Text

Proposed Tech Specs,Revising Definitions for Eccs,Isolation Sys & Reactor Protection Sys Response Times
	ML20137M654
Person / Time
Site:	Brunswick
Issue date:	03/27/1997
From:	; CAROLINA POWER & LIGHT CO.
To:	;
Shared Package
ML20137M645	List: BSEP-97-0125, Application for Amends to Licenses DPR-71 & DPR-62,revising Definitions for ECCS Response Time,Isolation Sys Response Time & Reactor Protection Sys Response Time; ML20137M654;
References
	NUDOCS 9704080147
	Download: ML20137M654 (73)
	v • d • e

ENCLOSURE 5 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 1

REQUEST FOR EMERGENCY / EXIGENT LICENSE AMENDMENTS INSTRUMENTATION RESPONSE TIME TESTING t

MARKED-UP TECHNICAL SPECIFICATION PAGES - UNIT 1 t

I 1

9704080147 970327 PDR ADOCK 05000324 P PDR

3/4.3 INSTRUMENTATION

-3/4.3L1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LSURVEILLANCE REQUIREMENTS

. 4.3.1.1 Each reactor arotection system instrumentation channel-sh'all be demonstrated OPERABLE ay the performance of the CHANNEL CHECK. CHANNEL-CALI8 RATION 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 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

. 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 reactor trip-function.

Neutrondetectorsareexemptfromresponsetimetesting.g ADD INSERT "A" 4

e

I, INSERT A

. The sensor response times for the following functions may be assumed to be the design sensor response time:

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

! tem 4, " Reactor Vessel Water Level - Low, Level 1" I '.

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INSTRUMENTATTON SURVEILLANCE REQUIREMENTS (Continued)

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( 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 I 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.

ADD J N S E RT " B " l l

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BRUNSWICK - UNIT 1 3/4 3-11 Amendment No.

INSERT B The sensor response tir. as 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"

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 I 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 tri consistent with the Trip Setpoint value.p setpoint adjusted

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. _ _

The ECCS RESPONSE TIME of each ECCS function shall be demonstrated to 4.3.3_3 e within [the 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 ECCS function.

- - ~ ~ ~ = - - '

= = _ _ _

_ _ _ =

'De t e h ect ,

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

I EMERGENCY CORE COOLING SYSTEMS i

. SURVEILLANCE REQUIREMENTS (Continued) d

2. Verifying that each valve (manual, power-operated, or l 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 l reactor pressure 21025 psig when steam is being supplied to the !

turbine at 1025, +20, -80, psig.

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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 operating sequence and verifying that each automatic valve in i the flow path actuates to its correct position. Actual ,

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

2. Verifying that the system develops a flow of at least 4250 gpm i for a system head corresponding to a reactor pressure of a 165 i psig when steam is being supplied to the turbine at 165, i 15, j psig. ,

3. Verifying that the suction for the HPCI system is automatically l 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. '

b'D D INGERT"C" BRUNSWICK - UNIT 1 3/4 5-2 Amendment No. l A

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INSERT C - 1 4, Verify,ing that the ECCS RESPONSE TIME for the HPCI system is within its ,

limit.

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' _' Instrumentation response time may be assumed to be the design - !

instrumentation response time. ,

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EMERCENCY CORE COOLING SYSTEMS 4

SURVEILLANCE REQUIREMENTS (Concinued) - .

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

. c. At least once per 92 days by:

I

1. Verifying that each CSS pump can be started from the control l room and develops a flow of at least 4625 spm on recirculation flow against a system head corresponding to a reactor vessel pressure of > 113 psig.

2. Performing a CEANNEL CALIBRATION of the core spray header AP

- instrumentation and verifying the set point to be 5, +1.5, psid l greater than the normal indicated AP.

' 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 this test.

ADD INSERT "D" r l

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INSERT D 1

e. At least once per 18 months by verifying the ECCS RESPONSE TIME for each ]

CSS subsystem is within its limit. "

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

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? W EMERGENCY CORE COOLING SYSTEMS l

. SURVEILLANCE REOUIREMENTS : !

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' 4.5 3.2 . Each- 1.PCI subsystem shall be demonstrated OPERABLE: ' j

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a. At' least once per 31 days byt

1. Verifying that the system piping from the pump discharge valve ,

to the system isolation valve is filled with water, l

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

in the flow path that is not locked, sealed, or otherwise :

j secured in position, is in its correct position, and

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3. Verifying that the subsystem cross-tie valve is closed with power removed from the valve operator.

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b. At least once per 92 days by verifying each pair of LPCI pumps i 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 120 psig. i At least once per '18 months

by performing a system functional test l

c. !

which includes simulated autouatic 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 jj injection of coolant into the reactor vessel is excluded from this test.

ADD I NS ERT " E " 1 i

For the performance of this system functional test scheduled to be completed 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 1 commence in March, 1981.

BRUNSWICK - UNIT 1 3/4'5-8

, ,-- - - .. ._ ....7. N

INSERT E

d. At least once per 18 months by verifying the ECCS RESPONSE TIME for each LPCI subsystem is within its limit. *

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

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1 f#, 3/4.3 INSTRUMENTATION l Ch BASES- j 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:

a. Preserve the integrity of the fud 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 I function even during periods when instrument channels may be out of service because of maintenance. When necessary, one channel may be made inoperable !

for brief 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 i Protection System." March 1988 and MDE-81-0485. Rev. 1. " Technical 1 Specification Im]rovement Analysis for the Reactor Protection System for !

