Text

Proposed Tech Specs Changing Corporate Name from PA Electric Co to PECO Energy Co & Removing Obsolete Info & Correcting Typos
	ML20135D096
Person / Time
Site:	Limerick
Issue date:	02/25/1997
From:	; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	;
Shared Package
ML20135D090	List: ML20135D086; ML20135D096;
References
	NUDOCS 9703050049
	Download: ML20135D096 (47)
	v • d • e

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ATTACHMENT 2 --

LIMERICK GENERATING STATION UNITS 1 AND 2 Docket Nos.

50-352 50-353 License Nos.

NPF-39 NPF-85 LICENSE CHANGE REQUEST No. 96-06-0 AFFECTED PAGES UNIT 1 UNIT 2 LICENSE: FOL FOL FOL FOL FOL FOL FOL . -----

FOL FOL FOL -------

FOL -------

APPENDIX A: x -x xix xix 3/43-8 3/43-8 3/4 3-103 _3/4 3-103 3/45-1 3/45-1 3/45-2 3/45-2 3/4 6-14 1 1

3/4 6-15 3/4 6-15 3/4 6-16 3/4 6-16 l 3/4 6-42 3/4 6-42 3/4 6-43 3/4 6-43 ;

3/47-1 3/47-1 3/47-3 3/47-3 3/4 8-15 --

6-8 6-8 ;

6-9 6-9 APPENDIX B: EPP Cover EPP Cover EPP 4-4 EPP 4-3 9703050049 970225 PDR ADOCK 05000352!

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,_ NUC'.EA A :.EGULATCRY CCMMISSION - -

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;0cKET No 5c.352 LIMERICX 3EhERAsiHG 1;ATION. UNIT 1 FACILE

OPERAinG L.;EN5E )

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License No. HPF-39 1

1.

~he that: Nuclear Regulatory Commission (the Comission or the NRC) has found ,

.PEdd .w..Q+ Q i Company 1

A.

The acolication for license filed by . .. Ab... l (the licensee) complies with the standards and reouirements of the Atomic Energy Act of 1954, as amenced (the Act), and the Comissien's )

regulations set forth in 10 CFR Chacter 1, and all recu1 rec notifica-tions to other agencies er cocles nave oeen duly made; B. 1 Construction of the Limerick Generating Station, Unit 1 (the facility) !

has been substantially comoleted in confomity with Construction Pemit No. CPPR-105 and the application, as amended, the provisions of the Act and the regulations of the Comission; i C.

The facility will operate in conformity with the application, as amended, the provisions of the Act, and the regulations of the Comissior (except as exempted from compliance.in Section 2.D. below); ,

D.

There is reasonable assurance: (i) that the activities authorized by this operating license can be conducted without endangering the nealth and safety of the public, and (ii) that such activities will be -

conducted in compliance with the Cemitsion's regulati:ns set f:rtn in 10 CFR Chapter I (except as exemotec from compliance in Section 2.3.

below);

E.

The licensee is technically cualified to engage in the activities authorized by this license in accordance with the Comission's reguia-tions set forth in 10 CFR Chapter I; F.

The licensee has satisfied the applicable provisions of 10 CFR Part 140, " Financial Protection Requirements and Indemnity Agreements', of the Comission's regulations; G.

The issuance of this license will not be inimical to the c: mon defense and security or to the health and safety of the puolic; g 0M' e 4 h

__72.__:. 2.3 o

3885125590 H. *fteryet;ning :ne envir:r entai, economic, tecnnical, anc ::ner !

ener1 s :r :ne racility a;ains: env1ren. ental anc etner :::.':s anc j

nsicer* 9 availacle siterna wes. :ne issuance :f this Fac11ity i Cperatin; License he. hPF-39, suoject :: :ne c:ncitiens f:r protec-i:n Of the environment te; for:n in tne Environmantal Protecti:n Plan j attacnec as Appenc1x B, is 1n acc:reance with 10 CFR Par 51 cf the '

Comissien's regulations anc all appiicacie requirements have eeen !

satisfied; and

. ~he recei:t, possession, anc use of source, byprecuct and s:ecial !

nuclear material as authert:ec by this license will be in accercance ;

with the Cccuission's regulations in 10 CFR Parts 20, 40 ano 70. :

2. Basec en the feregoing fincings, the Partial Initial Oecisions issued !

by*the Atomi: Safety anc Licensing Boaro catec Maren E,1983, August 29. j 1984, May 2, 19E5 ano July 22, 1985, anc the Cecision cf the Appeal Boarc !

cate: Septem:er 2F, '984, regarcing :nis fac11ity, anc approval by the Nuclear j

teguia::ry C:m1ssion in its Meterancum ano Orcer cateo August E, *985, . :ne l license for Fuel Loacing anc Low Power Testing, License No. NPF-E7, issueo cn C: .::er 25,1984, is superseced by Facili y Operating License NPF-35 hereby issuec to tne " - := ~~

nr4c Ccmpany (the licensee), :: rers as i follows: P Ed o "Coc q :

i A. This license applies to the Limerick Generating Station, Unit 1, a j boiling water nuclear reac :r and associated equipment, cwnec by '

peu arm . ;,..... r.. .c a Ccmpany. The facility is located en :he iicensee's

. psq site in Montgomery and Chester Counties, Pennsylvania on the banks of ne Schuylkill River approximately 1.7 miles southeast cf :ne city limits ,

of Pottstcwn, Pennsylvania and 21 miles northwes: of the etty limits of ;

Philaceiphia, Pennsylvania, and is uescribed in tne licensee's Final '

Safety Analysis Repcrt, as supplementec ano amencec, anc in :ne licensee's !

Eny1rencental Ee;cr:-Ocera:1ng Li:ense 5: age, as suppiements: anc ;

amencec.

E. Subject := the conditions and requirements incorporated herein, the Comission hereby licenses T.M ' -

t.-

~~

" Ccmpany: j CCobGM \

(1) Pursuant to Section 10 o et and 10 CFR Part 50, to pos- i sess, use, and operate the facility at the casignated location. in Montgomery and Chester Counties, Pennsylvania, in accordance with the procecures and limitations set forth in this license;

, (2) Pursuant to the Act and 10 CFR Part 70, to receive, pcssess and to use at any time special nuclear material as reac::r fuel, in acc:rdance with the limitations fcr storage and amounts required for reacter operation, as cescribed in the Final Safety Analys1s Repert, as supplemented and amended; L(os (

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1 O L- 1 AUS 9 55

I

. -3 (3) Pursuant to tne Act and 10 CFR Parts 30, 40 ano 70, to 3

! receive, ;ossess and use at any time any byprocutt, source I and soecial nuclear material as sealed neutron sources for '

reactor startup, sealed sources for reactor instrumentation i i

and raatation monitoring equipment :alibration, and as fission cetectors in amounts as required:- l (4)' Pursuant to the Act and 10 CFR Parts 30, 40 ano 70, to '

receive, possess, and use in amounts as requireo any byproouct, 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) Pursuant to the Act and 10 CFR Parts 30, 40 ano 70, to possess. :ut not separate, sucn cyprocutt ano soecial nuclear materials as may be procuteo by the operation of the f acility, anc to receive and oossess, but not separate, such source, byprocuct, and special nuclear materials as contained in the fuel assemolies and fuel channels from the Shorenam 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 provistens 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:

l (1) Maximum Dower *avel m-m ~. u.

. __ D . ed is author to operate the tsellity at reactor core power level not in excess of 3458 megawatts thermal (100% rated powe { in accordance with the conditions !

specified herein and in Attachment 1 to this license. The '

items identified in Attachinent 1 to this license shall be completed as specified. Attacnment 1 is hereby incorporated into this license. _.

(2) Technical Seecifications Pf2'O Bneen vv pmw_

(7 The Technical Specifica ons contained in Appendix A and the r.nvironmental Protecti Plan contained in Appencix B, as revised through Amen ent No. , are hereoy incorporated in the license. shall operate the facility in accordance with the Technical Specificatinns and the t

Environmental Protection Plan.

lgs l>W 70L xmenement m . 52, "5

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i j, (3) Fire Protection (Section 9.5. $$ER-2.-4)* !

l f(,d,0 Mf'M4hh E'--"4 r Company shall implement ano maintain in effect l

'g all provisions of the approved Fire Protection Program as described in the Updated Final Safety Analysis Report for the facility, and as i

' approved in the NRC Safety Evaluation Report dated August 1983 through Supplement 9. dated August 1989, and Safety Evaluation dated i November 20 , 1995, subject to the following provision:

i i

The licensee may make changes to the approved fire protection ;

l program without prior approval of the Commission only if those :

changes would not adversely affect the ability to achieve and '

{ maintain safe shutdown in the event of a fire. ;

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The parentnetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.

Amenoment No. 104 I E 2 0 1995 t G5 h I

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fc)1 n t'.e first re'uel.

ing ou se. ;revice 3 dseletionsignalste j the rest. erclesurt tv . - 811carc isolat' ion a

yalves in t suis,h anc return alttier tungs are ths crywe. ch111ec mater uttoarc .. Ives ir. the su;;1y anc ret'.:rn ines.

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{(14) fuelino Floor Volume Connection to Stanaby Gas Treatmen j Svhem (Section 6.2.3. SSER-2 ano SSER-3) '

Prior t any movement of irraciated fuel withi he refueling !

floor voi the licensee shall complete a ' test all modifi-cations req ' red to connect the refueli floor volume to standby gas t tment system. Duri ( !

the interim period, t the licensee sha not remove t eactor pressure vessel head l l prior to the NRC s ff revie nd approval.

l , (15) Emeraency Plannina Procedures Sub_iect to CFR Part 50 In the event t RC finds that the I k of progress in comple-l tion of the rocedures in the Federal E rgency Management Agency's inal rule, 44 CFR Part 350, is indication that a major ubstantive problem exists in achievi or maintaining an uate state of emergency preparedness, the ovisions of j (a CFR Section 50.54(s)(2) will apply. - -

D. The facility requires exemptions from certain reouirements of --

10 CFR Part 50. _ These include f(a) exempti from eral Design f Criteria ( C) 61 of Appendix A, operation o t portion of the

[ standby gas atment system (SGTS) that ve the refueling area until the first refueling (Section 6 . of SSE 2 and SSER-3), (b) exemption from G -56 of Append , the requirem t for additional automatic containme isola ' n valves for the hyd en recombiner lines and the requirement omatic isolation of exist isolation valves in the Drywell Chi ter (DCW) and the Reactor Enc sure Cooling Water (RECW) e ems unti rior to startup following th first refueling outage ( ion 6.2.4.2 of e SER, SSER-1 and SSER-3), (c exemption from -19 of Appendix A, as elated to the requirement fo redundant ote shutdown capability (Sec 'on 7.4.2.3 of SSER-3 and SSE 5), (d)

/ @) air locks at times when the containment integrity is not requiredxemption g (Section 6.2.6.1 of the SER and SSER-3), W exemption from the requirements of Appendix J, tne leak rate testing of thy Main Steam Isolation Valves (MSIVs) at the peak calculated containment pressure, Pa, and exemption from the reouirements of Appendix J that the measured MSIV leak rates be included in the summation for the local leak rate test (Section 6.2.6 of SSER-3),

( exemption from the requirement of Appendix J, the local leak rate testing of l the Traversing Incore Probe Shear Valves (Section 6.2.6 of the SER and SSER-3), l ig) : One tim: ::::pti:n 're +he rcre-ent f Sppedix J to perf;m.100 1 le3k- I h/g&C ua %t I foc

! l l l JAN 2 41997 Amenomen: No. *18

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,,, , ,,, 3 ; ) 6.1.6 ui :::?. ;, v.3 exemami i um i qu i, ....c.: :' . . . n r oi .

a .... , . . . . . . . . i c. ; J . .. . . . . .. ...u 6. . . .i.. . .. . . . . o u ca my

(% ;i .. ;.".; vi ;;;'. T, .;; (i) c._ .i;; '.;; :h: c: pi.......;. -'_

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l ;;;". I'. These exemotions are authorized by law and will not endanger

life or property or the conson defense and security and are othemise

! in the public interest. Therefore these exemptions are hereby granted

( pursuant to 10 CFR 50.12 ana 50.47(c). With the granting of these

exemptions the facility will operate, to the extent authorized herein,

! in confomity with the application, as amended, the provisions of the j Act, and the rules and regulations of the Comission.

