Text

Amends 131 & 92 to Licenses NPF-39 & NPF-85,respectively, Revising FOLs & Apps A,B & C for Unit 1 & Apps a & B for Unit 2 of Licenses to Reflect Corporate Change,Remove Obsolete Info & Correct Typographical Errors
	ML20155C560
Person / Time
Site:	Limerick
Issue date:	10/23/1998
From:	Capra R; NRC (Affiliation Not Assigned)
To:	;
Shared Package
ML20155C565	List: ML20155C560; ML20155C571;
References
	NUDOCS 9811020268
	Download: ML20155C560 (55)
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i UNITED STATES

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NUCLEAR REGULATORY COMMISSION

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WASHINGTON, D.C. 20066 4 001 PHILADELPHfA ELECTRIC COMPANY DOCKET NO. 50-352 LIMERICK GENERATING STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.131 i

License No. NPF-39 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Philadelphia Electric Company (the licensee) dated Februay 25,1997, as supplemented by letters dated September 8 and November 18,1997 and January 8 and July 2,1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; l

D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

9811020268 981023 PDR ADOCK 05000352 P

PDR

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2 2.

Accordingly, pages 1 through 5 and 7 and 8 of Facility Operating License No. DPR-39 are I

hereby amended by changing Philadelphia Electric Company to PECO Energy Company, 3.

Further, the license is, amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating i

j License No. NPF-39 is hereby amended to read as follows:

Technical Soecifications i

The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.

131, are hereby incorporated into this license. PECO Energy Company shall operate the facility in accordance with the Technical Specifications and the 1

Environmental Protection Plan.

i i

4.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days.

i f

FOR THE NUCLEAR REGULATORY COMMISSION T O G.C W Robert A. Capra, Director 4

j Project Directorate 12 Division of Reactor Projects - t/II Office of Nuclear Reactor Regulation Attachments: 1. Pages 1 through 5 and 7 and 8 of Facility Operating License NPF 39

2. Changes to the Technical Specifications

3. Cover page and page 4-4 of the Environmental Protection Plan

4. Appendix C, Additional Conditions, Page 1 Date of issuance: October 23, 1998

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'Pages 1 through 5, and 7 and 8, and are attached, for convenience, for the composite license to reflect this change.

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ATTACHMENT TO LICENSE AMENDMENT NO.131 FACILITY OPERATING LICENSE NO. NPF-39 DOCKET NO. 50-312 Replace the following pages of the Facility Operating License (FOL), the Appendix A Technical Specifications, the Appendix B Environmental Protection Plan (EPP), and Appendix C Additional Conditions, with the attached pages. The revised pages are identified by Amendment number and contain verticallines indicating the area of change.

Remove Insert FOL 1

1 2

2 3

3 4

4 5

5 6

7 7

8 8

Appendix A x

x xix xix 4

3/4 3 8 3/4 3 8 3/4357 3/4357 3/4 3-103 3/43103 3/4 5-1 3/4 5-1 3/4 5-2 3/4 5-2 3/4 6-15 3/4 6-15 3/4 6-16 3/4616 3/4 6-42 3/4642 3/4 6-43 3/4 6-43 3/4 7-1 3/4 7-1 3/4 7-3 3/4 7-3 3/4 8-15 3/4815 6-8 6-8 6-9 69 Appendix B EPP Cover EPP Cover EPP 4-4 EPP 4-4 Appendix C Additional Conditions Appendix C Additional Conditions Pagei Page 1

. j o atooq UNITED STATES

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,j NUCLEAR REGULATORY COMMISSION t

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WASHINteTON, D.C. 3084H001 l

PECO ENERGY COMPANY l

DOCKET NO 50-352 i

LIMERICK GENERATING STATION. IINIT 1 I

[A( RITY OPERATING LICE M License No. NPF-39 1

1.

The Nuclear Regulatory Comission (the Comission or the NRC) has found that:

A.

The application for license filed by PECO Energy Company (the licensee complies with the standards and requirements of the Atomic Energy) Act of 1954, as amended (the Act), and the C regulations set forth in 10 CFR Chapter I, and all required notifica-tions to other agencies or bodies have been duly made; B.

Construction of the Limerick Generating Station, Unit 1 (the facility) has been substantially completed in conformity with Construction Permit No. CPPR-106 and the application, as amended, the provisions of the Act and the regulations of the Comission; C.

The facility will operate in conformity with the application, as amended, the provisions of the Act, and the regulations of the Comission (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 co(nducted without endangering the and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I (except as exempted from compliance in Section 2.D.

below);

l E.

The licensee is technically qualified to engage in the activities authorized by this license in accordance with the Comission's regula-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 comon defense and security or the health and safety of the public; l

l Amendment No.131 l

H.

After weighing the environmental, economic, technical, and other benefits of the facility against environmental and other costs and considering available alternatives, the issuance of this Facility Operating License No. NPF-39, subject to the conditions for protection of the environment set forth in the Environmental Protection Plan attached as Appendix B, is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been.

satisfied; and I.

The receipt, possession, and use of source, byproduct and special nuclear material as authorized by this licensa will be in accordance with the Commission's regulations in 10 CFR F4rts 30, 40 and 70.

2.

Based on the foregoing findings, the Partial Initial Decisions issued by the Atomic Safety and Licensing Board dated March 8, 1983, August 29, 1984, May 2, 1985 and July 22, 1985, and the Decision of the Appeal Board dated September 26, 1984, regarding this facility, and approval by the Nuclear Regulatory Commission in its Memorandum and Order dated August 8,1985, the license for Fuel Loading and Low Power Testing, License No. NPF-27, issued on October 26, 1984, is superseded by Facility Operating License NPF-39 hereby l

issued to the PECO Energy Company (the licensee), to read as I

follows:

A.

This license applies to the Limerick Generating Station, Unit 1, a boiling water nuclear reactor and associated equipment, owned by PECO Energy Company. The facility is located on the licensee's site in Montgomery and Chester Counties, Pennsylvania on 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 in the licensee's Final l

Safety Analysis Report, as supplemented and amended, and in the licensee's Environmental Report-Operating License Stage, as supplemented and amended.

B.

Subject to the - conditions and requirements incorporated herein, the Commission hereby licenses PECO Energy Company:

(1)

Pursuant to Section 103 of the Act and 10 CFR Part 50, to pos-sess, use, and operate the facility at the designated location in Montgomery and Chester Counties, Pennsylvania, in accordance with the procedures and limitations set forth in this license; (2)

Pursuant to the Act and 10 CFR Part 70, to receive, possess and to use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended;

[

I i

Amendment No.131

[;

. (3)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility, and to receive and possess, but not separate, such source, byproduct, and special nuclear materials as contained in the fuel assemblies and fuel channels from the Shoreham Nuclear Power Station.

C.

This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I (except as exempted from corgliance in Section 2.D.

below) and is subject to all applicable proeisions 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:

(1)

Maximum Power Level PECO Energy Company is authorized to operate the facility at reactor core power levels not in excess of 3458 megawatts thermal (100% rated power) in accordance with the conditions specified herein and in Attachment I to this license. The items identified in Attachment 1 to this license shall be completed as specified. Attachment 1 is hereby incorporated into this license.

(2)

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.

, are hereby incorporated in the license. PECO Energy Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Amendment No. 6E,406,131

(3)

Fire Protection (Section 9.5. SSER-2.-41*

PECO Energy Company shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Updated Final Safety Analysis Report for the facility, and as approved in the NRC Safety Evaluation Report dated August 1983 through Supplement 9, dated August 1989, and Safety Evaluation dated November 20, 1995, subject to the following provision:

The licensee may make changes to the approved fire protection 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.

4 4

The parenthetical 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.

Amendment No. 464,131

~

The Information contained on FOL pages 5 and 6 were intentionally omitted.

Amendment No.131

070 (16)

Additional Conditions The Additional Conditions contained in Appendix C, as revised through Amendment No.128, are hereby incorporated into this license.

PECO Energy Capany shall operate the facility in accordance with 3

the AddWonalConditions.

D.

The facility requires exemptions from certain requirements of 10 CFR Part 50.

These include (a) exemption from the requirement of Appendix J, the testing of containment air locks at times when the containment integrity is not required (Section 6.2.6.1 of the SER and SSER-3), (b) exemption from the requirements of Appendix J, the I

leak rate testing of the Main Steam Isolation Valves (MSIVs) at the peak calculated containment pressure, Pa, and exemption from the requirements of Appendix J that the measured MSIV leak rates be (c) exemption from the requirement of Appendix J, the loc the Traversing Incore Probe Shear Valves (Section 6.2.6 of the SER and SSER-3).

