ML20137Y691

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Proposed Tech Specs,Revising Requirements for Instrumentation Response Time Testing Associated W/Rps, Isolation Actuation Sys & ECCS
ML20137Y691
Person / Time
Site: Brunswick  Duke Energy icon.png
Issue date: 04/11/1997
From:
CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML20137Y682 List:
References
NUDOCS 9704230181
Download: ML20137Y691 (45)
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Text

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i-j ENCLOSURE 3

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d I. BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 i NRC DOCKET NOS. 50-325 AND 50-324 OPERATING LICENSE NOS. DPR-71 AND DPR-62

) SUPPLEMENT TO REQUEST FOR EMERGENCY /EXtGENT LICENSE AMENDMENTS l INSTRUMENTATION RESPONSE TIME TESTING j .

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TYPED TECHNICAL SPECIFICATION PAGES - UNIT 1 1

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9704230181 970411 PDR ADOCK 05000324 p PM

INDEX

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BASES 1

l L SECTION PAGE 3/4.0 ' APPLICABILITY ... ..... .. . B 3/4 0-1

, 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN. .. .. . ... .. . .. .. . . . ..B 3/4 1-1 3/4.1.2 REACTIVITY ANOMALIES.. . ... . . .. . .. .B 3/4 1-1

! 3/4.1.3 CONTROL RODS.. ... . . . . .. .. ....... . .B 3/4 1-1 3/4.1.4 CONTROL ROD PROGRAM CONTROLS. . .. . . . ... . . ..B 3/4 1-3 '

3/4.1.5 STANDBY LIQUID CONTROL SYSTEM. . .. .. . .B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE. .... . .. .B 3/4 2-1 l 3/4.2.2 MINIMUM CRITICAL POWER RATIO.. . .. .B 3/4 2-2 i l l

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION. . . .... .B 3/4 3-1 l 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION. ..... .. .B 3/4 3-2

3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION..B 3/4 3-2  !

l l 3/4.3.4 CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION . B3/4 3-2a '

3/4.3.5 MONITORING INSTRUMENTATION . . .. .. B3/4 3-2a ,

1 3/4.3.6 j ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION... .B 3/4 3-6 3/4.3.7 REACTOR CORE ISOLATION COOLING SYSTEM. ACTUATION-INSTRUMENTATION . . .. . B 3/4 3-6 l

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM.. .. . .. . .B 3/4 4-1 l

3/4.4.2 SAFETY / RELIEF VALVES.. . . .. . .8 3/4 4-1 l 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE. .. . .. . .8 3/4 4-1 l l

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i BRUNSWICK - UNIT 1 X Amendment No. I l

. o j INDEX l

, BASES -

SECTION. PAGE i 3/4.4 REACTOR COOLANT SYSTEM (Continued) l 3/4.4.4 CHEMISTRY.. . .... . ... . . .. . .... .. B 3/4 4-2 3/4.4.5 SPECIFIC ACTIVITY. ..... . ..... .. . ... ... . B 3/4 4-2 l 3/4.4.6 PRESSURE / TEMPERATURE LIMITS.. ... . .. ... B 3/4 4-3 l 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES... .. . . .... B 3/4 4-7 3/4.4.8 STRUCTURAL INTEGRITY. ... . .. .. . .. B 3/4 4-7 3/4.5 EMERGENCY CORE COOLING SYSTEMS j 3/4.5.1 HIGH PRESSURE COOLANT INJECTION SYSTEM. . . .. . .. B 3/4 5-1 3/4.5.2 AUTOMATIC DEPRESSURIZATION SYSTEM (ADS). .. . . . B 3/4 5-lb 3/4.5.3 LOW PRESSURE COOLING SYSTEMS. ... . . ... B 3/4 5-3 3/4.5.4 SUPPRESSION POOL. . . . .. . . .... .. . . . . B 3/4 5-4 I 3/4.6 CONTAINMENT SYSTEMS l 3/4.6.1 PRIMARY CONTAINMENT. .. . . . . B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS. . . . .. . B 3/4 6-3  ;

3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES. .. . . . ... B 3/4 6-4 l 3/4.6.4 VACUUM RELIEF.... . ........ . .. .. ... ... . .. B 3/4 6-5 i 3/4.6.5 SECONDARY CONTAINMENT. .. ... ... . . .... B 3/4 6-5 3/4.6.6 CONTAINMENT ATMOSPHERE CONTROL. . . .. .. B 3/4 6-6 3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS. .. . . . . ... ... B 3/47-1 3/4.7.2 CONTROL ROOM EMERGENCY VENTILATION SYSTEM . . .. B 3/4 7-Ic o.

BRUNSWICK - UNIT 1 XI Amendment No.

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3/4.3 INSTRUMENTATION ,

3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION  !

SURVEILLANCE REQUIREMENTS 1

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4.3.1.1 Each reactor protection system instrumentation channel shall be i demonstrated OPERABLE by the performance of the CHANNEL CHECK. CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1-1.

4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of  !

all channels shall be performed at least once per 18 months and shall include j

] calibration of time delay relays and timers necessary for proper functioning  !

of the trip system.

4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip  !

l function' shall be demonstrated to be within its limit at least once per 18 i months. Each test shall include at least one logic train such that both logic l trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where l N is the total number of redundant channels in a specific reactor trip i function.

1 Neutron detectors are exempt from response time testing. The sensor I response times for the following functions may be assumed to be the j design sensor response time:

l l Item 3. " Reactor Vessel Steam Dome Pressure-High"

( Item 4. " Reactor Vessel Water Level-Low. Level 1" BRUNSWICK - UNIT 1 3/4 3-la Amendment No.

l

)* INSTRUMENTATION l

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SURVEILLANCE REQUIREMENTS (Continued) i 4.3.2.3 The ISOLATION SYSTEM RESPONSE TIME of each isolation function'shall 1 be demonstrated to be within its limit at least once per 18 months.

4 Each test

! shall include at least one logic train such that both logic trains are tested )

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at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of

[ redundant channels in a specific isolation function.

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' Radiation monitors are exempt from response time testing. The sensor response times for the following functions may be assumed to be the design sensor response time:

i Item 1.a.2 " Reactor Vessel Water Level-Low. Level 3" Item 1.c.3 " Main Steam Line Flow-High" Response time testing is not required for the functions noted in Table 4.3.2-1.

1 I BPUNSWICK - UNIT 1 o/4 3-11 Amendment No.

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n TABLE 4.3.2-1 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REOUIREMENTS 5

H CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH

- TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED

1. PRIMARY CONTAINMENT ISOLATION
a. Reactor Vessel Watp'r Level -
1. Low. Level 1 ) i Transmitter: NA(*) NA R( 1. 2. 3 Trip Logic: D 0 0 1. 2. 3
2. Low. Level 3 Transmitter: NA(*) NA R(b) 1. L 3 o Trip Logic: D 0 0 1. 2. 3 s
b. Drywell Pressure - High (9) I Y Transmitter: NA(') NA R(*' 1, 2. 3 E3 Trip Logic: D Q Q 1. 2. 3
c. Main Steam Line
1. (Deleted)
2. Pressure - Low ( I Transmitter: NA(*) NA R(b' 1 Trip Lo ic: D Q Q 1
3. Flow - Hi h Transmi ter: NA(*) NA R(b' 1

, rip Logic: D Q Q 1

d. Main Steam Line Tunnel (9) l Temperature - Hi NA Q R 1. 2. 3
e. Condenser Vacuum ghLow (9)

I g Transmitter: NA( NA R(b) 1. 2(*)

<u Trip Logic: D Q Q 1. 2(

a f. Turbine Building Area (S) g Temperature - High NA Q R 1. 2. 3 l 2

" g. Main Stack Radiation - High) NA Q R 1. 2. 3

h. Reactor Building Exhaust

& Radiation - High D 0 R 1, 2. 3 1

E 5;

c y TABLE 4.3.2-1 (Continued)
ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REOUIREMENTS

" CHANNEL .

