Text

Proposed Tech Specs Section 3/4.4.2 & TS Bases Sections B 3/4.4.2,B 3/4.5.1 & B 3/4.5.2 to Increase Allowable as-found Main Steam SRV Code Safety Function Lift Setpoint Tolerance from +1% to +3%
	ML20199G202
Person / Time
Site:	Limerick
Issue date:	01/12/1999
From:	; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	;
Shared Package
ML20136B419	List: ML20199G165; ML20199G202; ML20199G237;
References
	NUDOCS 9901220215
	Download: ML20199G202 (16)
	v • d • e

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

REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY / RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.2 The safety valve function of at least 12 of the following reactor coolant system l safety / relief valves shall be OPERABLE with the specified code safety valve function lift settings:*#

4 safet relief valves 91170 ps i3% 1 5 safet relief valves 91180 ps *3% ;

5 safet relief valves 9 1190 ps i3% ,

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION: 1

-a. ' With the safety valve function of one or more of the above required safety / relief valves inoperable be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the,next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. With one or more safety / relief valves stuck open, provided that suppression i pool average ~ water tem is less than 105'F close . the stuck open i safety / relief valve s ; peratureif unable to close the stuck, open valve s l minutes or if suppre(ss) ion pool average water temperature is 110*F (or) within 2 l greater, place the reactor mode switch in the Shutdown position.

l

c. With one or more safety / relief valve acoustic monitors inoperable, restore the inoperable acoustic monitors to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ..

SURVEILLANCE REQUIREMENTS 4.4.2.1 The acoustic monitor for each safety / relief valve shall be demonstrated OPERABLE with the setpoint verified to be 0.20 of the full open noise leve1## by performance of a:

a. CHANNEL FUNCTIONAL TEST at least once per 92 days, and a

b. CHANNEL CALIBRATION at least once per 24 months **.

4.4.2.2 At least 1/2 of the safety relief valves shall be removed, set pressure tested I and reinstalled or replaced with spares that have been previously set pressure tested and !

stored in accordance with manufacturer's recommendations at least once per 24 months, and !

they shall be rotated such that all 14 safety relief valves are removed, set pressure '

. tested and reinstalled or replaced with spares that have been previously set pressure tested and stored in accordance with manufacturer's recommendations at least once per 54 months. All safety valves will be recertification tested to meet a *1% tolerance prior to returning the valves to service.

- The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.

=** The provisions of Specification 4.0.4 are not applicable provided the Surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate to perform the test.

.# Up to 2 inoperable valves may be replaced with spare OPERA.BLE valves with lower setpoints until the next refueling

1. Initial setting shall be in accordance with the manufacturer's recommendation. Adjustment to the valve full open noise level shall be accomplished during the startup test program.

LIMERICK - UNIT 1 3/4 4-7

9901220215 990112- %

PDR ADCCK 05000352 fs e eDe y

REACTOR COOLANT SYST M BASES RECIRCULATION SYSTEM (Continued)

Plant specific calculations can be performed to determine an applicable region for monitoring neutron flux noise levels. In this case the degree of l conservatism can be reduced since plant to plant variability would be eliminated.

l In this case, adequate margin will be assured by monitoring the region which has a decay ratio greater than or equal to 0.8.

Neutron flux noise limits are also established to ensure early detection of limit cycle neutron flux oscillations. BWR cores typically operate with neutron flux noise caused by random boiling and flow noise. Typical neutron 1 flux noise levels of 1-12% of rated power l l

the range of low to high recirculation loop (peak-to-peak)both flow during single and dualhave been reporte recirculation loop operation. Neutron flux noise levels which significantly bound these values are considered in the thermal / mechanical design of GE BWR fuel and are found to be of negligible consequence. In addition stability testsatoperatingBWRshavedemonstratedthatwhenstabilityrelatedneutron flux limit cycle oscillations occur they result in peak-to-peak neutron flux limit cycles of 5-10 times the typical values. Therefore, actions taken to reduce neutron flux noise levels exceeding three (3) times the typical value are sufficient to ensure early detection of limit cycle neutron flux oscillations.