Brunswick Steam Electric Plant. Units 1 and 2." August 1994, as modified by l

[ *.. BWROG-92102. Letter from C. L. Tully (BWROG)'to B. K. Grimes (NRC). "BWR ;

Owners' Group (BWROG) Topical Reports on Technical Specification Improvement i Analysis for BWR Reactor Protection Systems - Use for Relay and Solid State {

Plants (NEDC-30844 and NEDC-30851P)." November 4, 1992, i 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 trio systems will produce a reactor scram. The system meets the intent of I 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) utilizing replacement sensors with certified response times.

ADD IN S E.RT " F "

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

I4 A DO f rq s ERT " 6 "

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

i INSERT F l As noted (Note #), neutron detectors are excluded from REACTOR PROTECTION [

SYSTEM RESPONSE TIME testing because the principles of detector operation j virtually ensure an instantaneous response time. In addition, this note states that the j response time of the sensors for item 3, " Reactor Vessel Steam 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 j response time. This is allowed since other surveillance testing (e.g., channel l calibration) and other techniques ensure detection of response time degradation before l performance is significantly affected (Reference 1). !

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References * ,

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

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6 -INSTRUMENTATION !

a/ BASES l l

3/4.3.2- ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used '

to mitigate the consequences of accidents by prescribing the trip settings for -l 1 solation of the reactor systems. When necessary, one channel may be i inoperable for brief intervals to conduct required surveillance. Some of the ;

trip settings have tolerances explicitly stated where both the high and low

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values are critical and may have a substantial effect on safety. The j 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 j from the normal o i systems involved.perating range to prevent inadvertent actuation of the

{

S ecified surveillance intervals and allowed out-of-service times were estab ished based on the 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, as modified by OG90-579-32A. i Letter to Millard L. Wohl (NRC)'from W. P. Sullivan and J. F. Klapproth (GE). j

'" Implementation Enhancements to Technical Specification Changes Given in i Isolation Actuation Instrumentation Analysis." June 25. 1990 and supplemented '

by GE letter report GENE-A31-00001-02. " Assessment of Brunswick Nuclear Plant l Isolation Actuation Instrumentation Against NEDC-31677P-A Bounding Analyses." :

August-1994. !

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 tie design i protection. Although the instruments are listed by system, in some cases the l same instrument is used to send the start signal to several systems at the same time. The out-of-service times for the instruments are consistent with the requirements of the specifications in Section 3/4.5.

Specified surveillance intervals and allowed out-of-service times were established based on the reliability analyses documented in GE reports NEDC-30936P-A. Carts 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 Actuat bn Instrumentation for Brunswick Steam Electric Plant. Units 1 & 2." August 1994, as modified by 0G90-319-32D. 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.

ADD ir4GERT " H "

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

INSERT H 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.1, " Reactor Vessel Water Level - Low, Level 1"; Item 1.a.2,

" Reactor Vessel Water Level - Low, Level 3"; Item 1.c.3, " Main Steam Line Flow - High";

Item 2.c, " Reactor Vessel Water level - Low, Level 2"; and Item 3.e, " Reactor Vessel Water level- Low, Level 2" 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. NEDO-32291-A," System Analyses for the Elimination of Selected Response ;

Time Testing Requirements," October 1995. j

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BASES.

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3/4.3.4 CONTROL R00 WITHDRAWAL BLOCK' INSTRUMENTATION

The' control rod block functions are provided consistent with the r uirements of !

, the specifications in Section 3/4.1.4. Rod Program Controls, and Sectio 3/4.2. i Power Distribution Limits. The trip lo :

,'the-inputs will result in a rod block, gic is arranged so that a trip in any one of j Specified surveillance intervals and allowed out-of-service times were l established based on the reliability analyses documented in GE report NEDC-30851P-A, ;

Supplement 1. " Technical Specification Improvement Analysis for BWR Control Rod j Block Instrumentation." October 1988.

i 3/4.3.5 MONITORING INSTRUMENTATION l

-3/4.3.5.1 SEISMIC MONITORING INSTRUMENTATION .

\

The OPERABILITY of the seismic monitorin instrumentation ensures that j sufficient capability is available to prompt y determine the magnitude of a seismic ;

event and evaluate the response of those fea ures important to safety. This i capability is required to permit comparison of the measured response to that used in '

the design basis for the facility. {

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BRUNSWICK - UNIT'l B 3/4 3-2a- l Amendment No. [

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1 3 /4 . 5 EMERGENCY CORE COOLING SYSTEM i I'

BASES 3/4.5.1 HIGH PRESSURE COOLANT INJECTION SYSTEM (Continued)

ACTIONS:

With the HPCI system inoperable, adequate core cooling is assured by the demonstrated operability of the redundant and diversified Automatic !

Depressurization system and the 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 pressures on a reactor low water level condition. The out-of-service period of 14 days is based on the demonstrated operability of redundant and diversified low pressure core cooling systems.

