Text

Proposed Tech Specs Supporting Planned Implementation of Power Rerate Program at Plant
	ML20058L449
Person / Time
Site:	Limerick
Issue date:	12/09/1993
From:	; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	;
Shared Package
ML19311B241	List: ML20058L448; ML20058L449;
References
	NUDOCS 9312160357
	Download: ML20058L449 (59)
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L. ATTACHMENT 2 LIMERICK GENERATING STATION Units 1 and 2 Docket Nos. 50-352 50-353 License Nos. NPF-39 NPF-85 PROPOSED OPERATING LICENSE CHANGES No. 93-24-0 List of Attached Pages E0j_t_1 Unit 2 Operating License (0L), Page 3 Operating License (OL), Page 3 OL, Appendix A OL, Appendix A

" Technical Specifications" " Technical Specifications" Pages: 1-6 Pages: 1-6, 2-4 ,

2-4 ,

3/4 1-20 3/4 1-20 3/4 3-18 3/4 3-18 3/4 3-19 3/4 3-19 3/4 3-20 3/4 3-20 3/4 3-44, 3/4 3-44, 3/4 3-60 3/4 3-60 3/4 4-1 3/4 4-1 3/4 4-2 3/4 4-2 3/4 4-3 3/4.4-3 l 3/4 4-7 3/4 4-7 l

3/4 4-18 (Unit 1 oaly) 3/4 4-20 3/4 4-20 3/4 4-22 3/4 4-22 3/4 5-4 l 3/4 5-4 3/4 6-46 3/4 6-46 3/4 6 ;

3/4 6-54 3/4 7-9 3/4 7-9 B 3/4 4-5 l i

B 3/4 4-5 B 3/4 4-7 B 3/4 4-7 B- 3/4 4-8 B 3/4 4-8 8 3/4 5 B 3/4.5-1 B 3/4 6-1 ;

B 3/4 6-1 B 3/4 6-2, .

B 3/4 6-2,- B 3/4 6-3 i B 3/4 6-3. B 3/4 6-5 B 3/4 6-5 B 3/4 6-6 (Unit 2 only) 5-A 9312160357 931209 T. '

5 -8 PDR ADOCK 05000352 ;-

,p PDR ,-

inese pages rettect propoie'd changes due to APRM Rod Block Monitor Technical. Specifications / Maximum Extended Load Line Limit Analysis (ARTS /MELLLA) implementation submitted to the NRC by letter dated August 27, 1993 prior to Power Rerate.

(3) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, !

possess and use at any time any byproducts, source and special !

nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation .

monitoring equipment calibration, and as fission detectors in l amounts as required; l (4) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or '

special nuclear material without restriction to chemical or !

physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Persuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, ;

but not separate, such byproduct and special nuclear materials as :

may be produced by the operation of the facility, and to receive and possess, but not separate, such source, byproduct, and special nuclear materials as contained in the fuel assemblies and ,

tuel channels from the Shoreham Nuclear Power Station.

C. This license shall be deemed to contain and is subject to the _,

conditions specified in the Commission's regulations set forth in 10 CFR Chapter I (except as exempted from compliance in Section 2.D.

below) and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below i

(1) Maximum Power level ,

E i The licensee is authorized to operate the facility at reactor !

core power levels not in excass of 3458 megawatts thermal (100% l >

rated power) in accordance with the conditions specified herein '

and in Attachment 1 to this license. The items identified in i Attachment 1 to this license shall be completed as specified.

Attachment 1 is hereby inccrporated into this license. '

(2) Technical Specifications t The Technical Specificatior c mt (ned in Appendix A and the [

Environmental Protection P1 r* tained in Appendix B, as revised thrcugh Amendment No. 63, are hereby incorporated in the license. PECo shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

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. DJFINITIONS ,

PURGE - PURGING 1.31' PURGE or PURGING shall be the controlled process of discharging air or ;

gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a mannt ' hat replacement air or gas is required to purify the confinement, j i

RATED THERMAL POWER 1.32 RATED THERMAL POWER :h11 be a total reactor core heat transfer rate to the reactor coohat of 3458 MWt. l ;

REACTOR ENCLOSURE SiCONDARY CONTAINMENT INTEGRITY 1.33 REACTOR ENCLISURE SECONDARY CONTAINMENT INTEGRITY shall exist when: [

I

a. All reactor enclosure secondary containment penetrations required to Fe closed during accident conditions are either: ;

1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or a

2. Closed by at least one manual valve, blind flange, slide gate l damper, or deactivated automatic valve secured in its closed ;

position, except as provided in Table 3.6.5.2.1-1 of i Specification 3.6.5.2.1.

b. All reactor enclosure secondary containment hatches and blowout panels t are closed and sealed. !

c. The standby gas treatment system is in compliance with the requirements ,

of Specification 3.6.5.3. j

d. The reactor enclosure recirculation system is in compliance with the !

requirements of Specification 3.6.5.4.

e. At least one door in each access to the reactor enclosure secondary i containment is closed. j

f. The sealing mechanism associated with each reactor enclosure secondary containment penetration, e.g., welds, bellows, or 0-rings, is OPERABLE.

9 The pressure within the reactor enclosure secondary containment is ;

less than or equal to the value required by Specification 4.6.5.1.la. l REACTOR PROTECTION SYSTEM RESPONSE TIME ,

1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel :

sensor until de-energization of the scram pilot valve solenoids. The l response time may be measured by any series of sequential, overlapping '

or total steps such that the entire response time is measured.

REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY-1.35 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY shall exist when:

a. All refueling floor secondary containment penetrations required to ,

be closed during accident conditions are either:

LIMERICK - UNIT 1 1-6

TABLE 2.2.1-1

[ REACTOP PROTECTION SYSTEM INSTRUMENTATION SETPOINTS g ALLOWABLE 5 FUNCTIONAL UNIT 1 RIP SETPOINT VALUES l p 1. Intermediate Range Monitor, Neutron Flux-High s 120/125 divisions s 122/125 divisions

, of full scale of full scale c 2. Average Power Range Monitor:

z a. Neutron Flux-Upscale, Setdown s 15% of RATED THERMAL POWER s 20% of RATED H THERMAL POWER

~ b. Neutron Flux-Upscale

1) During two recirculation loop operation:

a) Flow Biased s 0.66 W+ 62%, with $ 0.66 W+ 64%, with l

, a maximum of a maximura of b) High Flow Clamped s 115% of RATED $ 117% of RATED l i THERMAL POWER THERMAL POWER ,

2) During single recirculation loop operation:

a) Flow Biased s 0.66 W+ 57%, s 0.66 W+ 59%, l b) High Flow Clamped Not Requirea Not Required OPERABLE OPERABLE

c. Inoperative N.A. N.A.

7' d. Downscale 2 4% of RATED 2 3% of RATED

THERMAL POWER THERMAL POWER -

3. Reactor Vessel Steam Dome Pressure - High s 1096 psig s 1103 psig l.

4. Reactor Vessel Water Level - Low, level 3 2 12.5 inches above instrument 2 11.0 inches above ,

zero* instrument zero

5. Main Steam Line Isolation Valve - Closure s 8% closed s 12% closed

6. Main Steam Line Radiation - High 5 3.0 x full power background s 3.6 x full power background

7. Drywell Pressure - High s 1.68 psig 5 1.88 psig ,

8. Scram Discharge Volume Water Level - High

a. Level Transmitter s 260' 9 5/8" elevation ** s 261' 5 5/8" elevation !

b. Float Switch 5 260' 9 5/8" . elevation ** s 261' 5 5/8" elevation !'

9. Turbine Stop Valve - Closure s 5% closed s 7% closed

10. Turbine Control Valve Fast Closure, Trip 011 Pressure - Low 2 500 psig 2 465 psig

11. Reactor Mode Switch Shutdown Position- N.A. N.A. !

12. Manual Scram- N.A. N.A.

See Bases Figure B 3/4.3-1.

- Equivalent to-25.45 gallons / scram discharge volume.

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

?

REACTIVITY CONTROL' SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

' b. At least once per 31 days by: !

1. Verifying the continuity of the explosive charge. I

2. Determining by chemical analysis and calculation

that the ,

available weight of sodium pentaborate is greater than or equal !

to 3754 lbs; the concentration of sodium pentaborate in solution is less than or equal to 13.8% and within the limits '

of Figure 3.1.5-1 and; the following equation is satisfied:

t C E O

EI 13% wt. 29 atom % 86 gpm l

where C = Sodium pentaborate solution (% by weight) i Q = Two pump flowrate,. as determined per j surveillance requirement 4.1.5.c. i E = Boron 10 enrichment (atom % Boron 10)

3. 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, j

c. Demonstrating that, when tested pursuant to Specification 4.0.5, the _

minimum flow requirement of 41.2 gpm per pump at a pressure of greater j than or equal to 1230 25 psig is met. l

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d. At least once per 18 months during shutdown by:

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1. Initiating at least one of the standby liquid control system loops, including an explosive valve, and verifying that a flow path from the pumps to the reactor pressure vessel is available. :

by pumping demineralized water into the reactor vessel. The .

replacement charge for the explosive valve shall be from the 'l same manufactured batch as the one fired.or from another batch t which has been certified by having one of that batch success- '

fully fired. All injection loops shall be tested in 3 operating cycles. *

2. Verify all heat-traced piping between storage tank and pump suction is unblocked.** ;

e. Prior to addition of Boron to storage tank verify sodium pentaborate !

enrichment to be added is 2 29 atom % Boron 10. ,

1 !

oThis test shall also be performed anytime water or boron is added to the solu- .i tion or when the solution temperature drops below the limits of Figure 3.1.5-1 !

for the most recent concentration analysis, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after water or beror, i addition or solution temperature is restored.

ooThis test shall also be performed whenever suction piping temperature drops below the limits of Figure 3.1.5-1 for the most recent concentration analysis, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after solution temperature is restored. t LIMERICK - UNIT 1 3/4 1-20

-TABLE 3.3.2-2 .

~ ISOLATION ACTUATION INSTRUMENTATION SETPOINTS C ALLOWABLE M TRIP FUNCTION TRIP SETPOINT VALUE 2

'Q l. MAIN STEAM LINE ISOLATION

a. Reactor Vessel Water. Level

-5 1) Low, low - Level 2 2 - 38 inches

2 - 45 inches-

2) ' Low, Low, low - Level 1 2 - 129 inches

2 - 136-inches

b. Main' Steam Line s 3.0 x Full Power s 3.6 x Full Power Radiation - liigh Background. Background

-c. Main' Steam Line 2 756 psig 2 736 psig Pressure - Low.

d. Main Steam Line Flow - liigh s 122.1 psid s 123 psid l' i e. Condenser-Vacuum - Low 10.5 psia 210.1 psia /s 10.9 psia

. w

.L f.' Outboard MSIV Room

. Temperature - liigh 5 192*F s 200*F i

g. Turbine Enclosure.- Main Steam Line Tunnel Temperature - liigh s 165'F- s 175 F

, h. Manual initiation' N.A. N.A.

2. RilR SYSTEM SHUTDOWN COOLING MODE ISOLATION

a. React'or Vessel Water Level '

' Low - Level.3 2 12.5 inches

2 11.0 inches

' b. Reactor Vessel'(RilR Cut-in Permissive) Pressure - liigh s 75'psig s 95 psig 1

c.. Manual Initiation N.A. N.A.

i.