E. The licensee shall fully implement and maintain in effect all provisions i of the Commission-approvec physical security, guard training and i qualification, and safeguares contingency plans, including all ameno-i ments anc rev' ' ,s made pursuant to the authority of 10 CFR 50.90 ano i 10 CFR 50.. , which are part of the license. These plans, which i

contain Sve rds'Information protected under 10 CFR 73.21, are entitled

i " Limerick' ra ng Station Physical Security Plan," " Limerick Generating i Station Plan urity Personnel Training and Qualification Plan," and i

" Limerick . Generating Station Safeguards Contingency Plan."

F. Except as othemise provided in the Technical Specifications or Environmental Protection Plan, the licensee shall report any violations of the requirements contained in Section 2.C of this license in the following manner: initial notification shall be made within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the NRC Operations Center via the Emergency Notification System with written followup within thirty days in accoroance with the procedures describec in 10 CFR 50.73(b), (c), and (e). .

I G. The licensee shall have and maintain financial protection of such !

type and in such amounts as the Commission shall require in accord- I ance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.

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LIMITING CONDITIONS FOR OPERATTON AND SURVEII. LANCE RE0UIREMENTS

IEllDK EAE

II(STRUMENTATION (Continued)

Tdh : .; . 7. ;- ; ~ ' . . L... . ... ;... ..

................... ;/^ ; ;;

l 1

Loose-Part Detection System............................... 3/4 3-97

! The infomation from pages 3/4 3-98 through 3/4 3-101 has been intentionally omitted. Refer to note on page 3/4 3-98. . . . . . . . . . . . . . . . . .

3/4 3-98

.I' 1

Offgas Monitoring Instrumentation......................... 3/4 3-103 i Table 3.3.7.12-1 Offgas Monitoring Instrumentation. . . . . . . . . . . . . . . 3/4 3-104 Table 4.'3.7.12-1 Offgas Monttoring Instrumentation Surveill ance Requi rements . . . . . . . . . . . . . . . . 3/4 3-107 3/4.3.8 (Deleted) The information on pages 3/4 3-110 and l

3/4 3-111 has been intentionally omitted.

Refer to note on page 3/4 3-110................. 3/4 3-110 l 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION j INSTRUMENTATION....................................... 3/4 3-112

. Table 3.3.9-1 Feedwater/ Main Turbine Trip i

System Actuation Instrumentation......... 3/4 3-113

! Table 3.3.9-2 Feedwater/ Main Turbine Trip

) System Actuation Instruman-l tation Setpoints......................... 3/4 3-114 i

Table 4.3.9.1-1 Feedwater/ Main Turbine Trip i System Actuation Instrumenta-I tion Surveillance Require-l monts..................................... 3/4 3-115 j 3/4.4 REACTOR CDOLANT SYSTEM t

3/4.4.1 RECIRCULATION SYSTEM '

Recirculation Loops....................................... 3/4 4-1 4

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${f 21 1995 LIMERICK - UNIT 1 x Ananament 2. 48. ' M yu ..

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i BASES 1 SECTION g j i

' !NSTRUMENTATION (Continued) l (Deleted).............................................. .B 3/4 3-5 l

( Del e t e d ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 3-5 Remote Shutdown System Instrumentation and Controls.... B 3/4 3-5 l l Accident Monitoring Instrumentation.................... B 3/4 3-5 Source Range Monitors.................................. B 3/4 3-5 j a

l

( Del e t e d ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 3-6' '

Chlorine and Toxic Gas Detection Systems . . . . . . . . . . . . . . . B 3/4 3-6 a

(Deletec)................ ............ ............ ... B 3/4 3-6 Loose-Part Detecu an System............................ B 3/4 3-7 i

(Deleted).............................................. B 3/4 3-7 !

Offgas Monitoring Instrumentation...................... B 3/4 3-7 3/4.,. 3 ,...n...-

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. B 3/4 3 7 !

i 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION........................................ B 3/4 3-7 Bases Figure B 3/4.3-1 Reactor Vessel Water Level..................... B 3/4 3-8 3 /4 4 QEACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM... ............ ....... ......... B 3/4 4-1 3/4.4.2 SAFETY / RELIEF VALVES................................... B 3/4 4-2 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.............................. B 3/4 4-3 Operati onal Le a kag e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-3 3/4.4.4 CHEMISTRY.............................................. B 3/4 4-3 LIMERICK - UNIT 1 xix Amencment No. J2. 52. 59. 75. 72 Jul. '.1 19F.

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e-J A8tE 4.3.1.1-1 (Cont inued)

I,I REAC10R PR0liCl(UN_SyE][fLIN118tt![lflMl!!N_jlilfliYLil1 ANGLBEOUIREMENIS +< ;

OPERATIONAL g CilANNEL CllANNEL CONDITIONS FOR WillCil

$ CllANNLL FUNCTIONAL GUBRATION LURYULLANCE_MElBEi! ,

g CHECK _ TEST p fUNCI10NAL UN H R 1 N.A. Q

9. Turbine Stop Valve - Closure !

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x U 10. Turbine Control Valve fast

~ Closure, Irlp 011 N.A. Q R 1 Pressure - Low N.A. 1,2,3,4,5

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

N.A.

12. flanual Scram la)

(b) Neutron detectors entering OPtRATIONAl may be2 and CONDITION excluded the IRM and from APRM CilANNEl channels shall Call l 8RAIION.The be determined to over a if not performed within the previous 7 days. l~

decades during each centrolled shutdown, eg ,

(y (c) This calibration shall consist of the adjustment of the APRM channel DELETED Adjust totheconform APRii to m

(d) a heat balance during OPERAT10NAL CONDITJON I when THERNAL POWER 225% t d flow of RAT absolute difference is greater than 2% of RATED 1HERNAL POWER.

channel if the This calibration shall consist of the adjustment of the APRM flow hiased channel to conf y (c)

> g signal.

LPRMs shall be calibrated at least once per 1000 effective full power hours (EfPil). t the existing l l 8 (l) The Verify measured core flow (total core flow) to be greater than or equal to established core flow (g)

- Ior_, flow (APRM % flow). During the startup test program, data shall be recorded for the paramete

,' provide a basis for establishing the specified relationships. lion O

(h) the criteria listed shall commence upon the conclusion of the startup test 10.2.

progr 3.10.1. Not applicable to control rods removed per Specification 3.9.10.1 lled for or 3.9.

up to 2 With any control rod withdrawn.

??

(i)

( j) If the RPS shorting links are registred to be removed per Specification 3.9.2, they m P!

hours for required survelilance. i 3 be moved from its existing position.

Required to be OPERA 8LE only prior to and during shutdown margin demonstrations'

l (k) 3.10.3.

m. O

- i EiN

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

2, .

{

INSTRLMENTATION

{

OFFGAS GAS MONITORING INSTRUMENTATION l .

l LIMITING CONDITION FOR OPERATION

i 3.3.7.u The offgas' monitoring instruentation channels shown in Table 3.3.7.u-1 shall be OPERABLE with tnair alars/ trip setpoints set to ensure

! that the limits of Specifications 3.u.2.5 and 3.11.2.6 respectively, are not exceeded.

i .

1 APPLICABILITY: As shown in Table 3.3.7.12-1 f ACTION:

a. With an offgas monitoring instrumentation channel alars/ trip 1 setpoint less conservative than required by the above Specification,

' declare the channel inoperante, and taka the ACTION shown in Table 3.3.7.12-1.

b b.

i With less than the minimum neber of offgas monitoring instruentation

! channels OPERABLE, take the ACTION shown in Table 3.3.7.12-1. Restore the inoperable instraentation to OPERABLE status within the time

! specified in the ACTION or explain why this inoperability was not corrected in a timely manner in the next Radioactive Effluent j Release Report.

nnuet I

c. The provisions of Specifications 3.0.3 and .0.4 are not applicable.

I !

! SURVEILLANCE REQUIREMENTS 1

4. 3. 7. H Each offgas monitcring instroentation channel shall be demonstrated 1

OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3.7.12-1.

l l

d 3

i i

i l LIMERICK - UNIT 1 3/4 3-103 Amenonen: No. 48

. m 0215 9%

j .

3/4.5 EMERGENCY CORE COOLING SYSTEMS r

3/4.5.1 ECCS - OPERATING i LIMITING CONDITION FOR OPERATION 3.5.1 The emergency core cooling systems shall be OPERABLE with:

a. The core spray system (CSS) consisting of two subsystems with each i subsystem comprised of:

1. Two OPERABLE CSS pumps, and i 2. An OPERABLE flow patch capable of taking suction from the

! suppression chamber and transferring the water through the spray l

sparger to the reactor vessel.

1 b. The low pressure coolant injection (LPCI) system of the residual

, heat removal system consisting of four subsystems with each l subsystem comprised of:

1. One OPERABLE LPCI pump, and ,

2. An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.

c. The high pressure coolant injection (HPCI) system consisting of:

1. One OPERABLE HPCI pump, and )

2. An OPERABLE flow path capable of taking suction from the j suppression chamber and transferring the water to the reactor vessel.

d. The automatic depressurization system (ADS) with at least five-OPERABLE ADS valves.

APPLICABILITY: OPERATIONAL CONDITION 1, 2* ** f, and 3* ** ff .

The HPCI system is not required to be OPERABLE when reactor steam dome pressure is less than or equal to 200 psig.

- The ADS is not required to be OPERABLE when the reactor steam dome pressure is less that or equal to 100 psig.

See Special Test Exception 3.10.6.