Amendment No. 44e, +26,131 l

l

s '

These exemptions are authorized by law and vill not endanger life or property or the common defense and security and are otherwise in the public interest. Therefore these exemptions are hereby granted pursuant to 10 CFR 50.12 and 50.47(c). With the granting of these exemptions 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 Commission.

E.

The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).

The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled:

" Limerick Generating Station, Units 1 and 2, Physical Security Plan," with revisions submitted through December 7, 1987;

" Limerick Generating Station, Units 1 and 2, Plant Security Personnel Training and Qualification Plan," with revisions submitted through October 1, 1985; and " Limerick Generating Station, Units 1 and 2, Safeguards Contingency Plan," with revisions submitted through November 15, 1986. Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein.

F.

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 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 accordance with the procedures described in 10 CFR 50.73(b), (c), and (e).

G.

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

Amendment No. 9,131 h

INDil

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PAGE INSTRUMENTATION (Continued)

Loose-Part Detection System...............................

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 Offgas Monitoring Instrumentation.........................

3/4 3-103 Table 3.3.7.12-1 Offgas Monitoring Instrumentation...............

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 information 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 TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION.......................................

3/4 3-112 Table 3.3.9-1 Feedwater/ Main Turbine Trip System Actuation Instrumentation.........

3/4 3-113 Table 3.3.9-2 Feedwater/ Main Turbine Trip System Actuation Instrumen-tation Setpoints.........................

3/4 3-114 Table 4.3.9.1-1 Feedwater/ Main Turbine Trip System Actuation Instrumenta-tion Surveillance Require-ments....................................

3/4 3-115 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM Recirculation Loops.......................................

3/4 4-1 Amendment No. 46,400.131 LIMERICK - UNIT 1 x

m

INDEX

' BASES SECTION PgE JNSTRUMENTATION (Continued)

(Deleted)............................................

B 3/4 3-5 (0eleted).............................................

B 3/4 3-5 Remote Shutdown System Instrumentation and Controls..

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 Loose-Part Detection System...........................

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

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

B 3/4 3-7 3/4.3.8 (0eleted).............................................

B 3/4 3-7 3/4.3.9 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION.......................................

B 3/4 3-7 l

Bases Figure B 3/4.3-1 Reactor Vessel Water Leve1.....................

B 3/4 3-8 3/4.4 REACTOR 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 Operational Leakage...................................

B 3/4 4-3 3/4.4.4 CHEMISTRY..............................................

B 3/4 4-3 Amendment No. 46,63.69.76, LIMERICK - UNIT 1-xix

i TABLE 4.3.1.1-1 (Continued) g REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS r9 CHANNEL OPERATIONAL M

CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK TEST CALIBRATIONM SURVEILLANCE REQUIRED E

~1 9.

Turbine Stop Valve - Closure N.A.

Q R

1 10.

Turbine Control Valve Fast Closure, Trip 011 Pressure - Low N.A.

Q R

1

'I 11.

Reactor Mode Switch Shutdown Position N.A.

R N.A.

5,2,3,4,5 ll 12.

Manual Scram N.A.

W N.A.

1,2,3,4,5 l

(a)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b)

The IRM and SRM channels shall be determined to overlap for at least 1/2 decades during each startup after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for at least 1/2 w

'i decades during each controlled shutdown, if not performed wi..!n the previous 7 days.

(c)

DELETED w

a (d)

This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER 225% of RATED THERMAL POWER. Adjust the APRM

[

channel if the absolute difference is greater than 2% of RATED THERMAL f0FR.

(e)

This calibration shall consist of the adjustment of the APRM flow biases sr.annel to conform to a calibrated flow ij signal.

(f)

The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFPH).

(g)

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

g provide a basis for establishing the specified relationships. Comparisons of the actual data in accordance with g

the criteria listed shall commence upon the conclusion of the startup test program.

s (h)

This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.

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

5 (j) 9 ()

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

':tg hours for required surveillance. During this time, CORE ALTERATIONS shall be suspended, and no control rod shall P'

be moved from its existing position.

t M (k)

Required to be OPERABLE only prior to and during shutdown margin demonstrations

as performed per Specification 2d 3 3.10.3.

R$

i

INSTRUMENTATION 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.6.

The control rod block instrumentation channels shown in Table 3.3.6-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.6-2.

APPLICABILITY: As shown in Table 3.3.6-1.

ACTION:

With a control rod block instrumentation channel trip setpoint less a.

conservative than the value shown in the Allowable Values column of Table 3.3.6-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

b.

With the number of OPERABLE channels less than required by the Minimum i

OPERABLE Channels per Trip Function requirement, take the ACTION required by Table 3.3.6-1.

SURVEILLANCE RE0VIREMENTS 4.3.6 Each of the above required control rod block trip systems and instrumentation channels shall be demonstrated OPERABLE

by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.6-I.

A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for required surveillance without placing the trip system in the tripped condition, provided at least one other operable channel in the same trip system is monitoring that parameter.

i Amendment No. 70,131 LIMERICK - UNIT 1 3/4 3-57

i INSTRUMENTATION OFFGAS GAS MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.7.12 The offgas monitoring instrumentation channels shown in Table 3.3.7.12-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specifications 3.11.2.5 and 3.11.2.6 respectively, are not exceeded.

APPLICABILITY: As shown in Table 3.3.7.12-1 ACTION:

)

With an offgas monitoring instrumentation channel alarm / trip a.

setpoint less conservative than required by the above Specification, declare the channel inoperable, and take the ACTION shown in i

Table 3.3.7.12-1.

b.

With less than the minimum number of offgas monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3.7.12-1.

Restore the inoperable instrumentation to OPERAELE status within the time specified in the ACTION or explain why this inoperability was not corrected in a timely manner in the next Annual Radioactive Effluent Release Report.

The provisicas of Specifications 3.0.3 and 3.0.4 are not applicable.

c.

SURVEILLANCE RE0VIREMENTS 4.3.7.12 Each offgas monitoring instrumentation channel shall be demonstrated 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.

Amendment No. 46,131 LIMERICK - UNIT 1 3/4 3-103

,3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - OPERATING LIMITING CONDITION FOR OPERATION 3.5.1 The emergency core cooling systems shall be OPERABLE with:

The core spray system (CSS) consisting of two subsystems with each a.

subsystem comprised of:

1.

Two OPERABLE CSS pumps, and 2.

An OPERABLE flow patch capable of taking suction from the suppression chamber and transferring the water through the spray i

sparger to the reactor vessel.

b.

The low pressure coolant injection (LPCI) system of the residual heat remeval system consisting of four subsystems with each i

subsystem comprised of:

i 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.

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

c.

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 0')ERABLE ADS valves.

4 APPLICABILITY:

OPERATIONAL CONDITION 1, 2* ** #, and 3* ** ##.

The HPCI system is not required to be OPERABLE when reactor steam dome l

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.

ifTwo LPCI subsystems of the RHR system may be inoperable in that they are aligned in the shutdown cooling mode v; hen reactor vessel pressure is less than the RHR Shutdown cooling permissive setpoint.

LIMERICK - UNIT 1 3/4 S-1 4

Amendment No, sB,66,131 4

I n.

_ _... _,.4 i

j EMERGENCY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION:

I a.

For the core spray' system:

j 1.

With one CSS subsystem inoperable, provided that at least two LPCI subsystems are OPERABLE, restore the inoperable CSS subsystem to i

OPERABLE status within 7 days or be in at least HOT SHUTDOWN within thenext12hoursandinCOLDSHUTDOWNwithinthefollowing24 hours.l 2.

With both CSS 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 .

b.

For the LPCI system:

1.

With one LPCI subsystem inoperable, provided that at least one CSS subsystem is OPERABLE, restore the inoperable LPCI pump 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.

With one RHR cross-tie valve (HV-51-182 A or 8) open, or power not removed from one closed RHR cross-tie valve operator, close the open valve and/or remove power from the closed 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-51-182 A, B) closed, or power not removed from both closed RHR cross-tie valve operators, 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 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 rubsystems inoperable, provided that at least one CSS 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 the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

5.