SURVEILLANCE REOUIREMENTS:

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 dameJe and to provide cooling at the earliest moment.

ADD IN S E R T I " --

REFERENCES:

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

2. Brunswick Stemm 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. 1 ADD INSERT "J" l 3 /4 . 5 . 2 AITTOMATIC DEPRESSURI"ATION 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 provide adequate core cooling up to 150 psig.

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

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Surveillance Requirement 4.5.1.c.4 ensure', that the ECCS RESPONSE TIME for the HPCI system is less than or equal to the er ar.imum value assumed in the accident !

analysis. Response time testing acceptar.ce criteria are included in Reference 6. This ,

surveillance requirement is modified by a note that allows the instrumentation portion of 1

the response time to be assumed to be the design instrumentation response time.

j 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).

f INSERT J

, 6. Updated Final Safety Analysis Report, Section 6.3.3.7.

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7. NEDO-32291-A,

System Analyses for the Elimination of Selected Response

Time Testing Requirements," October 1995. ,

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)M. ' BASES l c ld CORE SPRAY SYSTEM (Continued) l 1

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 restriction on operations is not applicable. [

l The surveillance requirements provide adeouate assurance that the CSS will ;

be OPERAB1E 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 i

'is maintained full to prevent water hammer damage to piping .and to start l cooling at the earliest moment. j

--- Apo IN SERT' " M

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3/ 4. 5. 3. 2 LOW PRESSURE COOLANT INJECTION SfSTEM (LPCIS) t l The LPCIS is provided to assure that the core is adequately cooled l

. following a loss-of-coolant accident. Two loops each with two pumps provide l adequate core flooding for all break sizes from 0.2 ft2up to and including ]

4 - 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 because LPCIS is a primary source of water for flooding the core af ter the j 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 i OPERABILITY of the redundant LPCIS pumps or loop, and both CSS loops. The reduced redundancy justifies the specified 7-day out-of-service period.

The surveillance requirements provide adequate assurance that LPCI will be 3 OPERAB1E when required. Although all active components are testable and full )

e flow can be demonstrated by recirculation during reactor operation, a complete l functional test requires reactor shutdown. The pump discharge piping is 1 maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.

ADD I N S E R.T " L "

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INSERT K 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 maximum value assumed in the accident analysis. Response time testing acceptance criteria are included in i Reference 1. This surveillance requirement is modified by a note that allows the 5

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 othei techniques ensure . ,

detection of response time degradation before performance is signific.antly Effected (Reference 2).

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 Time Testing Requirements," October 1995, t 4

INSERT L ,

I 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 equal to the maximum value assumed in the accident analysis. Response time testing acceptance criteria are >

included in Reference 1. This surveillance requirement is modified by a nt 1 that allows the instrumentation portion of the response time to be assumed to be the design l

' i 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. Updated Final Safety Analysis Report, Section 6.3.3.7.

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

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ENCLOSURE 6 !

i 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 i I

REQUEST FOR EMERGENCY / EXIGENT LICENSE AMENDMENTS INSTRUMENTATION RESPONSE TIME TESTING l

l

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MARKED-UP TECHNICAL SPECIFICATION PAGES - UNIT 2 i

c 3/4.3 INSTRUMENTATTON

q 3/4 3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.1.1 Each reactor 3rotection system instrumentation channel shall be demonstrated OPERABLE Jy 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 i calibration of time delay relays and timers necessary for proper functioning '

i of the trip system.

t 4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip ,

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 4 i

N is the total number of redundant channels in a specific reactor trip function.

R i

Neutrondetectorsareexemptfromresponsetimetesting.g ado IN S E RT " A "

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

INSERT A ,

The sensor response times for the following functions may be assumed to be the design sensor response time:

Item 3, " Reactor Vessel Steam Dome Pressure - High" item 4, " Reactor Vessel Water Level - Low, Level 1" 1

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INSTRUMENTATION

~ SURVEILLANCE REQUIREMENTS (Continued)

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4 4.3.2.3 The ISOLATION SYSTEM RESPONSE TIME of each isolation function'shall be -j

-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 I !

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 i redundant channels in a specific isolation function. ;

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' Radiation monitors are exempt from response time testi i '

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i INSERT B !

l The sensor response times for the following functions may be assumed to be the design sensor response time: l i

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

ltem 1.c.4, " Main Steam Line Flow - High" ;

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INSTRUMENTATION !

3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3 The Emergency Core Coolin ystem (ECCS) actuation instrumentation channels shown in Table 3.3.3-1 s I be OPERABLE with their trip se oints I i

set consistent with the values shown in the Trip Setpoint column of ble '

3.3.3-2.

i APPLICABILITY- As shown in Table 3.3.3-1.