1. Two LPCI ubsystems of the RHR system may be inoperable in that they are aligned in the shutdown cooling mode when reactor vessel pressure is less than the r RHR Shutdown cooling permissive setpoint. p f.

y w f#ffDur e 14 day extended A0T period allowed for installa ommon ESW and RHRSk system Mo 1 ons P-0166, P0167, and P0 e Unit 2 third refueling outage (2R03), in addition e in and 'D' LPCI subsystems of the RHR system anc the inoperable 'B' the 'A' LPCI subsystem of the RHR system may be inoperabi it is aligned in . cooling mode when reactor vessel pressure i than the RHR Shutdown cooling permiss _

LIMERICK - UNIT 1 3/4 5-1 Amendment No. 23. 86 JAN 3 1 1995

___ _ __ _ __ __ __ _ . _ _ . _ _ - __ ___ _ ~ _ . _ _ __

r 4

PERGENCY CORE COOLING SYSTEMS

.!MITING CONDITION FOR OPERATICH ' Continued)

'.CTION:

a. For the core spray system:

1. With one CSS suosystem inoperable, provideo tnat at least two LPCI' subsystems are OPERABLE. restore the inoperacie CSS suosystem to

, OPERABLE status within 7 days or be in at '. east HOT SHUTDOWN withi

)

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

, , 2. With both CSS suosystems inoperable, be in at least HOT SHUTDOWN l

.~ .

within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within tne next ?4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . -

1

b. For the LPCI system:

1. With one LPCI suosystem inoperable, provided that at least one CSS subsystem is OPERABLE. restore the inoperaole LPCI pumo to OPERABLE status within 30 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 anc in COLD SHUTDOWN within .ne following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2. With one RHR cross-tie valve (HV-51-182 A or B) open or power not removed from one closed RHR cross-tie valve coerator. : lose the i open valve anc/or remove power from the closec valves operator within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . 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 next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . i

3. With no RHR cross-tie valves (HV-51-182 A, 2) closed. or power not removec from coth closed RHR cross-tie valve operators, or with one RHR cross-tie valve open and power :ot removec from the other P.HR cross-tie valve operator, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4. With two LPCI suosystems inoperable, provided that at least one CSS subsystem 1: OPEPABLE. restore at least threa LPCI suosystems to OPERABLE statu: within 7 days or ce in at '.esst HOT SHUTDOWN within the next '2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ; ano in COLD SHUTDOWN witnw tne following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months -

5. With three LPCI suosystems inoperacie, proviceo that both CSS subsystems are OPERABLE, restore at least two LPCI subsystems to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 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 .

6. With all four LPCI subsystems inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .*

Whenever botn snutdown cooling subsystems are inoperable, 'f unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature 4

_ as low as practical by use of alternate heat removal methocs.

. . . . -' W idge sAmth 6,rw a .e s Q T T.ne b' Lore dorav stf neve*cm =r- ~~T 'T

- "avs to ailow for nstallation of the comon LPCI suosystems may ve .~--

tions P-0166. P-016/, ano r- - - .ov u. ; '.' " Q

.1MERICK - UNIT '. 3/4 5-2 Amens =en: No. M. C4

o. .

._ JX. 0 7 1995

0NTsiNMENT Iv5T g SUPC:E5510N COOL SPRAY

~~TNG CCNDITION FOR OPERATION 3.5.2.2 The suppression pool :: ray moce :f the residuai heat removai (RHR) system shall be OPERABLE with two " cepencen loops, each 1000 consisting of:

a. One OPERABLE RHR : ump, anc

b. An OPERABLE flow cath cacaole of recirculating water from the suppression chamcer througn an RHR heat exchanger and the suppression pool spray soarger(s).

APPLICABILITV* OPERATIONAL CONDITIONS 1, 2, and 3. i ACTION:

a. With one suppression pooi : pray loop inoperable, restore the inoperable loop to GPERABLE :tatus within 7 days or be in at least HOT SHUTDOWN within the next '.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . l

o. With both suppression pool spray loops inoperable, restore at least one loop to OPED.ABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the next *2 . hours and in COLD SHUTDOWN

within the '

following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

"lRVE1LLANCE DEOUIREMENTS 4.6.2.2 The suppression pool :cray moce of the RHR system shall be demonstrated OPERABLE:

a. At least once :er 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not lockec, sealed, or otherwise secured in oosition, is in its correct position.

c. By verifying : .at eacn f tne reoutrec RHR oumo: ceveio:s a flow of at least 500 9:m on recirculation flow througn the RHR neat excnanger and the suppression pooi spray sparger wnen tested pursuant to Speci-fication 4.0.5.

Whenever coth RHR suosystems are inocertole, if unable to attain COLD SHUTDOWN as required by this ACTION. maintain reactor coolaat temoerature as low as practical by usE of alternate hcat removai metnoos.

ne d Outage Time (A0T) of the loop of the suooressien cool corav 3 moce of the ,

tem may be ext *~~ : .., 2 , oays to allow f or instaliation of the common t -

1 - ustem Mocifications P-0155, F-0167, and P-016S l d' . . . nit 2 third refuen.3 ---= (2R03) y

.IMERICr. - UNIT 1 1/4 6-15 amen: men: no. H AN 1 1 1995 l l

CONTsWENT D'I WS SUPP8Eii10N POOL COOLING L IMIT*':G CONDIT*0'l FOR ODEDATTON

2.6.2.2 The suppression pool cooiing moce of the residual heat removal (RHR) system snall be OPERABLE with two inoeoencen- ioops, eacn loop consisting of:

a. One OPERABLE RHR umo, ano

b. An OPERABLE flow path capable of recirculating water from the suppression

~

chamber through an RHR heat exchanger.

APPL!**BILITY: OPERATIONAL CONDITIONS 1, 2, ano 3.

ACT10'l:

a. With one suppression pool cooling loop inoperable, restore tne inocerable loop to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 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 anc in COLD SHUTDOWN within the followino 24 ~

hours

b. With both suppression pooi cooling loops inoperacle, be in at least HOT SHUTCOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ano in COLD SHUTDOWN

within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVE!L!ANCE ~EOUIREMENTS 4.6.2.2 The suppression pool cooling mooe of the RHR system snail be demonstrated OPERABLE:

a. At least once per 21 days by verifying that each valve (::anual, power-operated, or automatic) in the flow path that is not lockec, sealed, or otherwise s cured in position, is in its correct position,

b. By verifying that eacn of the required RHR pumos develops a flow of at least 10,000 g:n en recirculation flow througn the flow ca inciuding the RHR heat sxcnanger ino its assoc 1ateo :iosec : years valve. the suppression pool ano the full flow test line wnen testec :ursuant t:

Specification 4.0.5.

Whenever octn RHR suosystems are inoperacle, if unaole to attain COLD SHUTDOWN as requireo by this ACTION, maintain reactor coolant temperature as low as practical by use cf alternate heat removai methods.

wec Outage Time (A0T) of the 'E' loop of tne Suonroe '-- o m moce of ..

system may be extended - ts ;., cays to allow for instai'.ation cf tne common ESW a -

w noo1fications P-0166, P-0167 anc P-016E curing tne i- ...tro refueling 2R03).

L f 2/4 6-16 men: en: N: E7 EE- !

.1MER1 3 - UNIT 1 l 4 = -- 'JAlt 2 1 1995 !

TABLE 3.5.3-1 PRIMARY CONTAlhMENT .-OLAT**5l VALVES NoiAiich NOTE! (Continuto) 15.

Chact vaIve usec insteac of flow crifice.

! 15.

Denetration is sealec by a flange with coucle 0-ring seals. These seals are leanage rate testea by pressurizing entween tne 0-rings. Botn the T*P

' Purge Supply (Penetration 353) anc the TIP Drive Tubes (penetrations 35 C thru Glare weldec to their respective flanges. Leakage tnrougn these seals is inclucac in the Type C leakage rate total for snis penetration. The call valves (XV 141A thru E) are Type C testec. It is not practicaole to leat i

test tne snear va)ves (XV-140A thru E) because soutb firing is reautree for closure. Shear valves (XV-140A tnru E) are normally open.

i 17.

i Instrument line isolation provisions consist of an excess flow enact valve.

Because tne instrument line is connecteo to a closee cooling water system insice containment, no flow crifice is oroviced. The excess flow chect valves are suoject to operacility testing, out no Type C test is performee nor recuireo. The line coes not isolate curing a LOCA ano can laat only if tne line or instrument snould rupture. Leaktigntness of the line is verifiec curing the integrated leak rate test (Type A test).

18. In accition to couble "0" ring seals, this cenetration is testec by pres-surizing volume between doors per Specification 4.6.1.3. ,

i

19. The RHR system safety pressure relief valves which are flangen to facilisatt removal will be equipped with double 0 ring seal assemblies on the flange closest to primary containment. These seals will be leak rate testen by pressurizing eetween the 0-rings, anc tne results added into the Type C total for tais penetration.

20, see Scacification 3.3.2, Table 3.3.2-1, for a description of the PCRVICS isolation signal (s) that initiate closure of esen automatic isolation valve.

In accition, the following non-PCRVICS isolation signals also initiate closure of selectec valves: 1

~'

t S :::=n ..

n,,_,e. ,'u :

. 9 7 : :::r , t -- i ' N 1==k=7=

' N '

M

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

LFHP Vith HPCI pumps running, opens on low flow in associatec pipe, closes when flow is above setpoint LFRC With RCIC pump running, cuens on low flow in associatec pipe, closes when flow is above setpoint LFCH With CSS pump running, coens on low flow in associated pipe, closes when flow is above setpoint LFCC Steam supply valve fully closec or RCIC turnine sten valve fully closec All power coeratec isolation valves may be conneo or closeo remote manually.

LIMERICK - 'JNIT 1 3/4 6-42 Amenoment No. 23,23 DCT 3 0 EES m . .

l . . ,

. TAILE 1. -

i '

PRIMARY CONTAI 4 MENT TION VALVES

NOTATION

NOTES (Continued) i F. Automatic isolation s1gnal causes TIP to retract; ball valve closes when

{

probe is fully retracted.

l 22. Isolation barrier remains water filled or a water seal remains in the line post-LOCA. Isolation valve may be tested with water. Isolation

, valve leakage is not included in 0.60 La total Type B & C tests.

l 23. Valve does not receive an isolation signal. Valves will be open during

Type A test. Type test not required.

I 24. Both isolation signals required for valve closure.

25. Deleted ,

i 26. Valve stroke times listed are maximum times verified by testine per Spect-i fication 4.0.5 acceptance criteria. The closure times for isolation valves i in lines in which high-energy line breaks could occur are identified with a 1

single asterisk. The closure times for isolation valves in itnes which provide an open path from the containment to the environs are identified l with a double asterisk.

27. The reactor vessel heaa seal leak detection line (penetration 29A) excess I

flow eneck valve is not sub, ject to OPERABILITY testing. This valve will not be exposed to primary system pressure except unoer the unlikely con-ditions of a seal failure wnere it could be partially pressurized to reactor pressure. Any leakace path is restricted at the source; therefore, this valve need not be OPERABILITY tested.

28. (DELETED) .

29. Valve may be open during normal operation; capable of manual isolation from control room. Position will be controlled procedurally.

30. Valve normally open, closes on scram signal.

31. Valve 41-1016 is an outboard isolation barrier for penetrations X-9A, B and X-44. Leakage through valve 41-1016 is included in the total for penetration X-44 only.

32. Feedwater long-path recirculation valves are sealed closed whenever the reactor is critical ano reactor pressure is greater inan 600 psig. The valves are expectec to ne openea only in the following instances:

a. Flushing of the condensate and feedwater systems during plant startup.

b. Reactor pressure vessel hydrostatic testing, which is conducted follow-ing each refueling outage prior to commencing plant startup.

Therefore, valve stroke timing in accordance with Specification 4.0.5 is not required.