W1th three LPCI subsystems 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 lea'.t 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 one 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 .*

OWhenever both 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.

LIMERICK - UNIT 1 3/4 5-2 Amendment No. 86,94,131

CONTAINMENT SYSTEMS

SUPPRESSION POOL SPRAY t

LIMITING' CONDITION FOR OPERATION 3.6.2.2 The suppression sool spray mode of the residual heat removal (RHR) system shall be OPERABLE wit) two independent loops, each loop consisting of:

1 a.

One OPERABLE RHR pump, and b.

An OPERABLE flow path capable of recirculating water from the suppression chamber through an RHR heat exchanger and the suppression pool spray sparger(s).

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

i a.

With one suppression pool s) ray loop inoperable, restore the inoperable loo) to OPERABLE status wit 11n 7 days or be in at least HOT SHUTDOWN wit 1in 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 suppression pool 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 4

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 RE0VIREMENTS 4.6.2.2 The suppression pool spray mode of the RHR system shall be demonstrated OPERABLE:

1 At least once per 31 days by verifying that each valve (manual, power-a.

operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

b.

By verifying that each of the required RHR pumps develoss a flow of at least 50') gpm on recirculation flow through the RHR ieat exchanger i

and the suppression pool spray sparger when tested pursuant to Speci-fication 4.0.5.

4

Whenever both RHR 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.

4 4

Arnendnient No. 86,131 LIMERICK - UNIT 1 3/4 6-15

CONTAINMENT SYSTEMS l

SUPPRESSION POOL COOLING LIMITING CONDITION FOR OPERATION 3.6.2.3 The suppression po'o1 cooling mode of the residual heat removal (RHR) system shall be OPERABLE with two independent loops, each loop consisting of:

a.

One OPERABLE RHR pump, and b.

An OPERABLE flow path capable of recirculating water from the suppression chamber through an RHR heat exchenger.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

With one suppression pool cooling loop inoperable, restore the inoperable a.

loo) 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 wit 11n 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 su)pression pool cooling loops inoperable, be in at least HOT SHUTDOWN wit 11n 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 .

SURVEILLANCE REQUIREMENTS 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 otherwise secured in position, is in its correct position.

b.

By verifying that each of the required RHR pumps develops.a flow of at least 10,000 gpm on recirculation flow through the flow path including the RHR heat exchanger and its associated closed bypass valve, the suppression pool and the full flow test line when tested pursuant to Specification 4.0.5.

$Whenever both RHR 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.

Amendment No. 57,58,66,131 LIMERICK - UNIT 1 3/4 6-16

TABLE 3.6.3-1 PRIMARY CONTAINMENT ISOLATION VALVES NOTATION NOTES (Continued) 15.

Check valve used instead of flow orifice.

16.

Penetration is sealed by a flange with double 0-ring seals.

These seals are leakage rate tested by pressurizing between the 0-rings.

Both the TIP Purge Supply (Penetration 35B) and the TIP Drive Tubes (Penetrations 35 C i

thru G) are welded to their respective flanges.

Leakage through these seals l

is included in the Type C leakage rate total for this penetration. The ball l

valves (XV-141A thru E) are Type C tested.

It is not practicable to leak test the shear valves (XV-140A thru E) because squib firing is required for closure. Shear valves (XV-140A thru E) are normally open.

17.

Instrument line isolation provisions consist of an excess flow check valve.

l Because the instrument line is connected to a closed cooling water system inside containment, no flow orifice is provided. The excess flow check valves are subject to operability testing, but no Type C test is performed nor required.

The line does not isolate during a LOCA and can leak only if the line or instrument should rupture.

Leaktightness of the line is l

verified during the 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 which are flanged to facilitate removal will be equipped with double 0-ring seal assemblies on the flange closest to primary containment.

These seals will be leak rate tested by pressurizing between the 0-rings, and the results added into the Type C l

l total for this penetration.

t l

20.

See Specification 3.3.2, Table 3.3.2-1, for a description of the PCRVICS i

isolation signal (s) that initiate closure of each automatic isolation valve.

l In addition, the following non-PCRVICS isolation signals also initiate l

closure of selected valves:

LFHP With HPCI pumps running, opens on low flow in associated pipe, closes 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 pump running, opens on low flow in associated pipe, closes when flow is above setpoint LFCC Steam supply valve fully closed or RCIC turbine stop valve fully closed i

All power operated isolation valves may be opened or closed remote manually.

f LIMERICK - UNIT 1 3/4 6-42 Amendment No. 29,33,131

TABLE 3.6.3-1 PRIMARY CONTAINMEN" ISOLATION VALVES NOTA" ION "MEji (Continued) isolation signal causes TIP to retract; ball valve closes when Automatic probe is fully retracted,.

22.

Isolation barrier remaink 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.

23.

Valve does not receive an isolation signal.

Valves will be open during Type A test. Type C test not required.

24.

Both isolation signals required for valve closure.

25.

Deleted 26.

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

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

27.

The reactor vessel head seal leak detection line (penetration 29A) excess flow check valve is not subject to OPERABILITY testing.

This valve will not be exposed to primary system pressure except under the unlikely con-ditions of a seal failure where it could be partially pressurized to reactor pressure.

Any leakage path is restricted at the source; therefore, this valve need not be OPERABILITY tested.

28.

(DELETED)

I 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 and reactor pressure is greater than 600 psig.

The valves are expected to be opened only in the following instances:

Flushing of the condensate and feedwater systems during plant startup.

a.

~

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 Automatic Isolation Valve and a Refueling Area Secondary Containment Automatic Isolation Valve.

l 34.

Auto isolation signals have been removed from HV-087-124 A/B and 125 A/B.

Valves to be closed with associated circuit breakers locked open during OPCONs 1, 2, and 3.

LIMERICK - UNIT 1 3/4 6-43 Amendment No. EB,49B,131

3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS RESIDUAL HEAT REMOVAL SERVICE WATER SYSTEM - COMMON SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.1 At least the following independent residual heat removal service water (RHRSW) system subsystemh, with each subsystem comprised of:

a.

Two OPERABLE RHRSW pumps, and b.

An OPERABLE flow path capable of taking suction from the RHR service water pumps wet pits which are supplied from the spray pond or the cooling tower basin and transferring the water through one Unit 1 RHR heat exchanger, shall be OPERABLE:

3 a.

In OPERABLE CONDITIONS 1, 2, and 3, two subsystems.

b.

In OPERABLE CONDITIONS 4 and 5, the subsystem (s) associated with systems and components required OPERABLE by Specification 3.4.9.2, 3.9.11.1, and 3.9.11.2.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, 4, and 5.

ACTION:

a.

In OPERATIONAL CONDITION 1, 2, or 3:

1.

With one RHRSW pump inoperable, restore the inoperable pump to OPERAB.I 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 SHUTOOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i 2.

With one RHRSW pump in each subsystem inoperable, restore at least one of the inoperable RHRSW pumps to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.

With one RHRSW subsystem otherwise inoperable, restore the inoperable subsystem 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 4

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 .

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 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 .

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

LIMERICK - UNIT 1 3/4 7-1 Amendment No. 66,86,131 a

.v 9

--7----

,3y y

w

PLANT SYSTEMS EMERGENCY SERVICE WATER SYSTEM - COMMON SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least the following independent emergency service water system loops, I

with each loop comprised of:

Two OPERABLE emergency service water pumps, and a.

b.

An OPERABLE flow path capable of taking suction from the emergency service water pumps wet pits which are supplied from the spray pond or l

the cooling tower basin and transferring the water to the associated Unit 1 and common safety-related equipment, shall be OPERABLE:

a.

In OPERATIONAL CONDITIONS 1, 2, and 3, two loops.

b.

In OPERATIONAL CONDITIONS 4, 5, and *, one loop.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, 4, 5, and *.

ACTION:

l a.

In OPERATION CONDITION 1, 2, or 3:

1.

With one emergency service water pump inoperable, restore the l

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 SHUTOOWN 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 inoperable, restore at least one inoperable pump 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 .

3.

With one emergency service water system loop otherwise inoperable, declare all equipment aligned to the inoperable i

loop inoperable **, restore the inoperable loop to OPERABLE l

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 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 .

l

When handling irradiated fuel in the secondary containment.

- The diesel generators may be aligned to the OPERABLE emergency service water j

system loop provided confirmatory flow testing has been performed. Those diesel generators no aligned to the OPERABLE emergency service water system loop shall be declared inoperable and the actions of 3.8.1.1 taken.