' _ _ _ _ _ __._ _ _ _ _ _ .m___-.______.m.__m..,_,_.m__ , . . . . . . . , . . . . , , . . . ~ , . . . . - , . . _ _ . . . . , , . . _ . . . . . . , . . . . . . . . . . . , , , . . , . _

_ _ _ __.__r,., ., ., . . . . _ , ,

TABLE 3.3.2-2 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SETP0 tilts ALLOWABLE C TRIP FUTICTION TRIP SETP0 lilt VALUE E

5 3. REACTOR WATER CLEANUP SYSTEM ISOLATION R

, a. RWCS A Flow - liigh 5 54.9 gpm s 65.2 gpm C

5 w

b. RWCS Area Temperature - liigh s 142 F or 132*F** $ 147 F or 137 F** l

~

c. RWCS Area Ventilation A Temperature - High 5 32*F s 40*F

d. SLCS Initiation N.A. N.A.

e. Reactor Vessel Water Level -

Low, Low, - Level 2 2 -38 inches

2 -45 inches

f. Manual Initiation fl . A . fl . A .

4. HIGH PRESSURE C00LAfiT INJECTION SYSTEM ISOLATION

a. HPCI Steam Line a Pressure - High s 974" H O 2 s 984" H 2O l

b. HPCI Steam Supply Pressure - Low 2 100 psig 2 90 psig

c. HPCI Turbine Exhaust Diaphragm Pressure - High s 10 psig s 20 psig 4' d. HPCI Equipment Room Temperature - High 175*F 2 165 F, s 200*F

e. HPCI Equipment Room a Temperature - liigh s 80 F 5 88 F

f. HPCI Pipe Routing Area Temperature - High 175 F 2 165 F, s 200 F

g. Manual Initiation ti . A . fl . A .

h. HPCI Steam Line a Pressure - Timer 3 s r s 12.5 seconds 2.5 s 7 s 13 seconds

TABLE 3.3.2-2 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SETPOINTS C ALLOWABLE' TRIP FUNCTION TRIP SETPOINT VALUE E 5. REACTOR' CORE ISOLATION COOLING SYSTEM ISOLATION t

e a. RCIC Steam Line a

.} Pressure - High s 373" H,0 s 381" 'li,0 ll-

b. RCIC Steam Supply Pressure - Low 2 64.5 psig 2 56.5 psig

c. RCIC Turbine Exhaust Diaphragm Pressure - High 5 10.0 psig s 20.0 psig

d. RCIC Equipment Room Temperature - High 175'F. 2 165 F, s 200*F

e. RCIC Equipment Room a Temperature High s 80*F .s 88 F.

w f. RCIC Pipe Routing Area g Temperature - High 175'F 2 165*F,-'s 200*F

.g. Manual Initiation- N.A. N.A.

h. RCIC Steam Line a. Pressure Timer 3 s r~s 12.5 seconds 2.5 s r s 13-seconds-I

. . _ . _ . _ , . _ _ . . _ _ _ _ , . . . . ._.._-,__-..;.L.,- . , _ - . . , . . . - - . _ .._..a_..._.. -. . _ . . . . . . - . -

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

..._.1._.. . . _ _ _ _ _ _ _ _ _ _ _ _

TABLE 3.3.4.1-2 i

ATWS RECIRCULAT10N PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS TRIP ALLOWABLE TRIP FUNCTION SETPOINT VALUE

1. Reactor Vessel, Water Level -

Low Low, Level 2 a -38 inches * = -45 inches ;

2. Reactor Vessel Pressure - High s 1149 psig 5 1156 psig [

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See Bases Figure B3/4 3-1. l r

LIMERICK - UNIT 1 3/4 3-44

TABLE ~3.3.6-2 CONTROL R0D BLOCK INSTRUMENTATION SETPOINTS TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE r- 1. ROD BLOCK MONITOR E a. Upscale")

9 1)- Low Trip Setpoint (LTSP) * *

2) Intermediate Trip Setpoint (ITSP)

M

3) High Trip Setpoint (HTSP)

{ *

t si b. Inoperative N/A N/A ;

c. *

Z Downscale (DTSP)

. d. Power Range Setpoint N

1) Low Pcwer Setpoint (LPSP) 23% RATED THERMAL POWER 26% RATED TilERMAL POWER

2) Intermediate Power Setpoint (IPSP) 58% RATED THERMAL' POWER 61% RATED THERMAL POWER

3) High Power Setpoint (HPSP) 78% RATED THERMAL POWER 81% RATED THERMAL POWER

2. APRM

a. Flow Biased Neutron Flux - Upscale

1) During two recirculation loop s 0.66 W + 55% s 0.66 W + 59% l operation

2) During single recirculation loop $ 0.66 W + 50% $ 0.66 W + 54% l:

R operation

b. Inoperative N.A. N.A.

Y c. Downscale- 2 4% of RATED THERMAL POWER 2 3% of RATED THERMAL POWER

- g; d. Neutron Flux - Upscale, Startup s 12% of RaiED THERMAL POWER s 14% of RATED THERMAL POWER l 3. SOURCE RANGE MONITORS

a. Detector not full in N.A. N.A.

b. Upscale- 51 x 10' cps s 1.6 x 10' cps

c. Inoperative N.A. N.A.

d. Downscale ~ 2 3 cps ** 2 1.8 cps **

4. INTERMEDIATE RANGE MONITORS

a. Detector not full in N.A. N.A.

b. Upscale $ 108/125 divisions of s 110/125 divisions of full scale full scale l

c. Inoperative N.A. N.A.

l d. ~Downscale 2 5/125 divisions of full 2 3/125 divisions of full scale scale ,

5. SCRAM DISCHARGE VOLUME a, Water Level-High s 257' 5 9/16" elevation *** s 257' 7 9/16" elevation

a. Float Switch l.

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

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECTRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION l 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation i with: ,

a. Total core flow greater than or equal to 45% of rated core flow, or

b. THERMAL POWER within the unrestricted zone of Figure 3.4.1.1-1.

APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*.

ACTION:

a. With one reactor coolant system recirculation loop not in operation:

1. Within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s:

a. Place the recirculation flow control system in the Local i Manual mode, and

b. Reduce THERMAL POWER to 5 76.2% of RATED THERMAL POWER, and, l-I

c. Limit the speed of the operating recirculation pump to less :

than or equal to 90% of rated pump speed, and

d. Verify that the differential temperature requirements of ;

Surveillance Requirement 4.4.1.1.5 are met if THERMAL i POWER is 5 30% of RATED THERMAL POWER or the recirculation i loop flow in the operating loop is s 50% of rated loop i flow, or suspend the THERMAL POWER or recirculation loop i flow increase. .

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See Special Test Exception 3.10.4.

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4 LIMERICK - UNIT 1 3/4 4-1

REACTOR COOLANT SYSTER 1 SURVEILLANCE REQUIREMENTS 4.4.1.1.1 Each pump discharge valve shall be demonstrated OPERABLE by cycling i each valve through at least one complete cycle of full travel during each ;

startup* prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER. j 4.4.1.1.2 Each pump MG set scoop tube mechanical and' electrical stop shall be demonstrated OPERABLE with overspeed setpoints less than or equal to 109% and !

'107%, respectively, of rated core flow, at least once per 18 months.

4.4.1.1.3 Establish a baseline APRM and LPRM** neutron flux noise value within the regions for which monitoring is required (Specification 3.4.1.1, ACTION c) '

eithin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of entering the region for which monitoring is required unless baselining has previously been performed in the region since the last refueling ;

outage.

4.4.1.1.4 With one reactor coolant system recirculation loop not in operation, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months verify that:

a. Reactor THERMAL POWER is s 76.2% of RATED THERMAL POWER, l

b. The recirculation flow control system is in the Local Manual mode, and

c. The speed of the operating recirculation pump is s 90% of rated pump ~!

speed.

d. Core flow is greater than 39% when THERMAL POWER is within the !'

restricted zone of Figure 3.4.1.1-1.

4.4.1.1.5 With one reactor coolant system recirculation loop not in operation, within 15 minutes prior to either THERMAL POWER increase or recirculation loop flow increase, verify that the following differential temperature requirements are met if THERMAL POWER is s 30% of RATED THERMAL POWER or the recirculation loop flow in the operating recirculation loop is 5 50% of rated loop flow:

a. s 145 F between reactor vessel steam space coolant and bottom head ',

drain line coolant, -

b. s 50 F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel, and

c. 5 50 F between the reactor coolant within the loop not in operation ,

and the operating loop.

The differential temperature requirements of Specification 4.4.1.1.5b. and c.

do not apply when the loop not in operation is isolated from the reactor pressure vessel.

off not performed within the previous 31 days. [

" Detector levels A and C of one LPRM string per core octant plus detectors A and C of one LPRM string in the center of the core should be monitored.

LIMERICK - UNIT 1 3/4 4-2 p

[

r-5 70-5 n

X i 60 c Restricted h Zone 50

% / 45.8% Power

[ 40 g r /

M 33.5% Power A 2 30 i

u E Unrestricted Zone O

5 20 10 0

30~ 40 50 60 70 Core Flow (8 Rated)

Figure 3.4.1.1 - 1 THERMAL POWER VERSUS CORE FLOW

. o wc.w---,e,- -i-+% --m=w+w+.e+e< e=**+t-+- -----n e =m-*- n w n= w e --<'*---vw e e -*~w. r =w< e e - e +e e,.~ _ w%= -.r-=- = - +mw-.s--,-_-- ----- - - w ---___*_a ---

REACTOR COOLANT SYSTEM i

3/4.4.2 SAFETY /REllEF VALVES LIMITING CONDITION FOR OPERATION 3.4.2 The safety valve function of at least 11 of the following reactor coolant system safety / relief valves shall be OPERABLE with the specified code safety valve function lift settings:*#

4 safety / relief valves @ 1170 psig $1% '

5 safety / relief valves 01180 psig 21%

5 safety / relief valves @ 1190 psig 21%

MPLICABILITY: 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 H0T 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 pool average water temperature is less than 105 F, close the stuck open safety / relief valve (s) if unable to close the stuck open valve (s within 2 minutes or if suppress;on pool average water temperature is 110 F)or i

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 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 level ## by performance of a:

a. CHANNEL FUNCTIONAL TEST at least once per 31 days, and a

b. CHANNEL CALIBRATION at least once per 18 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 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.

The lif t 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 1 3/4 4-7 !

. REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM QM,ITINGCONDITIONFOROPERATION 3.4.6.1 The reactor coolant system temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4.6.1-1 (1) curve A and A' for .l hydrostatic or leak testing; (2) curve B for heatup by non-nuclear means, cool-down following a nuclear shutdown and low power PHYSICS TESTS; and (3) curve C for operations with a critical core other than low power PHYSICS TESTS, with:

a. A maximum heatup of 100 i in any 1-hour period,

b. A maximum cooldown of 100 F in any 1-hour period, '

c. A maximum temperature change of less than or equal to 20 F in any 1-hour period during inservice hydrostatic and leak testing opera- i tions above the heatup and cooldown limit curves, and

d. The reactor vessel flange and head flange temperature greater than or equal to 80 F when reactor vessel head bolting studs are under tension.

APPLICABILITY: At all times. i ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the reactor coolant system; determine that the reactor coolant system remains acceptable for continued operations or 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 following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE RE0VIREMENTS ;

4.4.6.'l.1 During system heatup, cooldown and inservice leak and hydrostatic !

testing operations, the reactor coolant system temperature and pressure shall be determined to be within the above required heatup and cooldown limits and to the right of the limit lines of Figure 3.4.6.1-1 curve A and A', B, or C as applicable, l at least once per 30 minutes.

