ML20092J396

Proposed TS Table 4.3.1.1-1, RPS Instrumentation SRs & TS Bases 3/4.3.1,changing Calibr Frequency for LPRM Signal from Every 1,000 EFPH to Every 2,000 Megawatt Days Per Std Ton

ML20092J396
Person / Time
Site:	Limerick
Issue date:	09/18/1995
From:	PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML20092J395	List: ML20092J394 ML20092J396
References
NUDOCS 9509220110
Download: ML20092J396 (5)

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,

2-ATTACHMENT 2 UMERICK GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-352 50-353  :

UCENSE NOS. NPF-39 NPF-85 TECHNICAL SPECIFICATIONS CHANGE REQUEST NO. 9541-0 UST OF AFFECTED PAGES UNIT 1 UNIT 2  ;

3/43-8 3/43-8 B 3/4 3-1 B 3/4 31 8

1 i

9509220110 950918  !

PDR ADOCK 05000352

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l TABLE 4.3.1 1-1 (Centinu;d)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS C CHANNEL OPERATIONAL E CilANNEL FUNCT10NAL ' CI'IANNEL CONDITIONS FOR WHICH CllECK TEST CALIBRATION SURVEILLANCE REQUIRED

$ FUNCTIONAL UNIT S 9. Turbine Stop Valve - Closure N.A. Q R 1 c

5 Turbine Control Valve Fast

[ 10. Closure, Trip 011 Pressure - Low N.A. Q R 1 1.1. - Reactor Mode Switch Shutdown Position N.A. R N.A. 1,2,3,4,5

12. Manual' Scram N.A. W N.A. 1 2,3,4,5 (a) Neutron detectors may be excluded from CllANNEL CALIBRATION.

(b) The IRM and SRM channels shall be determined to overlap for at least 1/2 decades during each startup after-entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to 6verlap for a least 1/2 w decades during each controlled shutdown, if not. performed within the previous 7 days.

2 (c) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup, if not performed within the previous 7 days, w,(d)' This calibration shall consist of the adjustment of the APRM channel to confork to the power values calculated by 4, a h' eat balance 'during OPERATIONAL CONDITION 1 when THERMAL POWER 225% of RATED THERMAL POWER. . Adjust the APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER. l This calibration hall consist of the adjustment _oy" RN Elow_hi 1 to conform to a calibrated flow i 4

(e) 206%bPUMT 20F WW M fM J signa).

The LPRMs shall be calibrated at least dnce pe@e; effe,.ti : f F ;r rr t rs (EF using the TIP system.

(f) Verify measured core flod total core flow) to be greater than or equal to established core flow at the existing

@ yg (g) loop flow (APRM % flow). ring the startup test program, data shall be recorded for the parameters listed to ce g- provide a basis for establishing the specified relationships. Comparisons of the actual data in accordance with

-g the criteria listed shall commence upon the conclusion of the startup test program, o "=(h) This function is not required to be OPERABLE when'the reactor pressure vessel head-is remoyed per Specification-3.10.1.

P (1) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

3 (j) If the RPS shorting links are required to be removed per Specification 3.9.2, they may be reinstalled for up to 2

- hours for required surveillance. During this time, CORE ALTERATIONS shall be suspended, and no control rod shall O be moved from its existing position. t t

- (k) Required to be OPERABLE only prior to and during shutdown margin demonstrations as performed per Specification  ;

g 3.10.3.

~

s .

t

S '

3/4.3 INSTRUMENTATION

, BASES.

Vw 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:

a. Preserve the integriidy 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

d. Prevent inadvertent criticality.

This specification provides the limiting conditions for operation necessary to preserve the 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 required surveillance.

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

There are usually four channels to monitor each parameter with two channels 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 trip systems will produce a reactor scram. The system meets the intent of IEEE-279 for nuclear power plant protection systems. Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P, " Technical Specification g Improvement Analyses for BWR Reactor Protection System, as a proved by the by letter to T. A.

• NRC Pickens andfrom documented A. Thadaniindated the NRC Safety July 15, Evaluation 1987. The basesReport for (SER)t e trip settings of RPS are discussed in the bases for Specification 2.2.1.

Automatic reactor trip upon receipt of a high-high radiation signal from the Main Steam Line Radiation Monitoring System was removed as the result of an analysis performed by General Electric in NED0-31400A. The NRC apprbved i the results of this analysis as documented in the SER (letter to George J. E,eck, BWR Owner's Group from A. C. Thadani, NRC, dated May 15,1991).