2 7

LIMERICK - UNIT 1 3/4 7-3 Amendment No. N,40,86,131

~'~

~ '. _ _ _.. _ i_. _

~~

^~

^^

~

^^

ELECTRICAL POWER SYSTEMS 3/4.8.3 ONSITE POWER DISTRIBUTION SYSTEMS DISTRIBUTION - OPERATING LIMITING CONDITION FOR OPERATION 3.8.3.1 The following power distribution system divisions shall be energized:

a.

A.C. power distribution:

1.

Unit 1 Division 1, Consisting of:

a) 4160-VAC Bus:

011 (10A115) b)

480-VAC Load Center:

D114 (10B201) c)

480-VAC Motor Control Centers:

Dll4-R-Cl (108219)

Dll4-R-C (10B213)

Dll4-R-G (108211)

Dll4-R-G1 (10B215)

D114-D-G (108515) d)

120-VAC Distribution Panels:

10Y101 10Y206 2.

Unit 1 Division 2, Consisting of:

a) 4160-VAC Bus:

D12 (10All6) b)

480-VAC Load Center:

D124 (10B202) c)

480-VAC Motor Control Centers:

D124-R-C1 (108220)

D124-R-C (108214)

D124-R-G (108212)

D1c,-7-G1 (108216)

D124-D-G (10B516) d)

120-VAC Distribution Panels:

10Y102 10Y207 3.

Unit 1 Division 3, Consisting of:

a) 4160-VAC Bus:

D13 (10All7) b)

480-VAC Load Center:

D134 (10B203) c)

480-VAC Motor Control Centers:

Ol34-R-H1 (10B221)

D134-R-H (108217)

D134-R-E (108223)

D134-C-8 (00B131) 0134-D-G (108517) d)

120-VAC Distribution Panels:

10Y103 10Y163 4.

Unit 1 Division 4, Consisting of:

a) 4160-VAC Bus:

D14 (10All8) b)

480-VAC Load Center:

D144 (108204)

Amendment No. N,131 LIMERICK - UNIT 1 3/4 8-15

_ m; _ _ _. _ _ _ _ _ _ _

ADMINISTRATIVE CONTROLS l

RESPONSIBILITIES 1

6.5.1.6 The PORC shall be responsible for:

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

programs or procedures required by specification 6.8 and requiring a 10 CFR j.

50.59 safety evaluation, and (3) proposed changes to programs or procedures required by Specification 6.8 and requiring a 10 CFR 50.59 safety l

evaluation; j

b.

Review of all proposed tests and experiments that affect nuclear safety; j

Review cf all proposed changes to Appendix A Technical Specifications; c.

d.

Review of cll proposed changes or modifications to unit systems or equipment that affect neclear safety; e.

DELETED.

f.

Investigation of all violations of the Technical Specifications, including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence, to the Vice President, Limerick i

Generating Station, Plant Manager, and to the Nuclear Review Board; g.

Review of all REPORTABLE EVENTS; h.

Review of unit operations to detect potential hazards 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;

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

Review of the Emergency Plan and implementing procedures and submittal of the recommended changes to the Nuclear Review Board.

l.

Review of every unplanned onsite release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation, recommendations 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 radwaste treatment systems.

Review of the Fire Protection Program and implementing procedures and the n.

submittal of recommended changes to the Nuclear Review Board.

6.5.1.7 The PORC shall:

a.

Recommend in writing to the Plant Manager approval or disapproval of items considered under Specification 6.5.1.6a. through d. prior to their implementation.

)

b.

Render determinations in writing with regard to whether or not each item considered under Specification 6.5.1.6b. through f. constitutes an unreviewed safety question.

LIMERICK - UNIT 1 6-8 Amendment No. %, G, %4,131

_ ADMINISTRATIVE CONTROLS RESPONSIBILITIES (Continued) 1 Provide written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the Vice President, c.

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 respo pursuant to Specifica,nsibility for resolution of such disagreements tion 6.1.1.

RECORDS l

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 i

i provided to the Vice President, Limerick Generating Station, Plant Manager, and the Nuclear Review Board.

l 6.5.2 NUCLEAR REVIEW BOARD (NRB) i FUNCTION 6.5.2.1 The NRB shall function to provide independent review and audit of designated activities in the areas of:

l a.

Nuclear power plant operations, b.

Nuclear engineering, c.

Chemistry and radiochemistry, d.

Metallurgy, e.

Instrumentation.and control, l

f.

Radiological safety, 1

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.

1 COMPOSITIO_N l

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 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.

l 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.

l Amendment No. 40,95,96,131 i

LIMERICK - UNIT 1 6-9

i

\\

l t

APPENDIX B 1

i TO FACILITY OPERATING LICENSE NO. NPF-39 LIMERICK GENERATING STATION UNITS 1 AND 2 l

PECO ENERGY COMPANY DOCKET NOS. 50-352, 50-353 l

l l

l ENVIRONMENTAL PROTECTION PLAN j

(NONRADIOLOGICAL) i Amendment No.131 l

1

sensitive land uses in the site vicinity (e.g., residences, schools, churches, cemeteries, ho,spitals, parks); and (3) previously conducted noise surveys in the site vicinity.

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

The results of the surveys conducted under this program shall be summarized, interpreted and reported in accordance with Section 5.4.1 of this EPP.

The final report of this program shall present a brief assessment by the licensee of the environmental impact of plant and supplemental cooling water system operation on the various offsite acoustic environments, and shall describe the mitigative measures, if any, that have been, or are to be taken to reduce the impact of plant or supplemental cooling water system noise levels on the offsite environments. This report shall also contain a list of noise-related complaints or inquiries received by PECO Energy Company concerning the Limerick Genersting Station or its supple-mental cooling water system subsequent to issuance of the operating license along with a description of the action taken by PECO Energy Company to resolve these complaints or inquiries.

This program shall terminate upon completion of the collection of the specified sound level data for each phase and submission of an acceptable final report.

4-4 Amendment No.131

l APPENDIX C ADDITIONAL CONDITIONS

' OPERATING LICENSE NO. NPF-39 PECO Energy Company shall comply with the following conditions on the schedules noted below:

Amendment Additional Conditions implementation Number Date 128 This amendment authorizes the licensees to incorporate 30 days from in the Updated Final Safety Analysis Report (UFSAR)

May 14,1998 j

i certain changes to the description of the facility.

implementation of this amendment is the incorporation of these changes as described in the licensee's application i

dated October 6,1997, as supplemented by letter dated February 2,1998, and evaluated in the safety evaluation dated May 14,1998.

128 The suppression pool floor and the low pressure ECCS 30 days from (RHR and Core Spray) suction strainers shall be visually May 14,1998 inspected for sludge accumulation and foreign material.

The visualinspection allows use of a remote camera in lieu of divers. The interval of these inspections shall be i

every other refueling outage. The inspection interval may be increased based on findings of two consecutive inspections. Should the licensee choose to increase the inspection interval, data supporting this increase shall be submitted to the NRC for review.

)

i l

i

{

< Amendment No. +26,131 l

l p2 aarg

[

t UNITED STATES g

j NUCLEAR REGULATORY COMMISSION WASHINoToN, D.C. 20666 4 001 o%...../

I i

PHILADELPHIA ELECTRIC COMPANY DOCKET NO. 50-353 LIMERICK GENERATING STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING UCENSE Amendment No. 92 Ucense No. NPF-85 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Philadelphia Electric Company (the licensee) dated February 25,1997, as supplemented by letters dated September 8 and November 18,1997 and January 8 and July 2,1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; I

B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied, t

4

2-2.

Accordingly, pages 1,2,3 and 5 of Facility Operating License No. NPF-85 are hereby amended by changing Philadelphia Electric Company to PECO Energy Company.

i 3.

Further, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-85 is hereby amended to read as follows:

Technical Soecifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.

92, are hereby incorporated in the license. PECO Energy Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

4.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION 3M o..Cy Robert A. Capra, Director Project Directorate I-?.