LIMERICK - UNIT 1 3/4 4-18 r

1

i l

1

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BOLTUP CURVES A*, B. C ARE VAUD j so*F , / FOR 12 EFPY OF OPERATION ' ;

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CURVE A IS VAUD

. FOR 9 EFPY OF OPERATION 0 , j i i O 100 200 300 400 500 600 ,

MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F)

MINIMUM REACTOR VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE i FIGURE 3.4.6.1-1 ,

i UMERICK- UNIT 1 3/4 4-20 i

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REACTOR COOLANT SYSTEM

. REACTOR STEAM DOME LIMITING CONDITION FOR OPERATION 3.4.6.2 The pressure in the reactor steam dome shall be less than:1053 psig. l APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*.

ACTION: .

With the reactor steam dome pressure exceeding 1053 psig, reduce the pressure to less than 1053 psig within 15 minutes or be in at least HOT SHUTDOWN within ,

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.6.2 The reactor steam dome pressure shall be verified to be less than 1053 psig at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . l.

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ONot applicable during anticipated transients. (

i t

LIMERICK - UNIT 1 3/4 4-22 .

t 9

EMERGENCY CORE'C00 LING SYSTEMS SURVEILLANCE RE0UIREMENTS 4.5.1 The emergency core cooling systems shall be demonstrated OPERABLE by:

a. At least once per 31 days: -

1. For the CSS, the LPCI system, and the HPCI system:

a) Verifying by venting at the high point vents that the ;

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

i b)

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

otherwise secured in position, is in its correct

position.

2. For the LPCI system, verifying that both LPCI system subsystem cross-tie valves (HV-51-182 A, B) are closed with power removed from the valve operators.

3. For the HPCI system, verifying that the HPCI pump flow controller is in the correct position. ,

4. For the CSS and LPCI system, performance of a CHANNEL FUNCTIONAL TEST of the injection header 4P instrumentation.

b. Verifying that, when tested pursuant to Specification 4.0.5: ;

1. Each CSS pump in each subsystem develops a flow of at least 3175 gpm against a test line pressure corresponding to a reactor vessel to primary containment differential pressure of a 105 psid .

plus head and line losses. ';

2. Each LPCI pump in each subsystem develops a flow of at least 10,000 gpm against a test line pressure corresponding to a reactor vessel to primary containment differential pressure of a 20 psid plus head and line losses.

3. The HPCI pump develops a flow of at least 5600 gpm against a test line pressure which corresponds to a reactor vessel pressure of 1040 psig plus head and line losses when steam is being supplied to the turbine at 1040, +13, -120 psig.**

c. At least once per 18 months: ;

1. For the CSS, the LPCI system, and the HPCI system, 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 t path actuates to its correct position. Actual injection of i coolant into the reactor vessel may be excluded from this test. t o Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in position for another mode of operation.

co 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. If OPERABILITY is not successfully demonstrated within the 12-hour period, reduce reactor steam dome pressure to less than 200 psig within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

LIMERICK - UNIT 1 3/4 5-4

CONTAINMENT SYSTEMS 1/4.6.5 SECONDARY CONTAINMENT REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY ,

LIMITING CONDITION FOR OPERATION 3.6.5.1.1 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. i LCTION: l Without REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY, restore REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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. j SURVEILLANCE RE0VIREMENTS 4.6.5.1.1 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall be demon- l strated by: f

a. Verifying at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months that the pressure within the

reactor enclosure secondary containment is greater than or equal to 0.25 inch of vacuum water gauge.

b. Verifying at least once per 31 days that: [

1. All reactor enclosure secondary containment equipment hatches and blowout panels are closed and sealed.

2. At least one door in each access to the reactor enclosure secondary containment is closed.

3. All reactor enclosure secondary containment penetrations not capable of being closed by OPERABLE secondary containment auto- r matic isolation dampers / valves and required to be closed during '

accident conditions are closed by valves, blind flanges slide gate dampers or deactivated automatic dampers / valves sec,ured in .

position. '

c. At least once per 18 months:

1. Verifying that one standby gas trei. ment subsystem will draw down ,

the reactor enclosure secondary containment to greater than or ;

equal to 0.25 inch of vacuum water gauge in less than or equal to 126 seconds with the reactor enclosure recirc system in operation, l and >

2. Operating one standby gas treatment subsystem for one hour and i maintaining greater than or equal to 0.23 inch of vacuum water ;

gauge in the reactor enclosure secondary containment at a flow .

rate not exceeding 1250 cfm with wind speeds of s 7.0 mph as measured on the wind instrument on Tower 1 elevation 30' or, l if that instrument is unavailable, Tower 2, elevation 159'. ,

i LIMERICK - UNIT 1 3/4 6-46 i

CONTAINMENT SYSTEMS SURVETtlANCE RE0VIREMENTS (Continued)

2. Verifying that the fan starts and isolation valves necessary to i draw a suction from the refueling area or the reactor enclosure recirculation discharge open on each of the following test signals:

a) Manual initiation from the control room, and b) Simulated automatic initiation signal.

3. Verifying that the temperature differential across each heater t is a 15 F when tested in accordance with ANSI N510-1980. .

1

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the inplace penetration i and leakage testing acceptanca criteria of less than 0.05% in accordance with ANSI N510-1980 while operating the system at a flow ,

rate of 3000 cfm 10%.

f. After each complete or partial replacement of a charcoal adsorber ,

bank by verifying that the charcoal adsorber bank satisfies the inplace penetration and leakage testing acceptance criteria of less than 0.05%

in accordance with ANSI N510-1980 for a halogenated hydrocarbon ~+

refrigerant test gas while operating the system at a flow rate of 3000 cfm : 10%. ;

i

g. After any major system alteration:

1. Verify that when the SGTS fan is running the subsystem flowrate !

is 2800 cfm minimum from each reactor enclosure (Zones I and i II) and 2200 cfm minimum from the refueling area (Zone III). '

2. Verify that one standby gas treatment subsystem will drawdown ,

reactor enclosure Zone I secondary containment to greater than or equal to 0.25 inch of vacuum water gauge in less than or equal to 126 seconds with the reactor enclosure recirculation l system in operation and the adjacent reactor enclosure and '

refueling area zones are in their isolation modes. ,

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' LIMERICK - UNIT 1 3/4 6-54 !

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PLANT SYSTEMS 3 /4. 7.3 REACTOR CORE ISOLATION COOLING SYSTEM _j LIMITING CONDITION FOR OPERATION 3.7.3 The reactor core isolation cooling (RCIC) system shall be OPERABLE with !

an OPERABLE flow path capable of automatically taking suction from the !

suppression pool and transferring the water to the reactor pressure vessel.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3 with reactor steam dome pressure greater than 150 psig. ;

ACTION: !

a. With the RCIC system inoperable, operation may continue provided the !

HPCI system is OPERABLE; restore the RCIC system to OPERABLE status within !

14 days. Otherwise, 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 i and reduce reactor steam dome pressure to less than or equal to 150 psig within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. In the event the RCIC system is actuated and injects water into the reactor coolant system, a Special Report shall be prepared and sub- ;

mitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total i accumulated actuation cycles to date. ,

SURVEILLANCE REQUIREMENTS 4.7.3 The RCIC system shall be demonstrated OPERABLE: !

a. At least once per 31 days by:

1. Verifying by venting at the high point vents 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. )

3. Verifying that the pump flow controller is in the correct position, l

b. At least once per 92 days by verifying that the RCIC pump develops a ,

flow of greater than or equal to 600 gpm in the test flow path with 1 a system head corresponding to reactor vessel operating pressure when steam is being supplied to the turbine at 1040, + 13, - 120 psig.* l !

i o 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. If OPERABILITY is not successfully demonstrated ,

within the 12-hour period, reduce reactor steam pressure to less than' 150 t psig within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . [

LIMERICK - UNIT.1 3/4 7-9 :

REACTOR COOLANT SYSTEM ;

' BASES PRESSURE / TEMPERATURE ~ LIMITS (Continued)

The operating limit curves of Figure 3.4.6.1-1 are derived from the fracture ,

toughness requirements of 10 CFR 50 Appendix G and ASME Code Section III, ;

Rppendix G. The curves are based on the RTer and stress intensity factor j information for.the reactor vessel components. Fracture toughness limits and i the basis.for compliance are more fully discussed in FSAR Chapter 5, Para- !

graph 5.3.1.5, " Fracture Toughness." ;

The reactor vessel materials have been tested to determine their initial I RTer. The results of these tests are shown in Table B 3/4.4.6-1. Reactor operation and resultant fast neutron, E greater than 1 MeV, irradiation will '

cause an increase in the RTet. Therefore, an adjusted reference temperature, based upon the fluence, nickel content and copper content of the material in question, can be predicted using Bases Figure B 3/4.4.6-1 and the recommenda-tions of Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials." The pressure / temperature limit curve, figu N 3.4.6.1-1, ,

curve A includes a shift in RTer for conditions at 9 EFPY. The A', B and C l.

limit curves are predicted to be bounding for all areas-of the RPV until 12 EFPY !

Twhen the beltline material's RTer will shift due to neutron fluence and :

the beltline curves will intersect the non-beltline discontinuity curves. !

The actual shift in RTet of the vessel material will be established period- ;

ically during operation by removing and evaluating, in accordance. with 10 CFR -

Part 50, Appendix H, irradiated reactor vessel flux wire and Charpy specimens ,

installed near the inside wall of the reactor vessel in the core area. Since the !

neutron spectra at the flux wires, Charpy specimens and vessel inside radius are i essentially identical, the irradiated Charpy specimens can be used with confi- 'i dence in predicting reactor vessel material transition temperature shift. -The ;

- operating limit curves of Figure 3.4.6.1-1 shall be adjusted, as required, on the basis of the flux wire and Charpy specimen data and recommendations of Regulatory Guide 1.99, Revision 2.

The pressure-temperature limit lines shown in Figures 3.4.6.1-1, curves-C, and A and A', for reactor criticality and for inservice leak and hydrostatic testing '

l have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR Part 50 for reactor criticality and for inservice leak and hydrostatic testing. ;

The number of reactor vessel irradiation surveillance capsules and the' frequencies for removing and testing the specimens in these capsules are pro-l vided in Table 4.4.6.1.3-1 to assure compliance with the requirements of Appendix H to 10 CFR Part 50. .

I l

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LIMERICK - UNIT 1 B 3/4 4-E i

BASES TABLE B 3/4.4.6-1 REACTOR VESSEL TOUGHNESSo l

HEAT / SLAB MIN. UPPER C BELTLINE WELD SEAM I.D. OR STARTING SHELF g COMPONENT OR MAT'l TYPE HEAT / LOT CU (%) Ni (%) RTwot ('F) ARTwo t * * ( ' F ) (LFT-LBS) ART (*F)'

j Plate SA-533 Gr B CL. 1 C 7677-1 .11 .5 +20 +69 NA +89 i Weld AB(Field Weld) 640892/ .09 1.0 -60 +114 NA +54 g J424B27AE a

~

NOTES:

Based on 110% of original rated power.