OPERATIONAL

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CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REOUIRED

2. SECONDARY CONTAINMENT ISOLATION
a. ReactorBuildingExp, gust Radiation - High D 0 R 1.2.3.5. and(" I
b. Drywell Pressure - High (S' .I Transmitter: NA NA R(*' 1. 2. 3 Trip Logic: D 0 0 1. 2. 3 a

2 c. ReactorVesselp*gterLevel-a Low. Level 2 I a

m Transmitter:

Trip Logic:

NA( NA R(b' 1. 2. 3 D 0 0 1. 2. 3

3. REACTOR WATER CLEANUP SYSTEM ISOLATION
a. A Flow - High (9' NA SA R 1. 2. 3 I
b. Area Temperature - High ( NA SA R 1. 2. 3 i ,
c. Areay,9ntilationATemperature- NA SA R 1. 2. 3 High I g d. SLCS Initiation (S) NA .R NA 1. 2 l

@ e. ReactorVesselp9gterLevel- '

& Low. Level 2 I

@ Transmitter: NA(*) NA R(b' 1. 2. 3  !

" Trip Logic: 0 0 0 1. 2. 3 z i-

f. A Flow - High - Time Delay (9' NA SA- R 1. 2. 3 I
g. PipingOutsideRWCUJoomsArea NA SA R 1. 2. 3 Temperature - High l .,

__ _ _ _ _ _ _ _ _ _ _ . _ - - - -__----____-__-__-_--______-__________-_J

to r

y TABLE 4.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REOUIREMENTS 5
  • CHANNEL OPERATIONAL

~ CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH i TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REOUIRED 1

4. CORE STANDBY COOLING SYSTEMS ISOLATION i
a. High Pressure Coolant Injection System Isolation
1. HPCI Steam Line Flow - High 'S' i I

Transmitter: NA NA R'b' 1. 2. 3 Trip Logic: D 0 0 1. 2. 3  :

o 2. HPCI Steam Line High Flow ) l 2 Time Delay Relay NA R R 1. 2. 3 a

A 3. HPCISp,9amSupplyPressure- NA 0 R 1. 2. 3 i e Low I  !

4. HPCI Steam Line Tunne,1,' I  ;

Temperature - High NA SA R 1, 2. 3 l

Bus Power Monitor 'S' S. NA R NA 1. 2. 3 4

6. HPCI Turbine Exhaust I t

Diaphragm Pressure .High (S' NA 0 0 1. 2. 3

7. HPCISteamLineAmbien,} l y

m Temperature - High NA SA R 1. 2. 3  ;

a 8. HPCI Steam Line Area 1  !

j A Temperature - High (S) NA SA R 1. 2. 3 e

2 9. HPCI Equipment Area l I o Temperature - High 'S' NA SA R 1. 2. 3 1

10. Drywell Pressure - High (S' Transmitter: NA) NA R'b) 1, 2. 3 Trip Logic: D 0 0 1. 2. 3 ,

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_ ._....___.__.._...-.__.__--._-______-.-.____,.-__..___,.v -

- , , , .w,- _ ,. u.,.._-...

TABLE 4.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS -

en E CHANNEL OPERATIONAL 5 CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH C0 ST BY COOLING SYSTEMS ISOLATION (Continued)

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b. Reactor Core Isolation Cooling System Isolation

- 1. RCIC Steam Line Flow - High (S' Transmitter: NA(*) NA R(b) 1. 2. 3 Trip Logic: D Q Q 1. 2, 3

2. RCIC Steam Line Flow - High (S) l Time Delay Relay NA R R 1. 2. ' 3
3. RCIC St9am Supply Pressure - NA 0 0 1. 2. 3 Low 'S l o 4. RCICSteamLineTung,91 g Temperature High NA SA R 1. 2. 3
5. Bus Power Monitor (S' NA R NA 1. 2. 3 1
6. RCICTurbineExhaupjDiaphragm Pressure - High NA 0 R 1. 2. 3 1
7. RCICSteamLineAmbieng Temperature - High NA SA R 1. 2. 3 1
8. RCIC Steam Line Area A Temperature - High (S' NA SA R 1. 2. 3 l
9. RCICEquipmentRoomAmyient Temperature - High " NA SA R 1. 2. 3 I E

3 10. RCIC Equipment Room

@ A Temperature - High (S) NA SA R 1. 2. 3 l-

<u A 11. RCIC Steam Line Tunn,9 1 Temperature - High g lime Delay Relay ' NA SA R 1. 2. 3 l

12. Drywell Pressure - High (S' 1 Transmitter: NA(*) NA R(b' 1. L 3 Trip Logic: D Q Q 1. 2. 3 .

Si .

E y TABLE 4.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS E

q CHANNEL OPERATIONAL

,_. CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED

5. SHUTDOWN COOLING SYSTEM ISOLATION 1
a. ReactorVesselWgterLevel-  !

Low. Level 1 (9 i Transmitter: NA) NA R.'b) LL3 Trip Logic: D Q Q 1. 2. 3 a b. Reactor Steam Dome Pressure - High (S) NA 0 R 1. 2. 3 1 2

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. _.__.____________________.______.__.___m____________._ __.___._._________.__.___._____.___._______m.___J

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TABLE 4.3.2-1 (Continued)

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ISOLATION' ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS NOTES (a) The transmitter channel check is satisfied by the trip unit channel check. A separate transmitter check is not required l (b) Transmitters are exempted from the quarterly channel calibration.

(c) Deleted.

(d) Deleted.

(e) When reactor steam pressure 2 500 psig.

l (f)- When handling irradiated fuel in the secondary containment.

(c) Response time testing of the function is not required. l l

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BRUNSWICK - UNIT 1 3/4 3-32 Amendment No. l l

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INSTRUMENTATION

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3/4 3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3 3.3 The Emergency Core Cooling System (ECCS) actuation instrumentation channels shown in Table 3.3.3-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table  ;

3.3.3-2. -

APPLICABILITY: As shown in Table 3.3.3-1.

ACTION:

a. With an ECCS actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of l Table 3.3.3-2. declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value
b. With one or more ECCS actuation instrumentation channels inoperable, take the ACTION required by Table 3.3.3-1.
c. The provisions of Specification 3.0.3 are not applicable in OPERATIONAL CONDITION 5.

SURVEILLANCE REQUIREMENTS 4.3.3.1 Each ECCS actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK. CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.3-1.

4.3.3.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months and shall include calibration of time delay relays and timers necessary for proper functioning of the trip system.

4.3.3.3 Deleted.

BRUNSWICK - UNIT 1 3/4 3-33 Amendment No. I

Et1ERGENCY CORE COOLING SYSTEMS

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SURVEILLANCE REQUIREMENTS (Continued)

2. Verifying that each valve (manual power-operated, or automatic)

-in the flow path that is not locked sealed. or otherwise secured in position. is in its correct position.

b. At least once per 92 days, by verifying that the system develops a flow of at least 4250 gpm for a system head corresponding to a reactor pressure 2 1025 psig when steam is being supplied to the turbine at 1025. +20. -80. psig.
c. At least once per 18 months by:
1. Performing a system functional test which includes simulated automatic actuation of the system throughout its emergency o]erating sequence and verifying that each automatic valve in t1e flow path actuates to its correct position. Actual injection of coolant into the reactor vessel is excluded from this test.
2. Verifying that the system develops a flow of at least 4250 gpm for a system head corresponding to a reactor pressure of 2 165

. psig when steam is being supplied to the turbine at 165. 15.

psig.

3. Verifying that the suction for the HPCI system is automatically transferred from the condensate storage tank to the suppression <

pool on a condensate storage tank low water level signal or l suppression pool high water level signal.