Division of Reactor Projects - 1/Il Office of Nuclear Reactor Regulation Attachments: 1. Pages 1,2,3, and 5 of Facility Operating License No. NPF-85

2. Changes to the Technical Specifications

3. Cover page and page 4-3 of the Environmental Protection Plan Date of Issuance: October 23, 1998 l

4

Pages 1,2,3, and 5 of the license are attached, for convenience, for the composite license to reflect this change.

i

ATTACHMENT TO LICENSE AMENDMENT NO, 92 FACILITY OPERATING LICENSE NO. NPF-85 DOCKET NO. 50-353 i

Replace the following pages of the Facility Operating Ucense (FOL), the Appendix A Technical Specifications and the Appendix B Environmental Protection Plan (EPP) with the attached pages.

The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

Remove jng,r.1 FOL 1

1 2

2 3

3 5

5 Appendh A x

x l

xix xix 3/4 3-8 3/4 3-8 3/4 3-103 3/4 3-103 3/4 5 1 3/4 5-1 W4 52 W4 52 3/4 6-14 3/4 6-14 3/4 6-15 3/4 6-15 3/4 6-16 3/4 6-16 3/4 6-42 3/4 6-42 3/4 6-43 3/4 6-43 3/4 6-45 3/4 6-45 3/4 7-1 3/4 7-1 3/4 7-3 3/4 7-3 3/4 6-4 3/4 8-4 6-8 6-8 6-9 6-9 Appendix B EPP Cover EPP Cover EPP 4-3 EPP 4-3

I fwo u

-t UNITED STATES E

NUCLEAR REGUL.ATORY COMMISSION

[

WASHINGTON, D.C. 2006s4001 PECO ENERGY COMPANY DOCKET NO. 50-353 LIMERICK GENERATING STATION. UNIT 2 FACILITY OPERATING LICENSE License No. NPF-85 1.

The Nuclear Regulatory Comission (the Comission or the NRC) has found that:

A.

The application for license filed by PECO Energy Company (the licensee complies with the standards and requirements of the Atomic Energy)Act of 1954, as amended (the Act), and the Comissio regulations set forth in 10 CFR Chapter I, and all required notifica-tions to other agencies or bodies have been duly made; B.

Construction of the Limerick Generating Station, Unit 2 (the facility) has been substantially completed in conformity with Construction Permit No. CPPR-107 and the application, as amended, the provisions of the Act and the regulations of the Comission; C.

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

D.

There is reasonable assurance:

1) that the activities authorized by this operating license can be co(nducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I (except as exempted from complianc.e in Section 2.D. below);

E.

The licensee is technically qualified to engage in the activities authorized by this license in accordance with the Commission's regulations 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,"

4 of the Commission's regulations; G.

The issuance of this license will not be inimical to the comon defense and security or to the health and safety of the public; i

Amendment No. 92 l

l l

H.

After weighing the environmental, economic, technical, and other i

benefits of the facility against environmental and other costs and i

considering available alternatives, the issuance of this Facility Operating License No. NPF-85, subject to the conditions for

{

protection of the environment set forth in the Environmental i

Protection Plan attached as Appendix B, is in accordance with.10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied; and I.

The receipt, possession, and use of source, byproduct and special nuclear material as authorized by this license will be in accordance i

with the Commission'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 Board, LBP-85-25, dated July 22, 1985, the Commission's Order 1

dated July 7, 1989, and the Commission's Memorandum and Order dated August 25, 1989, regarding this facility, Facility Operating License

')

NPF-85 is hereby issued to the PECO Energy Company (the l

licensee), to read as follows:

A.

This license applies to the Limerick Generating Station, Unit 2, a boiling water nuclear reactor and associated equipment, owned by PEC0 Energy Company. The facility is located on the licensee's site in Montgomery and Chester Counties, Pennsylvania on 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 Analysis Report, as supplemented and amended, and in the licensee's Environmental Report-Operating License Stage, as supplemented and amended.

l l

B.

Subject to the cenditions and requirements incorporated herein, the Commission hereby licenses PECO Energy Company:

(1)

Pursuant to Section 103 of the Act and 10 CFR Part 50, to possess, use, and operate the facility at the designated i

location in Montgomery and Chester Counties, Pennsylvania, in accordance with the procedures and limitations set forth in this license; (2)

Pursuant to the Act and 10 CFR Part 70, to receive, possess and to use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required i

for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source and special 1

nuclear material as sealed neutron sources for reactor startup, i

sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in I

amounts as required; 1

Amendment No. 92

\\

~.

(4)

Pursuant to the Act and 10 CFR Parts 30, 40, 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility, and to receive and possess, but not separate, such source, byproduct, and special nuclear materials as contained in the fuel assemblies and fuel channels from the Shoreham Nuclear Power Station.

C.

This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I (except as exempted from compliance in Section 2.0. below) and is subject to all applicable provisions of the Act and to the rules, regulations, and crders of the Commission now or hereafter in effect; and is subject to the l

additional conditions specified or incorporated below:

(1)

Eaximum Power Level PECO Energy Company is authorized to operate the facility at reactor core power levels of 3458 megawatts thermal (100 percent rated power) in accordance with the conditions l

specified herein.

(2)

Technical Soecifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.

, are hereby incorporated into this license.

PECO Energy Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Fire Protection (Section 9.5. SSER-2.-4)*

PECO Energy Cortpany shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the U;, dated Final Safety Analysis Report for the facility, and as approved in the NRC Safety Evaluation Report dated August 198:: tirough Supplement 9, dated August 1989, and Safety Evaluatio: dated November 20, 1995, subject to the following provisien:

The licensee may make changes to the approved fire protection 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.

l

The parenthetical notation following the title of license conditions denotes the I

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

Amendment No. 4, N, H,60,92

the local leak rata testing of the Traversing Incore Probe Shear Valves (Section 6.2.6.1 of the SER and SSER-3), and (d) an exemption l

from the schedule requirements of 10 CFR 50.33(k)(1) related to availability of funds for decommissioning the facility (Section 22.1, SSER 8).

The special circumstances regarding exemptions (a), (b) and (c) are identified in Sections 6.2.6.1 of the SER and SSER 3.

An exemption from tha criticality monitoring requirements of 10 CFR 70.24 was previously granted with NRC materials license No. SNM-1977 issued November 22, 1988, li.9 licensee is hereby exenipted from the requirements of 10 CFR M.24 insofar as this requirement applies to the handling and storage of ft.91 assemblies held under this license.

These exemption: are authoriz9d by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security. The exemptions in items a, b, c, and d above are granted pursuant to 10 CFR 50.12.

With these exemptions, 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 Commission.

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 made 4

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 accordance with the proce'Jures described in 10 CFR 50.73(b), (c),

and (e).

F.

The licensee shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.

Amendment No. 92

INDEX

' LIMITING CONDITIONS FOR OPERATION AND SVRVElllANCE RE0VIREMENTS SECTION PEif

\\

INSTRUMENTATION (Continued)

I loose-Part Detection System.............................

3/4 M 7 The information from pages 3/4 3-98 l

through 3/4 3-101 has been intentionally omitted. Refer to note on page 3/4 3-98................

3/4 3-98 Of fgas Monitoring Instrumentation.......................

3/4 3-103 l

Table 3.3.7.12-1 Offgas Monitoring Instrumentation.......................

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 information 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 TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION....................................

3/4 3-112 Table 3.3.9-1 Feedwater/ Main Turbine Trip System Actuation Instrumentation.......

3/4 3-113 Table 3.3.9-2 Feedwater/ Main Turbine Trip System Actuation Instrumen-tation Setpoints......................

3/4 3-114 Table 4.3.9.1-1 Feedwater/ Main Turbine Trip System Actuation Instrumenta-tion Surveillance Require-ments.................................

3/4 3-115 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATI0t! SYSTEM Recirculation loops........................................

3/4 4-1 Amendment No. 44,64,92 l

LIMERICK - UNIT 2 x

.. z _ :. _:

.=

.~-

1HQfl

BASES SECTION P.ME INSTRUMENTATION (Continued)

(Deleted).....................................................

B 3/4 3-5 l

(Deleted).....................................................

B 3/4 3-5 Remote Shutdown System Instrumentation and Controls...........

B 3/4 3-5 Accident Monitoring Instrumentation...........................

B 3/4 3-5 l

l Source Range Monitors.........................................

B 3/4 3-5 1

(Deleted).....................................................

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

B 3/4 3-6 l

(Deleted).....................................................

g 3,4 3.s Loose-Part Detection System..................................

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

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

B 3/4 3-7 3/4.3.8 (Deleted).....................................................

B 3/4 3-7 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 REACTOR 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 Operational Leakage...........................................