33. Valve also constitutes a Unit 2 Reactor Enclosure Secondary Containment Automatte Isolation Valve and a Refueling' Area Seconcary Containment Automatic isolation Valve.r - " t : Illaf'_-'

T 2 '.. :."..'..:.: ' ------*"^:';. W 34 Auto isolation signals have been removed from HV-057-124 A/B and 125 A/B.

Valves to De closec with assoc 1ated circuit breakers locked open during OPCONs 1, 2, and 3.

EN 171995 LIMERICK - tlNIT 1 3/4 6-43 Amencment fe. 22.102

.u .

l

1- ?!. TNT !fSTEMS 21 7.' ?E:VICE WATE: SYSTEMS

EEIDJAL -E T DEMOVAL EERVICE WTE: SYSTEu f.0MMON SYSTEM

.w! TING C 'iOIT10N FOR CPERATION l

.7.1.1 at least the followir.; inoeoencent residuai heat removai serv 1:a water

HRSW) :ysttm subsystems, with ea:n suosystem comprised of:

a. Two OPERABLE RHRSW oumes. and

. An OPERABLE flow cath ca:able of taking sucticn from I .s EHR service water pumps wet pits unich are supplied from tne spray ;:nd or the i cooling tower tasin ano transferring the water througn ene Unit 1 2

RHR heat exchanger, !

nall be OPERABLE:

i

t. In OPERABLE CONDITIONS *., 2, and 2, two subsystems.

b. In OPERABLE CONDITIONS and 5, the subsystem (s) associated with i systems and c:moonents reouired OPERABLE by 5:ecifican:n 3.4.9.2, 3.9.11.1, ano 1.3.11.2.

Dot!CABIL:': OPERATIONAL CONDIT:0NS 1, 2, 3, 4, ano 5.

1* TION: !

a. In OPERATIONAL CONDITION 1, 2, or 3: I

1. With one RHRSW pumo inoperable, r ucre the inocertole pump to OPERABLE :tatus within 30 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 . 1

2. With one RHRSW pumo in each subsystem inocerable, restore at least one of the inocerable RHRSW pumos to OPEFAELE status within 7 tays or te in at least HOT SHUTCOWN witnin the next 12 hour: 1:c in COL" SHUTDOWN within ::e f:ilowir; I4 hour:.

3. With one RHRSW suosystem otherwise inoceracle, rest:re the inoperaole suosystem to OPERABLE status with at least one OPERABLE RHRSW pump within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOWN within the nexL12 hours and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , g l

4. With both RHRSW suosystems otherwise inocerable restore at least ene subsystem to OPERABLE status within 2 h:urs or ce in at least HOT SHUTDOWN within the next '.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months anc in COLD SHUTDOWN

within tne following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Whenever coth RHRSW subsystems are inocerable, if unable : attain C;LD SHUTCOWN as reouired by the ACTION, maintain reactor cooiant temoertture as low as cractical by use of alternate heat removal methocs.

had Outaae Time ( A0T) f the '?' RHRW e"**"'"- :- ;; ....;... up to I ' c ays to a n " '--- " . . ... . . tne common ESW anc RHRSW systas Moci F - " :.. r-vi66. P-Olb . ... '.~.-_N"- W UMt : ". ire ref eiinc '

M i2R03). 7 3/4 7-; : ennmen n:. 53. 5

_IMERItr. - UNIT 1

' ' .. ~JAN 3 1 1995

* ;':~ -'! STEMS ME:GEiiCY eE:VICE 'JaTE: Iv5TEu _ -ieMon tv- gy

" TING C"NDITION F : DEDATION i

F. . '. 2 it least tne #:llowing 1 :eoencent amergency service water :ystem i::os,

,1tn eacn icco comortsec er:

. l

t. Two C?EPABLE emergency :ervice water cumos, and I

. An CPEPABLE flow cath casaole of taking suction from the emergency

' I l service water pumos wet :its which are supplied from the s: ray pano or the c:oling tower casin and transferring the water to the associatec <

Unit '. and common safety-related equipment,

nali be OPEPABLE:

1. In CFERATIONAL . NDITICNS *., 2, ano 1, two locos.

. In CFEPATIONAL ONDITICNS a. E, and *, one loop. ;

DO:',BIL: v- OPERATIONAL CONDITIONS '. 2, 2. 4. 5. ano *. I Ir.T* SN -

a. In C?EPATION C',"DITION ., 2, or 3: '

1. With one emergency :ervice water pumo inoperable, restcre the inoperable tumo to GPERABLE status within 45 days or te 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 COLC SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . l

2. With one emergency service water pump in each loop inocerable, ,

restore at least ene inoperable pump to OPERABLE status within !

30 days or be in at least HOT SHUTDOWN within the next '.2. hours and in COLD SHUTCO' N within Ine following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2. lith one e ergencv :erv1ce water :ystem i:co :tnerwise inoperaole. :eciare all ecutement aligneo to tne inoceraole loop inoperacle**, . estore the inoperable 1000 to OPEPABLE status with at least one OPERABLE pump within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOWN within the next 1L hours and in COLD SHUTDOWN within ins following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

'When nanoling trractated fuel in the secondary containment.

- The diessi generat:rs may be aligneo *.: the OPERABLE emergency service water '

system 1300 provice: confirmatcry ficw testing has been performeo. ~ hose i ciesei generators no aligned t: the OPERABLE emergency service water system l ioco sr. ail be ceciated inoperacle anc t .e actions of 3.8.1.1 taken.

ime l'- ~ n tage u_ Time (A01) cf tne 'E' 1000 of *% c-a---

_,,,,, m., ,, ,,M !

may oe extenceo . .. A a : "Ea .nstallation of th'e common ESW ano I i RHRSW :ystem MM'"--- L..a r vid5. ;-Olb , ..m . : : -' "' -- - - a U-M : -.ird g .., cutage :2R03)y

.'.F.ERICy. . NIT '. I!? 7 - : men: ment :. 27.

H q :. . m.

. . 3892300520 !

ELECTRICA1. POG SYSTEMS .

3'/4. 8. 3 ONSITI 20WER DISTRIBUTION SYSTEMS DISTRIBUTION - OPERATING 1

LIMITING CONDIT'ON FOR OPERATION i 3.8.3.1 The following power distribution system civisions shall be energizes:

4. A.C. ;ower distribution:

1. Unit 1 Division 1, Consisting of:

4 .

a) l

{ 4160-VAC Bus:

3

' b) 480-VAC Load Center: Du (10AuS) c) Du4 (108201)

480-VAC Motor Control Centers:

1 Du4-R-C1 (108219) t >

Ou4-R-C (108213)

Ou4-R-G (1082n) I 1

Ou4-R-G1 (108215) i

) 120-VAC ibution Panels: Du4-0-G (108515) 10Y101

$. I 10Y206 1

2. Unit Division 2. Consistia.1 of:

1 a) 4160-VAC Bus:

{; b) 480-VAC Load Center: Du (10AUE) c) Du4 (108202)

480-VAC Motor Control Centers: Du4-R-C1 (108220)

Du4-R-C (108214) '

Du4-R-G (108212)

Du4-R-G1 (108216) d) 120-VAC Distribution Panels: Du4-D-G (108516) 10n02 10Y207 g

3. Unit 1 Division 3, Consisting of: I a) 4160-VAC Bus:

b) O U (10A u7) 480-VAC Load Center: 0134 (108203) c) 480-VAC Motor Control Centers: 0134-R-H1 (108221) 0134-R-H (108217)

D134-R-E (101223)

D134-C-8 (008131) d) D134-0-G (108517) 120-VAC Distribution Panels: 10n03 I N

I

' ~

4. Unit 1 Division 4, Consisting of:

I a) 4160-VAC Sus:

014 (10A118) b) 480-VAC Load Center:

0144 (108204)

LIMERICK - UNIT '. 3/4 B-15 juli 2 2131 Amenament No. 24 N b 4 U

l . . . ,

, CMINis: :TIVE CONTD01.5 _

!!PONst?!L1 TIES 4 i.5.1.6 The all be resconsible for:

T j

a. Review of (1) Administrative Procedures and changes thereto, (2) new i

- programs or procedures requireo by specification 6.8 and requiring a 10 CFR 50.59 safety evaluatten, and (3) proposed changes to orograms or procedures required by Specification 6.8 and requiring a 10 CFR 50.59 safety I

evaluation; ;

b. Review of all proposeo tests and experiments that affect nuclear safety; j

c. Review of all proposeo changes to Appendix A Technical Specifications; f ~

d. Review of all proposed changes or modifications to unit systems or equipment that affect nuclear safety; l e. DELETED.

l

f. Investigation of all nolations of the Technical Soecifications, including the preparation and forwarcing of reports covering evaluation and j

recommencations to prevent recurrence to the Vice President, Limerick i Generating Station, Plant Manager, and to the Nuclear Review Boaro; I

g. Review of all REPORTABLE EVENTS;

h. Review of unit operations to detect potential hazaros to nuclear safety;

i. Performance of special reviews, investigations, or analyses and reports thereon as requested by the Vice President, Limerick Generating Station, plant Manager or the Chairman of the Nuclear Review Board; i

j. Review of the Security Plan and implementing procedures and submittal of recommended changes to the Nuclear Review Board; ano

k. Review of the Emergency Plan and implementing procteures and suomittal of the recommenced changes to the Nuclear Review Boarc. l

1. Review of every unplanned onsite release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation, recoussendations and disposition of the corrective action to prevent recurrence to the Vice President, Limerick Generating Station, Plant Manager, and to the Nuclear Review Board.

m. Review of changes to the PROCESS CONTROL PROGRAM, FFSITE DOSE CALCULATION MANUAL, and radwaste treatment systems. '

n. Review of the Fire Protection Program and implementing procedures and the submittal of recommended changes to the Nuclear Review Boarc.

6.5.1.7 The PORC shall:

a. Recommend in writing to the Plant Manager approval er disapproval of items considered under Spec 1fication 6.5.1.6a. through d. prior to their implementation. -

b. Render determinations in writing wita regard to wnether or not each item considereo under Specification 6.5.1.bb. througn f :enstitutes an unreviewed safety cuestion.

! LIMERICK - UNIT 1 6-8 Amenment N:. 4.47, na i m-4a e

A'DMINTSTRATIVE CONTR0ls -

RESPONSIBILITIES (Continued)

c. Provide written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the Vice President, Limerick Generating Station and the Nuclear Review Board of disagree-ment between the PORC and the Plant Manager; however, the Plant Man-ager shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.1.

RECORDS 6.5.1.8 The PORC shall maintain written minutes of each PORC meeting that, at a minimum, document the results of all PORC activities performed under the ;

responsibility provisions of these Technical Specifications. Copies shall be provided to the Vice President, Limerick Generating Station, Plant Manager, and the Nuclear Review Board.

6.5.2 NUCLEAR REVIEW BOARD (NRB)

FUNCTION 6.5.2.1 The NRB shall function to provide independent review and audit of designated activities in the areas of:

a. Nuclear power plant operations,

b. Nuclear engineering, .

c. Chemistry and radiochemistry, Posit.i o n #, m

d. Metall urgy,

e. Instrumentation and control,

f. Radiological safety,

g. Mechanical and electrical engineering, and

h. Quality assurance practices.

The NRB shall report to and advise the Senior Vice President and Chief Nuclear Officer on those areas of responsibility pertaining to NRB Review and Audits.

