantly affected by the removal of low steam line pressure safety injection. Therefore, the physical protection provide by the fuel

. cladding and reactor containment were not affected by this change. Accident acceptance criteria continued to be met without credit for the safety function. - The radiological consequences of accidents was not affected by the change.

The removal of the low steam line pressure safety injection did not significantly reduce the margin of safety, Statement of Environmentalimpact The removal of the low stearn pressure safety injection will not alter or change the routine operation of the plant and will not affect any radiological or non radiological effluent streams.

Radiological consequences of accidents remai ned unchanged by the removal of the low steam pressure safety injection. Therefore, there was no change in the environmental consequences of these accidents. There will be no change in normal or post accident radiation conditions or ,

amount of time that workers would be exposed to operating or accident radiatiosi doses.

Therefore, there is no affect on cumulative radiation exposure.

This change has negligible environmental impact with regard to 10 CFR 51.22 ( c ) ( 9 ).

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