4. Verifying that th withinitslimit.pECCSRESPONSETIMEfortheHPCIsystemis

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' Instrumentation response time may be assumed to be the design instrumentation response time.

BRUNSWICK - UNIT 1 3/4 5-2 Amendment No. I

EMERGENCY CORE COOLING SYSTEMS

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SURVEILLANCE REQUIREMENTS (Continued)

2. Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
c. At least once per 92 days by:
1. Verifying that each CSS pump can be started from the control I room and develops a flow of at lea.st 4625 gpm on recirculation flow against a system head corresponding to a reactor vessel  !

pressure of a 113 psig.

2. Performing a CHANNEL CALIBRATION of the core s] ray header AP instrumentation and verifying the set greater than the normal indicated AP. point to 3e 5, 11.5, psid
d. At least once per 18 months by performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position.

Actual injection of coolant into the reactor vessel is excluded from I this test.  ;

e. At least once per 18 months by verifying the ECCS RESPONSE TIME for each CSS subsystem is within its limit i

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. 1 i ' Instrumentation response time may be assumed to be the design instrumentation response time.

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BRUNSWICK - UNIT 1 3/4 5-6 Amendment No. I

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EMERGENCY CORE COOLING SYSTEMS

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SURVEILLANCE REQUIREMENTS 4.5.3.2 Each-LPCI subsystem shall-be demonstrated OPERABLE:

d. At least once per 31 days by:

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1. Verifying that the system piping from the pump discharge valve to the system isolation valve is filled with water,
2. Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position, and
3. ' Verifying that the subsystem cross-tie valve is closed with power removed from the valve operator.
b. At least once per 92 days by verifying each pair of LPCI pumps cisct.]rging to a common header can be started from the control room ind d3velops a total flow of at least 17,000 gpm against a system tead corresponding to a reactor vessel pressure of a 20 psig.
c. At ' east once per 18 months
  • by performing a system functional test .

u:nch includes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position.

Actual injection of coolant into the reactor vessel is excluded from  !

this test. ,

d. At least once per 18 months by verifying'the ECCS RESPONSE TIME for i each LPCI subsyster,i is within its limit.
  • For the performance of this system functional test scheduled to be completed i by February 25, 1981, a onetime-only exemption is allowed to extend this test  ;

until "before the completion of the Spring 1981 outage," scheduled to commence in March, 1981.

' Instrumentation response time may be assumed to be the design instrumentation response time. ,

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BRUNSWICK - UNIT 1 3/4 5-8 Amendment No. l l

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- l 3/4.3 INSTRUMENTATION BASES I 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to: l a, Preserve the integrity of the fuel cladding

b. Preserve the integrity of the reactor coolant system.
c. Minimize the energy which must be adsorbed following a loss-of-coolant accident, and prevent inadvertent criticality. '

This specification 3rovides the limiting conditions for operation  !

necessary to preserve t1e ability of the system to perform its intended l function even during periods when instrument channels may be out of service j because of maintenance. When necessary, one channel may be made inoperable  ;

for brief intervals to conduct the required surveillance tests. l S ecified surveillance intervals and allowed out-of-service times were {

estab ished based on the reliability analyses documented in GE re] orts NEDC-30851P-A, " Technical Specification Improvement ses for 3WR Reactor Anal.y' Technical Protection System," March 1988 and MDE-81-0485 Rev. 1.

Specification Im]rovement Analysis for the Reactor Protection System for Brunswick Steam Electric Plant, Units 1 and 2." August 1994, as modified by ,

i BWROG-92102, Letter from C. L. Tully (BWROG) to B. K. Grimes (NRC). "BWR l Owners' Group (BWROG) Topical Reports on Technical Specification Improvement l Analysis for BWR Reactor Protection Systems - Use for Relay and Solid State Plants (NEDC-30844 and NEDC-30851P)," November 4, 1992.

The reactor protection system is made up of two independent trip systems.

There are usually four channels to monitor each parameter, with two in each trip system. The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system. The tripping of both tri) systems will produce a reactor scram. The system meets the intent of IEEE-279 for nuclear power plant protection systems.

The measurement of response time at the specified frequencies provides  !

assurance that the 3rotective, isolation. and emergency core cooling functions associated with eac1 channel are completed within the time limit assumed in the accident analysis. No credit was taken for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, overlapping or-total channel test measurements, provided such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) inplace, onsite, or offsite test measurements, or 2) i utilizing replacement sensors with certified response times. l As noted (Note /). neutron detectors are excluded from REACTOR PROTECTION SYSTEM RESPONSE TIME testing because the principles of detector operation virtually ensu e an instantaneous response time. In addition, this note states that the response time of the sensors for Item 3. " Reactor Vessel Steam .

BRUNSWICK - UNIT 1 B 3/4 3-1 Amendment No. 1

. _ . _ . _ . _ _ = _ . . . _ _ _ ___ ___ __. __.

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3/4.3 INSTRUMENTATION I BASES l

3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (Continued) 1 Dome Pressure-High" and Item 4. " Reactor Vessel Water Level-Low. Level 1" may be

)

' assumed in the REACTOR PROTECTION SYSTEM RESPONSE TIME test to be the design {

sensor response time. This is allowed since other surveillance testing (e.g.,

channel calibration) and other techniques ensure detection of response time degradation before performance is significantly affected (Reference 1).

The bases for the trip settings of the reactor protection system are discussed in the bases for Specification 2.2.

REFERENCES:

1. NED0-32291-A, " System Analyses for the Elimination of Selected Response Time 4

Testing Requirements." Octobor 1995.

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BRUNSWICK - UNIT 1 8 3/4 3-la Amendment No. I t + ---

l INSTRUMENTATION BASES 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This s3ecification ensures the effectiveness of the instrumentation used to mitigate t1e consequences of accidents by prescribing the trip settings for isolation of the reactor systems. When necessary. ora channel may be inoperable for brief intervals to conduct required surveilhnce. Some of the trip settings have tolerances explicitly stated where both the high and low values are critical and may have a substanti 1 effect on safety. The setpoints of other instrumentation where oniy the high or low end of the setting has a direct bearing on the safety. are established at a level away from the normal operating range to prevent inadvertent actuation of the systems involved.

Specified surveillance intervals and allowed out-of-service times were established based on the reliability analyses documented in GE reports  !

NEDC-30851P-A, Supplement 2, " Technical S3ecification Improvement Analysis for  !

BWR Isolation Instrumentation Common to R)S and ECCS Instrumentation," March 1989 l and NEDC-31677P-A. " Technical Specification Improvement Analysis for SWR  !

I, solation Actuation Instrumentation," July 1990. as modified by OG90-579-32A, i Letter to Millard L. Wohl (NRC) from W. P. Sullivan and J. F. Klapproth (GE).

" Implementation Enhancements to Technical Specification Changes Given in Isolation Actuation Instrumentation Analysis," June 25. 1990 and supplemented by GE letter report GENE-A31-00001-02, " Assessment of Brunswick Nuclear Plant Isolation Actuation Instrumentation Against NEDC-31677P-A Bounding Analyses,"

August 1994.

As noted (Note #), radiation monitors are excluded from ISOLATION SYSTEM RESPONSE TIME testing because the principles of detector operation virtually ensure an instantaneous response time. in addition, this note states that the response time of the sensors for Item 1.a.2. " Reactor Vessel Water. Level-Low, Level 3": and Item 1.c.3. " Main Steam Line Flow-High" may be assumed in the ISOLATION SYSTEM RESPONSE TIME test to be the design sensor res3onse time. This is allowed since other surveillance testing (e.g. , channel cali] ration) and other techniques ensure detection of response time degradation before performance is significantly affected (Reference 1).