COMPOSITION 6.5.2.2 The Chairman, members, and alternates of the NRB shall be appointed in writing by the "' -" " #^"-"" '

' Chief Nuclear Officer, and shall have an academic l degree in an engineering or physical science field; and in addition, shall have a minimum of 5 years technical experience, of which a minimum of 3 years shall be in one or more areas given in Specification 6.5.2.1. The NRB shall be composed of no less than eight and no more than 12 members.

The members and alternates of the NRB will be competent in the area of Quality Assurance practice and cognizant of the Quality Assurance requirements of 10 CFR Part 50, Appendix B. Additionally, they will be cognizant of the corporate Quality Assurance Program and will have the corporate Quality Assurance organization available to them.

AUG 0 81995 LIMERICK - UNIT 1 6-9 Amendment No. 70,35,96

I l

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i APPENDIX B TO FACILITY OPERATING LICENSE NO. NPF-39 LIMERICK GENERATING STATION l

i UNITS 1 AND 2 l

1 l

1 PW ENEn-6v 1 rn mvi aiA E d.nw COMPANY !

DOCKET N05. 50-352, 50-353 j AUS 8 1985 ENVIRONMENTAL PROTECTION PLAN (NONRADIOLOGICAL)

l l

1

. l l

1 sensitive iano uses in the site vicinity (e.g., residences, schools, !

cnurenes, cemeteries, hospitals, parks); ano (3) previously concuttec noise surveys in the site vicinity.

The selection, calibration and use of equipment, conduct of the surveys, l ano the analysis and reporting of data shall confom to the provisions of 4 the applicable American National Standards Institute Standards.

The results cf the surveys conducted under this program shall be sumarizec.

interpretec anc reporteo in accorcance with Section 5.4.1 of this EFF.

The final report of this program shall present a brief assessment by the licensee of the environmental impact of plant and supplemental cooling l

water system operation on the various offsite acoustic environments, ano j 1

shall describe the mitigative measures, if any, that have been, or are to l

be taken to reduce the impact of plant or supplemental cooling water system noise levels on the offsite environments. This report shall also centa1A ff list cf noise-related complaints or inouiries received by PMiesci.. ie p

% Q$.-

Eleeme Company concerning the Limerick Generating Station or its supple?

I mental cooling water system subsecuent to issuance of the operating license along with a description of the action taken by PMiedelp.i: E'. e c t = !

Company to resolve these complaints or inquiries.

This program shall termainate upon completion of the collection of the specified sound level data for each phase ano submission of an accectable final report.

/(05IN I 44 A'.: a an E.PP b

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/,p c e o.,,\\

JNITED STATES TC273L~~**

~~'~'~~~~"

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NUCLEAR REGULATORY COMMISSION MSHINGTON. O. c. ::SS$ j

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{

PECO ENO2.(c-/ '

A A...T'I" C'.:-~ ~~ "CMPANY COCKET NO. ~ -353 LIMERICK GENERATING STATION, UNIT 2 FACILITY OPERATING C CENSE

i. -

License No. NPF-65 2.

1. The Nuclear Regulatory Comission (the :: mission or the NRC; has founc that:

A. The applicat1on fcr license filec ::y ""h MM.

?FLobem-

Company

'the iicensee; ccmoiies with the s ancares anc reouirements of tr.e Atomic Energy Act of 1954, as amencea (the Act), anc the Comission's

' regulations set forth in 10 CFR Chapter I, ano all recuired notifica-tions to other agencies or bodies nave been culy mace; B. Construction of the Limerick Generating Staticn, Unit 2 f the facility) i has oeen substantially completed in conformity with Construction Permit Mo. CPPR-1C7 and the application, as amended, the provisions ,

of the Act and the regulations of the Comission; I 4

C. The f acility will operate in conformity with the application, as amenced, the provisions of the Act, and the regulations cf the Coeaission (except as exempted from compliance in Section 2.D. belew);

1 2

C. ~bere is reasonable assurance: 'il that the activities authorizec by l this coerating license can be concuttee with::ut encangering the henir, are safety ci *ne Duolic, ec 'ii} trat sucn ac- "' ties w- '

ce ccrcuctec in comoliance with - e Comissien's regu".ations set fortn in 10 CFR Chacter ! (except as exemptec from ccmoiiance in Section 2.D. below); '

i j

E. The licensee is technically quali'ieo to engage in the activities authorized by this license in accercance with the Comission's regulations set forth in 10 CFR Chacter I; F. The licensee has satisfded the apolicable crevisions of 10 CFR Part 140, " Financial Protection Recutrenents ant' Indemnity Agreements.'

of the Comission's regulations; -

G. 'he issuance of this license will rot be inimical to the common cefense and security cr to the hecl? and safety uf the cublic; g;e m u .-

7 t ~.

... 7 7.f.,0

.-J 7 i '

i 4

i -2 1

i H.

! After weighing :ne environmental, economic, technical, anc other benefits of ite f acility against environmental and other costs anc consicering available alternatives, the issuance of this Facility Operating License No. !!PF-85, subject to the conditions for

.. protection cf t*,e environment set forth in the Environmental

. Protection F'an attached as Appendix B, is in accordance with 10 CFR Part 51 of the Consnission's regulations and all applicanle requirements nave been satisfied; and I.

The receipt, :ossession, and use of source, byproduct and special nuclear material as authorized by this license will be in accoraance with the Cemission's regulations in 10 CFR Parts 30, 40 and 70.

2.

Based on the foregoing findings and the Decision of the Atomic Safety and Licensing Boarc, .57-55-25, datec July 22, 1985, Ine Consnission's Order catec July 7, '989, ano the Commission's Memoranoum and Orcer cated August 25. 1989, regarcing this facility, facility Operating License NPF-85 is hereoy ytsued to the Phi'... '6 "' 1; company (the licensee), to rene as follows:

A Pedo Gw A.

This license applies to the Limerick Generating Station, Unit 2, a ;

boiling water nuclear reactor and associated equipment, owned by

?@ Phildhi. :';mic Company. The facility is locatec cn the l licensee's site in Montgomery and Chester Counties, Pennsylvania on gE the banks of the Schuylkill River approximately 1.7 miles southeast of the city limits of Pottstown,' Pennsylvania and 21 miles northwest ,

of the city limits of Philadelphia, Pennsylvania, and is described ~j in the licensee's Final Safety Analy. sis Report, as supplemented and amended, and in the licensee's Environmental Report-Operating License Stage, as succlemented 'anc amenced. j i

5. Iubject I: tre ::ncitions anc requirements corocratec nerein, Ine !

Consnission rerecy licenses Tbi':i:' MQ Company: !

(1) Pursuant to Section 103 o o CFR Part 50, to possess, use, and operate the facility at the cesignated location in Montgomery and Chester Counties, Pennsylvania, in accorcance with the procedures and limitations set forth in this license; (2) Pursuant t: the Act and 10 CFR Part 70, to receive, possess and to use at any time special nuclear material as reactor fuel, in accorcance with the limitations for storage and amounts reouirea for react:r operation, as described in tne Final Safety Analysis Report, as supplemented and americed; (3) Pursuant t: the Act anc 10 CFR Parts 20, 40 and 70, to receive, possess ana use at any time any byproeuct, source anc special (2 Jflg nuclear aterial as sealec neutren sources for react:r startup, sealed sources for reactor. instrumentation and raciation monitoring eouipment calibration, and as fission cetectors in L amount: as recuirec; m5=

,u .

j .I . .

1-i

'4) Pursuant to t e Act ano 10 CFR Parts 30, 40, ~3, to receive, j possess, ano use in amounts as recuireo any oyprocuct, source er

special nuclearJ.aterial without restriction to enemical or l physical form. :or samose analysis or instrument calibration or associated with racicactive apparatus or components; ana

'5)

Pursuant to the Act ano 10 CFR Parts 30, 40 ano 70, to possess, j but not separate. such byproduct and special nuclear materials as

may be produceo by the operation of the facility, and to receive and possess, but not separate, such source, bycrodcct, and special i nuclear materials as contained in the fuel unalies ano fuel l channels from the Shorenam Nuclear Power Sttt %

j (C) This license shall be deemed to contain and is subject to the conditions i

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 Cosenssion now or hereafter in effect; ano is suoject to the noditional conditions soecifiec or incorporated below

j :1) Maximum Power uvei i

j

W 9 T h Ihcr.; Compan

s authorized to operate the.

j facility at reactor core powr. levels of 3458 megawatts thermal a

(100 percent rated power) in cordance witn the conditions specified herein.

1 (2)

Technical Soec1 F eations ?EC.O6Mr% Og l The Technical Specification contain % in Appenoix A and the 1

Environmental Protection P n ecat.ained in Appencix B, as reviseo through Amenoment No. , are hereoy incorporated into this i

l license. P.thniphi; :.ecnic Company shall operate the facility in accordance with the Technical Specifications ano the

j Environmental Protection Plan. '

(3) Fire Protec"en (Section 9.5. SSER-2.-U* l l

' .tiliiphi; Ch;;. .i Company shall implement and maintain in effect all provisions of the approved Fire Protection Program as j described in the Updated Final Safety Analysis Report for the facility, and as approved in the NRC Safety Evalustion Report i

h dated August 1983 through Supplement 9, dated August 1989, and V Safety Evaluation dated November 20 , 1995, subject to the 1

following provision:

i The licensen may make changes to the approved fire protection i

program without prior approval of the Commission only if those j changes would not adversely affect the ability to achieve and g

maintain safe shutdown in the event of a fire.

}

The parenthetical notation following the title of license conditions cenotes the section cf the Safety Evaluation Report and/or its supplements wnerein the license

i. condition is discussed.

J Q 'E Amenoment No. 7.27.37.55 hb

{' , .-

C T .: ', : / - .

l i

.i.

the local leat rate test 1ng of the Traversing In re Proce Shear (d) an exemotion

valves (Section from the seneaule reovirements 6.2.5.1 cf50.33 cf 10 CFR the SER anc SSER.3)lk availability of funos for cecomissionine the facility (Section 22.1 SSER 8]/ n; is; u.mp u u.. . . vi . w. ,w .r x: : ;f .; ;TT r;rt)

F.2.u,;;;'ae-ttg:#

cr W :: l i t ; ' !:: t er ' 2 * " t:t.:;nt

' " r ; F.' :i:The; spec

.tM1alcit:r Sitif;1 circumstances regarding exemptions (a), (b) ano (c) are identified in Sections i

6.2.6.1 of the SER and SSER 3. /An ex clon from

)

e criticalit TEV i Emonnor~in equirements o 10 CFR 7 . 4 was pre - usly granted ith

' NRC ma als license i . SNM-197' issued Nov er 22, 1988 The lice ee is hereby e mpted fro the requi ents of 10 70.24 '

i far as this r irement a lies to t handling an storage Lue_1 assemblies eld uncer is licen .

J

'hese exemot1ons are author 1 zed by law, will not present an unoue

' risk to the puDlic health and safety, and are consistent with the comon defense and security. The exemotions in items a, b, c, d, ano l e above are granted pursuant to 10 CFR 50.12. With these exemottens, the facility will operate, to the extent authorized herein, in

' conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Comission.