Typically neutron flux noise levels show a gradiaal increase in absolute magnitude as cor,e flow is increased (constant control rod pattern) with two reactor recirculation loops in operation. Therefore, the baseline neutron flux noise level obtained at a specific core flow can be applied over a range of core flows. To maintain a reasonable variation between the low flow and qh flow end of the flow range, the range over which a specific baseline is applied should not exceed 20% of rated core flow with two recirculation loops in operation. Data from tests and operating plants indicate that a range of 20%

of rated core flow will result in approximately a 50% increase in neutron flux noise level during operation with two recirculation loops. Baseline data should be taken near the maximum rod line at which the majority of eration will occur. However, baseline data taken at lower rod lines (i.e.

will result in a conservative value since the neutron flux noise 1 ower power) el is

l i proportional to the power level at a given core flow.

3/4.4.2 SAFETY / RELIEF VALVES The safety valve function of the afety/ relief valves operates to prevent #

the reactor coolant system from being tr'essurized above the Safety 1.imit of

, 1325 psig in accordance with the ASME Code. A total of 12 OPERABLE safety / l l

relief valves is required to limit reactor pressure to within ASME III allow-able values for the worst case upset transient.

Demonstration of the safety / relief valve lift settings will occur only l during . shutdown. The safety / relief valves will be removed and either set pressure tested or replaced with spares which have been previously set pres-sure tested and stored in accordance with manufacturers recommendations in the specified frequency.

( -

LIMERICK - UNIT 1 B 3/4 4-2 l

3/4.5 EMERGENCY CORE COOLING SYSTEM BASES 3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHUTDOWN The core spray. system is provided to assure that th(e co)re is adequately cooled following a loss-of-CSS , toget coolant accident and provides adequate core cooling capacity for all break sizes up to and including the double-ended reactor recirculation line break, and for smaller breaks following depressurization by the ADS.

The CSS is a primary source of emergency core cc vessel is depressurized and a source for flooding of w,'ing coreafter the reactor in case of accidental draining.

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 through a test loop durin reactor o>eration, a complete functional test requires reactor shutdown.g The pump disc 1arge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.

The low pressure coolant injection i provided to assure that the core is adequa(LPCI) mode of the RHR system istely cooled follo coolant accident. Four subsystems, each with one pump, provide adequate core flooding for all break sizes up to and including the double-ended reactor ;

recirculation line break, and for small breaks following depressurization by !

the ADS.

The surveillance requirements provide adequate assurance that the LPCI system will be OPERABLE when required. Although all active components are !

testable and full flow can be demonstrated by recirculation through a test loop durin shutdown. Theg reactor pumpoperation, discharge apiping complete functionalfull is maintained test to requires reactor prevent water hammer damage to piping and to start cooling at the earliest moment.

The high pressure coolant injection i that the reactor core is adequately cooled (HPCI) system to limit is provided fuel clad to assure temperature in i the event of a small break in the reactor coolant system and loss of coolant which does not result in rapid depressurization of the reactor vessel. The HPCI system permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized. The HCPI system continues to operate until reactor vessel pressure is below the pressure at which CSS operation or LPCI mode of the RHR system operation maintains core cooling.

The capacity of the system is selected to provide the required core cooling.

The HPCI pump is designed to deliver greater than or equal to 5600 gpm at reactor

)ressures between 1182 and 200 psig and is capable of delivering at least 5000 gpm

>etween 1182 and 1205 psig. Initially, water from the condensate storage tank is used instead of injecting water from the suppression pool into the reactor, but no credit is taken in the safety analyses for the condensate storage tank water.