B 3/4 4-3 3/4.4.4 CHEMISTRY.....................................................

B 3/4 4-3a 1

LIMERICK - UNIT 2 xix Amendment No. 44,42, +7, Os,.

66,68,79,92

h JABLE 4.3.1.1-1 (Continued) 9 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVE M

NCE REOUIREMENTS

^

E CHANNEL q

FUNCTIONAL UNIT CHANNEL FUNCTIONAL Turbine Stop Valve - Closure

_ TEST CHANNEL OPERATIONAL

\\

CHECK 9.

CALIBRATION (a)

CONDITIONS FOR WHICH N.A.

SURVEILLANCE REQUIRE 0 10.

Q Turbine Control Valve Fast R

closure 1

Prc<.. ure, Trip Oil Low N.A.

11.

Reactor Mcde Switch Q

R Shutdown Position 1

N.A.

12.

Manual Scram R

~~

N.A.

Y N.A.

1,2,3,4,5 (a)

W Neutron E e~ctors may be excluded from CHANNEL CAL N.A.

(b)

The IRM and SRM channels shall be determined to ov IBRATION.

1,2,3,4,5 entering OPERATIONAL CONDITION 2 and the IRM and APRM erlap for at least 1/2 decades during e decades during each controlled shutdown (c)

DELETED

, if not performed within the previous 7 days.

er (d)

This calibration shall consist of the adjustment channel if the absolute difference is greater than 2%a heat balance d l

of the APRM channel to conform to the power values (e)

This calibration shall consist of the adjustment of RATED THERMAL POWER. POWER 225%

{

ae by signal.

Adjust the APRM l

f The Verify measured core flow (total core flow) to bLPRMs shall be ca

\\

((g))

(

rated flow loop flow (APRM % flow).

ective full power hours (EFPH).

(

a e greater During the startup test program,than or equal to established core flow at the provide a basis for establishing the specified relati

{

3 the criteria listed shall commence upon the con ldata shall be recorded for the parameters listed t

(

E (h) xs ng onships.

This function is not required to be OPERABLE when thc usion of the startup test program.ctu Comparisons of the a i

z o

3.10.1.

IP.*

((j)) With any control rod withdrawn.

i Not applicable to control rods removed per Specif j

If the RPS shorting links are re l- $

n hours for required surveillance. quired to be removed per Specification 3 9 2

%~I (k) cation 3.9.10.1 or 3.9.10.2.

be moved from its existing position.

Required to be OPERABLE only prior to and duringDuring this time, COR

-p 3.10.3.

n e, and no control rod shall shutdown margin demonstrations as performed per Spec on 3

TABLE 4.3.1.1-1 (Continued) h REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILtANCE REQUIREMENTS 9

7; x

CHANNEL OPERATIONAL g

CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH y

FUNCTIONAL UNIT CHECK TEST CALIBRATION (a)

SURVEILLANCE REQUIRED m

9.

Turbine Stop Valve - Closure N.A.

Q R

1-10.

Turbine Control Valve Fast Closure, Trip Oil Pressure - Low N.A.

Q R

1 11.

Reactor Mode Switch Shutdown Position N.A.

R N.A.

1,2,3,4,5 12.

Manual Scram N.A.

W N.A.

1,2,3,4,5 Y

(a)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b)

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

(c)

DELETED (d)

This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER 225% of RATED THERMAL POWER. Adjust the APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER.

(e)

This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a calibrated flow signal.

(f)

The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFPH).

(g)

Verify measured core flow (total core flow) to be greater than or equal to established core flow at the existing 3

loop flow (APRM % flow). During the startup test program, data shall be recorded for the parameters listed to a

provide a basis for establishing the specified relationships.

Comparisons of the actual data in accordance with 3

the criteria listed shall commence upon the conclusion of the startup test program.

5 (h)

This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.

z!"

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

g *

(i)

(j)

If the RPS shorting links are required to be removed per Specification 3.9.2, they may be reinstalled for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance.

During this time, CORE ALTERATIONS shall be suspended, and no control rod shall h (k) be moved from its existing position.

u Required to be OPERABLE only prior to and during shutdown margin demonstrations as performed per Specification

,3 3.10.3.

INSTRUMENTATION

^

OFFGAS GAS MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 1

3.3.7.12 The offgas monitoring instrumentation channels shown in Table 3.3.7.12-1 shall be OPERABLE with their alk-m/ trip setpoints set to ensure that the limits of Specifications 3.11.2.5 and 3.11.2.6 respectively, are not exceeded.

APPLICABILITY:

As shown in Table 3.3.7.12-1 ACTION:

With an offgas monitoring instrumentation channel alarm / trip setpoint a.

less conservative than required by the above Specification, declare the channel inoperable, and take the ACTION shown in Table 3.3.7.12-1.

b.

With less than the minimum number of offgas monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3.7.12-1.

Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION or explain why this inoperability was not l

corrected in a timely manner in the next Annual Radioactive Effluent Release Report.

c.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE0UIREMENTS 4.3.7.12 Each offgas monitoring instrumentation channel shall be demonstrated CPERABLE 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.

)

, 92 LIMERICK - UNIT 2 3/4 3-103

~

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - OPERATING LIMITING CONDITION FOR OPERATION 3.5.1 The emergency core cooli g systems shall be OPERABLE with:

The core spray system (CSS) consisting of two subsystems with each a.

subsystem comprised of:

1.

Two OPERABLE CSS pumps, and 2.

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

b.

The low pressure coolant injection (LPCI) system of the residual heat removal system consisting of four subsystems with each subsystem comprised of:

1.

Ont 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.

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

c.

1.

One OPERABLE HPCI pump, and 2.

An OPERABLE flow path capab'.e of taking suction from the suppression chamber and transferring the water to tne reactor vessel.

d.

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

APPLICABILI1Y:

OPERATIGTAL CONDITION 1, 2* ** #, and 3* ** ##.

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 subsystems 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 RHR Shutdown cooling permissive setpoint.

Amendment No. 70,92 LIMERICK - UNIT 2 3/4 5-1

. EMERGENCY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION:

a.

For the core spray system:

1.

With one CSS subsystem inoperable, provided that at least two LPCI subsystems are OPERABLE, restore the inoperable CSS subsystem to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

With both CSS subse tems inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 subsystem inoperable, provided that at least one CSS subsystem is OPERABLE, restore the inoperable LPCI pump 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.

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 and/or remove power from the closed valves operator within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or be in at least H0T SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.

With no RHR cross-tie valves (HV-51-282 A, B) closed, or power not removed from both closed RHR cross-tie valve operators, 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 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 subsystems inoperable, provided that at least one CSS 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 withir, the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

5.

With three LPCI subsystems 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 he:trs 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 koperable, 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 both shutdown cooling subrystems 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.

j

(

Amendment No. 66,70,92 LIMERICK - UNIT 2 3/4 5-2 i

CONTAINMENT SYSTEMS

SURVEILLANCE REQUIREMENTS (Continued)

c. By verifying at least 8 rnression pool watet temperature indicators in at least 8 locations,,0PadLE by performance of a:

1.

CHANNEL CHECK at leaf,t 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 CALIBRATION at least once per 24 months, with the temperature alarm setpoint for:

1.

High water temperature:

a) First setpoint s 95'F i

i b) Second setpoint s 105'F c) Third setpoint s 110*F d) Fourth setpoint s l'0*F 2

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 CALIBRATION at least once per 24* months, with the water level alarm setpoint for high water level s 24'l-1/2"

e. Drywell-to-suppression chamber bypass leak tests shall be conducted to coincide with the Type A test at an initial differential pressure of 4 psi and verifying that the A//k calculated 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.

f. By conducting a leakage test on the drywell-to-suppression chamber 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 area for an individual set of vacuum breakers is less than or equal to 12% of t

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.

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

LIMERICK - UNIT 2 3/4 6-14 Amendment No. H, se,34, M, H, 92

CONTAINMENT SMIGS

' SUPPRESSION POOL SPRAY LIMITING CONDITION FOR OPERATION 3.6.2.2 The suppression pool spray mode of the residual heat removal (RHR) system shall be OPERABLE with two independent loops, each loop consisting of:

l One OPERABLE RHR pump, and b.

An OPERABLE flow path capable of recirculating water from the suppression chamber through an RHR heat exchanger and the suppression pool spray sparger(s).