REFERENCES:

1. NED0-32?91-A. " System Analyses for the Elimination of Selected Response Time Testing Requirements." October 1995.

3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the operator's ability to control. This specification provides the tri) point settings that will ensure effectiveness of the systems to provide t1e design protection. Although the instruments are listed by system in some cases the same instrument is used to send the start signal to several systems at the same t

l time. The out-of-service times for the instruments are consistent with the requirements of the specifications in Section 3/4.5.

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BRUNSWICK - UNIT 1 B 3/4 3-2 Amendment No. 1

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INSTRUMENTATION BASES 3/4 3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION dontinu l Specified surveillance intervals and allowed out-of-service times were established based on the reliability analyses documented in GE reports NEDC-30936P-A Parts 1 and 2. "BWR Owners' Group Technical Specification Improvement Methodology (With Demonstration for BWR ECCS Actuation Instrumentation)." December 1988 and RE-011. Rev.1. " Technical Specification Improvement Analysis for the Emergency Core Cooling System Actuation Instrumentation for Brunswick Steam Electric Plant. Units 1 & 2." August 1994, as modified by 0G90-319-320. letter from W. P. Sullivan and J. F. Klaaproth (GE) to Millard L. Wohl (NRC). " Clarification of Technical Specification C1anges Given in ECCS Actuation Instrumentation Analysis." March 22, 1990.

3/4.3.4 CONTROL R00 WITH0RAWAL BLOCK INSTRUMENTATION The control rod block functions are provided consistent with the requirements of the specifications in Section 3/4.1.4. Rod Program Controls, and Section 3/4.2. Power Distribution Limits. The trip logic is arranged so that a trip in any one of t'le inputs will result in a rod block.

Specifiec surveillance intervals and allowed out-of-service times were established based on the reliability analyses documented in GE report f

NEDC-30851P-A. Supplement 1. " Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation." October 1988.

1/4.3 5 MONITORING INSTRUMENTATION l 3/4.3.5.1 H!SMIC MONITORING INSTRUMENTATION The OPEPABILITY of the seismic monitorin instrumentation ensures that sufficient capability is available to prompt y determine the magnitude of a seismic event and evaluate the response of t ose features important to safety. i This capability is required to permit comparison of the measured response to that used in the design basis for the facility.

p BRUNSWICK - UNIT 1 B 3/4 3-2a Amendment No.

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

l 3/4.5 EMERGENCY CORE COOLING SYSTEM i

l BASES 3/4.5.1 HIGHPRESSURECOOLANTINJECTIONSYSTEM(Continued)

ACTIONS:

With the HPCI system inoperable. adequaM core cooling is assured by the demonstrated OPERABILITY of the redundant ; iiversified Automatic i

Depressurization system and the low pressur. Joling systems. In addition, the Reactor Core Isolation Coonng (RCIC) system, a system for which no credit is taken in the safety analysis, will automatically provide makeup at reactor pressures on a reactor low water level condition. The out-of-service period of 14 days is based on the demonstrated operability cf redundant and diversified low pressure core cooling systems.

SURVEILLANCE REQUIREMENTS:

The surveillance requirements provide adequate assurance that the HPCI will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation during reactor operation, a complete functional test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage and to provide cooling at the earliest moment.

Surveillance Requirement 4.5.1.c.4 ensures that the ECCS RESPONSE TIME for the HPCI system is less than or equal to the acceptance criteria included in Reference 6. This surveillance requirement is radified by a note that allows the instrumentation portion of the response time to be assumed to be the design instrumentation response time. Therefore, the instrumentation response time is excluded from the ECCS RESPONSE TIME testing. This exception is allowed since other surveillance testing (e.g., channel calibration) and other techniques ensure detection of response time degradation before performance is significantly affected (Reference 7).

REFERENClS:

1. Brunswick Steam Electric Plant Updated FSAR. Section 6.3.2.2.1.
2. Brunswick Steam Electric Plant Updated FSAR. Section 15.1.3.
3. Brunswick Steam Electric Plant Updated FSAR. Section 15.2.5.
4. Brunswick Steam Electric Plant Updated FSAR. Section 15.2.6.
5. Brunswick Steam Electric Plant Updated FSAR. Section 15.5.2.
6. Updated Final Safety Analysis Report. Section 6.3.3.7.
7. NEDO-32291-A. " System Analyses for the Elimination of Selected Response Time Testing Requirements." October 1995.

BRUNSWICK - UNIT 1 B 3/4 5-la Amendment No. I

l 3/4.5 EMERGENCY CORE COOLING SYSTEM i

BASES 3/4.5.2 AUTOMATIC DEPRESSURIZATION SYSTEM (ADS)

Upon failure of the HPCIS to function properly after a small break loss-of-coolant accident, the ADS automatically causes the safety-relief valves to open, depressurizing the reactor so that flow from the low pressure cooling system can enter the core in time to limit fuel cladding temperature to less than 2200 F. ADS is conservatively required to be OPERABLE whenever reactor vessel pressure exceeds 113 psig even though low pressure cooling systems l

provide adequate core cooling up to 150 psig. ~

1 l

1 i

BRUNSWICK - UNIT 1 B 3/4 5-lb Amendment No. l

_7_

~

-EMERGENCY CORE COOLING SYSTEMS BASES l

CORE SPRAY SYSTEM (Continued)

When in CONDITION 4 or 5 with neither CSS loop OPERABLE, prohibition of all operations which have a potential for draining the reactor vessel minimizes the ]robability of emergency core cooling being required. The required OPERA 31LITY of at least one LPCI loop, or requiring tie reactor vessel to be flooded with the fuel pool gates removed, provides assurance of adequate core flooding, and the restrictions on operations are not applicable.

The surveillance requirements provide adequate assurance that the CSS will -

be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation during reactor operation, a i complete functional test requires reactor shutdown. The pump discharge piping is c.aintained full to preve nt water hammer damage to piping and to start cooling at the earliest mome,t.

Strveillance Requirement 4.5.3.1.e ensures that the ECCS RESPONSE TIME for each are spray system subsystem is less than or equal to the acceptance criteria included in Reference 1. This surveillance requirement is modified by a note that allows the instrumentation portion of the response time to be assumed to be the design instrumentation response time. Therefore, the instrumentation response time is excluded from the ECCS RESPONSE TIME testing.

This exception is allowed since other surveillance testing (e.g., channel calibration) and other techniques ensure detection of response time degradation before performance is significantly affected (Referer.ce 1).

l

REFERENCES:

l 1. Updated Final Safety Analysis Report, Section 6.3.3.7.

2. NEDO-32291-A, " System Analyses for the Elimination of Selected Response l l Time Testing Requirements," October 1995.  :

l l 3/4.5.3.2 LOW PRESSURE COOLANT INJECTION SYSTEM (LPCIS)

The LPCIS is provided to assure that the core is adequately cooled following a loss-of-coolant accident. Two loops each with two pumps 3rovide adequate core flooding for all break sizes from 0.2 ft' up to and inc~ uding the double-ended reactor recirculation line break, and for small breaks following depressurization by the ADS.

The LPCIS specifications are applicable during CONDITIONS 1, 2. and 3 I because LPCIS is a primary source of water for flooding the core after the i i reactor vessel is depressurized.

When in CONDITION 1. 2, or 3 with one LPCIS pum) inoperable, or one LPCIS loo) inoperable, adequate core flooding is assured )y the demonstrated OPEMBILITY of the redundant LPCIS pumps or loop, and both CSS loops. The reduced redundancy justifies the specified 7-day out-of-service period.

BRUNSWICK - UNIT 1 B 3/4 5-3 Amendment No. I

^

EMERGENCY CORE COOLING SYSTEMS I ~

BASES l i 3/4.5.3.2 LOW PRESSURE COOLANT INJECTION SYSTEM (LPCIS) (Continued)

The surveillance requirements provide adequate assurance that LPCI will be l
OPERABLE when required. Although all active components are testable and full 1 flow can be demonstrated by recirculation during reactor operation, a complete 4

functional test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling '

at the earliest moment.