A_CTION*

i

a. With an ECCS actuation instrumentation channel trip setpoint less l conservative than the value shown in the Allowable Values column of i Table 3.3.3-2. declare the channel inoperable until the channel is l l 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. l 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. j SURVEILLANCE REQUIREMENTS l 4.3.3.1 Each ECCS actuation instrumentation channel shall be demonstrated i OPERABLE by the performance of the CHANNEL CHECK. CHANNEL CALIBRATION. and l

('

CHANNEL FUNCTIONAL TEST operations during the OPERATIONAL CONDITIONS and at i s the frequencies shown in Table 4.3.3-1. l 4.3.3.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of i

< all channels shall be oerformed at least once per 18 months and shall include !

calibration of time delay relays and timers necessary for proper functioning ,

of the trip system. j 4.3.1.3/Tiie ECCS RESPONSE TIME of each ECCS function shall be demonstrated to i e within the limit at least once per 18 months. Each test shall include at least one locic train such that both logic trains are tested at least once per 36 months anc one channel per function such that all channels are tested at L least once every N times 18 months, where N is the total number of redundant nnels n a specific ECCS function.

D e l e+cd .

1 b i i

[

\ !

BRUNSWICK - UNIT 2 3/4 3-33 !

Amendment No. [

b l

EMERGENCY CORE COOLING SYSTEMS O ~

SURVEILLANCE REOUIREMENTS (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 gym for a system head corresponding to a reactor pressure > 1000 psig when steam is being supplied to the turbine at 1000, 720, -18, psig. .

c. At least once per 18 months by:

i

1. 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. {

2. Verifying that the system develops a flow of at least 4250 gpm i for a system head corresponding to a reactor pressure of > 165 ,

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

p 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. l 1

I ADD INS ERT "C "

MRUNSW1CK - UNIT 2 3/45-2 nmm Tre. Er:cSU

}'- y2 m. . - - ' 7%

I t

l; f

INSERT C

4. Verifying that the ECCS RESPONSE TIME for the HPCI system is wi+hin its i; limit.

I

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

I t

i i

i r

l l

i l

i I

l I

i

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

O EMERCENCY CORE C00LINC SYSTEMS ,

SURVEILLANCE REQUIREMENTS (Continued)

2. Verifying 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. ,

i

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 spa on recirculation flow against a system head corresponding to a reactor vessel pressure of > 113 psig. l

2. Performing a CHANNEL CALIBRATION of the core spray header AP instrumentation and verifying the setpoint to be 5, +1.5, paid [

greater than the normal indicated AP.

d. At least once per 18 months by performing a system functional test :'

which includes simulated automatic actuati'on 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 l p test.

ADD INSERT"D"- 4 1

1 l

i

\

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

1 I

i INSERT D ;

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

Instrumentation response time may be assumed to be the design ;

instrumentation response time, j r

t i

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f r

T 5

l t

l i

e i

p r

l i

?

= !

l l

,=% EMERCENCY CORE C00LINC SYSTEMS i

SURVEILLANCE REOUIREMENTS l t

l 4.5 1.2 Each LPCI subsystem shall be demonstrated OPERABLE: :

a. At least once per 31 days by:

1. Verifying that the system piping from the pump discharge valva !

to the system isolation valve is filled with water, j

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

h. 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 ],20 psig. ,

c. At least once per 18 months by performing a system functional test [

which includes simulated automatic actuation of the system throughout

(~'S Lts 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 th test. ;

Aon I N S E RT " E " - l l

l 1

5 nRUNSWICK - UNIT 2 3/4 5-8 a m orn reco. ::: g

@na na

n. -w ,a r ,

INSERT E

d. At least once per 18 months by verifying the ECCS RESPONSE TIME for each LPCI subsystem is within its limit. "

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

4 i

1 l

1 3/4.3 INSTRUMENTATION .

BASES t .

3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION 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. l

c. Minimize the energy which must be adsorbed following a loss-of- '

coolant accident, and prevent inadvertent criticality.

This specification arovides the limiting conditions for operation necessary to preserve tie ability of the system to perform its intended function even during periods when instrument channels may be out of service because of maintenance. When necessary, one channel may be made inoperable for brief 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 NEOC-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 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 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.

There are usually four channels to monitor each parameter with two in each trip system. The outputs of the cnannels ir c 'a 4 6tr 3ra combined in a logic so that either channel will trip that trip system. The tripping of both tria systems will produce a reactor scram. The system meets the intent of IEE b279 for nuclear power plant protection systems.

The measurement of response time at the specified frequencies provides assurance that the arotective, 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, overiapping 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 reolacement sensors with certified response times. ;

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

ADD I N S EAT " G " ------

I ADD INSERT " F "

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

i l

.i l

l'

- INSERT F i

l. As noted (Note #), neutron detectors are excluded from REACTOR PROTECTION l SYSTEM RESPONSE TIME testing because the principles of detector operation _

virtually ensure an instantaneous response time. In addition, this note states that the l

~

response time of the sensors for item 3, " Reactor Vessel Steam Dome Pressure - High" and item 4, " Reactor Vesse! Water Level- Low, Level 1" may be assumed in the !

