Text

Limerick Generating Station, Unit 2 - Enclosure 2: Bases
	ML14112A474
Person / Time
Site:	Limerick
Issue date:	04/10/2014
From:	; Exelon Generation Co
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML14112A484	List: ML14112A472; ML14112A473; ML14112A474;
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License No. NPF-85Limerick Generating Station,Unit No. 2Docket No. 50-353Issued by theU.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation LIMERICK GENERATING STATIONUNIT 2 OPERATING LICENSE NPF-85 PAGE REVISION LIST(Generated by Exelon -Not part of Operating License)Page/ Attachment Amendment No.1 thru 21 thru 344a5 thru 5a6166108166Revised by letter datedAugust 9, 2007149Original IssueILicense Amendment Change List (Generated by Exelon)PageiDated December 19, 1994LIMERICK

-UNIT 2-A -Amendment No. 166 K1Ek, UNITED STATESNUCLEAR REGULATORY COMMISSION

-. WASHINGTON, D.C. 20555-0001 EXELON GENERATION

COMPANY, LLCDOCKET NO. 50-353LIMERICK GENERATING

STATION, UNIT 2AMENDMENT TO FACILITY OPERATING LICENSEAmendment No. 171License No. NPF-85The U.S. Nuclear Regulatory Commission (the Commission) has found that:A. The application for amendment by Exelon Generation

Company, LLC (thelicensee) dated July 6, 2012, complies with the standards and requirements ofthe 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 theAct, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by thisamendment can be conducted without endangering the health and safety of thepublic, and (ii) that such activities will be conducted in compliance with theCommission's regulations; D. The issuance of this amendment will not be inimical to the common defense andsecurity or to the health and safety of the public; andE. The issuance of this amendment is in accordance with 10 CFR Part 51 of theCommission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications asindicated in the attachment to this license amendment, and paragraph 2.C.(2) of FacilityOperating License No. NPF-85 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 171, are hereby incorporated into this license.

Exelon Generation Companyshall operate the facility in accordance with the Technical Specifications and theEnvironmental Protection Plan.3. This license amendment is effective as of the date of its issuance and shall beimplemented within 30 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Jeremy S. Bowen, ChiefPlant Licensing Branch 111-2Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to theFacility Operating LicenseDate of Issuance:

June 20, 2013 UNITED STATESNUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION

COMPANY, LLCDOCKET NO. 50-353LIMERICK GENERATING

STATION, UNIT 2FACILITY OPERATING LICENSELicense No. NPF-85The Nuclear Regulatory Commission (the Commission or the NRC) has found that:A. The application for license filed by Exelon Generation

Company, LLC (ExelonGeneration Company or the licensee) complies with the standards andrequirements of the Atomic Energy Act of 1954, as amended (the Act), and theCommission's regulations set forth in 10 CFR Chapter I, and all requirednotifications to other agencies or bodies have been duly made;B. Construction of the Limerick Generating

Station, Unit 2 (the facility) has beensubstantially completed in conformity with Construction Permit No. CPPR-107and the application, as amended, the provisions of the Act and the regulations ofthe Commission; C. The facility will operate in conformity with the application, as amended, theprovisions of the Act, and the regulations of the Commission (except asexempted from compliance in Section 2.D. below);D. There is reasonable assurance:

(i) that the activities authorized by this operating license can be conducted without endangering the health and safety of thepublic, and (ii) that such activities will be conducted in compliance with theCommission's regulations set forth in 10 CFR Chapter I (except as exemptedfrom compliance in Section 2.D. below);E. The licensee is technically qualified to engage in the activities authorized by thislicense in accordance with the Commission's regulations set forth in 10 CFRChapter I;F. The licensee has satisfied the applicable provisions of 10 CFR Part 140,"Financial Protection Requirements and Indemnity Agreements,"

of theCommission's regulations; G. The issuance of this license will not be inimical to the common defense andsecurity or to the health and safety of the public;Amendment No. 92, 108 H. After weighing the environmental,

economic, technical, and other benefits of thefacility against environmental and other costs and considering available alternatives, the issuance of this Facility Operating License No. NPF-85, subjectto 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 ofthe Commission's regulations and all applicable requirements have beensatisfied; andThe receipt, possession, and use of source, byproduct and special nuclearmaterial as authorized by this license will be in accordance with theCommission'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 dated July 7, 1989,and the Commission's Memorandum and Order dated August 25, 1989, regarding thisfacility, Facility Operating License NPF-85 is hereby issued to the Exelon Generation Company (the licensee),

to read as follows:A. This license applies to the Limerick Generating

Station, Unit 2, a boiling waternuclear reactor and associated equipment, owned by Exelon Generation Company.