E l Surveillance Requirement 4.5.3.2.d ensures that the ECCS RESPONSE TIME for

, each low pressure coolant injection system subsystem is less than or eaual to the acceptance criteria included in Reference 1. This surveillance reqairement is modified by a note that allows the instrumentation portion of

the response time to be assumed to be the design instrumentation response time. Therefore, the instrumentation response time is excluded from the ECCS )

RES90NSE TIME testing. This exception is allowed since other surveillance '

. testing (e.g., channel calibration) and other techniques ensure detection of response time degradation before performance is significantly affected (Reference 2).

REFERENCES:

1. Updated Final Safety Analysis Report. Section 6.3.3.7.
2. NED0-32291-A, " System Analyses for the Elimination of Selected Response Time Testing Requirements." October 1995.

4 a

l 1

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BRUNSWICK - UNIT 1 B 3/4 5 3a Amendment No. I

5 e i

J ENCLOSURE 4 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 NRC DOCKET NOS. 50-325 AND 50-324

} OPERATING LICENSE NOS. DPR-71 AND DPR-62 SUPPLEMENT TO REQUEST FOR EMERGENCY / EXIGENT LICENSE AMENDMENTS INSTRUMENTATION RESPONSE TIME TESTING TYPED TECHNICAL SPECIFICATION PAGES - UNIT 2 t

.,. r-.

INDEX BASES SECTION PAGE 3/4.0 APPLICABILITY .............. . . .. .. . . ... ...B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN... ...... . ... .. . ..... .. .. ...B 3/4 1-1 3/4.1.2 REACTIVITY ANOMALIES.... . .. . .... ... ... .B 3/4 1-1 3/4.1.3 CONTROL RODS.... .. . . .. ... .. .. .... . .B 3/4 1-1 3/4.1.4 CONTROL ROD PROGRAM CONTROLS. . . .. .. ... .B 3/4 1-3 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM. . .... ... .. . .. ..B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.. .. .....B 3/4 2-1 3/4.2.2 MINIMUM CRITICAL POWER RATIO...... . .. .. . .B 3/4 2-2 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION. ......... ..B 3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION.. ... . . .. .B 3/4 3-2 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION..B 3/4 3-2 i i

3/4.3.4 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION .. . B 3/4 3-2a 3/4.3.5 MONITORING INSTRUMENTATION.... . . .. .. . ... . .B 3/4 3-2a 3/4.3.6 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION. . .8 3/4 3-6 3/4.3.7 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION., . .. . .. . ... ...B 3/4 3-7 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM. . . . , . . .. .B 3/4 4-1 3/4.4.2 SAFETY / RELIEF VALVES... .. . . . . . .. .B 3/4 4-1 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE. . . .,. . .B 3/4 4-1 BRUNSWICK - UNIT 2 X Amendment No.

4 INDEX

~

BASES SECTION A _E.

PAG 3/4.4 REACTOR COOLANT SYSTEM (r utinued) 3/4.4.4 CHEMISTRY. .. . .

.. ... B 3/4 4-2 3/4.4.5 SPECIFIC ACTIVITY. . . . . .. . .... .... . .. B 3/4 4-2 3/4.4.6 PRESSURE / TEMPERATURE LIMITS.. .... .. .. ... . . .. B 3/4 4 3

'3/4.4.7 MAIN STEAM LINE ISOLATION VALVES. . .. . . .. . .. B 3/4 4-7 3/4.4.8 STRUCTURAL INTEGRITY.. .. ... . .. . . .. .... .. B 3/4 4-7 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 HIGH PRESSURE COOLANT INJECTION SYSTEM...... . . . . B 3/4 5-1 3/4.5.2 AUTOMATIC DEPRESSLRIZATION SYSTEM.. .

B3/45-lbl 3/4.5.3 LOW PRESSURE COOLING SYSTEMS.. . .

... . . ...... ... B 3/4 5-3 3/4.5.4 SUPPRESSION POOL. . ... .. .. . . .. . ...... B 3/4 5-4 3/4.6 CONTAINMENT SYSTEMS  !

3/4.6.1 PRIMARY CONTAINMENT. . .. . . . .. .. B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS. . .. .. B 3/ 4 6-3 3/4,6.3 PRIMARY CONTAINMENT ISOLATION VALVES. . .. . .... B 3/4 6-4 i 3/4.6.4 VACUUM RELIEF .. .. . . .. .. .. .. . . . . . . . B 3/4 6-5 3/4.6.5 SECONDARY CONTAINMENT.. . . .. .. . . . .. B 3/4 6-5 i l 3/4.6.6 CONTAINMENT ATMOSPHERE CONTROL. . . . . B 3/4 6-6 l

l 3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS. ... . .. .. . .. . B 3/4 7-1 l 3/4.7.2 CONTROL ROOM EMERGENCY VENTILATION SYS~.EM.. . . .B 3/4 7-Ic l - BRUNSWICK - UNIT 2 XI Amendment No.

.7_ _

INSTRUMENTATION SURVEILLANCE REQUIREMENTS (Continued) 4.3.2.3 The ISOLATION SYSTEM RESPONSE TIME of each isolation function'shall be~ demonstrated to be within its limit at least once per 18 months. Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels ,

are tested at least once every N times 18 months, where N is the total number  !

of redundant channels in a specific isolation function.

l l

i l

Radiation monitors are exempt from response time testing. The sensor response times for the following functions may be assumed to be the design.

sensor response time:

Item 1.a.2, " Reactor Vessel Water Level-Low Level 3" Item 1.c.3, " Main Steam Line Flow-High" Item 1.c.4 " Main Steam Line Flow-High"  !

Response time testing is not required for the functions noted in Table 4.3.2-1, '

I l i

l  !

BRUNSWICK - UNIT 2 3/4 3-11 Amendment No. I i

m *

c y TABLE 4.3.2-1 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REOUIREMENTS E--
  • CHANNEL OPERATIONAL

" CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED

1. PRIMARY CONTAINMENT ISOLATION
a. Reactor Vessel WatpS'r Level -
1. Low. Level 1 I Transmitter: NA'*) NA R'b' 1. 2. 3 Trip Logic: D Q Q 1. 2. 3
2. Low. Level 3 Transmitter: NA) NA R* 1. 2. 3 y Trip Logic: D Q Q 1. 2. 3 Y

N

b. Drywell Pressure - High 'S' i

Transmitter: NA) NA R'b' L 2. 3 Trip Logic: D Q Q 1. 2. 3

c. Main Steam Line
1. (Deleted) 'S'
2. Pressure - Low I

Transmitter: NA) NA R'b) 1 i Trip Logic: 0 0 0 1

3. Flow - High Transmitter: NA NA R'b' 1 Trip Logic: D 0 0 1
4. Flow - High D Q Q 2. 3 3 d. Main Steam Line Tunnel  ;

@ Temperature - High NA Q R 1. 2. 3

& e. Condenser Vacuum - Low (9) l 1

9 Transmitter: NA( NA R(b) 1. 2)

Trip Logic: D Q Q 1. 2

= f.

Turbine Building Area'S' Temperature - High NA Q R 1. 2. 3 1

g. Main Stack Radiation - High h.