- These values are given only for the benefit of calculating the end-of-life (E0L/32 EFPY) RTwot NON-BELTLINE MT'l TYPE OR ilEAT/ SLAB OR HIGilEST STARTING COMPONENT WELD SEAM I.D. HEAT / LOT RTwot (*F)

Shell Ring SA 533, Gr. B, CL. 1 C7711-1 +20 Bottom llead Dome C7973-1 +12

Bottom Head Torus C7973-1 +12 w

1 Top ilead Dome A6834-1 +10 Top Head Torus B1993-1 +10

[

Top ilead Flange SA-508, CL. 2 123B195-289 0 Vessel Flange "

2Vl924-302 -30 Feedwater Nozzle Q2Q22W-412 -10 l Weld Non-Beltline All 0 l

LPCI Nozzle *** SA-508, CL. 2 Q2Q25W -6 l Closure Studs SA-540, Gr. B-24 All Meet requirements of 45 ft-lbs and 25 mils Lat. Exp. at +10'F Note: *** The design of the LPCI nozzles results in their experiencing an E0L fluence in excess of 10" N/Cm which predicts an EOL (32 EFPY) RTwot of +42*F. e

_ ._.___.._.,____u__..._.._.~.__ , _ . . , _ . _ . _ - _ . _ . . _ _ - _ _ _ _ - . ._ _ . _ ~ . - - _ . _ . . __ ._ .. _e

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Service Life (Years *) !

i BASES FIGURE B 3/4.4.6-1 FAST NEUTRON FLUENCE (E)1 tieV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE * . !

'* At' 90% of Rated Thermal Power and 90% availability I i

LIMERICK - UNIT 1 B 3/4 4-8 ft

i 3/4.5 EMERGENCY CORE COOLING SYSTEM-BASES I I

3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHUTDOWN 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 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 i 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 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 i 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.

l 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 during reactor operation, a complete functional test requires reactor i 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 provide 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 t 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. Initially, water from the condensate storage l tank is used instead of injecting water from the suppression pool into the reactor, but no credit is taken in the safety analyses for condensate storage tank *

. water. l LIMERICK - UNIT 1 B 3/4 5-1

~

3/4.6 CONTAINMENT SYSTEMS 1

BASES 3/4.6.1 PRIMARY CONTAINMENT l

3/4.6.1.1 PRIMARY CONTAINMENT INTEGRITY l l

PRIMARY CONTAINMENT INTEGRITY ensures that the release of radioactive mate- l rials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation, will limit the SITE B0UNDARY radiation doses to within the limits of 10 CFR Part 100 during accident conditions. !

3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE !

The limitations on primary containment leakage rates ensure that the total containment leakage solume will not exceed the value calculated in the safety i analyses for the peak accident pressure of 5 44 psig, Pa. As an added conserva-tism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L. during performance of the periodic tests to account for possible degradation of the containment leakage barriers between leakage tests.

Operating experience with the main steam line isolation valves has indicated that degradation has occasionally occurred in the leak tightness of

-the valves; therefore the special requirement for testing these valves.

The surveillance testing for measuring leakage rates is consistent with the requirements of Appendix J of 10 CFR Part 50 with the exception of exemptions granted for leak testing of the main steam isolation valves, the i airlock and TIP shear valves.

3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCK The limitations on closure and leak rate for the primary containment air lock are required to meet the restrictions on PRIMARY CONTAINMENT INTEGRITY and the primary containment leakage rate given in Specifications 3.6.1.1 and 3.6.1.2. The specification makes allowances for the fact that there may be long periods of time when the air lock will be in a closed and secured position during reactor operation. Only one closed door in the air lock -

is required to maintain the integrity of the containment.

3/4.6.1.4 MSIV LEAKAGE CONTROL SYSTEM :

Calculated doses resulting from the maximum leakage allowance for the main steamline isolation valves in the postulated LOCA situations would be a small fraction of the 10 CFR Part 100 guidelines, provided the main steam line system ,

from the isolation valves up to and including the turbine condenser remains !

intact. Operating experience has indicated that degradation has occasionally occurred in the leak tightness of the MSIVs such that the specified leakage requirements have not always been maintained continuously. The requirement for the leakage control syrtem will reduce the untreated leakage from the MSIVs when isolation of the primary system and containment is required. l

}

LIMERICK - UNIT 1 B 3/4 6-1 t - "" - ~ -

CONTAINMENT SYSTEMS :

BASES 3/4.6.1.5 PRIMARY CONTAINMENT STRUCTURAL INTEGRITY This limitation' ensures that the structural integrity of the containment .

will be maintained comparable to the original design standards for the life of '

the unit. . Structural integrity is required to ensure that the containment will t$ithstand the maximum calculated pressure in the event of a LOCA. A visual l H inspection in conjunction with Type A leakage tests is sufficient to demonstrate -

this capability. j 3/4.6.1.6 DRYWELL AND SUPPRESSION CHAMBER INTERNAL PRESSURE The limitations on drywell and suppression chamber internal pressure ensure that the calculated containment peak pressure does not exceed the design l pressure of 55 psig during LOCA conditions or that the external pressure differ-ential does not exceed the design maximum external pressure differential of.

5.0 psid. The limit of - 1.0 to + 2.0 psig for initial containment pressure l will limit the total pressure to 5 44 psig which is less than the design l

~

pressure and is consistent with the safety analysis.

3/4.6.1.7 DRYWELL AVERAGE AIR TEMPERATURE i

The limitation on drywell average air temperature ensures that the con-tainment peak air temperature does not exceed the design temperature of 340 F ;

during steam line break conditions and is consistent with the safety analysis.

3/4.6.1.8 DRYWELL AND SUPPRESSION CHAMBER PURGE SYSTEM The drywell and suppression chamber purge supply and exhaust isolation valves are required to be closed during plant operation except as required for inerting, deinerting and pressure control. The 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months per 365 day limit on purge valve operation is imposed to protect the integrity of the SGTS filters.

Analysis indicates that should a LOCA occur while this pathway is being utilized.

- the associated pressure surge through the (18 or 24") purge lines will- adversely affect the integrity of SGTS. This limit is not imposed, however, on the subject valves when pressure control is being performed through the 2-inch bypass line, since a pressure surge through this line does not threaten the OPERABILITY of SGTS.

7 e

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LIMERICK . UNIT 1 B 3/4 6-2 I

CONTA1NMENT SYSTEMS -

BASES j

3 /4. 6. 2 DEPRESSURIZATION SYSTEMS The specifications of this section ensure that the primary containment !

3ressure will not exceed the design pressure of 55 psig during primary system alowdown from full operating pressure.

The suppression chamber water provides the heat sink for the reactor coolant system energy release following a postulated rupture of the system. The suppression chamber water volume must absorb the associated decay and structural '

sensible heat released during reactor coolant system blowdown from rated conditions. Since all of the gases in the drywell are purged into the -

suppression chamber air space during a loss-of-coolant accident, the pressure of the suppression chamber air space must not exceed 55 psig. The design volume of the suppression chamber, water and air, was obtained by considering that the total volume of reactor coolant is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.

Using the minimum or maximum water volumes given in this specification, suppression pool pressure during the design basis accident is below the design ;

pressure. Maximum water volume of 134,600 ft' results in a downcomer submergence !

of 12'3" and the minimum volume of 122,120 ft' results in a submergence approximately 2'3" less. The majority of the Bodega tests were run with a submerged length of 4 i feet and with complete condensation. Thus, with respect to the downcomer submergence, !

this specification is adequate. The maximum temperature at the end of the i blowdown tested during the Humboldt Bay and Bodega Bay tests was 170 F and this !

is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperature above 170 F. !

Should it be necessary to make the suppression chamber ineperable, this shall

. only be done as specified in Specificction 3.5.3. l 1

Under full power operating conditions, blowdown through safety / relief valves j assuming an initial suppression chamber water temperature of 95 F results in a bulk water temperature of approximately 140 F immediately following blowdown l which is below the 190 F bulk temperature limit used for complete condensation :

via T-quencher devices. At this temperature and atmospheric pressure, the available NPSH exceeds that required by both the RHR and core spray pumps, thus [

there is no dependency on containment overpressure during the accident injection t phase. If both R'IR loops are used for containment cooling, there is no dependency on containment overpressure for post-LOCA operations.

h i

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9 f

LIMERICK - UNIT I B 3/4 6-3 l i

CONTAINMENT SYSTEMS BASES Es.6.5 SECONDARY CONTAINMENT Secondary containment is designed to minimize any ground level release of radioactive material which may result from an accident. The Reactor Enclosure and associated structures provide secondary containment during normal operation when the drywell is sealed and in service. At other times the drywell may be '

open and, when required, secondary containment integrity is specified.

Establishing and maintaining a vacuum in the reactor enclosure secondary containment with the standby gas treatment system once per 18 months, along with the surveillance of the doors, hatches, dampers and valves, is adequate to ensure that there are no violations of the integrity of the secondary containment.

The OPERABILITY of the reactor enclosure recirculation system and the stanuby gas treatment systems ensures that sufficient iodine removal capability will l be available in the event of a LOCA or refueling accident (SGTS only). The reduction in containment iodine inventory reduces the resulting SITE BOUNDARY radiation doses associated with containment leakage. The operation of this system and resultant iodine removal capacity are consistent with the assumptions used in the LOCA and refueling accident analyses. Provisions have been made to a continuously purge the filter plenums with instrument air when the filters are !

! not in use to prevent buildup of moisture on the adsorbers and the HEPA filters. l Although the safety analyses assumes that the reactor enclosure secondary containment draw down time will take 140 seconds, these surveillance require- ,

ments sporify a draw down time of 126 seconds. This 14 second difference is I due to the diesel generator starting and sequence loading delays which is not part of this surveillance requirement.

The reactor enclosure secondary coatainment draw down time analyses assumes a starting poini. of 0.25 inch of vacuum water gauge and worst case SGTS dirty j filter flow rate of 2800 cfm. The surveillance requirements satisfy this as-sumption by starting the drawdown from ambient conditions and connecting the adjacent reactor enclosure and refueling area to the SGTS to split the exhaust l flow between the three zones and verifying a minimum flow rate of 2800 cfm from the test 7ene. This simulates the worst case flow aligiaent and verifies ade-quate flow is available to drawdown the test zone within the required time.

The Technical Specification Surveillance Requirement 4.6.5.3.b.3 is intended to be a multi-zone air balance verification without isolating any test zone.

The SGTS fans are sized for three zones and therefore, when aligned to a single zone or two zones, will have excess capacity to more quickly drawdown the affected zones. There is no maximum flow limit to individual zones or pairs of zones and the air balance and drawdown time are verified when all three zones are connected to the SGTS.

The three zone air balance verification and drawdown test will be done after any major system alteration, which is eny modification which will have an effect on the SGTS flowrate such that the ability of the SGTS to drawdown the reactor enclosure to greater than or equal to 0.25 inch of vacuum water gage in less than or equal to 126 seconds could be affected. l LIMERICK - UNIT I B 3/4 6-5

-DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE (Continued)

b. For a pressure of:

1. 1250 psig on the suction side of the recirculation pump.

2. 1500 psig from the recirculation pump discharge to the outlet !

side of the discharge shutoff vahe. '

a

3. 1500 psig from the discharge shutoff valve to the jet pumps. '

c. For a temperature of 575 F. [

VOLUME l 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 22,400 cubic feet at a nominal steam dome saturation temperature of 552 F. .

5.5 FUEL STORAGE l l

CRITICAllTY i

5.5.1.1 The spent fuel storage racks are designed and shall be maintained with: ;

a. A k,,, equivalent to less than or equal to 0.95 when flooded with !

unborated water, including all calculational uncertainties and !

biases as described in Section 9.1.2 or the FSAR.

b. A nominal 6.625 inch center-to-center distance between fuel assemblies. r placed in the storage racks. .

5.5.1.2 The k,,, for new fuel for the first core 1 ceding stored dry in- the i spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation j is assumed.