The facility is located on the licensee's site in Montgomery andChester Counties, Pennsylvania on the banks of the Schuylkill Riverapproximately 1.7 miles southeast of the city limits of Pottstown, Pennsylvania and 21 miles northwest of the city limits of Philadelphia, Pennsylvania, and isdescribed in the licensee's Final Safety Analysis Report, as supplemented andamended, and in the licensee's Environmental Reporf-Operating License Stage,as supplemented and amended.B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses Exelon Generation Company:(1) Pursuant to Section 103 of the Act and 10 CFR Part 50, to possess, use,and operate the facility at the designated location in Montgomery andChester Counties, Pennsylvania, in accordance with the procedures andlimitations set forth in this license;(2) Pursuant to the Act and 10 CFR Part 70, to receive, possess and to useat any time special nuclear material as reactor fuel, in accordance withthe limitations for storage and amounts required for reactor operation, asdescribed in the Final Safety Analysis Report, as supplemented andamended;(3) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possessand use at any time any byproduct, source and special nuclear materialas sealed neutron sources for reactor startup, sealed sources for reactorinstrumentation and radiation monitoring equipment calibration, and asfission detectors in amounts as required; Amendment No. 92, 108 (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 nuclearmaterial without restriction to chemical or physical form, for sampleanalysis 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 notseparate, such byproduct and special nuclear materials as may beproduced by the operation of the facility, and to receive and possess, butnot separate, such source, byproduct, and special nuclear materials ascontained in the fuel assemblies and fuel channels from the ShorehamNuclear 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 asexempted from compliance in Section 2.D. below) and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified orincorporated below:(1) Maximum Power LevelExelon Generation Company is authorized to operate the facility atreactor core power levels of 3515 megawatts thermal (100 percent ratedpower) in accordance with the conditions specified herein.(2) Technical Specifications The Technical Specifications contained in Appendix A, and theEnvironmental Protection Plan contained in Appendix B, as revisedthrough Amendment No. 171, are hereby incorporated into this license.Exelon Generation Company shall operate the facility in accordance withthe Technical Specifications and the Environmental Protection Plan.(3) Fire Protection (Section 9.5, SSER-2. -4)*Exelon Generation Company shall implement and maintain in effect allprovisions of the approved Fire Protection Program as described in theUpdated Final Safety Analysis Report for the facility, and as approved inthe NRC Safety Evaluation Report dated August 1983 throughSupplement 9, dated August 1989, and Safety Evaluation datedNovember 20, 1995, subject to the following provision:

The licensee may make changes to the approved fire protection programwithout prior approval of the Commission only if those changes would notadversely affect the ability to achieve and maintain safe shutdown in theevent of a fire.*The parenthetical notation following the title of license conditions denotes the sectionof the Safety Evaluation Report andjor its supplements wherein the license condition isdiscussed.

Amendment No. 171 (4) Physical Security and Safeauards Exelon Generation Company shall fully implement and maintain in effectall provisions of the Commission-approves physical

security, training andqualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments andSearch Requirements revisions to 10 CFR 73.55 (51 FR 27817 and27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p).

Thecombined set of plans1, submitted by letter dated May 17, 2006, isentitled:

"Limerick Generating Station Security Plan, Training andQualification Plan, and Safeguards Contingency Plan, Revision 2.' Theset contains Safeguards Information protected under 10 CFR 73.21.Exelon Generation Company shall fully implement and maintain in effectall provisions of the Commission-approved cyber security plan (CSP),including changes made pursuant to the authority of 10 CFR 50.90 and10 CFR 50.54(p).

The Exelon Generation Company CSP was approvedby License Amendment No. 166.(5) Exelon Generation Company shall provide the Director of the Office ofNuclear Reactor Regulation a copy of any application, at the time it isfiled, to transfer (excluding grants of security interests or liens) fromExelon Generation Company to its direct or indirect parent, or to anyother affiliated

company, facilities for the production, transmission ordistribution of electric energy having a depreciated book value exceeding ten percent (10%) of Exelon Generation Company's consolidated netutility plant, as recorded on Exelon Generation Company's book ofaccounts.

(6) Exelon Generation Company shall have decommissioning trust funds forLimerick, Unit 2, in the following minimum amount, when Limerick, Unit 2,is transferred to Exelon Generating Company:Limerick, Unit 2 $59,687,081 (7) The decommissioning trust agreement for Limerick, Unit 2, at the timethe transfer of the unit to Exelon Generation Company is effected andthereafter, is subject to the following:

(a) The decommissioning trust agreement must be in a formacceptable to the NRC.(b) With respect to the decommissioning trust fund, investments in thesecurities or other obligations of Exelon Corporation or affiliates

thereof, or their successors or assigns are prohibited.

Except forinvestments tied to market indexes or other non-nuclear sectormutual funds, investments in any entity owning one or morenuclear power plants are prohibited.

1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to theSecurity Plan.Amendment No. W 8-1-, 408, 166Ro ..o ..y latt ,r dated O ctob or 28, 2001V16-co by ett datod Doceambor1,20 Re.io b etr ae ay3,20

-4a-(c) The decommissioning trust agreement for Limerick, Unit 2, mustprovide that no disbursements or payments from the trust shall bemade by the trustee unless the trustee has first given the Directorof the Office of Nuclear Reactor Regulation 30 days prior writtennotice of payment.

The decommissioning trust agreement shallfurther contain a provision that no disbursements or paymentsfrom the trust shall be made if the trustee receives prior writtennotice of objection from the NRC.(d) The decommissioning trust agreement must provide that theagreement can not be amended in any material respect without30 days prior written notification to the Director of the Office ofNuclear Reactor Regulation.

(e) The appropriate section of the decommissioning trust agreement shall state that the trustee, investment

advisor, or anyone elsedirecting the investments made in the trust shall adhere to a'"prudent investor'

standard, as specified in 18 CFR 35.32(a)(3) ofthe Federal Energy Regulatory Commission's regulations.