E. Except as otherwise provided in the Technical Specifications or Environmental Protection Plan, the licensee shall report any violations of the requirements contained in Section 2.C of this license in the following manner: initial notification shall be mace within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the NRC Operations Center via the Emergency i

Notification System with written followup within thirty cays in accordance with the procecures described in 10 CFR 50.73(b), (c),

, and (e).

5 he licensee snall have ano maintain financial orotectler :f suen type and in sucn amounts as the Ccmission shall require in accordance with Section 170 cf the Atomic Energy Act of 1354, as i amended, to cover public liability claims.

i 1

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L&S U i 2-F:o L--

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

1GG IMITING CONDITTONS FOR OPERATION AND SURVEILLANCE *E00fREMENTS 2CIlE P.AE

'MSTRUMENTATION (Continued)

. .. .-... ,i r uw v.6.u.un inm - u n.m............. . , , , , -

Loose-Part Detection Systas...............................

3/4 3-97 The information from pages 3/4 3-98 through 3/4 3-101 has been intentionally omitted. Refer to note on page 3/4 3-98.................. 3/4 3-98 -

Offga: Monitoring Instrumentation......................... 3/4 3-103 Table 3.3.7.12-1 Offgas Moni tori ng Instrumentati on. . . . . . . . . . . . . . . 3/4 3-104 Table 4.3.7.12-1 Offgas Monitoring Instrumentation Surveillance Requirements................ 3/4 3-107 3/4.3.8 (Deleted) The infomation on pages 3/4 3-110 and 3/4 3-111 has been intentionally omitted.

Refer to note on page 3/4 3-110................. 3/4 3-110 3/4.3.9 FEEDWATER/ MAIN TUR8INE TRIP SYSTEM ACTUATION INSTRIMENTAT!0N. . . . . . . . . . . . . . . . . . f . . . . . . . . . . . . . . . . . . . . 3/4 3-112 Table 3.3.9-1 Feedaater/ Main Turbine Trip System Actuation Instrumentation......... 3/4 3-113 Table 3.3.9-2 Feedwater/ Main Turbine Trip System Actuation Instrumen-tation Setpo1nts......................... 3/4 3-114 Table 4.3.9.1-1 Feedwater/ Main Turbine Trip System Actuation Instrumenta-tion Surveillance Require-seats..................................... 3/4 3-115 3/4.A REACTOR tootANT SYSTEM 3/4.4.1 RECIRCUI.ATION SYSTEN Recirculation Loops....................................... 3/4 4-1 SEP 21 1995 LIMERICK - UNIT 2 x jaenomen: So. II, M

.u - -

. l!!D.El USES IECTION

.P.3.91

"'ISTRUMENTATION (Continued)

'(Deleted)..................................................... B 3/4 3-5

( D el e t e d ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 3-5 Remote Shutdown System Instrumentation and Contreis........... B 3/4 3-5 Accident Monitoring Instrumentation........................... B 3/4 3-5 Source Range Monitors......................................... B 3/4 3-5 (Deleted)...................................................... B-3/4 3-6 Chlorine and Toxic Gas Detection Systems...................... B 3/4 3-6 (Deleted)..................................................... B 3/4 3-6 ,

i Loose-Part Detect :n i/ stem. .................... . .......... B 3/4 3-7 !

(Deleted)..................................................... B 3/4 3 7 Offgas Monitoring Instrumentation............................. B 3/4 3-7 3 /4.'t .B ?"'."!NE C" n!"EED T20TtCTi0ii Pf6Tt!T. .bf/F.. .,........... B 3/4 3-7

.' .9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION............................................... B 3/4 3-7 .

i Bases Figure B 3/4.3-1 Reactor Vessel Water Level........................... B 3/4 3-8 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM. . . . . . . . . . . . . . . . . . . . ... ... ....... B 3/4 4-1 )

l 3/4.4.2 SAFETY / RELIEF VALVEI........................... ....... ... B 3/4 4-2 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems..................................... B 3/4 4-3 l

l Operati on al Le a kag e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-3 3/4.4.4 CHEMISTRY..................................................... B 3/4 4-3a ,

l i

f LIMERICK - UNIT 2 xix Amencmr . . . ;;. 12, 17. 32. 35. 5B.

JLR. I 1 E

s .

P . <

TA8tE 4.3.1.1-1 (Continued) . t REACTOR PROIECIION SYSIEM INSTRUMENIA110N SUHVLjllANCE RLOUIREMENil OPERATIONAL C CilANNEL CONDITIONS FOR WillCil CilANNEL FUNCTIONAL S!8RVfill A!!!1 RfD!!IBril M CHANNEL

_.IfST _ EAlIBRAUallf al '

5 _[Il((!L I p JUNC110NAl UN[1 Q

R N.A.

9. Turbine Stop Valve - Closure h

2 M 10. Inrbine Control Valve fast R I

to Closure, Trip Oil N.A. Q Pressure - tow ,

N.A.

1,2,3,4,5

11. Reactor Mode Switch M.A. R N.A.

I,2,3,4,5 Shutdown Position H.A. W

12. flanual Scram l' during each startup after

[a) Neutron detectors may be excluded from tilANNEL CAtl8 RATION. i d to overlap for least 1/2 (b) The IRN and SRM channels shall be determined to overlap for at l :

decades during each controlled shutdown, the power values calculated by

,t *

(t.)

(d)DEIETED This calibration shall consist of the adjustment of the APHMEDthannel THERMAL POWER. Adjust the APRM to <onf orm to i' t a heat balance during OPERATIONAL CONDITION h I when 1HERMAL POWER

= channel if the (e) This calibration shall consist of the adjustment of the APRM flow hiased c anne l >

signal. IPRMs shall be calibrated at least once per 1000 effective fulli power hours (EFPil).h d core flo 3 (f) The (g) Verify measured core flow (total core flow) 9 loop flow (APRM X flow). During the startup testtoprogram, be greater thanshall data or equal be recorded to estab ns forethe par ,

t 3 provide a basis for establishing the specified relationships.

" Specification g

the criteria IIsted shall commence upon the conclusion of the startup test p i

-i (h) 3.10.1. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.1

~

(i) With any control rod withdrawn.

O (J) If the RPS shorting links are required to be removed per Specification 3.9.2, they ma i

' hours for required surveillance. tions as performed per Spetilication

~

XI be moved from its existing position. ,.

- (L) Hequired to be OPlRABIE uniy prior to .ind during shutdown margini desieuristra i

,'"' ' s 3.10.3. i

! j.

N I

N

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

~ I' STRUu!NTATION N

i j OFFGAS GAS MONITORING INSTRUMENTATION '

l LIMITING CONDITION FOR OPERATION -

(

i j

i 3.3.7.2 The offgas monitoring instrumentation enannels shown in Table 3.3.7.12-1 j

shall be OPERABLE with their alare/ trip setpoints set to ensure that the limits i

of Specifications 3.n.2.5 and 3.n.2.6 respectively, are not exceedea. '

i

APPLICABILITY: As shown in Table 3.3.7. u -1

_ ACTION:

a.

With an offgas monitoring instrumentation channel alara/ trip setpoint less conservative than required by the above Specification :eclare the enannel inoperable, and take the ACTION shown in Table 3.3.7.u-1.

b. With less than the minizia number of effgas monitoring instrumentation channels OPERABLE, taka the ACTION shown in Table 3.3.7.u-1. Restore i

the inoperable instrumentation to OPERABLE status within the time l specified in the ACTION or explain wny this inoperability was not corrected in a timely manner in the next Radioactive j Effluent Release Report.

um t j c.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

4 SURVEILLANCE REQUIREMENTS i

j 3

4.3.7.u Each offgas monitoring instrumentation channel shall be demonstrated ,

OPERABLE by perfarmance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, 3

i and CHANNEL FUNCTIONAL TEST operations at the frecuencies shown in Tabl e 4. 3. 7. u-1.

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LIMERICK - UNIT 2 3/4 3-103 I Amenamen: No.11 JAX02ltt b

~

'A.5 EMEPGENCY CORE ~~ SLING ?*'! 35 i /4.5.1 ECCS - OPEPAT*';G

.!MITING CONDITION F0D OPEPATiON 1.5.1 The emergency core cooling systems shall be OPERABLE with:

a.

The c:re spray system (CSS) consisting of two subsystems with eacn subsystem comoriseo of:

1. Two OPERABLE CSS pumps, and 3

2. An OPERABLE flow patch capable of taking suction from the suppression chamber and transferring the water through the spray sparger to the reactor vessel.

b. Th'e iow pressure coolant injection (LPCI) system of the residual heat removal system consisting of four subsystems with each subsystem comprised of:

1. One OPERABLE LPCI oump, and

2. An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.

c. The high pressure coolant injection (HPCI) system consisting of:

1. One OPERABLE HPCI pump, and

, 2. An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.

! d. The automatic depressurization system (ADS) with at least five OPERABLE ADS valves.

PPLICABIL: V- OPERATIONAL ~0NDITION 1, 2* ** r, ano 2+ ** rr l

The HPCI system is not required to be OPERABLE when reactor steam done pressure.is less than or equal to 200 psig. .

- The ADS is not required to be OPERABLE when the reactor steam dome pressure is less 'that or equal to 100 psig.

See Special Test Exception 3.10.6.

L

1. Two LPCI subsystems of the RHR system may be inoperaole in that they are aligned

' in the shutdown c:oling mode when reactor vessel pressure is less than the RHR Shutdown cooling permissive setpoint. ff the 14 day extended A0T period allowed for installation of th- -- z.. m ano RHISW

'(_

system ns P-0166, P-0167, and P-0168 durin- + sixth refueling outage (1R06), in addition 6u ...; "wrabl e 's' ' ' e r I subsystems of the RHR system and the ' A' CSS subsystem. the 'e' LToi suu,y.x af the RHR system may be inoperable in that

i it is ali e " . .ne snutdown cooling mode when reactor vom o-...; : '"" *%n the 4

1 '".,, =

snutcown Cooling permissive setpoint. 7

.1MERICK - UNIT 2 3/4 5-1 .4nen:::nent tio. D Dtc 2 7 2

NED.GENCY CORE COOLING SYSTEMS

. _IMITING CONDITION FOR OPERATION Pentinum ACTION:

a. For the core spray system:

1. !

With one CSS suosystem inoperable, provided that at least two LPCI l subsystems are OPERABLE. restore the inoperable CSS suosystem to

,3 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

2. With both CSS suosystems inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

- b.

For the LPCI system:

1. With one LPCI suosystem inoperable, provided that at least one CSS subsystem is OPERABLE, restore the inoperable LPCI pump to OPERABLE status within 10 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 ano in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2. With one RHR cross-tie valve (HV-51-282 A or B) open, or power not removed from one closed RHR cross-tie valve operator, close the open valve ano/or remove power from the closeo valves operator within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , 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 next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3. With no RHR cross-tie valves (HV-51-282 A, 8) closed or power not removed from both closed RHR cross-tie valve opera, tors, or with one RHR cross-tie valve open and power not removed from the other RHR cross-tie valve operator, be in at least HOT '

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

4 With two LPCI subsystems inoperable, provided that at least one CSS l subsystem is OPERABLE, restore at least three LPCI subsystems 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 u.e following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

5. With three LPCI suosystems inoperable, provided that both CSS subsystems are OPERABLE, restore at least two LPCI subsystems to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 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 .

6. With all four LPCI subsystems inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .*

Whenever cath shutdown cooling subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.