DRAINAGE ]

f 5.5.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 346'0".

l CAPACITY 5.5.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 2040 fuel assemblies, j 5.6 COMPONENT CYCLIC OR TRANSIENT LIMIT j 5.6.I'The components identified in Table 5.6.1-1 are designed and shall be !

maintained within the cyclic or transient limits of Table 5.6.1-1. !

f 1

LIMERICK - UNIT 1 5-8 I

)

a

7

~

(4) Pursuant to the Act and.10 CFR Parts 30, 40.and 70, to. receive, ;

possess, and use in amounts as required any byproduct, source i or special nuclear material without restriction to chemical or :[

physical . form, for sample analysis or instrument calibration or !

associated with radioactive apparatus or components; and l (5) Pursuant to the Act and 10 CFR. Parts 30, 40 and 70, to posse'ss, l' but not separate, such byproduct and special nuclear materials as may_be~prodzed by the operation of the facility, and to receive j and possess, but not separate, such source, byproduct, and special nuclear materials as contained in the fuel assemblies and i fuel channels from the Shoreham Nuclear Power Station, j l

. C. This license shall be deemed to contain and is subject to the !

conditions specified in the Commission's regulations set forth.in 10 - l CFR Chapter I (except as exempted from compliance in Section 2.D. -

l below) and is subject to all applicable provisions of the Act and to l the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or j incorporated below: :

(1) Maximum Power 1.evel j Philadelphia Electric Company is authorized to operate.the , :

facility at reactor core power levels of 3458 megawatts tnermal l l (100 percent rated power) in accordance with the conditions .

specified herein. ~

(2) Technical Specifications

-l The Technical Specifications contained in Appendix A and the !

Environmental Protection Plan contained in Appendix B, as .revi. sed ;

through Amendment No. 28, are hereby incorporated into this !

license. Philadelphia Electric Company shall operate the - ;

facility in accordance with the Technical Specifications and the ;

Environmental Protection Plan.

(3) Fire Protection (Section 9.5. SSER-2)* f The licensee shall maintain in effect all provisions of the l

approved fire protection program as described in the Final Safety Analysis Report for the 'acility through Revision 58 and~ as .j approved in the SER through Supplement 9, and in the Fire ,

Protection Evaluation Report through Revision 12, subject _to the following provisions a and b below: !

a. The licensee shall make no change to features of the approved [

fire: protection program which would decrease the level of !

fire protection in the plant without prior approval of.the- 1 Commission. To make such a change the licensee must= submit l an application for license amendment pursuant to 10 CFR 1 50.90. l l

. oThe parenthetical notation following the-title of license conditions denotes

~

the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.

DEFINITIONS PURGE .- PURGING 1.31 PURGE or PURGING shall be the controlled process of discharging. air or

gas from a confinement to maintain temperature, pressure, humidity, ,

concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement. ;

RATED THERMAL POWER l' 1.-32 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3458 MWt. l REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY [

1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall exist when: :

I

a. All reactor enclosure secondary containment penetrations required to !

be closed during accident conditions are either ,

1. Capable of being closed by an OPERABLE secondary containment !

automatic isolation system, or

2. Closed by at least one manual valve, blind flange, slide gate i damper or deactivated automatic valve secured in its closed ;

position, except as provided in Table 3.6.5.2.1-1 of j Specification 3.6.5.2.1.

~

b. All reactor enclosure seconda,y containment hatches and blowout panels I are closed and sealed. ;

c. The standby gas treatment system is in compliance with the -requirements I of Specification 3.6.5.3. ,

d. The reactor enclosure recirculation system is in compliance with the f requirements of Specification 3.6.5.4.

4

e. At least one door in each access to the reactor enclosure secondary -

containment is closed. [

f. The sealing mechanism associated with each reactor enclosure secc-'dary !

containment penetration, e.g., welds, bellows, or 0-rings, is OPERABLE. 4

+

g. The pressure within the reactor enclosure secondary containment is '

less than or equal to the value required by Specification 4.6.5.I.la.

REACTOR PROTECTION SYSTEM RESPONSE TIME 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valva solenoids. The l response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured. ;

REFORING FLOOR SECONDARY CONTAINMENT INTEGRITY 1.35 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY shall exist when:

a. All refueling floor secondary containment penetrations required to be closed during accident conditions are either:

LIMERICK - UNIT 2 1-6 l

w TABLE 2.2.1-1 C

, M REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS

. iS ALLOWABLE R FUNCTIONAL UNIT TRIP SETPOINT VALUES h 1. Intermediate Range Monitor, Neutron Flux-High s 120/125 divisions s 122/125 divisions l :: of full scale of full scale

2. Average Power Range Monitor:

m

a. Neutron Flux-Upscale, Setdown s 15% of RATED THERMAL POWER s 20% of RATED THERMAL POWER

b. Neutron Flux-Upscale

1) During two recirculation loop operation:

a) Flow Biased s 0.66 "4+ 62%, with s 0.66 W+ 64%, with l a maximam of. a maximum of b) High Flow Clamped s 115% -of RATED s 117% of RATED TilERMAL POWER THERMAL POWER

2) During single recirculation loop operation:

i 7

a) Flow Biased s 0.66 W+ 57%, s 0.66 W+ 59%, -l.

b) liigh Flow Clamped Not Regaired Not Required OPERA 3LE OPERABLE

c. Inoperative N.A. N.A.

d. Downscale 2 4% of RATED 2 3% of RATED TilERMAL POWER TilERMAL POWER

, 3. Reactor Vessel Steam Dome Pressure - High' 's 1096 psig s 1103 psig

4. _ Reactor Vessel Water Level - Low, Level 3 2 12.5 inches above instrument 2 11.0 inches above zero* . instrument zero

5. Main Steam Line Isolation Valve - Closure s-8% closed s 12% closed

, 6. Main Steam Line Radiation - liigh s'3.0 x full power background 5 3.6 x full power background

7. Drywell Pressure - liigh s 1.68 psig s 1.88 psig

8. Scram Discharge Volume Water level'- liigh

a. Level Transmitter s 261' 1 1/4" elevation ** s 261' 9 1/4" elevation

b. Float Switch s 261'-i 1/4" elevation ** 5 261' 9 1/4" elevation

See Bases Figure B 3/4.3-1.

- Equivalent'to 25.58 gallons / scram, discharge volume.

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

. - . . , . . - . . - . - - . . - . . , - . ~ . . . . .- . .. . .,. -...--....-.--.-..--....-..--:..a

REACTIVITY CONTROL SYSTEMS-SURVEILLANCE RE0UIREMENTS (Continued)

(

b. At least once per 31 days by:

1. Verifying the continuity of the explosive charge.

2. Determining by chemical analysis and calculation *~ that the available weight of sodium pentaborate is greater than or equal to 3754 lbs; the concentration of ' sodium pentaborate in solution s less than cr equal to 13.8% and within the limits of.

r'igure 3.1.5-1 and; the following equation is satisfied:

C x E x 0 21-13% wt. 29 atom % 86 gpm .

where C - Sodium pentaborate solution (% by weight)

Q = Two pump flowrate, as determined per surveillance requirement 4.1.5 c.

E = Baron 10 enrichment (atom % Boron 10)

3. 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. Demonstrating that, when tested pursuant to Specification-4.0.5, the minimum flow requirement of 41.2 gpm per pump at a pressure of' greater than or equal to 1230 25 psig is met.

d. At least once per 18 months during shutdown by:

1. Ir .ating at least one of the standby liquid control system loops, including an explosive valve, and verifying that a' flow -

path from the pumps to the . reactor ' pressure vessel -is .available by pumping demineralized water into the reactor vessel. -The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch which has been certified by having one of that batch success-fully fired. All injection loops shall be tested in 3 operating cycles.

2. Verify all heat-treated piping between storage tank and pump suction is ' unblocked.**

e. Prior to addition of Boron to storage tank verify sodium pentaborate enrichment to be added is 2 29 atom % Boron 10.

This test.shall also be performed anytime water or boron is added to the solu--

tion or when the solution temperature drops below the limits of Figure 3.1.5-1 for the most recent concentration analysis, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after water or~ boron addition or solution temperature is restored.

- This test shall also be performed whenever suction piping temperature drops below-the limits of Figure 3.1.5-1 for the most recent concentration . analysis, within _

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after solution temperature is restored.

LIMERICK . UNIT 2 3/4 1-20

TABLE 3.3.2-2 >

7_

ISOLATION ACTUATION INSTRUMENTATION SETPOINTS 3!

m ALLOWABLE h{ TRIP FUNCTION TRIP SETPOINT VALUE n

i

1. MAIN STEAM LINE ISOLATION

'C h a. Reactor Vessel Water level s, 1) Low, low - Level 2 2 - 38 inches

2 - 45 inches

2) Low, Low, low - Level 1 2 - 129 inches

2 - 136 inches

b. Main Steam Line s 3.0 x Full Power s 3.6 x Full Power Radiation - liigh Eackground Background

c. Main Steam Line Pressure - Low 2 756 psig 2 736 psig

d. Main Steam Line Flow - liigh s 12E.1 psid s 123 psid y e. Condenser Vacuum - Low 10.5 psia 210.1 psia /s 10.9 psia

f. Outboard MSIV Room Temperature - liigh s 192*F s 200 F

g. Turbine Enclosure Main Steam Line Tunnel Temperature - liigh s 165 F s 175 F

h. Manual Initiation N.A. N.A.

2. RilR SYSTEM SilVTDOWN COOLING MODE ISOLATION

a. Reactor Vessel Water Level l Low - Level 3 2 12.5 inches

2 11.0. inches l

l b. Reactor Vessel (Ri1R Cut-in Permissive) Pressure - liigh sl75 psig s 95 psig

c. Manual Initiation N.A. N.A.

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

.= . . _- - .

, TABLE 3.3.2-2 (Continued)

.y ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ,

g ALLOWABLE

, g TRIP FUNCTION TRIP SETPOINT VALUE R

n 3. REACTOR WATER CLEANUP SYSTEM ISOLATION.

I g a. RWCS a Flow - liigh s 54.9 gpm s 65.2 gpm

[ b. RWCS Area Temperature - liigh 5 142*F or 132*F** s 147*F or 137 F** ' [ j

c. RWCS Area Ventilation .;

i a Temperature - liigh s 32*F s 40*F ,

t

d. SLCS Initiation N.A. N.A. ;

e. Reactor Vessel Water Level - ;

Low, Low, - Level 2 2 -38 inches

2 -45 inches

f. Manual Initiation N.A. N.A. ~;

y 4. IIIGli PRESSURE COOLANT INJECTION SYSTEM ISOLATION

.g ,

a. IIPCI Steam Line a Pressure - liigh s 974" 11,0 5 984" lij0 l. ,

b. IIPCI Steam Supply Pressure - Low 2 100 psig 2 90 psig

c. IIPCI Turbine Exhaust Diaphragm Pressure - liigh s 10 psig s 20 psig

d. IIPCI Equipment Room Temperature -'lligh 175 F 2 165'F, s 200*F

e. IIPCI Equipment Room a Temperature - liigh $ 80*F s 88 F ,

f. IIPCI Pipe- Routing Area Temperature - fligh 175 F 2 165 F, s 200* F - -"

g. Manual Initiation N . A ., N.A.

t

h. IIPCI Steam Line a Pressure - Timer 3 s r s 12.5 seconds 2.5 s r s 13 seconds

TABLE 3.3.2-2 (Continued)

'i ISOLATION ACTUATION INSTRUMENTATION SETPOINTS.

C 4 ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE-i 5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION !

- c- -

5 a. RCIC Steam Line a

[ Pressure - liigh s 373" 11,0 s 381" H,0 !. '

b. RCIC Steam Supply Pressure - Low 2 64.5 psig 2 56.5 psig :

c. RCIC Turbine Exhaust Diaphragm .