(8) Exelon Generation Company shall take all necessary steps to ensure thatthe decommissioning trust is maintained in accordance with theapplication for approval of the transfer of Umerick, Unit 2, license and therequirements of the Order approving the transfer, and consistent with thesafety evaluation supporting the Order.(9) Mitigation Strategy License Condition Develop and maintain strategies for addressing large fires and explosions and that include the following key areas:(a) Fire fighting response strategy with the following elements:

1. Pre-defined coordinated fire response strategy and guidance2. Assessment of mutual aid fire fighting assets3. Designated staging areas for equipment and materials

4. Command and control5. Training of response personnel (b) Operations to mitigate fuel damage considering the following:

1. Protection and use of personnel assets2. Communications

3. Minimizing fire spread4. Procedures for implementing integrated fire response strategy5. Identification of readily-available pre-staged equipment

6. Training on integrated fire response strategy7. Spent fuel pool mitigation measures(c) Actions to minimize release to include consideration of:1. Water spray scrubbing

2. Dose to onsite responders Amendment No. 408Revised by letter dated August 9, 2007 (10) The licensee shall implement and maintain all Actions required byAttachment 2 to NRC Order EA-06-137, issued June 20, 2006, exceptthe last action that requires incorporation of the strategies into the sitesecurity plan, contingency plan, emergency plan and/or guard trainingand qualification plan, as appropriate.

(11) Upon implementation of Amendment No. 149 adopting TSTF-448, Revision 3,the determination of control room envelope (CRE) unfiltered air inleakage asrequired by SR 4.7.2.2.a, in accordance with TS 6.16.c.(i),

the assessment ofCRE habitability as required by Specification 6.16.c.(ii),

and the measurement of CRE pressure as required by Specification 6.16.d, shall be considered met.Following implementation:

(a) The first performance of SR 4.7.2.2.a, in accordance with Specification 6.16.c.(i),

shall be within the specified Frequency of 6 years, plus the18-month allowance of SR 4.0.2, as measured from September 16,2004, the date of the most recent successful tracer gas test, as statedin the December 10, 2004 letter response to Generic Letter 2003-01, orwithin the next 18 months if the time period since the most recentsuccessful tracer gas test is greater than 6 years.(b) The first performance of the periodic assessment of CRE habitability, Specification 6.16.c.(ii),

shall be within 3 years, plus the 9-monthallowance of SR 4.0.2, as measured from September 16, 2004, the dateof the most recent successful tracer gas test, as stated in the December10, 2004 letter response to Generic Letter 2003-01, or within the next 9months if the time period since the most recent successful tracer gastest is greater than 3 years.(c) The first performance of the periodic measurement of CRE pressure, Specification 6.16.d, shall be within 24 months, plus the 180 daysallowed by SR 4.0.2, as measured from September 16, 2004, the date ofthe most recent successful pressure measurement test, or within 180days if not performed previously.

D. The facility requires exemptions from certain requirements of 10 CFR Part 50and 10 CFR Part 70. These include (a) exemption from the requirement ofAppendix 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 leak rate testing of the Main SteamIsolation Valves (MSIVs) at the peak calculated containment

pressure, Pa, andexemption from the requirements of Appendix J that the measured MSIV leakrates be included in the summation for the local leak rate test (Section 6.2.6.1 ofSSER-3),

(c) exemption from the requirement of Appendix J,the local leak rate testing of the Traversing Incore Probe ShearValves (Section 6.2.6.1 of the SER and SSER-3),

and (d) an exemption Amendment No. 92, 408, 443, 149Revised by letter dated August 9, 2007

-5a -from the schedule requirements of 10 CFR 50.33(k)(1) related toavailability of funds for decommissioning the facility (Section22.1, 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. Anexemption from the criticality monitoring requirements of 10 CFR 70.24was previously granted with NRC materials license No. SNM-1 977 issuedNovember 22, 1988. The licensee is hereby exempted from therequirements of 10 CFR 70.24 insofar as this requirement applies to thehandling and storage of fuel assemblies held under this license.These exemptions are authorized by law, will not present an unduerisk to the public health and safety, and are consistent with thecommon defense and security.

The exemptions in items a, b, c, and dabove are granted pursuant to 10 CFR 50.12. With these exemptions, the facility will operate, to the extent authorized herein, inconformity with the application, as amended, the provisions of theAct, and the rules and regulations of the Commission.

E. DeletedF. The licensee shall have and maintain financial protection of suchtype and in such amounts as the Commission shall require inaccordance with Section 170 of the Atomic Energy Act of 1954, asamended, to cover public liability claims.Amendment No. 92, 408, 443,149Revised by letter dated August 9, 2007 3893262360 6 -G. This license is effective as of the date of issuance and shallexpire at midnight on June 22, 2029.FOR THE NUCLEAR REGULATORY COMMISSION Thomas E. Murley, DirectorOffice of Nuclear Reactor Regulation

Enclosures:

1. Appendix A -Technical Specifications (NUREG-1376)

2. Appendix B -Environmental Protection PlanDate of Issuance:

August 25, 1989A 11S 2 5 1989 UNIT 2 OPERATING LICENSE NPF-85LICENSE AMENDMENT CHANGE LIST(Generated by PECO from NRC Issued Amendments)

License Section 2.C.(2)The License and the Technical Specifications contained in Appendix A havebeen revised through the Amendment number on the cover sheet of theLicense.