APPLICABILITY:

OPERATIONAL CONDITIONS 1. 2,'and 3.

ACTION:

With one suppression pool spray loop inoperable, restore the inoperable a.

loop to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

I b.

With both suppression pool 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 l

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

SURVEILLANCE REQUIREMENTS 4.6.2.2 The suppression pool spray mode of the RHR system shall be demonstrated OPERABLE:

a.

At least once per 31 days by verifying that each valve (manual, power-1 operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

b.

By verifying that each of the required RHR pumps develops a flow of L

at least 500 gpm on recirculation flow through the RHR heat exchanger l

and the suppression pool spray sparger when tested pursuant to Speci-j fication 4.0.5.

Whenever both RHR 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.

Amendment No. es,90,92 i

LIMERICK - UNIT 2 3/4 6-15

.. _ _ _. _ _. _ __..n._ _ _ _

CONTAINMENT SYSTEMS

' SUPPRESSION POOL COOLING LIMITING CONDITION FOR OPERATION 3.6.2.3 The suppression pool tooling mode of the residual heat removal (RHR) system shall be OPERABLE with two independent loops, each loop consisting of:

a.

One OPERABLE RHR pump, and b.

An OPERABLE flow path capable of recirculating water from the suppression chamber through an RHR heat exchanger.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

With one suppression pool cooling loop inoperable, restore the inoperable a.

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 i

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 .

l b.

With both suppression pool cooling loops 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 .

SURVEILLANCE RE0VIREMENTS 4.6.2.3 The suppression pool cooling mode of the RHR system shall be demonstrated OPERABLE:

At least once per 31 days by verifying that each valve (manual, power-a.

operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position, b.

By verifying that each of the required RHR pumps develops a flow of at least 10,000 gpm on recirculation flow through the flow path including the RHR heat exchanger and its associated closed bypass valve, the suppression pool and the full flow test line when tested pu.suant to Specification 4.0.5.

1 l

Whenever both RHR 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, i

LIMERICK - UNIT 2 3/4 6-16 Amendment No, Se, fe, 92

-l

TABLE 3.6.3-1 PRIMARY CONTAINMENT ISOLATION VALVES NOTATION NOTES (Continued) 15.

Check valve used instead of flow orifice.

16.

Penetration is sealed by a flange with double 0-ring seals. These seals are leakage rate tested by pressurizing between the 0-rings.

Both the TIP Purge Supply (Penetration 35B) and the TIP Drive Tubes (Penetrations 35 C thru G) are welded to their respective flanges.

Leakage through these seals is included in the Type C leakage rate total for this penetration.

i The ball valves (XV-241A thru E) are Type C tested.

It is not practicable to leak test the shear valves (XV-240A thru E) because squib firing is required for closure.

Shear valves (XV-240A thru E) are normally open.

17.

Instrument line isolation provisions consist of an excess flow check valve.

Because the instrument line is connected to a closed cooling water system inside containment, no flow orifice is provided.

The excess flow check valves are subject to operability testing, but no Type C test is performed nor required.

The line does not isolate during a LOCA and can leak only if the line or instrument should rupture.

Leaktightness of the line is verified during the 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 flanged to facilitate removal and are equipped with double 0-ring seal assemblies on the flange closest to primary containment. These seals will be leak rate tested by pressurizing between the 0-rings, and the results added into the Type C total for this penetration.

20.

See Specification 3.3.2, Table 3.3.2-1, for 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:

LFHP With HPCI pumps running, opens on low flow in associated pipe, closes 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 pump running, opens on low flow in associated pipe, closes when flow is above setpoint LFCC Steam supply valve fully closed or RCIC turbine stop valve fully closed All power operated isolation valves may be opened or closed remote manually.

LIMERICK - UNIT 2 3/4 6-42 Amendment No. 92

. _ _. _.._ _. ~.. _. _ _ _. _. _. _ _ _ _

l TABLE 3.6.3-1 PRIMARY CONTAINMENT ISOLATION VALVES NOTATION

. HQIES (Continued) 21.

Automatic isolation signal causes TIP to retract; ball valve closes when probe is fully retracted.

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.

23.

Valve does not receive an isolation signal. Valves will be open during Type A test. Type C test'not requirad.

24.

Both isolation signals required for valve closure.

25.

Deleted 26.

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

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

27.

The reactor vessel head seal leak detection line (penetration 29A) excess flow check valve is not subject to OPERABILITY testing. This valve will not be exposed to primary system pressure except under the unlikely con-ditions of a seal failure where it could be partially )ressurized to reactor pressure. Any leakage path is restricted at tie source; therefore, this valve need not be OPERABILITY tested.

28.

(CELETED) 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.

t 1

31.

Valve 41-2016 is an outboard isolation barrier for penetrations X-9A, B and X-44.

Leakage through valve 41-2016 is included in the total for penetration X-44 only.

32.

Feedwater long-path recirculation valves are sealed closed whenever the reactor is critical and reactor pressure is greater than 600 psig. The valves are expected to be opened only in the following instances:

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

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.

l 33.

Valve also constitutes a Unit 1 Reat. tor Enclosure Secondary Containment Automatic Isolation Valve and a Refueling Area Secondary Containment Automatic i

Isolation Valve.

34..

Isolation signal causes recombiner to trip; valve closes when recombiner is not operating.

35.

Auto isolation signals have been removed from HV-087-2?4 A/B and 225 A/B.

Valves to be closed with associated circuit breakers lockec open during

'OPCONs 1, 2, and 3.

LIMERICK - UNIT 2 3/4 6-43 Amendment No. 67,92 i

_ -.

4 CONTAINMENT SYSTEMS i

SURVEILLANCE RE0VIREMENTS 4.6.4.1 Each suppression chamber - drywell vacuum breaker shall be:

Verified closed at least once per 7 days.

a.

b.

Demonst. rated OPERABLE:

1.

At least once per 31 days and within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after any discharge of steam to the suppression chamber from the safety / relief valves, by cycling each vacuum breaker through at least one complete cycle of full travel.

2.

At least once per 31 days by verifying both position indicators OPERABLE by observing expected valve movement during the cycling 1

test.

3.

At least once per 24 months by:

a) Verifying each valve's opening setpoint, from the closed position, to be 0.5 psid i 5%, and b) Verifying both position indicators OPERABLE by performance of a CHANNEL CAllBRATION.

c) Verifying that each outboard valve's position indicator is capable of detecting disk displacement 20.050", and each inboard valve's position indicator is capable of detecting disk displacement 20.120".

92 LINERICK - UNIT 2 3/4 6-45

~

3/4.7 PLANT SYSTEMS 1/4.7.1 SERVICE WATER SYSTDil RESIDUAL HEAT REMOVAL SERVICE WATER SYSTEM - COMMON SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.1 At least the following inde)endent residual heat removal service water (RHRSW) system subsystems, with each su) system comprised of:

a.

Two OPERABLE RHRSW pumps, and b.

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

In OPERATIONAL CONDITIONS 1, 2, and 3, two subsystems, a.

b.

In OPERATIONAL CONDITIONS 4 and 5, the subsystem (s) associated with systems and components required OPERABLE by Specification 3.4.9.2, 3.9.11.1, and 3.9.11.2.

j APPLICABILITY:

OPERATIONAL CONDITIONS I, 2, 3, 4, and 5.

ACTION:

a.

In OPERATIONAL CONDITION 1, 2, or 3:

1.

With one RHRSW pump inoperable, restore the inoperable sump to OPERABLE status within 30 days, or be in at least HOT SiUTDOWN 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 I

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

2.

With one RHRSW pump in each subsystem inoperable, restore at least one of the inoperable RHRSW pumps to OPERABLE status within 7 days or be in at least H0T 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 .

3.

With one RHRSW subsystem otherwise inoperable, restore the inoperable subsystem 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 H0T SHUTDOWNwithinthenext12hoursandinCOLD the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.

With both RHRSW subsystems otherwise inoperable, restore at i

least one subsystem 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 .

"Whenever both RHRSW 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.

l LI!!ERICK - UNIT 2 3/4 7-1 Amendment No. BO,70,92

~.._.-

- - - -.. ~.

.>LANT SYSTEMS ER(dNCY SERU CE WATER SYSTEM - COMMON SYSTEM

.IMIT :NG COND

'10N FOR OPERATION 3.7.1.2 At least the following independent emergency service water system loops, with each loop comprised of:

Two OPERABLE emergency service water pumps, and a.

b.