REACTOR PROTECTION SYSTEM RESPONSE TIME test to be the design sensor i response time, This is allowed since other surveillance testing (e.g., channel j calibration) and other techniques ensure detection of response time degradation before !

performance is significantly affected (Reference 1). I l

i INSERT G l 4

References:

I

)

1. NEDO-32291-A, " System Analyses for the Elimination of Selected Response ,

Time Testing Requirements," October 1995. j l

)

l

-r 1 l

l l

l 1

l l

)

9 INSTRUMENTATI@

BASES v

3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the trip settings for isolation of the reactor systems. When necessary, one channel ma inoperable for brief intervals to conduct required surveillance. y Some be of the trip settings havi tolerances explicitly 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 o systems involved.perating range to prevent inadvertent actuation of the 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 S)ecification Improvement Analysis for BWR Isolation Instrumentation Common to R)S and ECCS Instrumentation." March 1989 and NEDC-31677P-A. " Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation. July 1990, as modified by OG90-579-32A.

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.

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

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 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.

ADD NSERT " H "

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

INSERT H 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.1, " Reactor Vessel Water Level- Low, Level 1"; 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"; Item 2.c, " Reactor Vessel Water level - Low, Level 2"; and item 3.e, " Reactor Vessel Water level - Low, Level 2" 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. NEDO-32291-A," System Analyses for the Elimination of Selected Response Time Testing Requirements," October 1995.

4

]

INSTRUMENTATION l 1

-BASES-

.{- 1 3/4.3.4 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION The control rod block functions are provided consistent with the :

recuirements of the specifications in Section 3/4.1.4, Rod Program Controls anc 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 l BWR Control Rod B1ock Instrumentation," October 1988. t 2/4.3.5 MONITORING INSTRUMENTATION l

3/4.3.5.1 SEISHIC HONITORING INSTRUMENTATION !

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

seismic event and evaluate the response of those features important to :

safety. This capability is required to permit comparison of the measured l response to that used in the design basis for-the facility, i l

l l

I BRUNSWICK - UNIT 2 8 3/4 3-2a Amendment No. [

3/4.5 EMERGENCY COPE COOLING SYSTEM BASES 3/4.5.1 HIGH DRESSURE COOLA!fr INJEC"rION SYSTEM (Continued)

ACTIONS:

With the HPCI system inoperable, adequate core cooling is assured by the-demonstrated operability of the redundant and diversified Automatic Depressurization system and the 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 pressures on a reactor low water level condition. The out-of-service period of 14 days is based on the demonstrated operability of redundant and diversified low pressure core cooling systems.

SURVEILLANCE REOUIPEMENTS:

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.

ADD IN S E. R.T "3

REFERENCES:

'. 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 L5.5.2.

--- ADI) f r4S E R.T " J " -

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 CPERABLE whenever reactor vessel pressure exceeds 113 psig even though low pressure cooling systems provide adequate core cooling up to 150 psig.

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

INSERT I Surveillance Requirement 4.5.1.c.4 ensures that the ECCS RESPONSE TIME for the HPCI system is less than or equal to the maximum value assumed in the accident analysis. Response time testing acceptance criteria are 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 ensure detection of response time degradation befors performance is significantly affected (Reference 7).

INSERT J

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.

_ . _ _ _ _ _= ,.

EMERGENCY CORE COOLING SYSTEMS p

BASES 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 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.

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 cooling at the earliest moment. ,

ADD fNSE AT " K " -

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 yith two pumps provide ;

g adequate core flooding for all break sizes from 0.2 ft up to and including 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 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 out-of-service period.

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 tnd to start cooling at the earliest moment.

ADD IN SE RT " L " -

/

BRUNSWICK - UNIT 2 B 3/4 5-3 R TT":::: TEC". S?? " .

f":ld:::d fp Mn ^--aA 71^

l I

INSERT K 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 maximum value assumed in '

the accident analysis. Response time testing acceptance criteria are 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 (Reference 2).

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 Time Testing Requirements," October 1995. .

INSERT 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 equal to the maximum value assumed in the accident analysis. Response time testing acceptance criteria are 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 (Reference 2). .

References:

1. Updated Final Safety Analysis Report, Section 6.3.3.7.

2. NEDO-32291-A, " System Analyses for the Elimination of Selected Response !

Time Testing Requirements," October 1995.

i i

1 h

ENCLOSURE 7 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 REQUEST FOR EMERGENCY /EX1 GENT LICENSE AMENDMENTS INSTRUMENTATION RESPONSE TIME TESTING TYPED TECHNICAL SPECIFICATION PAGES - UNIT 1 Y

3/4.3 INSTRUMENTATION' 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION t

SURVEILLANCE REQUIREMENTS i 4'3.1.1 Each reactor 3rotection system instrumentation channel shall be

^

demonstrated OPERABLE ay 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 i 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 function'shall be demonstrated to be within its limit at least once per 18

- t

. 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 moriths where !

N is the total number of redundant channels in a specific reactor trip !

function. !

i f

i I

' Neutron detectors 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 3,'" Reactor Vessel Steam Dome Pressure-High" Item 4. " Reactor Vessel Water Lovel-Low. Level 1"

^

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

TNSTRUMENTATION i 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 1 solation function.

r 1

I i

' 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" l Item 1.c.3. " Main Steam Line Flow-High" l l

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1 i

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BRUNSWICK - UNIT 1 3/4 3-11 Amendment No. I i

)

1NSTRUMENTATION 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 c.hannels 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 i 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. I BRUNSWICK - UNIT.1 3/4 3-33 Amendment No. I

EMERGENCY CORE C00LfNG 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.

b. At least once per 92 days by verifying that the system develops a flow of at _least 4250 gpm for a systera 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 a 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 suppression pool high water level signal.