The Technical Specifications as contained in Appendix A arehereby incorporated into this license.

PECO Energy shall operate thefacility in accordance with the Technical Specifications and theEnvironmental Protection Plan.LIMERICK

-UNIT 2-i -12/19/94 Technical Specifications Limerick Generating Station,Unit No. 2Docket No. 50-353Appendix "A" toLicense No. NPF-85Issued by theU.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation LIMERICK GENERATING STATIONUNIT 2 TECHNICAL SPECIFICATION PAGE REVISION LISTIndex Amendment Nos.i 48ii 153iii 48iv Original Issuev 4vi 48vii 147viii 147ix 147x 147xi 160xii 107xiii 135xiv 150xv 170xvi 95xvii 11xviii 48xix 117xx 160xxi 149xxii 95xxiii 11xxiv 11xxv Original Issuexxvi 138xxvii 153xxviii 149Section 1.0 Definitions 1-1 Original Issue1-2 1461-3 481-4 1531-5 1071-6 1631-7 1481-8 1481-9 341-10 112LIMERICK

-UNIT 2-A-LIMERICK GENERATING STATIONUNIT 2 TECHNICAL SPECIFICATION PAGE REVISION LISTIndex Amendment Nos.Section 2.0 Safety Limits and Limiting Safety System Settings2-1 1622-2 Original Issue2-3 1092-4 1632-4a Original IssueBases for Section 2.0B 2-1 ECR LG 12-00035B 2-2 ECR LG 12-00035B 2-3 Original IssueB 2-4 Original IssueB 2-5 Original IssueB 2-6 139B 2-7 Associated with Amendment 163B 2-7a Associated with Amendment 163B 2-8 52B 2-9 Original IssueB 2-10 139Section 3.0 and 4.0 Limiting Conditions for Operation and Surveillance Requirements 3/40-1 1323/4 0-2 1553/4 0-3 1553/4 1-1 Original Issue3/4 1-2 Original Issue3/4 1-3 1403/4 1-4 1473/4 1-5 1473/4 1-6 1473/4 1-7 Original Issue3/4 1-8 1323/4 1-9 1053/4 1-10 1473/4 1-11 1323/4 1-12 Original Issue3/4 1-13 1323/4 1-14 1473/4 1-15 Original IssueLIMERICK

-UNIT 2-B -

LIMERICK GENERATING STATIONUNIT 2 TECHNICAL SPECIFICATION PAGE REVISION LISTAmendment Nos.Index3/4 1-163/4 1-173/4 1-183/4 1-193/4 1-203/4 1-213/4 1-223/42-13/4 2-23/4 2-3 thru 3/4 2-6a3/4 2-73/4 2-83/4 2-93/4 2-103/4 2-10a thru 3/4 2-113/4 2-123/43-13/4 3-1a3/4 3-23/4 3-33/4 3-43/4 3-53/4 3-63/4 3-73/4 3-83/4 3-8a3/4 3-93/4 3-103/43-113/4 3-123/4 3-133/4 3-143/4 3-153/4 3-163/4 3-173/4 3-183/4 3-193/4 3-203/4 3-213/4 3-223/4 3-233/4 3-243/4 3-25Original IssueOriginal Issue147163163Original IssueOriginal Issue1474Deleted48481474Deleted1471391471395216116313916316316313214752Original IssueOriginal Issue747414610752164517474939393LIMERICK

-UNIT 2-C-LIMERICK GENERATING STATIONUNIT 2 TECHNICAL SPECIFICATION PAGE REVISION LISTAmendment Nos.Index3/4 3-263/4 3-273/4 3-283/4 3-293/4 3-303/4 3-313/4 3-323/4 3-333/4 3-343/4 3-353/4 3-363/4 3-36a3/4 3-373/4 3-383/4 3-393/4 3-403/43-413/4 3-423/4 3-433/43-443/43-453/4 3-463/43-473/4 3-483/4 3-493/4 3-503/4 3-513/4 3-523/4 3-533/4 3-543/4 3-553/4 3-563/4 3-573/4 3-583/4 3-593/4 3-603/4 3-60a3/4 3-60b3/4 3-613/4 3-623/4 3-633/4 3-643/4 3-65107147147147147147147OriginalOriginal17120120OriginalOriginal931471471473351147163147163OriginalOriginal1471471717IssueIssueIssueIssueIssueIssueOriginal Issue1471471391091631393147147147146146LIMERICK

-UNIT 2 LIMERICK GENERATING STATIONUNIT 2 TECHNICAL SPECIFICATION PAGE REVISION LISTAmendment Nos.Index3/4 3-663/4 3-673/4 3-683/4 3-69 thru 3/4 3-723/4 3-74 thru 3/4 3-753/4 3-763/4 3-773/4 3-783/4 3-793/4 3-803/4 3-813/4 3-823/4 3-833/4 3-843/4 3-853/4 3-863/4 3-873/4 3-883/4 3-893/4 3-903/4 3-913/4 3-923/4 3-92a thru 3-963/4 3-973/4 3-983/4 3-99 thru 3/4 3-1023/4 3-1033/4 3-1043/4 3-1053/4 3-1063/4 3-1073/4 3-1083/4 3-1093/4 3-1103/43-1113/4 3-1123/43-1133/4 3-1143/4 3-115147147153DeletedDeleted147OriginalOriginalOriginal47IssueIssueIssueOriginal IssueOriginal Issue1471471521351521477914714768Deleted11711Deleted14711111114714711153Deleted14755Original Issue147LIMERICK