LIMERICK - UNIT 1 B 3/4 5-1

REACIORGQptANT. SYSTEM 2 3 /4. 4. 2 SAFETY /RELI'EF VALVES g

LIMITING C0EITION FOR OPERATION 3.4.2 The safety valve function of at least of the following reactor coolant system safety settings:*f / relief valves shall be OPERABLE with he specified code safety valve function lift 4

5 safeth safet relief reliefvalves valves991170 1180 N gg i3%

5 safety relief valves 91190 ps g $

APPLICABILITY: OPERATIONAL C0EITIONS 1, 2, and 3.

gTig: .

a. With the safety valve function of one or more of the above required safety and in/ COLD relief valves SWTDOWNinoperable, within be the in at least' next HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 24 hours.

b. With one or more safety / relief alves stuck o pool average water t rature is less than Ib,provided thatstuck close the suppression open safety / relief valve s if unable to close the stuck open valve minutes or if suppre(ss) on pool average water temperature is 110{s) F or within 2 greater, place the reactor mode switch in the Shutdown position.

c. With one or more safety / relief valve acoustic monitors inoperable, restore the inoperable acoustic monitors to OPERABLE status within 7 days or be'in at least NOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . ,.

SURVEILLANCE REQUIREMENTS 4.4.2.1 The acoustic monitor for each safety / relief valve shall be demonstrated OPERA 8LE

, with the setpoint verified to be 0.20 of the full open noise levelN by performance of a:

~

a. CHANNEL FUNCTIONAL TEST at least once per g2 days, and a

b. CHANNEL CALIBRATION at least once per 24 months".

4.4.2.2 At least 1/2 of the safety relief valves shall be removed, set pressure tested and reinstalled or replaced with spares thWhave been previously set pressure tested and stored in accordance with manufacturer's recommendations at least once per 24 months, and they shall be rotated such tha 1 ty lief v lves removed, set pressure tested and rei st ave en p ous tested a ed in accordance with manu acturer's recommendations at east once per 54 months.$/V{.5 Whtig, WiU be rEccriffication te5kd 40 mett 8 If/o hierance orioc += refurnina %e valves de servke.

I lift setting' pressure shall corFespond to ambient conditions of the olv t non para temperatu s and sures

- The prov of ficat .. a ap e p 4 . the Surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is i adaquate to perfore the test.

Up to 2 inoperable valves may be replaced with spare OPERABLE valves with lower setpoints until the next refueling

8 Initial setting shall be in accordance with the manufacturer's recossendation. Adjustment to the valve fe?1 open noise level shall be accomplished during the startup test program.

LIMERICK - UNIT 1 3/4 4-7 Amendment No. 56.70,7f.106 FD 12 1996

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

REACTOS COOLANT SYSTEM t

! BASES RECIRCULATION SYSTEM (Continued)

Plant specific calculations can be performed to detemine an applicable region for monitoring neutron flux noise levels. In this case the degree of conservatism can be reduced since plant to plant variability would be eliminated.

In this case, adequate margin will be assured by monitoring the region which has a decay ratio greater than or equal to 0.8. .

Neutron flux noise limits are also established to ensure early detection of limit cycle neutron flux oscillations. BWR cores typically operate with neutron flux noise caused by randos boiling and flow noise. Typical neutron flux noise levels of 1-12% of rated power (peak-to peak) have been reported for the range of low to high recirculation loop flow during both single and dual recirculation loop operation. Neutron flux noise levels which significantly bound these values are considered in the thermal / mechanical design of GE BWR fuel and are found to be of negligible consequence. In addition, stability tests at operating BWRs have demonstrated that when stability related neutron j

flux limit cycle oscillations occur they result in peak-to peak neutron flux limit cycles of 5-10 times the typical values. Therefore, actions taken to reduce neutron flux noise levels exceeding three (3) times the typical value are sufficient to ensure early detection of limit cycle neutron flux oscillations.

Typically, neutron flux noise levels show a gradual increase in absolute magnitude as core flow is increased (constant control rod pattern) with two 4

reactor recirculation loops in operation. Therefore, the baseline neutron flux noise level obtained at a specific core flow can be applied over a range of core flows. To maintain a reasonable variation between the low flow and high 1

flow end of the flow range, the range over which a specific baseline is applied 1

should not exceed 20% of rated core flow with two recirculation loops in j operation. Data from tests and operating plants indicate that a range of 20%

of rated core flow will result in approximately a 50% increase in neotron flux 3 noise level during operation with two recirculation loops. Baseline data should be taken near the maximum rod line at which the majority of operation will occur. However, baseline data taken at lower rod lines (i.e. Iower power) e will result in a conservative value since the neutron flux noise level is proportional to the power level at a given core flow.