NA Q R 1. 2. 3 Reactor Building Exh, gust Radiation - High D Q R 1. 2. 3 l

n _ . . .

di E

c Si z

y TABLE 4.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE RE0VIREMENTS E

" CHANNEL OPERATIONAL

" CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH  :

TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED

2. SECONDARY CONTAINMENT ISOLATION
a. ReactorBuildingExp, gust Radiation - High D Q R 1.2.3.5. and(" I
b. Drywell Pressure - High 'S' l

Transmitter: NA) NA R*' 1. 2. 3 '

Trip Logic: D 0 0 1. 2. 3  !

a 2 c. Reactor Vessel 'S' W4ter Level -

w Low. Level 2 I a

m Transmitter: NA( NA R*) 1. 2. 3 Trip Logic: D Q Q 1. 2. 3

3. REACTOR WATER CLEANUP SYSTEM ISOLATION
a. A Flow - High(9) NA SA R 1. 2. 3 I
b. Area Temperature - High NA SA R 1. 2, 3 l
c. AreaV9ntilationATemperature- NA SA- R 1. 2. 3 i High 'S '

I

d. SLCS Initiation (9) NA R NA 1. 2 1 E

m e. Reactor Vessel 'SWgter Level -

E Low. Level 2 I .

8 Transmitter: NA( NA R*) 1. 2. 3 a Trip Logic: D Q Q 1. 2. 3 5 f. A Flow - High - Time Delay ) NA SA R 1. 2. 3 l

g. PipingOutsideRWCUJoomsArea NA SA R 1. 2. 3 Temperature - High l

J Y ,

y TABLE 4.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REOUIREMENTS 5
  • CHANNEL OPERATIONAL

" CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REOUIRED l

4. CORE STANDBY COOLING SYSTEMS ISOLATION
a. High Pressure Coolant Injection System Isolation
1. HPCI Steam Line Flow - High (0 1 Transmitter: NA) NA R* 1. 2. 3 Trip Logic: D 0 0 1. 2. 3 i

i L

a 2. HPCI Steam Line Flow - High (0 1 2 Time Delay Relay NA R R 1. 2. 3 w

a e

3. HPCIS(9amSupplyPressure- NA Q R 1. 2. 3 Low l 4.

HPCISteamLineTunne)0 Temperature - High NA SA R 1. 2. 3 1

5. Bus Power Monitor (S) NA R NA 1. 2. 3 1
6. HPCI Turbine Exhaust i Diaphragm Pressure - High (* NA Q Q 1. 2. 3 1
7. HPCISteamLineAmbieg}

Temperature - High NA SA R 1. 2. 3 I E 8. HPCI Steam Line Area

[ A Temperature - High (* NA SA R 1. 2. 3 l  !

a

@ 9. HPCI Equipment Area

" Temperature - High (0 2

NA SA R 1. 2. 3 1 t

10. Drywell Pressure - High (O I Transmitter: NA( NA R(b' 1. 2, 3 Trip Logic: D 0 0 1. 2. 3 t

.l H

TABLE 4.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REOUIREMENTS 5 CHANNEL .

OPERATIONAL Q CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH

, TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REOUIRED i E 4. CORE STANDBY COOLING SYSTEMS ISOLATION (Continued)

[ b. Reactor Core Isolation Cooling System Isolation

1. RCIC Steam Line Flow - High (S) l Transmitter: NA(*) NA R) 1. 2. 3 Trip Logic:

D 0 0 1. 2. 3

2. RCIC Steam Line High - Flow Time Delay Relay S) NA R R 1. 2. 3 1 1
3. RCICSp,9amSupplyPressure- NA 0 0 1. 2. 3 Low 1 E' 4. RCIC Steam Line Tunnel

^ Temperature - High S) w NA SA R 1. 2. 3 Q 5. Bus Power Monitor (S) NA R NA 1. 2. 3 1

6. RCICTurbineExhaup,} Diaphragm Pressure - High NA 0 R 1. 2. 3 1 ,
7. RCIC Steam Line Airbien,}

Temperature - High NA SA R 1. 2. 3 l {

8. RCIC Steam Line Area '

A Temperature - High ) NA SA R 1. 2. 3 l l

9. RCICEquipmentRoomAmyient '

g Temperature - High 'S NA SA R 1. 2. 3 1 e

a 10. RCIC Equipment Room g A Temperature - High (S) NA SA R 1. 2. 3 l o '

11.

& RCIC Steam Line Tunnel Temuera-(S) ture - High Time DelayJelay NA SA R 1. 2. 3 I

12. Drywell Pressure - High i

I Transmitter: NA) NA R(b) 1. 2. 3 l Trip Logic: D 0 0 1. 2. 3 1  !

_ _ - _ _ _ _ _ _ _ _ ____ ______ _______-________-_____ _ _-___ -__-_ _ _ _ _ _ _ _ _ _ _ __2

o.

-t E

C" F; TABLE 4.3.2-1 (Continued) 7; ISOLATION ACTUATION INSTRUMENTATION SURVEllLANCE REOUIREMENTS c

5

  • CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED
5. SHUTDOWN COOLING SYSTEM ISOLATION  ;
a. ReactorVesselpSgterLevel-Low. Level 1 1 Transmitter: NA* NA P* 1. 2. 3 Trip Logic: D Q Q 1. 2. 3 I t
b. Reactor,,}teamDomePressure- NA 0 R 1. 2. 3 I i y High l 6

g t i

o i

i 8  :

8 e >

O t

TABLE 4.3.2-1 (Continued)

~

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REOUIREMENTS j NOTES I

i (a) The transmitter channel check is satisfied by the trip unit channel check. 1 j ,A separate transmitter check is not required

-(b) Transmitters are exempted from the quarterly channel calibration.

(c) Deleted.

. (d) Deleted.

l (e) When reactor steam pressure a 500 psig.

l (f) When handling irradiated fuel in the secondary containment.

(g) Response time testing of the function is not required. I J

3 I

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J i

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BRUNSWICK - UNIT 2 3/4 3-32 Amendment No.

., . 1 INSTRUMENTATION

~

3/4 3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3 The Emergency Core Cooling System (ECCS) actuation instrumentation channels shown in Table 3.3.3-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.3-2.

APPLICABILITY: As shown in Table 3.3.3-1.

ACTION:

a. With an ECCS actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.3-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.  ;
b. With one or more ECCS actuation instrumentation channels inoperable, take the ACTION required by Table 3.3.3-1.

l

c. The provisions of Specification 3.0.3 are not applicable in OPERATIONAL CONDITION 5.

SURVEILLANCE REQUIREMENTS 4.3.3.1 Each ECCS actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK. CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations during the OPERATIONAL CONDITIONS and at the l frequencies shown in Table 4.3.3-1. '

4.3.3.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months and shall include calibration of time delay relays and timers necessary for proper functioning of the trip system.

4.3.3.3 Deleted. I l

BRUNSWICK - UNIT 2 3/4 3-33 Amendment No. I

-. . 1

)

EMERGENCY CORE COOLING SYSTEMS

~

SURVEILLANCE REQUIREMENTS (Continued)

2. Verifjmg that each valve (manual, power-operated, or automatic) I in the flow path that is not locked, sealed, or otherwise

. secured in position, is in its correct position.

b. At least once per 92 days, by verifying that the system developc a flow of at least 4250 gpm for a system head correspondinc to a reactor pressure a 1000 psig when steam is being suppliec to the turbine at 1000. +20, -18, psig.
c. At least once per 18 months by:  !
1. Performing a system functional test which includes simulated automatic actuation of the system throughout its emergency o)erating sequence and verifying that each automatic valve in t1e flow path actuates to its correct position. Actual injection of coolant into the reactor vessel is excluded from this test. I
2. Verifying that the system develops a flow of at least 4250 gpm for a system head corresponding to a reactor pressure of a 165 psig when steam is being supplied to the turbine at 165, 15.

ps19

3. Verifying that the suction for the HPCI system is automatically transferred from the condensate storage tank to the suppression pool on a condensate storage tank low water level signal or suppression pool high water level signal.
4. Verifying that th within its limit.p ECCS RESPONSE TIME for the HPCI system is 4

I

' Instrumentation response time may be assumed to be the design instrumentation response time.