A0Ts1 of the 'A' Core Spray subs s * '

e' LPCI subsystems may be extence ESW and RHRSW s r installation of the conson o ons r-0166, P-0167, an -

o e th t ref age (IR06).,

LIMERICK - UNIT 2 3/4 5-2 *"~"'t 2 I3. 3 DEC 2 7 7J95

CONTAINMENT SYSTEMS ,

l SURVEILLANCE REQUIREMENTS (Continued)

c. By verifying at least 8 suppression pool water temperature indicators in !

at least 8 locations, OPERABLE by performance of a:

1. CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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

3. CHANNEL cAllBRATION at least once per 24 months, with the .emperature alarm setpoint for:

1. High water temperature:

a) First setpoint s 95'F b) Second setpoint s 105'F c) Third setpoint s 110*F d) Fourth setpoint s 120*F

d. By verifying at least two suppression chamber water level indicators OPERABLE by performance of a:

1. CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

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

3. CHANNEL CAllBRATION at least once per 24* months, with the water level alarm setpoint for high water le 24'l-1/2"

e. Drywell-to-suppression chamber bypass le kN tests shal conducted to coincide with the Type A test at an in ial differential pressure of 4 psi and verifying that the A/(k calculat from the measured leakage is within the specified limit.

If any drywell-to-suppression chamber bypass leak test fails to meet the specified limit, the test schedule for subsequent tests shall be reviewed and approved by the Commission. If two consecutive tests fail to meet the specified limit, a test shall be performed at least every 24 months until two consecutive tests meet the specified limit, at which time the test schedule may be resumed. }

j

f. By conducting a leakage test on the drywell-to-suppression chamber I i

vacuum breakers at a differential pressure of at least 4.0 psi and !

verifying that the total leakage area A//k contributed by all vacuum breakers is less than or equal to 24% of the specified limit and the leakage 4

i area for an individual set of vacuum breakers is less than or equal to 12% of the specified limit. The vacuum breaker leakage test shall be conducted during each refueling outage for which the drywell-to-suppression chamber bypass leak test in Specification 4.6.2.1.e is not conducted.

The CHANNEL CAllBRATION for level transmitters LT-55-2N062B, -2N062F shall be performed at least once per 18 months.

LIMERICK - UNIT 2 JAN 2 41997 3/4 6-14 Amendment No. 347 -337-347-81

. . . l Z NTAINMENT r'!TEMS IUpPRESSION :0OL ScRAY l

!MITING CCNDITION AD GPERATION

)

2.6.2.2

'he sunoression pool scray moce of the residual heat removal (RHR) system snali :e OPERABLE with two indepencent loops, eacn loop consisting of:

{

s. One OPERABLE RHR pump, and b.

An OPERABLE flow path capable of recinulating water from the suppression chamber througn an RHR hte. exchanger and the suppression pool spray sparger(s).

'PPLICABILI M OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a.

With one suppression pool spray loop inoperable, restore the inomaele 1000 to OPERABLE status within 7 days or be in at least HOT SHUT 00WN within the next 12 h urs and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months g

o. With both suppression cool spray loops inoperable, restore at least one loop to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-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 DEOUIREMENTS 4.6.2.2 ine suppression pool spray mooe or tne kHR system snail, De cemonstrateo OPERABLE:

a. i At least once per 31 days by verifying that each valve (manud, ower- !

operated, or automatic) in the flow path that is not locked, sealej.

or otherwise secured in position, is in its correct position.

. By verifying that eacn of the reautreo RHR pumos deveicos a flow of at least 500 gpm on recirculation flow tnrougn the RHR heat exenangr.r and the suppression pool spray sparger when tested pursuant to Spect-fication 4.0.5.

Whenever :stn RHR subsystems are inocerable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methocs.

r Ine a ace time (A0T) of the ' A' 1000 of e

mode of the RHR sy .

a oa cays to allow for installation W system o P-0167 and P-0168 ouring (of the. .como sixth refueling outage (IR06). t LIMERICK - JIT 2 3/4 6-15 W "t M D' U nc 2 7 55

l SITAINMENT SYSTEMS I

""JPPDESSION POOL COOLI'M

'u! TING CONOTTION F0D POEDATION 1.5.2.3 The suppression pool cooling mooe of the resioual heat removai (RHR)

ystem snail be OPERABLE with two inoepencent loops, asen loop consisting of:

a. One OPERABLE RHR cump, and

b. An OPERABLE flow path capable of recirculating water from the suppression chamoer througn an RHR heat exchanger.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a. With one suppression pool cooling 1000 inoperaole, restore the inoperable loop to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 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

b. With both suppress 1on pool cooling locos inocerable, 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 tne next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE PEOUIREMENTS l

4.6.2.3 The suppression pool cooling mode of the RHR system shall be demonstrated {

OPERABLE: !

a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or cthentise secured in position, is in its correct position.

b. By verifying that each of the reouireo RHR pumos develoos a flow of at least 10.000 gpm on recirculation flow througn the ficw cath including the EHR heat excnanger ano its associatec closea bypass vaive, the suppression pool ano the full flow test line wnen testeo cursuant to Specification 4.0.5.

Whenever oath RHR suosystems are inoperable, if unable to attain COLD SHUTDOWN as reauired by this ACTION, maintain reactor coolant temperature as low as '

practical by use of alternate heat removal methods. -

3 l

The n Time (A0T) of the ' A' loop of th e' cooiing mode of the RHR syst a . *

. ays to allow for installation of the common ESW a . stem to ... D-0166. P-0167 and P-0168 curing the Uni refueling outage (1R06).

LIMERICK - UNIT 2 3/4 6-16 .wan=nenc 2. 22, 70 D I.C 2 7 2

c. - ~

TABLE 3. 5. 3-1 PRIMARY 00NTAlhMEhi 150LATION VALVE 5 NOTATION NOTES (Continued)

15. Check valve used insteam cf flow orifice.

16. Penetration is sealed cy a flange with couble 0-ring seals. These seals are leakage rate testeo ey pressurizing between the 0-rings. Both tne TIP Purge Supply (Penetratien 2SB) anc the TIP Drive Tubes (Penetrations 35C thru G) are welded to their respective flanges. Leakage througn these seals is included in the Type C leakage rate total for this penetration.

The ball valves (XV-241A nru E) are Type C testad. It is not cracticaole to leak test the snear valves (XV-240A thru E) because souib firing is required for closure. Shear valves (XV-240A thru E) are nomally open.

17. Instrument line isolation provisions consist of an excess flow enact valve.

Because the instrument line is connectec to a closeo cooling water system inside containment, no flow orifice is provided. The excess flow eneck valves are subject to operacility testing, but no Type C test is performed nor requiren. The line does not isolate curing a LOCA and can leak only if the line or instrument snould rupture. Leantightness of tne line is verified curing tne integrated leak rate test (Type A test).

18. In addition to double "0" ring seals, this penetration is tested by pres-surizing volume between doors per Specification 4.6.1.3.

19. The RHR system safety pressure relief valves are flangen to facilitate rsmoval and are eouippee _with double 0-ring seal asseanlies on the flange closest to primary containment. These seals will be leak rate tested by pressurizing between the 0-rings, and the results anden into the Type C total for this penetration.

20. See Specification 3.3.2, Table 3.3.2-1, fa a description of the PCRVICS isolation signal (s) that initiate closure of each automatic isolation valve.

In addition, the following non-PCRVICS isolation signals also initiate closure of selected valves: g-n u.,_ .. __ ,,__

ur,u , - - a, ...

.:_m _ _ _ _ . . . . . u _m a,

u... . . . . . . , , _

4 d4 #

LFHP With HPCI pumps running, cDens on low flow in associaten pipe, closes l when flow is above setpoint LFRC with RCIC pump running, opens on low flow in associated pipe, closes-when flow is above setpoint LFCH With CSS puso running, opens on low flow in associated pipe, closes when flow is above setpoint LFCC Steam supply valve fully closed or RCIC turbine sten valve fully closed All power operated isolation valves may be opened or closeo remote manually.

LIHERICK . UNIT 2 3/4 6-42 AUG 2 51989 w& . - -

. TaEM ? .b.3 1 4

' "

PRIMaRV CONTa W ENT 150 LATIN valve 5

iOTATION i NOTES (Continued)

! IT".- automatic isolation signal causes TIP to retract; ball valve closes wnen prone is fully retractea.

l 22. Isolation barrier remains water filled or a water seal remains in the line post-LOCA. Isolation valve may be tested with water. Isolation 2

valve leakage is not included in 0.60 La total Type B & C tests.

! 23. Valve does not receive an isolation signal. Valves will be open during Type A test. Type C test not required

' Both isolation signals required for valve closure.

24.

i '

I l 15. Deleted

26. Valve stroke times listed are maximum times verified by testing per Speci-fication 4.0.5 acceptance criteria. The closure times for isolation valves i i in lines in which high4nergy line breaks could occur are identified with a ;

single asterisk. The closure times for isolation valves in lines which J

provide an open path from the containment to the environs are identified ,

with a double asterisk.

! 27. The reactor vessel head ses1 leak detection line (peretration 29A) excess i

flow checa valve is not sub,iect to OPERABILITY testihg. This valve will not be exposed to primary system pressure except unoer the unlikely con- ,

4 ditions of a seal failure wnere it could be partially pressurizec to l i reactar pressure. Any leakaoe oath is restricted at the source; therefore, i this valve need not be OPERABILITY testsd. 1 I

l 28. (DELETED)

29. Valve may be open during normal operation; capable of manual isolation 1 i

from control room. Position will De controlled procedurally.

30. Valve nomally open, closes on scram signal. ,

i 31. Valve 41-2016 is an outboard isolation barrier for penetrations X-gA, B and X-44. Leakage through vaive 41-2016 is included in the total for

! penetration X-44 only.

32. path recirculation valves are sealed closed whenever the

- Feedwater reactor is cri longtical and reactor pressure is greater than 600 psig. The valves are expected to be openeo only in the following instances:

a. Flushing of the concensate and feeowater systems curing piant startuo.

i b. Reactor pressure vessel hydrostatic testing, which is conducted follow-ing each refueling outage prior to connancing plant startuo.

Therefore, valve stroke timing in acecrdance with Specification 4.0.5 is not required.

33. Valve also constitutes a Unit 1 Reactor Enclosure Secondary Containmen:

Automatic isolation Valve ano a Refueling Area Secondary Containment Au.cmati:

Isol ation Valve .r - t - - ~i : . ' . ; . .. . . ... T.;/. . ^. . ; . ^. . .; ; .

34. Isolation signal causes recomciner to trip; valve closes when recombine-is not coerating.

35. Auto isolation signals have oten removed from Hi'-087-224 A/B and 225 A/E, Valves to oe closeo with associated circuit creakers lockea open curing

~

OPCONs ;, 2, and 3.

l$ I7 E 1'

LIMERICK - UNIT 2 3/4 6-43 Amencmen; i.e E'

3/4.7 8 ANT "v5TEMS i 3/4.7. TERVICE WATER SYSTEMS i

i RESIDUAL -EAT :EMOVAL SERVICE WATER SYSTEMr:MMON SYSTEM f

LIMITTUG CONOTTION FOR OPEDATTON 3.7.1.1 At least the following indecencent residual heat removai :ervice water 4

(RHRSW) system suosystems, with eacn suosystem ccmortseo of:

a. Two OPERABLE RHRSW pumos, and

b.