-UNIT 2-E -

LIMERICK GENERATING STATIONUNIT 2 TECHNICAL SPECIFICATION PAGE REVISION LISTAmendment Nos.Index3/44-13/4 4-1a3/4 4-23/4 4-33/4 4-43/4 4-4a3/4 4-53/4 4-63/4 4-73/4 4-83/4 4-8a3/4 4-93/4 4-103/44-113/4 4-123/4 4-13 thru 3/4 4-143/4 4-153/44-163/4 4-173/44-183/4 4-193/4 4-203/4 4-213/4 4-223/4 4-233/4 4-243/4 4-253/4 4-263/45-13/4 5-23/4 5-33/4 5-43/4 5-53/4 5-63/4 5-73/4 5-83/4 5-9163139163139157157147Original Issue147169167144147144136Deleted136Original Issue1471471471251301471321601471471539213214714759147Original Issue147LIMERICK

-UNIT 2-F -

LIMERICK GENERATING STATIONUNIT 2 TECHNICAL SPECIFICATION PAGE REVISION LISTAmendment Nos.Index3/46-13/4 6-23/4 6-33/4 6-43/4 6-53/4 6-63/4 6-73/4 6-83/4 6-93/4 6-103/46-113/4 6-123/4 6-133/4 6-143/4 6-153/4 6-163/4 6-173/4 6-183/4 6-193/4 6-20 th3/4 6-443/4 6-453/4 6-463/4 6-473/4 6-483/4 6-493/4 6-503/4 6-513/4 6-51a3/4 6-523/4 6-52a3/4 6-533/4 6-543/4 6-553/4 6-563/4 6-573/4 6-583/4 6-593/47-13/4 7-1a3/4 7-23/4 7-3ru 3/4 6-43a147107146811321475381147153147Original Issue147147147165153147107Deleted914714714714715314715315314614714786147147135147147165165147165LIMERICK

-UNIT 2 LIMERICK GENERATING STATIONUNIT 2 TECHNICAL SPECIFICATION PAGE REVISION LISTAmendment Nos.Index3/4 7-43/4 7-53/4 7-63/4 7-6a3/4 7-73/4 7-83/4 7-93/4 7-103/47-113/4 7-11a3/4 7-11b3/4 7-123/4 7-133/4 7-143/4 7-153/4 7-163/4 7-173/4 7-183/4 7-193/4 7-20 thru 3/4 7-323/4 7-333/48-13/4 8-1a3/4 8-23/4 8-2a3/4 8-33/4 8-4.3/4 8-53/4 8-63/4 8-73/4 8-7a3/4 8-83/4 8-93/4 8-103/4 8-10a3/4 8-113/4 8-123/4 8-133/4 8-143/4 8-14a3/4 8-151471471491531491491471471515151542191919147Original Issue68Deleted147165150165150150150147147147150150154126126147147126126126Original IssueLIMERICK

-UNIT 2-H -

LIMERICK GENERATING STATIONUNIT 2 TECHNICAL SPECIFICATION PAGE REVISION LISTAmendment Nos.Index3/4 8-163/4 8-16a3/4 8-173/4 8-183/4 8-18a3/4 8-193/4 8-203/4 8-213/4 8-22 thru 3/4 8-263/4 8-273/4 8-283/49-13/4 9-23/4 9-33/4 9-43/4 9-53/4 9-63/4 9-73/4 9-83/4 9-93/4 9-103/49-113/4 9-123/4 9-133/4 9-143/4 9-153/4 9-163/4 9-173/4 9-183/4 10-13/4 10-23/4 10-33/4 10-43/4 10-53/4 10-63/4 10-73/4 10-83/4 10-93/4 11-13/4 11-2 thru 3/4 11-63/4 11-73/4 11-83/4 11-9 thru 3/4 11-14Original Issue102147Original IssueOriginal Issue102147153Deleted170147112147147147147Original Issue14788147147147Original Issue147Original Issue147147147147Original Issue147147147147Original IssueOriginal Issue9511Deleted14711DeletedLIMERICK