An OPERABLE flow path capable of taking suction from the emergency service water pumps wet pits which are sup)1ied from the spray pond or the cooling tower basin and transferring tle water to the associated Unit 2 and common safety-related equipment, shall be OPERABLE:

In OPERATIONAL CONDITIONS 1, 2, and 3, two loops.

a.

b.

In OPERATIONAL CONDITIONS 4, 5, and *, one loop.

l APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, 5, and *.

ACTION:

a.

In OPERATION CONDITION 1, 2, or 3:

l 1.

With one emergency service water pump inoperable, restore the inoserable pump to OPERABLE status within 45 days or be in at least HOT SHUM0WN 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 one emergency service water pum) in each loop inoperable I

least one inoperable pump to OPERABLE status within 30 days or, restore at 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 .

3.

With one emergency service water system loop otherwise inoperable, declare all equipment aligned to the inoperable loop inoperable **, restore the inoperable loop to OPERABLE 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 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHU1DOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i

When handling irradiated fuel in the secondary containment.

- The diesel generators may be aligned to the OPERABLE emergency service water system loop provided confirmatory flow testing has been performed.

Those diesel generators not aligned to the OPERABLE emergency service water system loop shall be declared inoperable and the actions of 3.8.1.1 taken.

I i

i

}

Amendment No. 44,79. 92 LIHERICK - UNIT 2 3/4 7-3

~

..:..=

-=...x-I 1

j ELECTRICAL POWER SYSTEMS 1

SVRVElllANCE RE0VIREMENTS (Continued) b.

By removing accumulated water:

i i

1) From the day tank at least once per 31 days and after each occa-sion when the diesel is operated for greater than I hour, and

2) From the storage tank at least once per 31 days.

l c.

By sampling new fuel oil in accordance with ASTM D4057-81 prior to addition to the storage tanks and:

1) By verifying in accordance with the tests specified in ASTM D975-81 prior to addition to the storage tanks that the sample has:

a) An API Gravity of within 0.3 degrees at 60*F or a specific gravity of within 0.0016 at 60/60*F, when compared to the supplier's certificate or an absolute specific gravity at 60/60*F of greater than or equal to 0.83 but less than or equal to 0.89 or an API gravity at 60'F of greater than or l

equal to 27 degrees but less than or equal to 39 degrees.

I b) A kinematic viscosity at 40*C of greater than or equal to 1.9 centistokes, but less than or equal to 4.1 centistokes, if gravity was not determined by comparison with the supplier's certification, c) A flash point equal to or greater than 125*F, and l

d) A clear and bright appearance with proper color when tested in accordance with ASTM D4176-82.

2) By verifying within 31 days of obtaining the sample that the other properties specified in Table 1 of ASTM D975-81 are met when tested in accordance with ASTM 0975-81 except that the analysis for sulfur may be performed in accordance with ASTM D1552-79 or ASTM D2622-82.

d.

At least once every 31 days by obtaining a sample of fuel oil from the storage tanks in accordance with ASTM D2276-78, and verifying that total particulate contamination is less than 10 mg/ liter when checked in accordance with ASTM D2276-78, Method A, except that the filters specified in ASTM D2276-78, Sections 5.1.6 and 5.1.7, may have a nominal pore size of up to three (3) microns.

e.

At the following frequency by:

l

1) Every 18 months subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for this class of standby service.

2) Every 24 months verify each diesel generator's capability to reject a load of greater than or equal to that of its single largest post-accident load while maintaining voltage at 4285 i 420 volts and i.

frequency at 60 i 1.2 hz and after steady state conditions are reached, voltage is maintained at 4280 i 120 volts.

Amendment No. 84,65,85,92 LIMERICK - UNIT 2 3/4 B-4

ADMINISTRATIVE CONTROLS t

RESPONSIBILITIES 6.5.1.6 The PORC shall be responsible for:

a. Review of (1) Administrative Procedures and changes thereto, (2) new programs or procedures required by Specification 6.8 and requiring a 10 CFR 50.59 safety evaluation, and (3) proposed changes to programs or procedures required by Specification 6.8 and requiring a 10 CFR 50.59 safety evaluation;

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

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

d. Review of all proposed 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 and forwarding of reports covering evaluation and recomendations to prevent recurrence, to the Vice President, Limerick Generating Station, Plant Manager, and to the Nuclear Review Board;

g. Review of all REPORTABLE EVENTS;

h. Review of unit operations to detect potential hazards 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;

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

k. Review of the Emergency Plan and implementing procedures and submittal of the recomended changes to the Nuclear Review Board.

1. Review of every unplanned onsite release of radioactive material to the environs including the pre;aration 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 radwaste treatment systems.

n. Review of the Fire Protection Program and implementing procedures and the submittal of recomended changes to the Nuclear Review Board.

6.5.1.7 The PORC shall:

a. Recomend in writing to the Plant Manager approval or disapproval of items considered under Specification 6.5.1.6a. through d. prior to their implementation.

b. Render determinations in writing with regard to whether or not each item considered under Specification 6.5.1.6b. through f. constitutes an unreviewed safety question.

LIHERICK - UNIT 2 6-8 Amendment No. 40,66. 92

. 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-i ment between the PORC.and the Plant Manager; however, the Plant i

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

RECORDS i

i 6.5.1.8 The PORC shall maintain written minutes of each PORC meeting that, i

at a minimum, document the results of all PORC activities performed under the responsibilit provisions of these Technical Specifications. Copies shall be provided to t e 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:

I f

a. Nuclear power plant operations,

b. Nuclear engineering,

c. Chemistry and radiochemistry,

d.. Metallurgy,

e. Instrumentation and con' trol, i

f. Radiological safety,

g. Mechanical and electrical engineering, and

h. Quality assurance practices.

The NR8 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 i

6.5.2.2 The Chairman, members, and alternates of the NRB shall be appointed in writin by the Chief Nuclear Officer, and shall have an academic egree 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 j

be composed of no less than eight and no more than 12 members.

l The members and alternates of the NRB will be competent in the area of Quality Assurance practice and cognizant of the Qualit Assurance requirements of 10 CFR 1

. Part 50, Appendix B.

Additionally, they will e cognizant of the corporate Quality Assurance Program and will have the corporate Quality Assurance organization available to them.

4 Amendment No. 9,60,92 LINERICK - UNIT 2 6-9 c

,e-

..,--.v---

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a..

, - =,.,, - -

,,n.

r,

I i

APPENDIX B TO FACILITY OPERATING LICENSE NO. NPF-85 LIMERICK GENERATING STATION UNITS 1 AND 2 PECO ENERGY COMPANY DOCKET NOS. 50-352 and 50-353 ENVIRONMENTAL PROTECTION PLAN (NON-RADIOLOGICAL)

August 25, 1989 Amendment No gg

=

\\

The selection, calibration and use of equipment, conduct of the surveys, and the analysis and reportirig.of data shall conform to the provisions of the applicable American National Standards Institute Standards.

The results of the surveys conducted under this program shall be summarized, interpreted and reported in accordance with Section 5.4.1 of this EPP.

The final report of this program shall present a brief assessment by the licensee of the environmental impact and supplemental cooling water system operation on the various offsite acoustic environments, and shall describe the mitigative measures, if any, that have been, or are to be taken to reduce the impact of plant or supplemental cooling water system noise levels on the offsite environments. This report shall also contain a list of noise-related complaints or inquiries received by PECO Energy Company concerning the Limerick Generating Station or its supplemental cooling water system subsequent to issuance of the operating license along with a description of the action taken by PECO Energy Company to resolve these complaints or inquiries.

This program shall terminate upon completion of the collection of the specified sound level data for each phase and submission of an acceptable final report.

i 4.2.4.2 Point Pleasant Pumphouse An ASLB ruling (LBP-83-ll; March 8,1983) requires that the licensee conduct a one-time field study after the transformers are placed in operation at Point Pleasant.

The noise from operation of the transformers shall be reduced to a level so that the transformer core tones will be inaudiblo (i.e., not above the masking level, as defined below) at the site boundary.

The licensee shall determine, based on onsite measurements, the delta L(ex)

(i.e., the noise level in excess of the masking level) for each tone. The masking level is defined as "N" dB above the ambient spectrum level, where "N" is defined as follows:

Amendment No. 92 4-3

.-

ML20155C560

Text

_ -.

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