4. Verifying that thp ECCS RESPONSE TIME for the HPCI system is within its limit

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

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

EMERGENCY CORE COOLING SYSTEMS >

-SURVEILLANCE REQUIREMEN1$ (Continued) !

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

automatic) in the flow path that is not locked. sealed. or i 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 i instrumentation and verifying the setpoint to 3e 5, 1.5, psid greater than the normal indicated AP,

d. At least once per 18 months by performing a system functional test I which includes simulated automatic actuation of the system t throughout its emergency operating sequence and verifying that each ;

automatic valve in the flow path actuates to its correct position. l Actual injection of coolant into the reactor vesse'l is excluded '

from this test.

i

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

i I

l

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

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

EMERGENCY CORE C00LlNG 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 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 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 verifying'the ECCS RESPONSE TIME for each LPCI subsystem is within its limit.

For the performance of this system functional test scheduled to be completed 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. l

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

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

3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION 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. ,

c. Minimize the energy which must be adsorbed following a loss-of-coolant accident, and prevent inadvertent criticality. .

This specification ]rovides the limiting conditions for operation necessary to preserve t1e ability of the system to perform its intended ;

function even during periods when instrument channels may be out of service because of maintenance. When necessary, one channel may be made inoperable i for brief 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 re) orts NEDC-30851P-A. " Technical Specification Improvement Analyses for 3WR Reactor Protection System." March 1988 and MDE-81-0485. Rev. 1. " Technical Specification Improvement Analysis for the Reactor Protection System for 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 P1 ants (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 monitor each parameter, with two in each trip system. The outputs of the channels in a trip system are combined in a i logic so that either channel 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. I The measurement of response time at the specified frequencies provides ,

assurance that the ]rotective, isolation. and emergency core cooling functions i associated with eac1 channel are completed within the time limit assumed in j 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) utilizing replacement sensors with certified response times.

As noted (Note #). neutron detectors are excluded from REACTOR PROTECTION ,

SYSTEM RESPONSE TIME testing because the principles of detector operation l 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 1 B 3/4 3-1 Amendment No. I 4

f 3/4.3 INSTRUMENTATION BASES .

3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (Continued)

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

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

JNSTRUMENTATION BASES ,

3/4 3.2 ISOLATION ACTUATION INSTRUMENTATION ;

This specification 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, one channel may be inoperable for brief intervals ~ to conduct required surveillance. Some of the trip settings have tolerances explicitly 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 i 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 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. as modified by OG90-579-32A, H 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 l Isolation Actuation Instrumentation Against NEDC-31677P-A Bounding Analyses." i 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 respon.se time of the sensors for Item 1.a.1, " Reactor Vessel Water Level-Low, level 1", Item 1.a.2. " Reactor Vessel Water Level-Low, level 3": Item 1.c.3.

" Main Steam Line Flow-High"; Item 2.c, " Reactor Vessel Water Level-Low. Level 2",

and Item 3.e. " Reactor Vessel Water Level-Low Level 2" may be assumed in the l ISOLATION SYSTEM RESPONSE TIME test to ha the design sensor res)onse time. This l is allowed since other surveillance testing (e.g.. channel cali) ration) and other i techniques ensure detection of response time degradation before performance is 1 significantly affected(Reference 1).

REFERENCES:

j

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 system actuation instrumentation is provided to l ioltiate 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 i 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 ,

time. .The out-of-service times for the instruments are consistent with the l requirements of the specifications in Section 3/4.5. i BRUNSWICK - UNIT 1- B 3/4 3-2 Amendment No. I l

i INSTRUMENTATION BASES 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION (Continued;

-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 Daners' Group Technical Specification i

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 0G90-319-32D. 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 R0D WITHDRAWAL BLOCK INSTRUMENTATION The control rod block functions are provided consistent with the requirements lof 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. l 3/4.3.5 MONITORING INSTRUMENTATION

[

3/4.3.5.1' SEISMIC MONITORING INSTRUMENTATION 1

The OPERABILITY of the seismic monitoring instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety.

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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BRUNSWICK - UNIT 1 8 3/4 3-2a Amendment No. 175 1

3/4.5 EMERGENCY CORE COOLING SYSTEM BASES- l

'3/4 5.1 HIGH PRESSURE COOLANT INJECTION SYSTEM (Continued) !

ACTIONS: ;

With the HPCI system inoperable, adequate core cooling is assured by the demonstrated OPERABILITY of the redundant and diversified Automatic Depressurization system and the 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 pressures on a reactor low water level condition. The out-of-service period of 14 days is based on the demonstrated operability of redundant and diversified low pressure core cooling systems.

SURVEILLANCE RE0VIREMENTS: ,

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. ,

i i

Surveillance Requirement 4.5.1.c.4 ensures that the ECCS RESPONSE TIME for the HPCI system is less than or equal to the maximum value assumed in the accident analysis. Response time testing acceptance criteria are included in e 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 j 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).