BRUNSWICK - UNIT 2 3/4 5-2 Amendment No. I

EMERGENCY CORE COOLING SYSTEMS

~

SURVEILLANCE REQUIREMENTS (Continued)

2. Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

) c. At least once per 92 days by:

1. Verifying that each CSS pump can be started from the control room and develops a flow of at least 4625 gpm on recirculation flow against a system head corresponding to a reactor vessel pressure of a 113 psig.
2. Performing a CHANNEL CALIBRATION of the core s] ray header AP instrumentation and verifying the set De 5, 1.5, psid greater than the normal indicated AP. point to
d. At least once per 18 months by performing a system functional test which includes simulated automatic actuation of the system throughout its emer ency operating sequence and verifying that each automatic valve in he flow path actuates to its correct position.

Actual injection of coolant into the reactor vessel is excluded from this test

e. At least once per 18 months by verifying the ECCS RESPONSE TIME for each CSS subsystem is within its limit i

' Instrumentation response time may be assumed to be the design instrumentation response time, i

BRUNSWICK - UNIT 2 3/4 5-6 Amendment No. I

4 #

o EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.3.2~ Each LPCI subsystem shall be demonstrated OPERABLE:

a. At least once per 31 days by:
1. Verifying that the system Diping from the pump discharge valve i to the system isolation valve is filled with water.
2. Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position. and
3. Verifying that the subsystem cross-tie valve is closed with '

power removed from the valve operator.

b. At least once per 92 days by verifying each pair of LPCI pumps  !

discharging.to a common header can be started from the control room and develops a total flow of at least 17.000 gpm against a system head corresponding to a reactor vessel pressure of a 20 psig.

c. At least once per 18 months by performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position.

Actual injection of coolant into the reactor vessel is excluded from this test.

, d. At least once per 18 months by verif ing,the ECCS RESPONSE TIME for '

each LPCI subsystem is within its li l

' Instrumentation response time may be assumed to be the design instrumentation l response time.

i l

BRUNSWICK - UNIT 2 3/4 5-8 Amendment No. l i

~ --

i 3/4.3 INSTRUMENTATION i

~

i BASES f

3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION 2

~

The reactor protection system automatically initiates a reactor scram to:

a. Preserve the integrity of the fuel cladding.
b. Preserve the integrity of. the reactor coolant system, i c. Minimize the energy wnich must be adsorbed fellowing a loss-of-coolant j accident, and prevent inadvertent criticality.

4 This specification provides the limiting conditions for operation necessary i

to preserve the ability of the system to perform its intended function even  !

1 during periods when instrument channels may be out of service because of i maintenance. When necessary, one channel may be made inoperable for brief l

intervals to conduct the required surveillance tests.

Specified surveillance intervals and allowed out-of-service times were established based on the reliability analyses documented in GE reports i NEDC-30851P-A " Technical Specification Improvement Analyses for BWR Reactor Protection System," March 1988 and MDE-81-0485. Rev. 1. " Technical Specification Improvement Analysis for the Reactor Protection System for i Brunswick Steam Electric Plant. Units 1 and 2." August 1994, as modified by BWROG-92102. Letter from C. L. Tully (BWROG) to B. K. Grimes (NRC). "BWR Owners' Group (BWROG) Topical Reports on Technical Specification Improvement Analysis for BWR Reactor Protection Systems - Use for Relay and Solid State Plants (NEDC-30844 and NEDC-30851P)." November 4. 1992.

The reactor protection system is made up of two independent trip systems, i There are usually four channels to ~,anitor each parameter with two in each trip system. The outauts of the channel in a trip system are combined in a logic so that either clannel will trip that trip system. The tripping of both trip systems will produce a reactor scram. The system meets the intent of IEEE-279 for nuclear power plant protection systems.

.! The measurement of se time at the specified frequencies provides I assurance that the aroi ., isolation. and emergency core cooling functions associated with eac1 channel are com31eted within the time limit assumed in the  !

accident anaiy sis. No credit was ta(en for those channels with response times indicated as clot applicable.

Response tiine may be demonstrated by any series of sequential, overlapping or total channel test measurements, provided such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) inplace, onsite, or offsite test measurements, or 2) utilizing replacement sensors with certified response times. l As noted (Note #). neutron detectors are excluded from REACTOR PROTECTION SYSTEM RESPONSE TIME testing because the principles of detector operation virtually ensure an instantaneous response time. In addition, this note states that the response time of the sensors for Item 3. " Reactor Vessel Steam BRUNSWICK - UNIT 2 B 3/4 3-1 Amendment No. I

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3/4.3 INSTRUMENTATION l BASES i

3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (Continued)

Dome Pressure-High" and Item 4. " Reactor Vessel Water Level-Low. Level 1" may i be assumed in the REACTOR PROTECTION SYSTEM RESPONSE TIME test to be the design i

sensor response time. This is allowed since other surveillance testing (e.g., i channel calibration) and other techniques ensure detection of response time '

degradation before performance is significantly affected (Reference 1).

The bases for the trip settings of the reactor protection system are discussed in the bases for Specification 2.2.

REFERENCES:

1. NEDO-32291-A. " System Analyses for the Elimination of Selected Response Time Testing Requirements, " October 1995.

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INSTRUMENTATION

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BASES 1

3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This s)ecification ensuras-the effectiveness of the instrumentation used to mitigate t1e consequences of accidents by prescribing the trip settings for isolation of the reactor systens. When necessary, one channel ma inoperable for brief intervals to conduct required surveillance. y Some be of the trip settings have tolerances e<plicitly stated where both the high and low values are critical and may have a substantial effect on safety. The setpoints of other instrumentation where only the high or low end of the setting has a direct bearing on the safety, are established at a level away from the normal operating range to prevent inadvertent actuation of the systems involved.

Specified surveillance intervals and allowed out-of-service times were established based on reliability analyses documented in GE reports NEDC-30851P-A Supplement 2. " Technical Specification Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation." March 1989 and NEDC-31677P-A. " Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation." July 1990, a modified by OG90-579-32A, Letter to Millard L. Wohl (NRC) from W. P. Sullivan and J. F. Klapproth (3E).

" Implementation Enhancements to Technical Specification Changes Given in Isolation Actuation Instrumentation Analysis." June 25, 1990 and supplemented by GE letter report GENE-A31-00001-02 " Assessment of Brunswick Nuclear Phnt i Isolation Actuation Instrumentation Against NEDC-31677P-A Bounding Analyses, August 1994.

As noted (Note #) radiation monitors are excluded from ISOLATION SYSTEM RESPONSE TIME testing because the principles of detector operation virtually ensure an instantaneous response time. In addition, thi.s note states that the response time of the sensors for Item 1.a.2. " Reactor Vessel Water Level-Low. l Level 3". Item 1.c.3 " Main Steam Line Flow-High": and Item 1.c.4. " Main Steam Line Flow-High" may be assumed in the ISOLATION SYSTEM RESPONSE TIME test to be the design sensor response time. This is allowed since other surveillance testing (e.g., channel calibration) and other techniques ensure detection of response time degradation before performance is significantly affected (Reference 1).

REFERENCES:

1. NED0-32291-A " System Analyses for the Elimination of Selected Response Time Testing Requirements " October 1995.

3/4 3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling systern actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the operator's ability to control. This specification provides the tri] point settings that will ensure effectiveness of the systems to provide t1e design protection. Although the instruments are listed by system, in some cases the same instrument is used to send the start signal to several systems at the same 1 time. The out-of-service times for the instruments are consistent with the requirements of the specifications in Section 3/4.5.  !