-UNIT 2-I-LIMERICK GENERATING STATIONUNIT 2 TECHNICAL SPECIFICATION PAGE REVISION LISTAmendment Nos.Index3/4 11-153/411-163/4 11-173/4 11-183/4 11-19 thru 11-203/4 12-13/4 12-2 thru 12-14Original Issue1471111Deleted11DeletedBases for Sections 3.0 and 4.0B 3/4 0-1B 3/4 0-2B 3/4 0-3B 3/4 0-3aB 3/4 0-3bB 3/4 0-3cB 3/4 0-4B 3/4 0-5B 3/4 0-6B 3/4 1-1B 3/4 1-2B 3/4 1-2aB 3/4 1-3B 3/4 1-4B 3/4 1-5B 3/4 2-1B 3/4 2-2B 3/4 2-3B 3/4 2-4B 3/4 2-5B 3/4 3-1B 3/4 3-1aB 3/4 3-1bB 3/4 3-1cB 3/4 3-1dB 3/4 3-1eB 3/4 3-1fB 3/4 3-2B 3/4 3-3B 3/4 3-3aB 3/4 3-4B 3/4 3-5B 3/4 3-5aOriginal IssueOriginal Issue132132132132124132155Original Issue131140147Associated with Amendment 163147484814ECR LG 12-0003548147Associated with Amendment 163139139Associated with Amendment Associated with Amendment Associated with Amendment 147147163163163Associated with Amendment 163147ECR LG 09-00585ECR LG 09-00585LIMERICK

-UNIT 2 LIMERICK GENERATING STATIONUNIT 2 TECHNICAL SPECIFICATION PAGE REVISION LISTAmendment Nos.IndexB 3/4 3-6B 3/4 3-7B 3/4 3-8B 3/4 3-9B 3/4 4-1B 3/4 4-2B 3/4 4-3B 3/4 4-3aB 3/4 4-3bB 3/4 4-3cB 3/4 4-3dB 3/4 4-3eB 3/4 4-4B 3/4 4-5B 3/4 4-6B 3/4 4-6aB 3/4 4-7B 3/4 4-8B 3/4 5-1B 3/4 5-2B 3/4 5-2aB 3/4 6-1B 3/4 6-2B 3/4 6-3B 3/4 6-3aB 3/4 6-4B 3/4 6-4aB 3/4 6-5B 3/4 6-5aB 3/4 6-6B 3/4 6-7B 3/4 7-1B 3/4 7-1aB 3/4 7-1bB 3/4 7-1cB 3/4 7-2B 3/4 7-3B 3/4 7-3aB 3/4 7-4B 3/4 7-5147139Original Issue109Associated with Amendment 157Associated with Amendment 157Associated with Amendment 169Associated with Amendment 169Associated with Amendment 169Associated with Amendment 167Associated with Amendment 167Associated with Amendment 167132130Associated with Amendment 1608211151ECR LG 00-00177147116ECR 11-003951475131147110ECR LG 09-00052ECR LG 09-00052861351491491491491542Original Issue6816LIMERICK

-UNIT 2-K-LIMERICK GENERATING STATIONUNIT 2 TECHNICAL SPECIFICATION PAGE REVISION LISTAmendment Nos.IndexB 3/4 8-1B 3/4 8-1aB 3/4 8-1bB 3/4 8-1cB 3/4 8-1dB 3/4 8-1eB 3/4 8-2B 3/4 8-2aB 3/4 8-2bB 3/4 8-3B 3/4 9-1B 3/4 9-2B 3/4 9-2aB 3/4 10-1B 3/4 10-2B 3/4 11-1B 3/4 11-2B 3/4 11-3B 3/4 11-4B 3/4 11-5B 3/4 12-1B 3/4 12-2ECR 05-00297ECR 09-00284ECR 09-00284150126ECR 09-00284147147126Associated with Amendment 170Original Issue82ECR LG 01 -00386Original Issue13011Associated with 14811111111DeletedSection 5.0 Design Features5-15-25-35-45-55-65-75-85-911Original IssueOriginal IssueOriginal IssueOriginal Issue11Original Issue51Original IssueSection 6.0 Administrative Controls6-16-26-36-46-56-660159159260171IlLIMERICK

-UNIT 2 LIMERICK GENERATING STATIONUNIT 2 TECHNICAL SPECIFICATION PAGE REVISION LISTIndex Amendment Nos.6-7 1386-8 1386-9 1386-10 1386-11 606-12 1386-12a 1386-13 1386-14 1296-14a 1586-14b 116-14c 1516-14d 1476-15 1376-16 1376-17 1006-18 116-18a 1616-19 1386-20 1386-20a 1006-21 1006-21a 1386-22 1496-23 149LIMERICK

-UNIT 2 INDEX THIS PAGE INTENTIONALLY LEFT BLANK INDEXDEFINITIONS SECTION1.0 DEFINITIONS PAGE1 .1 ACT ION ........................................................

1-11.2 AVERAGE PLANAR EXPOSURE

.......................................

1-11.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE ....................

1-11.4 CHANNEL CALIBRATION

...........................................

1-11.5 CHANNEL CHECK .................................................

1-11.6 CHANNEL FUNCTIONAL TEST .......................................

1-11.7 CORE ALTERATION

...............................................

1-21.7A CORE OPERATING LIMITS REPORT ..................................

1-21.8 CRITICAL POWER RATIO ..........................................

1-21.9 DOSE EQUIVALENT 1-131 .........................................

1-21.9a DOWNSCALE TRIP SETPOINT (DTSP) ................................

1-21.10 E-AVERAGE DISINTEGRATION ENERGY ..............................

1-21.11 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME ............

1-21.12 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME ..... 1-31.13 (DELETED)

.....................................................

1-31.14 (DELETED)

.....................................................

1-31.15 FREQUENCY NOTATION

............................................

1-31.15a HIGH (POWER) TRIP SETPOINT (HTSP) .............................