N

~

3/4.4.2 SAFETY / RELIEF VALVES The safety valve function of the safety / relief valve operates to prevent the reactor coolant system free being pressurized above the Safety Limit of

, 1325 psig in accordance with the ASME Code. A total of OPERA 8LE safety /

relief' valves is required to limit reactor pressure to within ASME III allow-able values for the worst case upset transient.

. Demonstration of the safety / relief valve lift settings will occur only during shutdown. The safety / relief valves will be removed and either set pressure tested or replaced with spares which have been previously set pres-i.

sure tested and stored in accordance with manufacturers recommendations in the specified frequency. -

LIMERICK - UNIT 1 B 3/4 4-2 Amenoment No. 30 g 3 0 G3 1

- ..- - - - .- - - - - --- - - - - ~ - - - -

3/4.5 EMERGENCY CORE CnntING SYSTEM W

l aAsEs

} 3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHtfTDOWN !

I

-The core spray system (CSS), together with the LPCI mode of the RHR system, is provided to assure that the core is adequately cooled following a loss-of- .

] coolant accident and provides adequate core cooling capacity for all break sizes up to and including the double-ended reactor recirculation line break,

, and for smaller breaks following depressurization by the ADS. i l

The CSS is a primary source of emergency core cooling after the reactor i vessel is depressurized and a source for flooding of the core in case of

accidental draining.

}

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 through a test loop during .

! reactor operation, a complete functional test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage to

]i piping and to start cooling at the earliest moment.

! The low pressure coolant injection (LPCI) mode of the RHR system is l

' provided to assure that the core is adequately cooled following a loss-of-coolant accident. Four subsystems, each with one pump, provide adequate core i i

flooding for all break sizes up to and including the double-ended reactor '

recirculation line break, and for small breaks following depressurization by i the ADS.

The surveillance requirements provide adequate assurance that the LPCI j

system will be OPERABLE when required. Although all active components are j

testable and full flow can be demonstrated by recirculation through a test j

loop during reactor operation, a complete functional test requires reactor j shutdown. The pump discharge piping is maintained full to prevent water a4 hammer damage to piping and to start cooling at the earliest moment.

The high pressure coolant injection (HPCI) system is provided to assure j that the reactor core is adequately cooled to limit fuel clad temperature in the event of a small break in the reactor coolant system and loss of coolant which does not result in rapid depressurization of the reactor vessel. The t

HPCI system permits the reactor to be shut down while maintaining sufficient

! reactor vessel water level inventory until the vessel is depressurized. The

HCPI system continues to operate until reactor vessel pressure is below the

[ pressure at which CSS operation or LPCI mode of the RHR system operation 4 maintains core cooling.

4 i The capacity of the system is selected to provide the required core cooling.

1 The HPCI pump is designed to deliver reater than or equal to 5600 gpa at reactor

!' pressures between 1182 and 200 psi Initially, water from the condensate storage l tank is used instead of injecting ter from the suppression pool into the reactor, but no credit is taken in the safety analyses for the condensate storage tank water.

4

and is capabie. e{ ddWeinj ap l

Q ea Q 5000 bekacq W.ad i

LIMERICK - UNIT 1 YLM PM 8 3/4 5-1

Amenament No. 106 li FEB 1 2 W l

4

REACTOR COOLANT SYSTEM j 3/4.4.2 SAFETY / RELIEF VALVE 1 LIMITING CONDITION FOR OPERATION

. 3.4.2 The safety ' valve function of at least 12 of the following reactor coolant system l l safety / relief valves shall be OPERABLE with the specified codo safety valve function lift 1 settings:*#

4 safet relief valves 91170 ps g i3%

5 -safet relief valves 9 1180 ps g i3% ,

5 safet relief valves 9 1190 ps g i3% l APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a. With the safety valve function of one' or more of the above required safety / relief valves ino)erable, be in at least HOT SHUTDOWN withir,12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN wit 11n the next.24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. With one or E, ore safety / relief valves stuck open, provided that suppression pool average water temperature is less than 105'F, close the stuck open safety / relief valve ;ifunabletoclosethestuckopenvalvejs within 2 l minutes or if su) pre (s) ss ion pool average water temperature is 110 F)or greater, place tie reactor mode switch in the Shutdown position.