Pressure - liigh s 10.0 psig s 20.0 psig-

d. RCIC Equipment Room .

Temperature - liigh 175'F 2 165 F, s 200*F ;

w e. RCIC Equipment Room 1 a Temperature - liigh s 80*F s 88'F w

E f. RCIC Pipe Routing Area Temperature - liigh 175'F 2165 F, s 200*F

g. Manual Initiation N.A. N.A.

h. RCIC Steam Line a Pressure Timer 3s7 s 12.5 seconds 2.5 s r s 13 seconds L

t 4

I1 TABLE 3'3.4.1-2 ATWS RECTRCULATION PUMP TRIP SYSTEM'1NSTRUMENTATION SETPOINTS ,

.i TRIP ALLOWABLE -!

TRIP FUNCTION SETPOINT- VALUE l

1. Reactor Vessel, Water Level - -!

Low Low, level 2 a -38 inches

a -45 inches !

2. Reactor Vessel Pressure - High : 1149 psig 5 1156 psig '

l

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P 3/4.4- REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM- 1 r

i RECIRCULATION LOOPS i l

LIMITING CONDITION FOR OPERATION f I

'3.4.1.1 Two reactor coolant system recirculation loops shall be in operation I with: [,

a. Total core flow greater than or equal to 45% of rated core flow, or ~!

I

b. THERMAL POWER within the unrestricted zone of Figure 3.4.1.1-1. l l

APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*.

l ACTION: j

a. With one reactor coolant system recirculation loop not in operation: i

1. Within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s: ;

i

a. Place the recirculation flow control system in the Local Manual mode, and

b. Reduce THERMAL POWER to 5 76.2% of RATED THERMA'L POWER, and, i f

c. Limit the speed of the operating recirculation pump to- ,

less than or equal to 90% of rated pump speed, and ;j i

d. Verify that the differential temperature requirements of :)

Surveillance Requirement 4.4.1.1.5 are met if THERMAL L POWER is 30% of RATED THERMAL POWER or the recirculation-loop flow in the operating loop is s 50% of. rated ' loop- !

flow, or suspend the THERMAL POWER or recirculation loop flow increase. '

i l

l 7 *See Special Test Exception 3.10.4.

1 i

LIMERICK - UNIT 2 3/4 4-1

._ - _ _ . - __ .. . . J

1

-REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS ,

-i 4.4.1.l.1 Each pump discharge valve shall be demonstrated OPERABLE by ' cycling I each valve through at least one complete cycle c f full travel during each.

startup* prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER-4.4.1.1.2 Each pump MG set scoop tube mechanical and electrical- stop shall be ;

demonstrated OPERABLE with overspeed setpoints_less than or equal to 109% and ;

107%, respectively, of rated core flow, at least once per 18 months. !

4.4.1.1.3 Establish a baseline .APRM and LPRM** neutron flux noise value within !

the regions for which monitoring is required (Specification 3.4.1.1,. ACTION c) ;

within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of entering the region for which monitoring is required unless ;

baselining has previously been performed in the region since the last refueling ,

outage. l 4.4.1.1.4 With one reactor coolant system recirculation loop not in ;

operation, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months verify that: '!

a. Reactor THERMAL POWER is 76.2% of RATED THERMAL POWER,

b. The recirculation flow control system is in the Local Manual mode, and

c. The speed of the operating recirculation pump is s 90% of rated pump speed. a i

d. Core flow is greater than 39% when THERMAL POWER is within the ;

restricted zone of Figure 3.4.1.1-1. !

4.4.1.1.5 With one reactor coolant system recirculation loop not in ;

operation, within 15 minutes prior to either THERMAL POWER increase or !

recirculation loop flow increase, verify that the following difierential .:

temperature requirements are met if THERMAL POWER is s 30% of RATED THERMAL ['

POWER or the recirculation loop flow in the operating recirculation loop is '

s 50% of rated loop flow:

a. s 145 F between reactor vessel steam space coolant and bottom head I drain line coolant,

b. 5 50 F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel, and

c. 5 50 F between the reactor coolant within the loop not in operation and the operating loop. '

' The differential temperature requirements of Specification 4.4.1.1.5b. and c. ,

do not apply when the loop not in operation is isolated ~ from the reactor pressure vessel.

3

If not performed within the previous 31 days. !

- Detector levels A and C of one'LPRM string per core octant plus detectors A and C of one LPRM string in the center of the core should be monitored.

LIMERICK - ijNIT 2 3/4 / -2 ,

L r- 1 5 70 I s

7

.i i 60 g Restricted r4 Zone to 50

$ 45 .8% Power

/

40 g i b

d $ 33.5% Power A a 30

^ E

" Unrestricted Zone g

5 20 10 f

0 30 40 50 60 70 Core Flow (% Rated)

Figure 3.4.1.1-1 THERMAL POWER VERSUS CORE FLOW

_ _ _ _ _ _ _ _ . _ . . . _ . . . _ _ _ . _ . - _ _ _ . _ _ - . . _ _ _..-__-_._..-._.-_..~...___,..._.._._.._____.;.._._._-__.._,.._... -

d REACTOR COOLANT SYSTEM j

1 3 /4. 4.' 2 SAFETY / RELIEF VALVES )

. LIMITING CONDITION FOR OPERATION :j 3.4.2 The safety valve function of at least 11 of the-following reactor coolant system !

safety / relief valves shall be OPERABLE with the specified code safety valve function lift !

settings:*#

}

4 safety / relief valves @ 1170 psig 21% i 5 safety / relief valves @ 1180 psig 21% '

5 safety / relief valves @ 1190 psig 21% -l APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3 ACTION

a. With the safety valve function of one or more of the above required i 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 ; i and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . 1 I

~

b. With one or more safety / relief valves stuck open, provided that suppression pool average water tem !

safety / relief valve (s)perature

if unableis to less thanthe close 105 F, close stuck openthe stuck valve (sopen within 2 i minutes or if su pression pool average water temperature is 110 F)or ;

greater, place t e reactor mode swit h 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 -r 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 '

?4 houn, >

, SURVEILLANCE RE0VIREMENTS ]

1 4.4.2.1 The acoustic monitor for each safety / relief valve shall be demonstbated OPERABLE i with the setpoint verified to be 0.20 of the full open noise level ## by performance of a: 1 q

a. CHANNEL FUNCTIONAL TEST at least once per 31 days, and a l

b. CHANNEL CALIBRATION at least once per 18 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 recommendations at least once per 24 months, and they shall be rotated such that 01 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 -l months.

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 'i adequate to perform the test. ,

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

1. Initial setting shall be in accordance with the manufacturer's 4 recommendation. Adjustment to the valve full open noise level shall be accomplished during the startup test program.

LIMERICK - UNIT 2 3/4 4-7 :(

i

i r

1600 i

- 'I A A'B B'C C' I 1400 l g ;

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$ 800 [

, SHIFT FROM AN INITIAL .

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% 600 ,/

/ j f WITH FUELIN VESSEL r

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/ B - NON-NUCLEAR HEATUP/ !

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@ 400 ,f D VESSEL DISCONTINUITY U UMITS su rsia .- :f

CORE BELTUNE WITH I E 53 oF SHIFT 200 !

BOLTUP -

CURVES A',B',C' ARE VAUD l 70Y / FOR 10 EFPY OF OPERATION

[

/ CURVES A.B.C ARE VAUD l FOR 1 EFPY OF OPERATION l l

0 ; 3 i !

O 100 200 300 400 500 600 l

MINIMUM REACTOR VESSEL METAL TEMPERARJRE ("F)

MINIMUM REACTOR VESSEL METALTEMPERATURE VS. REACTOR VESSEL PRESSURE FIGURE 3.4.6.1-1 ,

.t UMERICK- UNIT 2 3/4 4-20

i REACTOR COOLANT SYSTEM. !

cFEACTOR STEAM DOME i

' LIMITING CONDITION FOR OPERATION l 3.4.6.2 The pressure in.the reactor steam dome shall be less.than 1053 psig. ll [

APPLICABillTY: OPERATIONAL CONDITIONS 1* and 2*.

iACTION: j

Hith 'the reactor steam dome pressure exceeding 1053 psig, reduce the pressure -

to less than 1053 psig within 15 minutes or be in at least HOT SHUTDOWN within ;

212 hours0.00245 days 0.0589 hours 3.505291e-4 weeks 8.0666e-5 months . ,

i SURVEILLANCE REOUIREMENTS l 4.4.6.2 The reactor steam dome pressure shall be verified to be less

than 1053 psig at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . l. ;

i 3

l r

Not applicable during anticipated transients. ;

.;-i i

LIMERICK-- UNIT 2 3/4 4-22 1

i'

i EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.1. The emergency core cooling systems shall be demonstrated OPERABLE by:

a. At least once per 31 days: l

1. For the CSS, the LPCI system, and the HPCI system:

a) Verifying by venting at the high point vents that the ,

system piping from the pump discharge valve to the system (

isolation valve is filled with water. >

b)

Verifying)that automatic each valve (manual, power-operated, orin the flow'!p otherwise secured in position, is in its correct

position.

I

2. For the LPCI system, verifying that both LPCI system subsystem l cross-tie valves (HV-51-282 A, B) are closed with power removed from the valve operators. j;

3. For the HPCI- system, verifying that the HPCI pump flow controller is in the correct position. !

i

4. For the CSS and LPCI system, performance of a CHANNEL FUNCTIONAL [

TEST of the injection header 6P instrumentation. ;

1

b. Verifying that, when tested pursuant to Specification 4.0.5:

j

1. Each CSS pump in each subsystem develops a flow of at least- !

3175 gpm against a test line pressure corresponding to a reactor vessel to primary containment differential pressure of a 105 psid-plus head and line losses. ,

t

2. Each LPCI pump in each subsystem develops a flow of at least j 10,000 gpm against a test line pressure corresponding to a ;

reactor vessel to primar containment differential. pressure of- t t 20 psid plus head and ine losses. l

3. The HPCI pump develops a flow of at least 5600 gpm against a test line pressure which corresponds to a reactor vessel pressure of 1040 psig plus head and line losses when steam is i being supplied to the turbine at 1040, +13, -120 psig.** i

c. At least once per 18 months: 1;

1. For the CSS, the LPCI system, and the HPCI system, performing a-system functional test which includes simulated automatic ,

actuation of the system throughout its emergency operatina ,

sequence and verifying that each automatic valve in the flow '

path actuates to its correct position. Actual injection of -

coolant into the reactor vessel may be excluded from this test.

i

Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in position for another mode of operation. ;

i

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. If-OPERABILITY is not successfully demonstrated i within the 12-hour period, reduce reactor steam dome pressure to less than 200 psig within the following 72-hours. ;

LIMERICK - UNIT 2 3/4 5-4 i l

t

CONTAINMENT SYSTEMS 3/4.6.5 SECONDARY CONTAINMENT REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION-3.6.5.1.1 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

. ACTION:

Without REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY, restore REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 4.6.5.1.1 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall be demon-strated by:

a. Verifying at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months that the pressure within the reactor enclosure secondary containment is greater than or equal to 0.25 inch of vacuum water gauge.

b. Verifying at least once per 31 days that:

1. All reactor enclosure secondary containment equipment hatches and blowout panels are closed and sealed.

2. At least one door in each access to the reactor enclosure ' secondary containment is closed.

3. All reactor enclosure secondary containment penetrations not capable of being closed by OPERABLE secondary containment auto-matic isolation dampers / valves and required to be closed during accident conditions are closed by valves, blind flanges, slide gate dampers or deactivated automatic dampers / valves secured in position,

c. At least once per 18 months:

1. Verifying that one standby gas treatment subsystem will draw down the reactor enclosure secondary containment to greater than' or equal to 0.25 inch of vacuum water gauge in less than or equal to 126 seconds with the reactor enclosure recirc system in operation, l and

2. Operating one standby gas treatment subsystem for one hour and maintaining greater than or equal to 0.25 inch of vacuum water, gauge in the reactor enclosure secondary containment at a flow rate not exceeding 1250 cfm with wind speeds of 5 7.0 mph as measured on the wind instrument on Tower I elevation 30' or, if that instrument is unavailable, Tower 2, elevation 159'.