1-31.16 IDENTIFIED LEAKAGE ............................................

1-31.16a INTERMEDIATE (POWER) TRIP SETPOINT (ITSP) .....................

1-31.17 ISOLATION SYSTEM RESPONSE TIME ................................

1-31.18 LIMITING CONTROL ROD PATTERN ..................................

1-31.19 LINEAR HEAT GENERATION RATE ...................................

1-31.20 LOGIC SYSTEM FUNCTIONAL TEST ..................................

1-4LIMERICK

-UNIT 2iAmendment No. 4, 48 INDEXDEFINITIONS SECTIONDEFINITIONS (Continued)

PAGE1.20a LOW (POWER) TRIP SETPOINT (LTSP) ..............................

1-41.21 (DELETED)

....................................................

1-41.22 MEMBER(S)

OF THE PUBLIC ......................................

1-41.22a MAPFAC(F)

-(MAPLHGR FLOW FACTOR) ............................

1-41.22b MAPFAC(P)

-(POWER DEPENDENT MAPLHGR MULTIPLIER)

.............

1-41.23 MINIMUM CRITICAL POWER RATIO (MCPR) ..........................

1-41.24 OFFSITE DOSE CALCULATION MANUAL ..............................

1-41.25 OPERABLE

-OPERABILITY

.......................................

1-41.26 OPERATIONAL CONDITION

-CONDITION

............................

1-51.27 PHYSICS TESTS ................................................

1-51.28 PRESSURE BOUNDARY LEAKAGE ....................................

1-51.29 PRIMARY CONTAINMENT INTEGRITY

................................

1-51.30 PROCESS CONTROL PROGRAM ......................................

1-51.31 PURGE -PURGING ..............................................

1-61.32 RATED THERMAL POWER ..........................................

1-61.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY

............

1-61.34 REACTOR PROTECTION SYSTEM RESPONSE TIME ......................

1-61.35 RECENTLY IRRADIATED FUEL .....................................

1-61.36 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY

..............

1-61.37 REPORTABLE EVENT .............................................

1-71.37a RESTRICTED AREA ..............................................

1-71.38 (DELETED)

....................................................

1-71.39 SHUTDOWN MARGIN ..............................................

1-71.40 SITE BOUNDARY

................................................

1-71.41 SOURCE CHECK .................................................

1-7LIMERICK

-UNIT 2i iAmendment No. 4-,-1,48,-1-3, 168 INDEXDEFINITIONS SECTIONDEFINITIONS (Continued) 1.42 STAGGERED TEST BASIS ...................

1.43 THERMAL POWER ..........................

1.43A TURBINE BYPASS SYSTEM RESPONSE TIME ....1.44 UNIDENTIFIED LEAKAGE ...................

1.45 UNRESTRICTED AREA ......................

1.46 VENTILATION EXHAUST TREATMENT SYSTEM...

1.47 VENTING ................................

Table 1.1, Surveillance Frequency Notation

.......Table 1.2, Operational Conditions

................

............

PAGE1-81-81-81-81-81-81-81-91-10LIMERICK

-UNIT 2ii iAmendment No. 48 INDEXSAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGSSECTION PAGE2.1 SAFETY LIMITSTHERMAL POWER, Low Pressure or Low Flow ...........................

2-1THERMAL POWER, High Pressure and High Flow ........................

2-1Reactor Coolant System Pressure

...................................

2-1Reactor Vessel Water Level ........................................

2-22.2 LIMITING SAFETY SYSTEM SETTINGSReactor Protection System Instrumentation Setpoints

...............

2-3Table 2.2.1-1 Reactor Protection SystemInstrumentation Setpoints

.....................

2-4BASES2.1 SAFETY LIMITSTHERMAL POWER, Low Pressure or Low Flow ...........................

B 2-1THERMAL POWER, High Pressure and High Flow ........................

B 2-2Intentionally Left Blank ..........................................

B 2-3Intentionally Left Blank ..........................................

B 2-4Reactor Coolant System Pressure

...................................

B 2-5Reactor Vessel Water Level ........................................

B 2-52.2 LIMITING SAFETY SYSTEM SETTINGSReactor Protection System Instrumentation Setpoints

...............

B 2-6LIMERICK

-UNIT 2i v INDEXLIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE3/4.0 APPLICABILITY

....................................................

3/4 0-13/4.1 REACTIVITY CONTROL SYSTEMS3/4.1.1 SHUTDOWN MARGIN ...............................................

3/4 1-13/4.1.2 REACTIVITY ANOMALIES

..........................................

3/4 1-23/4.1.3 CONTROL RODSControl Rod Operability

.......................................

3/4 1-3Control Rod Maximum Scram Insertion Times .....................

3/4 1-6Control Rod Average Scram Insertion Times .....................

3/4 1-7Four Control Rod Group Scram Insertion Times ..................

3/4 1-8Control Rod Scram Accumulators

................................

3/4 1-9Control Rod Drive Coupling

....................................

3/4 1-11Control Rod Position Indication

...............................

3/4 1-13Control Rod Drive Housing Support .............................

3/4 1-153/4.1.4 CONTROL ROD PROGRAM CONTROLSRod Worth Minimizer

...........................................

3/4 1-16Rod Block Monitor .............................................