4 c. With one or more safety / relief valve acoustic monitors inoperable, restore the inoperable acoustic monitors to OPERABLE status within 7 days or be in at least !

HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 1 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .. l SURVEILLANCE REQUIREMENTS l

'4.4'.2.1 The acoustic monitor for each safety / relief valve shall be demonstrated OPERABLE with the setpoint verified to be 0.20 of the full open noise leve1H by performance of a: l

.a. CHANNEL FUNCTIONAL TEST at least once per 92 days, and a

b. . CHANNEL CALIBRATION at least once per 24 months **. ;

)

i 4.4.2.2 At least 1/2 of the safety relief valves shall be removed, set pressure tested and -reinstalled or replaced with spares that have been previously set pressure tested and 4 stored in accordance with manufacturer's recommendations at least once per 24 months, and !

they shall be rotated such that all 14 safety relief valves are removed, set pressure tested and reinstalled or replaced with spares that have been previously set pressure tested and stored in accordance with manufacturer's recommendations at least once per 54 months. All safety valves will be recertification tested to meet a il% toletance prior to returning the valves to service.

The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.

- The provisions of Specification 4.0.4 are not applicable provided the Surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate to perform the. test.

Up to 2 inoperable valves may be replaced with spare OPERABLE valves with lower setpoints until the next refueling

1. Initial setting shall be in accordance with the manufacturer's recommendation. Adjustment to the valve full open noise level shall be accomplished during the startup test program.

. LIMERICK - UNIT 2_ 3/4 4-7

REACTOR COOLANT SYSTEM BASES RECIRCULATION SYSTEM (ContinJed)

Plant specific calculations can be performed to determine an applicable region for monitoring neutron flux noise levels. In this case the degree of conservatism can be reduced since plant to plant veriability would be eliminated. In this case, adequate margin will be assured by monitoring the region which has a decay ratio greater than or equal to 0.8.

Neutron flux noise limits are also estabiished to ensure early detection of limit cycle neutron flux oscillations. BWR cores typically operate with neutron flux noise caused by random boiling and flow noise. Typical neutron flux noise levels of 1-12% of rated power the range of low to high recirculation loop (peak-to-peak)both flow during single and dualhave been repo recirculation. loop operation. Neutron flux noise levels which significant bound these values are considered in the thermal / mechanical design of GE BNl I fuel and are found to be of negligible consequence. In addition stability i testsatoperatingBWRshavedemonstratedthatwhenstabilityrelatedneutron l flux limit cycle oscillations occur they result in peak-to-peak neutron flux i

, limit cycles of 5-10 times the typical values. Therefore, actions taken to l reduce neutron flux noise levels exceeding three (3) times the typical value are sufficient to ensure'early detection of limit cycle neutron flux oscillations. l 1

Typically, neutron flux noise levels show a gradual increase in absolute 2- magnitude as core flow is increased (constant control rod pattern) with two j reactor recirculation loops in operation. Therefore, the baseline neutron flux 1 noise level obtained at a specific core-flow can be applied over a range of I

. core flows. To maintain a reasonable variation between the low flow and high flow end of the flow range, the range over which a specific baseline is applied i should not exceed 20% of rated core flow with two recirculation loops in

. operation. Data from tests and operatino plants indicate that a range of 20%

of rated core flow will result in approximately a 50% increase in neutron flux noise level during operation with two recirculation loops. Baseline data i should be taken near the maximum rod line at which the majority of operation l will occur. However, baseline data taken at lower rod lines (i.e. lower power) ,

will result in a conservative value since the neutron flux noise level is- '

proportional to the power level at a given core flow.