LIMERICK - UNIT 2 3/4 6-46

CONTAINMENT SYSTEMS '!

y SURVEILLANCE REQUIREMENTS (Continued)

2. Verifying that the fan starts and isolation valves necessary to draw a suction from the refueling area or the reactor enclosure . ;

recirculation discharge open on each of the following test signals: i a) Manual initiation from the control room, and :

b) Simulated automatic initiation signal. f 1

3. Verifying that the temperature differential across each heater [

is e 15 F when tested in accordance with ANSI N510-1980. !

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the.inplace penetration ;

and leakage testing acceptance criteria of less than 0.05% in !

accordance with ANSI N510-1980 while operating the system at a flow ;

rate of 3000 cfm 10%.

f. After each complete or partial replacement of a charcoal adsorber :

bank by verifying that the charcoal adsorber bank satisfies the inplace l penetration and leakage testing acceptance criteria of less than 0.05% -

in accordance with ANSI N510-1980 for a halogenated hydrocarbon ,

refrigerant test gas while operating the system at ~a flow rate of '

3000 cfm 2 10%. .

t

g. After any major system alteration:

1. Verify that when the SGTS fan is running the subsystem flowrate ;

is 2800 cfm minimum from each reactor enclosure (Zones I and !

II) and 2200 cfm minimum from the-refueling area (Zone III).. .

.i

2. Verify that one standby gas treatment subsystet will drawdown i reactor enclosure Zone 11 secondary containment to greater'than i.

or equal to 0.25 inch of vacuum water gauge in less than or >.

equal to 126 seconds with the reactor enclosure recirculation l system in-operation and the adjacent reactor enclosure and !

refueling area zones are in their isolation modes. !

B-l t

LIMERICK - UNIT 2 3/4 6-54 i

PLANT SYSTEMS ,

3 /4. 7 ~. 3 REACTOR CORE-ISOLATION COOLING SYSTEM i

. LIMITING CONDITION FOR OPERATION }

s 3.7.3 The reactor core isolation cooling (RCIC) system shall' be OPERABLE with ,

an OPERABLE flow path capable of automatically taking suction from the i suppression. pool and transferring the water to the reactor pressure vessel. .

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3 with reactor steam dome 'I pressure greater than 150 psig. j ACTION:

a. With the RCIC system inoperable, operation may continue provided the l HPCI system is OPERABLE; restore the RCIC system to OPERABLE status within l 14 days. Otherwise, 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 reduce reactor steam dome pressure to less than or equal to 150 psig j within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. In the event the RCIC system is actuated and injects water'into the reactor coolant system, a Special Report shall be prepared and sub- .i mitted to the Commission pursuant to Specification 6.9.2 within ;

90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. -7 SURVEILLANCE REQUIREMENTS t I

4.7.3 The RCIC system shall be demonstrated OPERABLE ,

a. At least once per 31 days by: !

, 1. Verifying by venting at the high point vents 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) l in the flow path that is not locked, sealed, or otherwise secured '

in position, is in its correct position. i

3. Verifying that the pump flow controller is in the. correct position. .l

b. At least once per 92 days by verifying that the.RCIC pump develops' a i flow of greater than or equal to 600 gpm in the test flow path with ,

a system head corresponding to reactor vessel operating pressure when .i steam is being supplied to the turbine at 1040, + 13, - 120 psig.* j. .l

-l 1

o 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. If OPERABILITY is not successfully demonstrated within the 12-hour period, reduce reactor steam dome pressure to less than- !

150 psig within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

LIMERICK - UNIT 2 3/4 7-9 i

i REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

'The operating limit curves of Figure 3.4.6.1-1 are derived from the fracture ~

~ toughness requirements of 10 CFR 50 Appendix G and ASME Code Section III, i Appendix G. The curves are based on the RTwot and stress intensity factor

information for the reactor vessel components. Fracture toughness limits and the basis for compliance are more fully discussed in FSAR Chapter 5, Para-graph 5.3.1.5, " Fracture Toughness." l The reactor vessel materials have been tested to determine-their initial RTeot . The results of these tests are shown in Table B 3/4.4.6-1. Reactor !

operation and resultant fast neutron, E greater than 1 MeV, irradiation will- i cause an. increase in the RTet. Therefore, an adjusted reference temperature, ?

based upon the fluence, nickel content and copper content of the material in question, can be predicted using Bases Figure B 3/4.4.6-1 and the recommenda- ,

tions of Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor i Vessel Materials." The pressure / temperature limit curve, Figure 3.4.6.1-1, curves A', B' and C', includes an assumed shift in RTwoT for the conditions at 10.EFPY. The A, B and C limit curves are predicted to be bounding for all

= areas of the RPV until 1 EFPY when the beltline materials RTwot will shift, due to neutron fluence, and the beltline curves will intersect the '

non-beltline discontinuity curves.

The actual shift in RTeT of-the vessel material will be established period-ically during operation by removing and evaluating, in accordance with 10 CFR Part 50, Appendix H, irradiated reactor vessel flux wire and charpy specimens 3 installed near the inside wall of ie reactor vessel in the core area'. Since 4 the neutron spectra at the charpy specimens and vessel inside radius are !

essentially- identical, the irradiated charpy specimens can be used with' con-

-fidence in predicting reactor vessel material transition temperature shift. .,

The operating limit curves of Figure 3.4.6.1-1 shall be adjusted, as required,-

on the basis of the flux wire and charpy specimen data and recommendations of Regulatory Guide 1.99, Revision 2.

The_ pressure-temperature limit lines shown in Figures 3.4.6.1-1, cu'rves- l C, and C', and A and A', for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum l temperature requirements of Appendix G to 10 CFR Part 50 for reactor criticality .. !

and for inservice leak and hydrostatic testing. !

6 The number of reactor vessel irradiation surveillance capsules and the -

frequencies for. removing and testing the specimens in these capsules are pro- ,

vided in Table 4.4.6.1.3-1 to assure compliance with the requirements of !

Appendix H to 10 CFR Part 50.

i 5

LIMERICK - UNIT 2 B 3/4 4-5 l

1

' BASES TABLE B 3/4.4.6-1 ,

REACTOR VESSEL TOUGHNESS

l ,

LIMITING MIN. UPPER h HEAT / SLAB g BELTLINE WELD SEAM 1.D. OR STARTING SHELF ,

COMPONENT OR MAT'l TYPE HEAT / LOT CU (%) Ni (%1 RTwot (*F) aRTwot **(*F) (LFT-lBS) . ART (*F) ,

7 Plate SA-533 Gr. B,CL. 1 B 3416-1 .14 .65 +40 +82 NA +122- :

E Weld AB (Field Weld) 640892/ .09 1.0 -60 +114 NA +54 Q J424B27AE.

NOTES:

Based on 110% of original power.

- These values are given only for the benefit of calculating the end-of-life (EOL/32 EFPY) RTwot j NON-BELTLINE MT'l TYPE OR- IIEAT/ SLAB OR llIGilEST STARTING :

COMPONENT WELD SEAM 1.0. HEAT / LOT RTwot (*F)

Top Shell Ring SA 533, Gr. B, CL. 1 C9800-2 -16 Bottom Head Dome C9306-2 +26

, Bottom Head Torus C9362-2 +28 R' Top Head Torus "

C9646-2 -20 ,

. ~ Top Head Flange SA-508, Cl. 2 123B300 +10 Vessel Flange "

2L2058 -20 .

Feedwater Nozzle Q2Q29W 0 l

Weld Non-Beltline All -12 i LPCI Nozzle *** SA-508, CL. 2 Q2Q33W -4' .l l Closure Studs SA-540,'Gr. B-24 All Meet requirements of 45 ft-lbs !

and 25 mils Lat. Exp. at +10'F. -

-l

- - The desig'n of the LPCI nozzles results in their. experiencing an EOL fluence in excess of 10" N/Cm which predicts an EOL (32 EFPY) RTwot of +35'F. ;

}

t t

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i 1.2 i

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o t E 0.4 -

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0.0 ;

10 20 30 40 t

Service Life (Years *) ,

1 i

BASES FIGURE B 3/4.4.6-1 f FAST NEUTRON FLUENCE (E> l MeV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE * ,

i

At 90% of Rated Thermal Power and 90% availability .

LIMERICK - UNIT 2 B 3/4 4-8 ;

3/4.5 EMERGENCY CORE COOLING SYSTEM r i

BASES .

.3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHUTDOWN i

The core spray system (CSS), together with the LPCI mode of the RHR system, l 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 f sizes up to and including the double-ended reactor recirculation line break, j and for smaller breaks following depressurization by the ADS. ,

i 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. l The surveillance requirements provide adequate assurance that the CSS will ;

be OPERABLE when required. Although all active components are testable and i full flow can be demonstrated by recirculation through a test loop during !

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

i The low pressure coolant injection (LPCI) mode of the RHR system is i 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 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 i shutdown. The pump discharge piping is maintained full to prevent water !

' hammer damage to piping and to start cooling at the earliest moment. l The high pressure coolant injection (HPCI) system.is provide to assure h that the reactor core is adequately cooled to limit fuel clad 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 i HPCI system permits the reactor to be shut down while maintaining sufficient- -!

reactor vessel water level inventory until the vessel is depressurized. The. l HCPI system continues to operate until reactor vessel pressure is- below the i pressure at which CSS operation or LPCI mode of the RHR system operation j maintains core cooling. 1 t

'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. Initially, water from the condensate storage l ;

tank is used instead of injecting water from the suppression pool into the reactor, .

but no credit is taken in-the safety analyses for condensate storage tank j water. l LIMERICK - UNIT 2 B 3/4 5-1 ,

'3/4.6' CONTAINMENT SYSTEMS ,

BASES-3/4.6.1 PRIMARY CONTAINMENT 3 /4. 6.1.1 PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY ensures that the release of radioactive mate-

- rial s from the containment atmosphere will be restricted to those leakage paths ,

and associated leak rates assumed in the safety analyses. This restriction, 't in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the limits of 10 CFR Part 100 during accident conditior.m 3/4.6.1.2 PRIMARY-CONTAINMENT LEAKAGE -;

i The limitations on primary containment leakage rates ensure that the total i containment leakage volume will not exceed the value calculated in the _ safety i analyses for the peak accident pressure of 5 44 psig, P . As an added conserva- '!

tism, the measured overall integrated leakage rate is further limited to less .

than or equal to 0.75 L during performance of the periodic tests to account for (

possible degradation of the containment leakage barriers between leakage tests. l Operating experience with the main steam line isolation valves has

' indicated that degradation has occasionally occurred in the leak tightness of j the valves; therefore the special requirement for testing these valves. ;

i The surveillance testing for measuring leakage rates is consistent with the requirements of Appendix J of 10 CFR Part 50 with the exception of exemptions granted for leak testing of the main steam isolation valves, the airlock and TIP shear valves. .