The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are completed within the time limit assumed in the safety analyses. 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 measurement, provided such tests demonstrale the total channel response time as defined. Sensor response time verification may be 1

demonstrated by either (t) sensors _with certified. response-times.inplace, (2) utilizing replacemen s onsite or offsit

'555 LPRH eaft'braScm fYefMcy j af N W # N hendeml 0" NK b "E er 2000 Muub/S T , l Sh - MoFI/ CORE as fI'e *0Ye

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FEB 161995 LIMERICK - UNIT 1 B 3/4 3-1 Amendment No. 53, 89

~

TABLE 4.3.1.1-1 (Ccntinued) . . .

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS C

M CHANNEL OPERATIONAL E -

CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHTCH R FUNCTIONAL UNIT _ CHECK TEST CALIBRATION SURVEILLANCE REQUIRED e 9. Turbine Stop Valve - Closure H.A. Q R 1 5

10. Turbine Control Valve Fast Closure, Trip 011 Pressure - Low N.A. Q R 1

11. Reactor Mode Switch Shutdown Position N.A. R N.A. 1,2,3,4,5

12. Manual Scram

~

N.A. W N.A. 1,2,3,4,5 9

w 3

Y (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b) The IRM and SRM channels shall be determined to overlap for at least 1/2 decades during each startup after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall .be determined to overlap for a least I/2 decades during each controlled shutdown, if not performed within the previous 7 days.

(c) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup, if not performed within the, previous 7 days.

(d) This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER ;2: 25% of RATED THERMAL POWER. Adjust the APRN channel if the absolute difference is greater than 2% of RATED THERMAL POWER.

fm (e) This calibration shalLconsistof the_adjustmen fdht_AERM E1= hinead channel to conform to a c liiii signa 1. Q000M*9'4caIThF Per 5fd Jcl 7avu (Moub/STD wJ (f) The LPRMs shall be calibrated at least once pe Q^^^ Off :tPc: !;" : : trE FEE"5fiTsing the TIP system.

3 (g) Verify measured core flow (total core flow) to be greater than or equal to established core flow at the existing loop flow (APRM % flow). During the startup test program, data shall be recorded for the parameters listed to gh ,

y provide a basis for establishing the specified relationships. Comparisons of the actual data in accordance with the criteria listed sha?1 commence upon the conclusion of the startup test program.

(h) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification Ij 3.10.1. t (1) With any control rod withdrawn. Not ap'plicable to control rods remnved per Specification 3.9.10.1 or 3.9.10.2.

$; (j) If the RPS shorting links are required to be removed per Specification 3.9.2, they may be reinstalled for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance. During this time, CORE ALTERATIONS shall be sus' pended, .and no control rod shall be moved from its existing position. '

(k) Required to be OPERABLE only prior to and during shutdown margin demonstrations as performed per Specification 3.10.3. g ,

< a e

4 4 3/4.3 INSTRUMENTATION

i. BASES

y

_ 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

d. Prevent inadvertent criticality.

.This specification provides the limiting conditions for operation necessary to preserve the 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 required surveillance.

The reactor protection system is made up of two in'ependent d trip systems.

There are usually four channels to monitor each parameter with two' channels in

. each trip system. The outputs of the channels in a trip system are combined j in a 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. Specified - ,

surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P, " Technical Specification p~W Improvement Analyses for BWR Reactor Protection System," as a proved by the NRC and documented in the NRC Safety Evaluation Re) ort (SER) letter to T. A.

Pickens from A. Thadani dated July 15, 1987. The )ases for t e trip settings l of RPS are discussed in the bases for Specification 2.2.1.

. Automatic reactor trip upon receipt of a high-high radiation signal /

from the Main Steam Line Radiation Monitoring System was removed as the result of an analysis performed by General Electric in NED0-31400A. The NRC approved the results of this analysis as documented in the SER (letter to George J. Beck, BWR Owner's Group from A. C. Thadani, NRC, dated May 15,1991).

The measurement of response time at the specified frequencies provides

assurance that the protective functions associated with each channel are completed within the time limit assumed in the safety analyses. No credit was -

taken for those channels with tesponse times indicated as not applicable.

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurewnt, 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 l

. (2) utilizing replacement _ sensors _witLcertif.ied_ response _t.ime_s. -

A mm cawtim peyey $ a[1%%uk dme p zocc Nw1/sT k depsuded n y 3h-mwicORE as H,e me nmdr / M-C FE8161995 LIMERICK - UNIT 2 B 3/4 3-1 Amendment No. 77, 52 1