3/4.4.2 SAFETY / RELIEF VALVES l' The safety valve function of the safety / relief valves operates to prevent ,

the reactor coolant system from being pressurized above the Safety Limit of 1325 psig in accordance with the ASME Code. A total of 12 OPERABLE safety / l 4

relief valves is required to limit reactor pressure to within ASME III allow-

- able values for the worst case upset transient.

Demonstration of the safety / relief valve lift settings will occur only during shutdown. The safety pressure tested or replaced w/ relief valves will be removed and either setith spares whic sure tested and stored in accordance with manufacturers recommendations in the specified frequency.

i LIMERICK - UNIT 2 B 3/4 4-2

3/4.5 EMERGENCY CORE COOLING SYST_E_M BASES 3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHUTDOWN The core spray system :

is provided to assure that th(CSS), togethee e core is adequately with the cooled LPCI mode of theaRHR following system, loss-of- '

coolant accident and provides adequate core cooling capacity for all break ,

sizes up to and incluoing the double-ended reactor recirculation line break, 1 and.for smaller breaks following depressurization by the ADS. )

The CSS is a primary source of emergency core cooling after the reactor vessel is depressurized and a source for flooding of the core in case of accidental draining.

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 through a test loop during reactor operation, a complete functional test requires reactor shutdown. The pump discharge piring is maintained full to prevent water hammer damage to

' piping and to s y t cooling at the earliest moment.

The low pressure coolant injection provided to assure that the core is adequa(LPCI) mode of the RHR system istely cooled follow coolant accident. Four subsystems, each with one pump, provide adequate core flooding for all break sizes up to and including the double-ended reactor recirculation line break, and for small breaks following depressurization by

,the ADS. !

The surveillance re I systemwillbeOPERABLEwhuirementsbrovideadequateassurancethattheLPCI en require . Although all active components are l testable and full flow can be demonstrated by recirculation through a test loop 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.

The high pressure coolant injecticn that the reactor core is adequately cooled (HPCI? to 19mitsystem is provided fuel clad to assure temperature in the event of a small break in the reactor coolant system and loss of coolant which does not result in rapid depressurization of the reactor vessel. The HPCI system permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized. The HCPI system continues to operate until reactor vessel pressure is below the

-pressure at which CSS operation or LPCI mode of the RHR system operation maintains core cooling.

The capacity of the system is selected to provide the required core cooling.

The HPCI pump is designed to deliver greater than or equal to 5600 gpm at reactor pressures between 1182 and 200 psig and is capable of delivering at least 5000 gpm between 1182 and 1205 psig. Initially, water from the condensate storage tank is used instead of injecting water from the suppression pool into the reactor, but no credit is taken in the safety analyses for the condensate storage tank water.

LIMERICK - UNIT 2 B 3/4 5-1

i I REACTOR COOLANT SYSTEM -

Q 3 /4.4. 2 SAFETY / RELIEF VALVES

' ' ' LIMITING CONDITION FOR OPERATION 3.4.2 The safet valve function of at least kof the following reactor coolant system safety / relief valves shall be OPERABLE with the specified code safety valve function lift settings:*f 4 safet relief valves 91170 ps g 't "

'[,

5 safet relief valves 91180 ps g + ,

5 safat relief valves 91190 ps g >

APPLICABILITY: . OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

- a. With the safety valve function of one or more of the above required safety / relief valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

With one or more pool average watersafety t / relief valves stuck openrature is less close the than stuck 105 I,provided open safety / relief valve if unable to close the stuck open valve minutes or if suppre(s)ssonpoolaveragewatertemperatureis110{s)within2 F or greater, place tue reactor mode switch in the Shutdown position.

c. With one or more safety / relief valve acoustic monitors inoperable, restore the inoperable acoustic monitors to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following

-24 hours.