An OPERABLE flow path capable of taking suction from the RHR service water pumps wet pits which are supplieo'from the spray pond or the cooling tower basin and transferring the water through one Unit 2 RHR heat exchanger, shall be OPERABLE:

a. :n OPERATIONAL CONDITIONS 1, 2, ano 2, two suosystems.

i

b. In OPERATIONAL CONDITIONS 4 and 5. the subsystem (s) associated with l

' :ystems and components reoutrea OPERABLE by toecification 2.4.9.2.

2.9.11.1, ano 3.9.11.2.

! t APPLIC BILITV- OPERATIONAL CONDITIONS 1, 2, 3, 4, and 5.

ACTION:

a. In OPERATIONAL CONDITION 1, 2, or 3: I 8 l

1. With one RHRSW pump inoperable, restore the inoperaole pumo to OPERABLE status within 30 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 .

2. I

With one RHRSW pump in each subsystem inoperable, restore at '

least one of the inoperable RHRSW pumos 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 ano in COLD SHUTDOWN othin tne followinc 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2. With one RHRSW subsystem otherwise inoperable, restore the inoperable subsystem to OPERABLE status with at least one OPERABLE RHRSW pumo within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT l

{

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 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months sh 4 With both RHRSW subsystems otherwise inoperable, restore at least one subsystem to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or ce 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 i SHUTDOWN

within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . '

'Whenever :otn RHRSW suosystems are inoperable, if unaole to attain COLD SHUTDOWN as recuireo by this ACTION, maintain reactor c:olant temperature as low as practical by use of alternate heat removal metnods.

em "- OTl of th

i

. . . ay oe extended *.:o to 14 days to allow for *-

o . .

-~'

MRtW system Modifications

( P-0J1 L 7- iw anc P-0168 during the Unit 1 :1xth refueling outage unA..

LIMERICK - UNIT 2 3/4 7-1 ;mmament No. 23, 3 DI,C 2 7 E

. I SLANT SYSTEMS EMERGENCY SERVICE UATEP. !v; gu _ ecuMON SYSTEu LIMITING CONDIT*0N FOR OP9tT;0N 3.7.1.2 At least the folicwing noecencent amergency serv' e water system loops, with eacn 1000 c:moriseo cf: ;

a. Two OPERABLE emergency service water pumos, :nd

b. An OPERABLE low pain cacaole of taking sucti:n fr m tne emergency service water pumps wet cits wnich are supplied from .ne spray pond or

)

the cooling tower basin ano transferring the water I: the associated Unit i 2 and common safety-related eouipment, shall be OPERABLE:

a. In OPERATIONAL CONDIT:0NS 1, 2, and 3, two icops.

b. In OPERATIONAL CONDITIONS 4, E, and *, one loop.

'PPl!CABili m OPERATIONAL CONDITIONS 1. 2, 3, 4, E. and

CTION:

a. In OPERATION CON 0! TION 1. 2, or 3:

1. With one emergency service water pump inoperabie, restore the inoperable pump to OPERABLE status within 45 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 ite following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . ]

2. With one emergency service water pumo in each loco inoperable, restore at least one inoperaole pump to OPERABLE :tatus within 30 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 ano in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3. With one emergency service water system ioop cinerwise ino erable, declare all eautoment alioneo to the inoperaole 1000 rooeraole* restore the inoceraole loco .i CPEPABLE status witt at least :ne OPEPABLE pumo w1 thin 7*

hours or ce in at least HOT SHUTDOWN wunin tne text 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ano in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

When nanaling trraciatea fuel in the secondary c:ntainment.

- The diesel generators may be aligneo to the OPERABLE emergency service water system loop provided confirmatory flow testing has been :erfomec. Those ciesel generators ,

not aligned to the OPERABLE emergency service water system ioop snall be declared inoperable and the actions of 3.B.1.1 taken.

r w=ine n.b Oc++7m Time f A0T1 of the ' A' 1000 of 9. r-- ;;. , 3, v u.e water system may be extended up to 14 o m - T. nstasiation :f the connon ESW ano RHRSW system Modifi w%.. r-0165, ?-0167 and P-0168 :.ony

. . . . * ' sixth refueling

- }

l

'1MERICK - UNIT 2

. 3/4 7-3 Ame~ m 2. IS, 70 Di,C 2 7 1995 i

g s =*

2 ,. , GMINIST:Att'/E r0NTD0ls

EEPONSfET E ES i.5.1.6 The nail be resconsible for:

( .

a. i Review of (1) Admin 1strative' Procacures and changes thereto. (2) new programs or procecures reoutree by Specification 6.8 anc reautring a 10 CFR 50.59 safety evaluation, and (3) proposeo changes to programs or procedures requireo by Spectft:stion 6.3 anc requiring a 10 CFR 50.59 safety evaluation; b.

Review of all proooseo tests ano experiments that affect nuclear safety;

c. Review of all processo changes to Appendix A Technical Specifications; d.

Review of all prooosed changes or modifications to unit systems or equipment that affect nuclear safety;

e. DELETED.

f.

Investigation of all violations of the Technical Specifications, including the preparation ano fomercing of reports covering evaluation and reconsnencations : Orevent recurrence, to the Vice President. Limerick Generating Station. .:lant Manager, and to the Nuclear Review Board; e

9 Review of all REPORTABLE EVENTS;

h. Review of unit operations to detect potential hazards to nuclear safety; i.

Performance of special reviews, investigations, or a'nalyses and reports thereon as requested by the Vice President, Limerick Generating Station, plant Manager or the Chaiman of the Nuclear Review Board; )

j. Review of the Security Plan and implementing procedures and submittal of i

recommended changes to the Nuclear Review Board; and j

k. Review of the Emergency Plan and implementing procedures and submittal of l the recommended enanges to the Nuclear Review Boaro.

1 Review of every uncianned onsite release of raoicactive material to the environs including the preparation and forwarding of reports covering

. evaluation, recomendations and disposition of the corrective action to prevent recurrence to the Vice President, Limerick Generating Station, Plant Manager, and to the Nuclear Review Board.

m. Review of changes to the PROCESS CONTROL PROGRAM, OFFSITE DOSE CALCULATION MANUAL, and raowaste treatment systems.

n. Review of the Fire Protection Program and implementing procedures and the submittal of recomenced changes to the Nuclear Review Boaro.

5.5.1.7 The PORC shall:

a. Recommend in writing to the ?lant Manager approval or disapproval of items considered under 5:ecification 6.5.1.6a. througn d. prior to their implementation.

b. Render determinatt:ns in writing with regard to wnether or not each item considered uncer 5:ecification 6.5.1.6b. througn f. :enstitutas an unreviewea sarety tuestion.

.lMERICK - UNIT 2 6-8 :mencment '.:. 4.U

,u .

., g,

. , ADMINISTRATIVE CONTROLS RESPONSIBILITIES (Continued)

' c. Provide written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the Vice President, Limerick Generating Station and the Nuclear Review Board of disagree-4 ment between the PORC and the Plant Manager; however, the Plant !

Manager shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.1.

RECORDS

$ 6.5.1.8 The PORC shall maintain written minutes of each PORC meeting that, at a minimum, document the results of all PORC activities performed under the responsibility provisions of these Technical Specifications. Copies shall be

' provided to the Vice President, Limerick Generating Station, Plant Manager, I and the Nuclear krt'.ew Board.

[

6.5.2 NUCLEAR REVIEW BOARD (NRB)

FUNCTION !

6.5.2.1 The NRB shall function to provide independent review and audit of i designated activities in the areas of:

a. Nuclear power plant operations,

b. Nuclear engineering, e

c. Chemistry and radiochemistry, .

Positei o n *g - -

d. Metallurgy,

e. Instrumentation and control,

f. Radiological safety,

g. Mechanical and electrical engineering, and

h. Quality assurance practices.

The NRB shall report to and advise the Senior Vic; 'N.sident and Chief Nuclear Officer on those areas of responsibility pertaining to NRO Mew and Audits.

COMPOSITION 6.5.2.2 The Chairman, members, and alternates of the NRB shall be appointed in writing by the D ' "i b-:" " Chief Nuclear Officer, and shall have an .-

academic degree in an engineering or physical science field; and in addition, shall have a minimum of 5 years technical experience, of which a minimum of 3 years shall be in one or more areas given in Specification 6.5.2.1. The NRB shall be composed of no less than eight and no more than 12 members.

The members and alternates of the NRB will be competent in the area of Quality Assurance practice and cognizant of the Quality Assurance requirements of 10'CFR Part 50, Appendix B. Additionally, they will be cognizant of the corporate ,

Quality Assuranc.e Program and will have the corporate Quality Assurance '

organization available to them.

J LIMEkICK - UNIT 2 6-9 Amendment No. 2,60 l AUG 08 1995 l

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APPENDIX B '

TO FACILITY OPERATING LICE!iSE NO. NPF-55 LIMERICK GENERATIllG STATION

~ UNITS 1 AND 2 P6co EM6/2 t

' .LO f' M f.n C COMPANY !

DOCKET N05. 50-352 AND 50-353 1

ENVIRONMENTAL PROTECT:0N PLAN (NON-RADIOLOGICAL)

August 25, 1999 l

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,, ' 0 7 0 ; ; t. 2 . s l .he select 1on, :slibratt:n ano use of ecutoment. ::nouct Of the . surveys, anc

{ the analysis ano recorting of data snall cenfom :: the crovisions :f the accli:aole American Nati:nal Stancarcs Institute itancaros.

i The results of the surveys conouc:no unoer --is :rogram snail be sumarizac.

inter reted ano reporteo in accercance with Sectt: n 5.4.1 of this EFP.

1 l The final report of this :rogram snall present a orief assessment by the i licensee of the environmental impact ano supplemental cooling water system I operation on the various offsite acoustic environments, and shall describe the l mitigattve measures, if any, that have been, r are to be taken to reduce One imoac: :f plant or supplemental cooling water system noise levels en the

cffsite environments. This report shall also contain a list of noise-relateo

cmolaints or enouiries receiveo ey ""h?.simr9esens Company c:ncerning tre Limerick Generating Station or 1:s s piemental cooling water system suesecuent to issuance of the coerating license along with a desert: tion of j the action taken by "5i h;;i3 h U K:i . c:mpany to, resolve these ::splaints or inouiries. Qg Qg i

This ;rogram snall terminate upon completion of the collection of.the specifito l sovne level data for each phase and submissien cf an acceptable final report.

4.2.4.2 Point Pleasant Dumonouse in ; ;E ruitr.g ;LSP-83-P.; Maren E, '-52) eoutres ::at Sne iscensee conou:: a .

one-time fitid study af ter the transformers are :iaceo in operat1on at Point Pleasant. The noise from operation of the transfermers shall be reouced t: a level so that the transformer core tones will be inaucible (i.e., r.et above the masting level, as definea below) at the site councary.

The licensee shall determine, caseo on ensite reasurements, the celta L(ex!

li.e., the coise level in excess of the maszing '.evel) for eacn tone. The masting levei is cefinec as "':" cB aoove tne amolent ::ectrum level, wnere ":"

is cefinea as follows:

a.:

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ML20135D096

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