3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCK t

The limitations on closure' and leak rate for the primary containment air lock are required to meet the restrictions on PRIMARY CONTAINMENT INTEGRITY and the primary containment leakage rate given in Specifications 3.6.1.1 and ,

3.6.1.2. The specification makes allcwances for the fact that there may be i long periods of time when the air lock will be in a closed and secured :

position during reactor operation. Only one closed door in the air loc'k is required to maintain the integrity of the containment. l 3/4.6.1.4 MSIV LEAKAGE CONTROL SYSTEM j I

Calculated doses resulting from the maximum leakage allowance for the main steamline isolation valves in the postulated LOCA ~ situations would be a small 'I fraction of the 10 CFR Part 100 guidelines, provided the main steam line system from the isolation valves up to and including the turbine condenser remains :

intact. Operating experience has indicated that degradation has occasionally }'

occurred in the leak tightness of the MSIVs such that the specified leakage requirements have not always been maintained continuously. The requirement for :

the leakage control system will reduce the untreated leakage from the MSIVs when isolation of the primary system and containment is required. g i

I LIMERICK - UNIT 2 B 3/4 6-1 l i

i

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

CONTAINMENT SYSTEMS BASES 3/4.6.1.5 PRIMARY CONTAINMENT STRUCTURAL INTEGRITY j This limitation ensures that'the structural integrity of the containment

will be maintained comparable to the original design standards for the life of' ;

the unit. Structural integrity is required to ensure that the containment will :

withstand the maximum calculated pressure in the event of a LOCA. A visual .b inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability.

3/4.6.I.6 DRYWELL AND SUPPRESSION CHAMBER INTERNAL PRESSURE-The limitations on drywell and suppression chamber internal pressure ensure ,

that the calculated containment peak pressure does not exceed the design l _.

pressure of 55 psig during LOCA conditions or that the external pressure differ- !

ential does not exceed the design maximum external pressure differential of .

5.0 psid. The limit of - 1.0 to + 2.0 psig for initial containment pressure '

will limit the-total pressure to s 44 psig which is less than the design l [

pressure and is consistent with the safety analysis. [

3/4.6.1.7 DRYWELL AVERAGE AIR TEMPERATURE The limitation on drywall average air temperature 'ensores that the con- '

tainment peak air temperature does not exceed the design temperature of 340 F ,

during steam line break conditions and is consistent with the safety analysis.

-3/4.6.1.8 DRYWELL AND SUPPRESSION CHAMBER PURGE SYSTEM e

The drywell and suppression chamber purge supply and exhaust isolation !

valves are required to be closed.during plant operation except as required for -j inerting, deinerting and pressure control. The 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months per 365 day limit on purge valve operation is imposed to protect the integrity of the SGTS- filters. .'

Analysis indicates that should a LOCA occur while this pathway is being utilized,

- the associated pressure surge through the (18 or 24") purge lines will adversely affect the integrity of SGTS. This limit is not imposed,.however, on the subject valves when pressure control is being performed through the 2-inch bypass line, ,

since a pressure surge through this line does not threaten the OPERABILITY _of ;

SGTS.

i

LIMERICK - UtlIT 2 B 3/4 6-2 l

1

i l CONTAINMENT SYSTEMS ';

BASES ,

3/4.6.2~ DEPRESSURIZATION SYSTEMS The specifications of- this section ensure that the primary containment '

pressure _will not exceed the design pressure of 55 psig during primary system blowdown from. full operating pressure.

The suppression chamber water provides the heat sink for the reactor coolant ;

system energy release following a postulated rupture of the system. The suppression chamber water volume must absorb the associated decay and structural sensible heat released during reactor coolant system blowdown from rated conditions. -

Since all of the gases in the drywell are purged into the suppression chamber air space during a loss-of-coolant accident, the pressure of the suppression chamber ;

air space must not exceed 55 psig. The design volume of the suppression chamber, :

water and air, was obtained by considering that the total volume of reactor ,

coolant is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber. :

Using the minimum or maximum water volumes given in this specification, ;

suppression pool pressure during the design basis accident is below the design pressure. Maximum water volume of 134,600 ft' results in a downtomer submergence of 12'3" and the minimum volume of 122,120 ft' results in a submergence approximately-2'3" less. The majority of the Bodega tests were run with a subirierged length of 4 :

feet and with complete condensation. Thus, with respect to the downcomer submergence, !

this- specification is adequate. The maximum temperature at the end of the r blowdown tested during the Humboldt Bay and Bodega Bay tests was 170 F and this is conservatively taken to be the limit for complete condensation of the reactor :

coolant, although condensation would occur for temperatures above 170 F. :

Should it be necessary to make the suppression chamber inoperable, this shall only be done as specified in Specification 3.5.3. ;

Under full power operating conditions, blowdown through safety / relief valves ;

assuming an initial suppression chamber water temperature of 95 F results in a -l bulk water temperature of approximately 140 F immediately following blowdown l 1 which is below the 190 F bulk temperature limit used for complete condensation via T-quencher devices. At this temperature and atmospheric pressure, the !

available NPSH exceeds that required by both the RHP, and core spray pumps, thus there is no dependency on containment overpressure during the accident injection phase. If both RHR loops are used for containment cooling, there is no dependency on containment overpressure for post-LOCA operations.

4 s

L.lMERICK.- UNIT 2 B 3/4 6-3

CONTAINMENT SYSTEMS ;

BASES' i 3 /4. 6. 5 SECONDARY CONTAINMENT I Secondary containment is designed to minimize any ground level release of (

radioactive material which may result from an accident. The Reactor Enclosure -l and associated structures provide secondary containment during normal operation ,

ehen the drywell is sealed and in service. At other times the drywell may be '!

open and, when required, secondary containment integrity is specified. l Establishing and maintaining a vacuum in the reector enclosure secondary I containment with the standby gas treatment system once per 18 months, along with :i the surveillance of the doors, hatches, dampers and valves, is adequate to ensure

-that there are no violations of the integrity of the secondary containment. :

The OPERABILITY of the reactor enclosure recirculation system and the standby gas treatment systems ensures that sufficient iodine removal capability will ;

be available in the event of a -LOCA or refueling accident (SGTS only). The reduction in containment iodine inventory reduces the resulting SITE B0UNDARY ;

radiation doses associated with containment leakage. The operation of this i system and resultant iodine removal capacity are consistent with the assumptions !

used in the LOCA and refueling accident analyses. Provisions have'been made to continuously purge the filter plenums with instrument air when the filters are ;

not in use to prevent buildup of moisture on the adsorbers and the HEPA filters. l t

Althour- the safety analyses assumes that the reactor enclosure secondary -;

containment draw down time will take 140 seconds, these surveillance require- !

meats specify a draw down time of 126 seconds. This 14 second difference is- i due to the diesel generator starting and sequence loading delays which is not l; part of this surveillance requirement.

i

.The reactor enclosure secondary containment draw down time analyses assumes 1 a starting point of 0.25 inch of vacuum water gauge and worst case SGTS dirty ,

filter flow rate of 2800 cfm. The surveillance requirements satisfy this as- j sumption by starting the drawdown from ambient conditions and~ connecting the ;

adjacent reactor enclosure and refueling area to the SGTS to. split the exhaust 1 flow between the three zones and verifying a minimum flow rate of 2800 cfm from ;

the test zone. This simulates the worst case flow alignment and verifies ade- ;

quate flow is available to drawdown the test zone within the required; time.

The Technical Specification Surveillance Requirement.4.6.5.3.b.3 is intended j to be a multi-zone air balance verification without isolating any test zone. 1

'The SGTS is common to Unit I and 2 and consists of two independent- 1 subsystems. The power supplies for the common portions of the subsystems are 1 from Unit 1 safeguard busses, therefore the inoperability of these. Unit 1 i supplies are addressed in the SGTS ACTION statements in order to ensure adequate !

oasite. power sources to SGTS for its Unit 2 function during a loss of offsite j power event. The allowable out of service times are consistent with those in ;

the Unit 1 Technical Specifications for SGTS and AC electrical power supply out i of service condition combinations. l l

1 LIMERICK - UNIT 2 B 3/4 6-5 ;

' CONTAINMENT SYSTEMS ,

6ASES SECONDARY CONTAINMENT (Continued)

The SSTS fans are sized for three zones and-therefore, when aligned to a i single zone or two zones, will have excess capacity to more quickly drawdown -l

-the affected zones. There is no maximum flow limit to individual zones or ,

pairs of zones and the air balance and drawdown time are verified when all

three zones are connected to the SGTS. :

The three zone air balance verification and drawdown test will be done after any major system alteration, which is any modification which will have an effect on the SGTS flowrate such that the ability of the SGTS to drawdown i the reactor enclosure to greater than or equal to 0.25 inch of vacuum water gage in less than or equal to 126 seconds could be affected. l :

The field tests for bypass leakage across the SGTS charcoal adsorber and HEPA filter banks are performed _ at a flow rate of 3000 e 10% cfm. This flow :

rate corresponds to the maximum overall three zone inleakage rate of 3264 cfm- !

l The SGTS filter train pressure drop is a function of air flow rate and q filter conditions. Surveillance testing is performed.using either the SGTS or i drywell purge fans to provide operating convenience.

Each reactor enclosure secondary containment zone and refueling area l secondary containment zone is tested independently to verify the design leak tightness. A design leak tightness of 1250 cfm or less for each reactor .j enclosure and 764 cfm or less for the refueling area at a 0.25 inch of vacuum >

eater gage will ensure that containment integrity is maintained at an acceptable - -;

level if all zones are connected to the SGTS at the same time. i The post-LOCA offsite dose analysis assumes a reactor enclosure secondary l containment post-draw down leakage rate of 1250 cfm and certain post-accident X/Q values. While the post-accident X/0 values represent a statistical inter- .

pretation of historical meteorological data, the highest ground level wind i speed which can be associated with these values is 7 mph (Pasquill-Gifford.

' stability Class G for a ground level release). Therefore, the surveillance- i

. requirement assures that the reactor enclosure secondary containment is verified i under meteorological conditions consistent with the assumptions utilized in the ~'

design basis analysis. Reactor Enclosure Secondary Containment leakage tests '

that are successfully performed at wind speeds in excess of 7 mph .would'also satisfy the leak rate surveillance requirements, since it shows compliance with more conservative test conditions.

l LIMERICK - UNIT 2 B 3/4 6-6 ;

i v > - e . , e .a m ._--___--.---__-----_-_--.-_---A__N._--_--_

. DESIGN FEATURES _

DESIGN PRESSURE AND TEMPEPATURE (Continued)

b. For a pressure of:

1. 1250 psig on the suction side of the recirculat'on pump.

2. 1500 psig from the recirculation pump discharge .to the outlet side of the discharge shutoff valve.

3. 1500 psig from the discharge shutoff- valve to the jet pumps.

c. For a temperature of 575 F.

VOLUME 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 22,400 cubic feet at a nominal steam dome saturation temperature of 552 F.

5.5 FUEL STORAGE 1 CRITICALITY 5.5.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. A k,,, equivalent to less than or equal to 0.95 when flooded with unborated water, including all calculational uncertainties and biases as described in Section 9.1.2 or the FSAR.

b. A nominal 6.625 inch center-to-center distance between fuel assemblies placed in the storage racks.

5.5.1.2 The k,,, for new fuel for the first core loading _ stored dry in the spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is assumed.

DRAINAGE 5.5.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent oraining of the' pool below elevation 346'0".

CAPACITY 5.5.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 2040 fuel assemblies.

5.6 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.6.1 The components identified in Table 5.6.1-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.6.1-1.

LIMERICK - UNIT 2 5-8

. _ . . -. . .- . - .

ML20058L449

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