SURVEILLANCE REQUIREMENTS The acoustic monitor for each safety / relief valve shall be demonstrated OPERABLE

. 4.4.2.1' with the setpoint verified to be 0.20 of the full open noise levelff by performance of a:

a. CHANNEL FUNCTIONAL TEST at least once per 92 days, and a

b. CHANNEL CALIBRATION at least once per 24 months **.

4.4.2.2 At least 1/2 of the safety relief valves shall be removed, set pressure tested and reinstalled or replaced with spares that have been previously set pressure tested and stored in accordance with manufacturer's recomunendations at least once per 24 months, and e nd tested stored in accordance with manufacturer's recommendahns at least 4encege month . All sa% vakres will be ev.cerWication hsied -te meet a 1. t le Mecance, p6cHe r Woias k V81Ves 40 seWice.

= lift setting pressure shall correspond to ambient conditions of t val - W atin eratu and pres res.

- .The provisions of Speci cation .. ar pi e provided the Surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate ~to perform the test.

'f Up to 2 inoperable valves may be replaced with spare OPERABLE valves with 4

lower Initial setpoints setting shall until' bethe next refuelingith in accordance w the manufacturer's ff recommendation. Adjustment to the valve full open noise level shall be accomplished during the startup test program.

LIMERICK - UNIT 2 3/4 4-7 Amendrer.t No. 2I, 33, 34, 51 FEB1619M

REACTOR COOLANT SYSTEM BASE 5 RECIRCULATION SYSTEM (Continued)

Plant specific calculations can be performed to determine an applicable region for monitoring neutron flux noise levels. In this case the degree of conservatism can be reduced since plant to plant variability would be eliminated. In this case, adequate margin will be assured by monitoring the region which has a decay ratio greater than or equal to 0.8.

Neutron flux noise limits are also established to ensure early detection of limit cycle neutron flux oscillations. BWR cores typically operate with neutron flux noise caused by random boiling and flow noise. Typical neutron flux noise levels of 1-12% of rated power (peak-to peak) have been reported for the range of low to high recirculation loop flow during both single and dual recirculation loop operation. Neutron flux noise levels which significantly bound these values are considered in the thermal / mechanical design of GE BWR fuel and are found to be of negligible consequence. In addition, stability tests at operating BWRs have demonstrated that when stability related neutron flux limit cycle oscillations occur they result in peak-to peak neutrob flux limit cycles of 5-10 times the typical values. Therefore, actions taken to reduce neutron flux noise levels exceeding three (3) times the typical value are sufficient to ensure early detection of limit cycle neutron flux oscillations.

Typically, neutron flux noise levels show a gradual increase in absolute magnitude as core flow is increased (constant control rod pattern) with two reactor recirculation loops in operation. Therefore, the baseline neutron flux noise level obtained at a specific core flow can be applied over a range of core flows. To maintain a reasonable variation between the low flow and high flow end of the flow range, the range over which a specific baseline is applied should not exceed 20% of rated core flow with two recirculation loops in operation. Data from tests and operating plants indicate that a range of 20%

of rated core flow will result in approximately a 50% increase in neutron flux noise level during operation with two recirculation loops. Baseline data sholild be taken near the maximum rod line at which the majority of operation will occur. However, baseline data taken at lower rod lines (i.e. lower power) will result in a conservative value since the neutron flux noise level is proportional to the power level at a given core flow.

I2 3/4.4.2 SAFETY / RELIEF VALVES The safety valve function of the safety / relief valves operates to prevent the reactor coolant system from being pressurized above the Safety Limit of 1325 psig in accordance with the ASME Code. A total of OPERABLE safety /

relief valves is required to limit reactor pressure to within ASME III allow-able values for the worst case upset transient.

Demonstration of the safety / relief valve lift settings will occur only during shutdown. The safety / relief valves will be removed and either set pressure tested or replaced with spares which have been previcusly set pres-sure tested and stored in accordance with manufacturers recommendations in the specified frequency.