BRUNSWICK - UNIT 2 B 3/4 3-2 Amendment No. I

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Specified surveillance intervals and allowed out-of-service times were established based on the reliability analyses documented in GE reports NEDC-30936P-A. Parts 1 and 2. "BWR Owners' Group Technical Specification Improvement Methodology (With Demonstration for BWR ECCS Actuation Instrumentation)." Deember 1988 and RE-011. Rev.1. " Technical Specification Improvement Analysis for the Emergency Core Cooling System Actuation Instrumentation for Brunswick Steam Electric Plant. Units 1 & 2." August 1994, as modified by OG90-319-320. letter from W. P. Sullivan and J. F. Klapproth (GE) to Millard L. Wohl (NRC). " Clarification of Technical Specification Changes Given in ECCS Actuation Instrumentation Analysis." March 22, 1990.

3/4 3.4 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION The control rod block functions are provided consistent with the requirements of the specifications in Section 3/4.1.4. Rod Program Controls and Section 3/4.2. Power Distribution Limits. The trip logic is arranged so that a trip in any one of the inputs will result in a rod block.

Specified surveillance intervals and allowed out-of-service times were established based on the reliability analyses documented in GE report NEDC-30851P-A, Supplement 1. " Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation." October 1988.

3/4.3.5 MONITORING INSTRUMENTATION 3/4.3.5.1 SEISMIC MONITORING INSTRUMENTATION l

The OPERABILITY of the seismic monitoring instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a i

seismic event and evaluate the response of those features important to sa fety. This capability is required to permit comparison of the measured

! response to that used in the design basis for the facility.

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3/4.5 EMERGENCY CORE COOLING SYSTEM 1 BASES

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3/4.5.1 HIGH PRESSURE COOLANT INJECTION SYSTEM (Contiriued)

ACTIONS:

With the HPCI system inoperable, adequate core cooling is assured by l the demonstrated operability of the redundant and diversified Automatic Depressurization system and low pressure cooling systems. In addition,  !

the Reactor Core Isolation Cooling (RCIC) system. a system for which no credit  :

is taken in the safety analysis, will automatically provide makeup at reactor l pressures on a reactor low water level condition. The out-of-service period of i 14 days is based on the demonstrated operability of redundant and diversified low pressure core cooling systems.

SURVEILLANCE REQUIREMENTS:  !

The surveillance requirements provide adequate assurance that the HPCI system will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation during reactor operation, a complete functional test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage and to  :

provide cooling at the earliest moment.

Surveillance Requirement 4.5.1.c.4 ensures that the ECCS RESPONSE TIME for the HPCI system is less than or equal to the acceptance criteria included in Reference 6. )

This surveillance requirement is modified by a note that allows '

the instrumentation portion of the response time to be assumed to be the design  !

instrumentation response time. Therefore, the instrumentation response time is

{

excluded from the ECCS RESPONSE TIME testing. This exception is allowed since  !

other surveillance testing (e.g., channel calibration) and other techniques l ensure detection of response time degradation before performance is significantly affected (Reference 7).

REFERENCES:

1. Brunswick Steam Electric Plant Updated FSAR, Section 6.3.2.2.1.

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2. Brunswick Steam Electric Plant Updated FSAR. Section 15.1.3.
3. Brunswick Steam Electric Plant Updated FSAR. Section 15.2.5.
4. Brunswick Steam Electric Plant Updated FSAR. Section 15.2.6.
5. Brunswick Steam Electric Plant Updated FSAR. Section 15.5.2.

6, Updated Final Safety Analysis Report. Section 6.3.3.7.

7. NED0-32291-A. " System Analyses for the Elimination of Selected Response Time Testing Requirements." October 1995.

BRUNSWICK - UNIT 2 B 3/4 5-la Amendment No. I

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3/4.5 EMERGEN(,Y CORE COOLING SYSTEM

- l BASES 3/4.5.2 AUTOMATIC DEPRESSURIZATION SYSTEM (ADS)

Upon failure of the HPCIS to function properly after a small break loss-of-coolant accident, the ADS automatically causes the safety-relief valves to open, depressurizing the reactor so that flow frcm the low pressure cooling system can enter the core in time to limit fuel cladding temperature to_less than 2200 F. ADS is conservatively required te be OPERABLE whenever reactor vessel pressure exceeds 113 psig even though law pressure cooling systems provide adequate core cooling up to 150 psig.

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BRUNSWICK - UNIT 2 8 3/4 D"-lb Amendment No. l I

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EMERGENCY CORE C00LINC SYSTEMS

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BASES )

CORE SPRAY SYSTEri (Continued)

When in CONDITION 4 or 5 with neither CSS loop OPERABLE. prohibition of all operations which have a potential for draining the reactor vessel minimizes the probability of emergency core cooling being required. The required OPERABILITY of at least one LPCI loop, or requiring the reactor vessel to be flooded with the fuel pool gates removed, provides assurance of adequate core flooding and <

the restrictions on operations are not applicable.

1 The surveillance requirements provide adequate assurance that the CSS will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation during reactor operation. a complete functional test requires reactor shutdown. The pump discharge piping ,

is maintained full to prevent water hammer damage to piping and to start l cooling at the earliest moment. I Surveillance Requirement 4.5.3.1.e ensures that the ECCS RESPONSE TIME for  !

each core spray system subsystem is less than or equal to the acceptance  !

criteria included ir, Reference 1. This surveillance requirement is modified by 1 a note that allows tihe instrumentation portion of the response time to be l assumed to be the design instrumentation response time. Therefore, the )

instrumentation response time is excluded from the ECCS RESPONSE TIME testing. l This exception is allowed since other surveillance testing (e.g. channel calibration) and other techniques ensure detection of response time degradation before performance is significantly affected (Reference 2). l l

REFERENCES:

1. Updated Final Safety Analysis Report. Section 6.3.3.7.

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2. NEDO-32291-A, " System Analyses for the Elimination of Selected Response l Time Testing Requirements." October 1995.  ;

3/4 5.3.2 LOW PRESSURE COOLANT INJECTION SYSTEM (LPCIS)

The LPCIS is provided to assure that the core is adequately cooled following a loss-of-coolant accident. Two loops each with two pumps provide i adequate core flooding for all break sizes from 0.2 ftz up to and including the double-ended reactor recirculation line break, and for small breaks following dep'essurization by the ADS.

The LPCIS specifications are applicable during CONDITIONS 1. 2. and 3 because LPCIS is a primary source of water for flooding the core after the reactor vessel is depressurized.

When in CONDITION 1. 2. or 3 with one LPCIS pump inoperable, or one LPCIS loop inoperable. adequate core flooding is assured by the demonstrated OPERABILITY of the redundant LPCIS pumps or loop, and both CSS loops. The reduced redundancy justifies the specified 7-day 0.t-of-service period. l BRUNSWICK - UNIT 2 B 3/4 5 ' Amendment No. I I

. e, EMERGENCY CORE COOLING SYSTEMS

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BASES 3/4.5.3.2 LOW PRESSURE COOLANT INJECTION SYSTEM (LPCIS) (Continued)

The surveillance requirements provide adequate assurance that LPCI will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation during reactor operation, a complete functional test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.

I Surveillance Requirement 4.5.3.2.d ensures that the ECCS RESPONSE TIME for i each low pressure coolant injection system subsystem is less than or equal to i the acceptance criteria included in Reference 1. This surveillance requirement j is modified by a note that allows the instrumentation portion of the response i time to be assumed to be the design instrumentation response time. Therefore, the instrumentation response time is excluded from the ECCS RESPONSE TIME testing. This exception is allowed since other surveillance testing (e.g.,

channel calibration) and other techniques ensure detection of response time degradation before performance is significantly affected (Reference 2).

REFERENCES:

1. Undated Final Safety Analysis Repuit. Section 6.3.3.7.
2. NEDO-32291-A, " System Analyses for the Elimination of Selected Response Time Testing Requirements " Jctober 1995.

I BRUNSWICK - UNIT 2 B 3/4 5-3a Amendment No. I

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