REFERENCES:

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

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

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1 3/4.5 EMERGENCY CORE COOLING SYSTEM BASES !

i 3/4.5.2 AUTOMATIC DEPRESSURIZATION SYSTEM (ADS) f Upon failure of the HPCIS to function properly after a small break loss- i of-coolant accident, the ADS automatically causes the safety-relief valves to ,

open, depressurizing the reactor so that flow from the low pressure cooling i system can enter the core in time to limit fuel cladding temperature to less ;

than 2200*F. ADS is ccnservatively required to be OPERABLE whenever reactor vessel pressure exceeds 113 psig even though low pressure cooling systems ;

provide adequate core cooling up to 150 psig.

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

EMERGENCY CORE COOLING SYSTEMS BASES 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 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.

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 cooling at the earliest moment.

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 maximum value assumed in the accident analysis. Response time testing acceptance criteria are 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 (Reference 1).

REFERENCES:

1. Updated Final Safety Analysis Report. Section 6.3.3.7.

2. NE00-32291-A. " System Analyses for the Elimination of Selected Response Time Testin~g 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 )rovide 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 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 pum) inoperable, or one LPCIS loop inoperable, adequate core flooding is assured )y the demonstrated OPERABILITY 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

L EMERGENCY CORE COOLING SYSTEMS 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.

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 equal to the maximum value assumed in the accident analysis. Response time testing ,

' acceptance criteria are 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 cclibration) 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.

l

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

4 4

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

ENCLOSURE 8 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 REQUEST FOR EMERGENCY / EXIGENT Ll CENSE AMENDMENTS INSTRUMENTATION RESPONSE TIME TESTING TYPED TECHNICAL SPECIFICATION PAGES - UNIT 2 -

I I

1 1

.- ...--.4 . .,- - ---

' 3/4.3 INSTRUMENTATION

~ 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.1.1 Each reactor )rotection system instrumentation channel shall be demonstrated OPERABLE )y 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 ,

calibration of time delay relays and timers necessary for proper functioning of the trip. system.

t 4.3.1.3 'The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip 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 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 reactor trip ,

function.

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' Neutron detectors 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 3. " Reactor Vessel Steam Dome Pressure-High" Item 4. " Reactor Vessel Water Level-Low. Level 1" l

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

INSTRUMENTATION SURVEILLANCE REQUIREMENTS (Continued) 4.3.2.3 The ISOLATION SYSTEM RESPONSE TIME of each isolation function'shall i 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 l 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.

k j

' 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" P

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

e

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 shalT 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.

i 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, i l

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 CONDITIJN 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.

J 4.3.3.3 Deleted.

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

EMERGENCY CORE C00 LING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 2.

Verifying)that automatic each valve (manual, power-operated, orin the flow path t 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 a 1000 psig when steam is being supplied to the turbine at 1000. +20. -18, psig. j

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 '

injectionofcoolantintothereactorvesselisexcludedfrom 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 a 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 j suppression pool high water level signal. t

4. Verifying that thp ECCS RESPONSE TIME for the HPCI system is l !'

within 1 s limit l

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

BRUNSWICK - UNIT 2 3/4 5-2 Amendment No, l !

.. l

l EMERGENCY CORE C001.ING SYSTEMS ~

l 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: l

1. Verifying that each CSS pump can be started f,-om the control room and develops a flow of at least 4625 gpm on recirculation 1 flow against a s stem head corresponding to a reactor vessel i pressure of a 11 psig l l

2. Performing a CHANNEL CALIBRATION of the core s3 ray header AP instrumentation and verifying the set ]e 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 sich includes simulated automatic actuation of the system thrcighout its emer ency operating sequence and verifying that each automatic valve in khe 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

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

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

t EMERGENCY CORE COOLING SYSTffz !

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 piping from the pump discharge valve to the system isolation valve is filled with water. :

i 2.

Verifying)that automatic each valve (manual, power-operated, orin the flow pati 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 -

i 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. j

c. At least once per 18 months by performing a system functional test ;

which includes simulated automatic actuation of the system throughout its emerfency automatic valve in he flowoperating sequence path actuates to its and verifying correct that each 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 !

each LPCI subsystem is within its limit.

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

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

3/4.3 INSTRUMENTATION r BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION 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.

c. Minimize the energy which must be adsorbed following a loss-of- -

coolant accident, and prevent inadvertent criticality.

This specification ]rovides the limiting conditions for operation necessary to preserve tie ability of the system to perform its intended function even during periods when instrument channels may be out of service 3 because of maintenance. When necessary, one channel may be made inoperable ,

for brief 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 re) orts ses for 3WR Reactor NEDC-30851P-A.

Protection System,"" Technical March 1988 Specification Improvement and MDE-81-0485, Analy' Technical Rev. 1.

Specification Imarovement 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.

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 )rotective, 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) utilizing replacement sensors with certified response times.

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

5 3/4.3 INSTRUMENTATION BASES 3/4.3.1 ' REACTOR PROTECTION SYSTEM INSTRUMENTATION (Continued) <

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 l in the bases for Specification 2.2.

REFERENCES:

1. NED0-32291-A. " System Analyses for the Elimination of Selected Response '

Time Testing Requirements. " October 1995.