LIMERICK - UNIT 2 B 3/4 4-2 146 2 5 1389

3/4.5 DIEREDICY CORE COOLING SYSTDI ma_ers 3/4.E.1 and 3/4.5.2 ECCS - OPERATING and 9Emnai The core spray system (CSS), together with the LPCI mode of the RHR system, is provided to assure that the core is adequately cooled following a loss-of-

" coolant accident and provides adequate core cooling capacity for all break sizes up to and including the double-ended reactor recirculation line break, and for smaller breaks following depressurization by the ADS.

The CSS is a primary source of emertency core cooling after the reactor vessel is depressurized and a source for flooding of the core in case of accidental draining.

The surveillance requirements provide adequate assurance that the CSS will

- be OPERA 8LE when required. Although all active components arp testable and full flow can be demonstrated by recirculation through a test loop during reactor omration, a complete functional test requires reactor shutdown. The pop disciarge piping is maintained full to prevent water hammer daange to piping and to start cooling at the earliest moment.

The low pressure coolant injection (LPCI mode of the RHR system is provided to assure that the core is adequately) cooled following a loss-of-coolant accident. Four subsystems, each with one pump, provide adequate core flooding for all break sizes up to and including the double-ended reactor recirculation line break, and for small breaks following depressurization by the ADS.

. The surveillance requirements provide adequate assurance that the LPCI system will be OPERA 8LE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop 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.

The high pressure coolant injection (HPCI) system is provided to assure that the reactor core is adequately cooled to limit fuel clad temperature in the event of a small break in the reactor coolant system and loss of coolant which does not result in rapid depressurization of the reactor vessel. The HPCI system permits the reactor to be shut down white maintaining sufficient reactor vessel water level inventory until the vessel is depressurized. The HCPI system continues to operate until reactor vessel pressure is below the pressure at which CSS operation or LPCI mode of the RHR system operation maintains core cooling.

The capacity of the system is selected to provide the required core cooling.

The HPCI pump is designed to. deliver reater than or equal to 5600 gpa at reactor pressures between 1182 and 200 psi Initially, water from the condensate storage tank is used instead of injecting ter from the suppression pool into the reactor, but no credit is taken in the safety nalyses for the condensate storage tank water.

nd is capabk o(delivedng a+

leasj sccoggm be{weert iiM md F2.05 psig .

LIMERICK - UNIT 2 B 3/4 5-I = .. - t No. 51 FEB 1 6 25

i l

l l

d i

ATTACHMENT 3 j l

PECO Energy Affidavit for Limerick Generating Station, Unite 1 and 2 Technical Specifications Change Request No. 98-08-0 1

(.

)-'

l COMMONWEALTH OF PENNSYLVANIA :

i

ss. l COUNTY OF CHESTER :

1 i

l J. J. Hagan, being first duly swom, deposes and says:

r That he is Vice President of PECO Energy, the Applicant herein; that he has read the foregoing application for amendment to Facility Operating License Nos. NPF-39 and NPF-85 for Limerick i

Generating Station, Units 1 and 2, conceming Technical Specifications Change Request No. 98-08-0, I l

Increuse Safety Relief Valve Setpoint Tolerance From 11% to 13%," and knows the contents thereof; and ;

l that the statements and matters set forth therein are true and correct to the best of his knowledge, information and belief.

Vi r)sident l

~

Subscribed and swom to ;

before me this /d day 1 of 99.

L()%

. [-

~

Not/ry Public NOTARIAL SEAL CAROL A. WALTON. Notary Pubte City of Philadelphia, Phila. County My Commission Empires May 28. 2001

J

, ENCLOSURE 1 General Electric (GE) Report NEDC-32645P, "Limenck Generating Station Units 1 and 2, SRV Setpoint Tolerance Relaxation Licensing' Report,"

Revision 2, December 1998 PROPRIETARYDOCUMENT

ENCLOSURE 3 General Electric (GE) Report NEDC-32645P,

" Limerick Generating Station Units 1 and 2, SRV Setpoint Tolerance Relaxation Licensing Report" NON-PROPRIETARY VERSION

ML20199G202

Text

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