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

INSTRUMENTATION l BASES i

3/4.3 2 ISOLATION ACTUATION INSTRUMENTATION j This specification ensures the effectiveness of the instrumentation used l to mitigate the consequences of accidents by prescribing the trip settings for isolation of the reactor systems. When necessary, one channel ma ,

inoperable for brief intervals to conduct required surveillance. Someyofbe the trip settings have tolerances explicitly 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 o systems involved.perating range to prevent inadvertent actuation of the j i

I 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 S)ecification Improvement Analysis for BWR Isolation Instrumentation Common to R)S and ECCS Instrumentation " March 1989 and NEDC-31677P-A, " Technical Sp*ecification Improvement Analysis for BWR Isolation Actuation Instrumentation. July,1990, as modified by OG90-579-32A, 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 .

l response time of the sensors for Item 1.a.1. " Reactor Vessel Water Level-Low, Level 1", 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".

Item 2.c. " Reactor Vessel Water Level-Low, Level 2": and Item 3.e. " Reactor Vessel Water Level-Low Level 2" 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 Anal Time Testing Requirements.ysesOctober 1995.for the Elimination of Selected Response .

1 3/4.3 3 EMERGENCY CORE COOLING SYSTEM ACTUAfl0N 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 sig'nal to several systems at.the same 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. l

INSTRUMENTATION BASES 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION (Continued)

Specified surveillance intervals and allowed out-of-service times were established based on the reliability analy'ses 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 D e 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. 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 R00 WITHDRAWAL BLOCK INSTRUMENTATION The control rod block functions are provided consistent with the recuirements of the specifications in Section 3/4.1.4. Rod Program Controls anc 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 INSTRUMENTATIQN The OPERABILITY 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. This capability is required to permit comparison of the measured response to that used in the design basis for the facility.

BRUNSWICK - UNIT 2 B 3/4 3-2c Amendment No. l

1 3/4.5 EMERGENCY CORE COOLING SYSTEM !

l

.BA3ES :

.3/4.5.1 HIGH PRESSURE COOLANT INJECTION SYSTEM (Continued)

ACTIONS:

l With the HPCI system inoperable, adequate core cooling is assured by 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 i is taken in the safety analysis, will automatically )rovide makeup at reactor i pressures on a reactor low water level condition. T1e out-of-service period .

of 14 days is based on the demonstrated operability of redundant and diversified low pressure core cooling systems. j 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 i discharge piping is maintained full to prevent water hammer damage and to l provide cooling at the earliest moment. l i

Surveillance Reguirement 4.5.1.c.4 ensures that the ECCS RESPONSE TIME for the HPCI system is less than or equal to the maximum value assumed in the accident analysis. Response time testing acceptance criteria are 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 resJonse time. Therefore, the instrumentation response l time is excluded from the ECCS RESPONSE TIME testing. This exception is i allowed since other surveillance testing (e.g., channel calibration) and other ;

technigues 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. i l

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. l S. Brunswick Steam Electric Plant Updated FSAR. Section 15.5.2.

6. Updated Final Safety Analysis Report. Section 6.3.3.7. I l

7. NED0-32291-A " System Anal ses for the Elimination of Selected Response Time Testing Requirements y' October 1995.

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

3/4.5 EMERGENCY CORE COOLING SYSTEM-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 OPEMBLE whenever reactor vessel pressure exceeds 113 psig even though low pressure cooling systems provide adequate core cooling up to 150 psig.

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BRUNSWICK - UNIT 2 B 3/4 5-lb Amendment No. l

EMERGENCY CORE COOLING SYSTEMS BASES CORE SPRAY SYSTEM (Continued)

When in CONDITION 4 or 5 with neither CSS loop OPERABLE prohibition of all cperations 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.

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 cooling at the earliest moment.

Surveillance Requirement 4.5.3.1.e ensures that the ECCS P'SPONSE c TIME for each core spray system subsystem is less than or equal to tho maximum value assumed in the accident analysis. Response time testing arceptance criteria are 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) degradationand otherperformance before techniques ensure detectionaffected is significantly of resp (onse time 2).

Reference

REFERENCES:

1. Updated Final Safety Analysis Report. Section 6.3.3.7.

2. NED0-32291-A " System Anal ses for the Elimination of Selected Response Time Testing Requirements.y' October 1995.

3/4.5.3 2 LOW PRESSURE COOLANT INJECTION SYSTEM (LPCIS)

The LPCIS is following a loss provided of-coolant to assure that the core is adequately cooledTwo l accident.

adequate core flooding for all break sizes from 0.2 ft up to and including 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 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 ram) inoperable, or one LPCIS loo) inoperable. adequate core flooding is assured )y the demonstrated OPERABILITY 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 2 B 3/4 5-3 Amendment No. l

EMERGENCY CORE COOLING SYSTEMS BASES i

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

The surveillance re OPERABLE when required.quirements provide adequate Although all active components are testableassurance and full that LPC; flow can be demonstrated by recirculation during reactor operation. a complete i 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 each low pressure coolant injection system subsystem is less than or equal to the maximum value assumed in the accident analysis. Response time testing :

acceptance criteria are 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.. charnel 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. NE00-32291-A, " System Anal.y'ses for the Elimination of Selected Response !

Time Testing Requirements. October 1995. l i

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