3/4 1-183/4.1.5 STANDBY LIQUID CONTROL SYSTEM .................................

3/4 1-19Figure 3.1.5-1 Sodium Pentaborate SolutionTemperature/Concentration Requirements

............................

3/4 1-21Figure 3.1.5-2 (LEFT BLANK INTENTIONALLY)

..............

3/4 1-223/4.2 POWER DISTRIBUTION LIMITS3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE ....................

3/4 2-1Information on pages 3/4 2-2 thru 3/4 2-6a has beenINTENTIONALLY

OMITTED, refer to note on page 3/4 2-2 ....... 3/4 2-2LIMERICK

-UNIT 2vAmendment No. 4 1 INDEXLIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PAGEPOWER DISTRIBUTION LIMITS (Continued) 3/4.2.2 (DELETED)

.....................................................

3/4 2-73/4.2.3 MINIMUM CRITICAL POWER RATIO ..................................

3/4 2-8Table 3.2.3-1 Deleted.Information on pages 3/4 2-10 thru 3/4 2-11 has beenINTENTIONALLY

OMITTED, refer to Note on page 3/4 2-10 .........

3/4 2-103/4.2.4 LINEAR HEAT GENERATION RATE ...................................

3/4 2-123/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION

.....................

3/4 3-1Table 3.3.1-1 Reactor Protection SystemInstrumentation

.............................

3/4 3-2Table 3.3.1-2 Reactor Protection SystemResponse Times ..............................

3/4 3-6Table 4.3.1.1-1 Reactor Protection SystemInstrumentation Surveillance Requirements

................................

3/4 3-7LIMERICK UNIT -2viAmendment No. 4, 48 INDEXLIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGEINSTRUMENTATION (Continued) 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION

...........................

3/4 3-9Table 3.3.2-1 Isolation Actuation Instrumentation

...... 3/4 3-11Table 3.3.2-2 Isolation Actuation Instrumentation Setpoints

................

3/4 3-18Table 3.3.2-3 Isolation System Instrumentation Response Time ............................

3/4 3-23Table 4.3.2.1-1 Isolation Actuation Instrumentation Surveillance Requirements

..............

3/4 3-273/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION

...............................................

3/4Table 3.3.3-1 Emergency Core Cooling SystemActuation Instrumentation

................

3/4Table 3.3.3-2 Emergency Core Cooling SystemActuation Instrumentation Setpoints

...... 3/4Table 3.3.3-3 Emergency Core Cooling SystemResponse Times ...........................

3/4Table 4.3.3.1-1 Emergency Core Cooling SystemActuation Instrumentation Surveillance Requirements

..............

3/43/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATWS Recirculation Pump Trip System Instrumentation

...........

3/4Table 3.3.4.1-1 ATWS Recirculation Pump TripSystem Instrumentation

.................

3/4Table 3.3.4.1-2 ATWS Recirculation Pump TripSystem Instrumentation Setpoints

..............................

3/4Table 4.3.4.1-1 (Deleted)

..............................

3/4End-of-Cycle Recirculation Pump Trip SystemInstrumentation

...............................................

3/43-323-333-373-393-403-423-433-443-453-46LIMERICK

-UNIT 2viiAmendment No. 147 INDEXLIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTIONINSTRUMENTATION (Continued)

Table 3.3.4.2-1 End-of-Cycle Recirculation PumpTrip System Instrumentation

......Table 3.3.4.2-2 End-of-Cycle Recirculation PumpTrip Setpoints

...................

Table 3.3.4.2-3 End-Of-Cycle Recirculation PumpTrip System Response Time ........Table 4.3.4.2.1-1 (Deleted)

......................

3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION

.........................................

Table 3.3.5-1 Reactor Core Isolation CoolingSystem Actuation Instrumentation...

Table 3.3.5-2 Reactor Core Isolation CoolingSystem Actuation Instrumentation Setpoints

..........................

Table 4.3.5.1-1 (Deleted)

........................

3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION

.......................

Table 3.3.6-1 Control Rod Block Instrumentation..

Table 3.3.6-2 Control Rod Block Instrumentation Setpoints

..........................

Figure 3.3.6-1 SRM Count Rate VersusSignal-to-Noise Ratio ..............

PAGE...... 3/4 3-48...... 3/4 3-49...... 3/4 3-50...... 3/4 3-51...... 3/4 3-52...... 3/4 3-53...... 3/4 3-55...... 3/4 3-56...... 3/4 3-57...... 3/4 3-58...... 3/4 3-60.3/4 3-60bTable 4.3.6-1 Control Rod Block Instrumentation Surveillance Requirements

................

3/4 3-613/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation

..........................

3/4 3-63Table 3.3.7.1-1 Radiation Monitoring Instrumentation

........................

3/4 3-64LIMERICK

-UNIT 2viiiAmendment No. 4, 147 INDEXLIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PAGEINSTRUMENTATION (Continued)

Table 4.3.7.1-1 Radiation Monitoring Instrumentation Surveillance Requirements

.....................

The information from pages 3/4 3-68through 3/4 3-72 has been intentionally omitted.

Refer to note on page 3/4 3-68 ...........

The information from pages 3/4 3-73through 3/4 3-75 has been intentionally omitted.

Refer to note on page 3/4 3-73 ...........

Remote Shutdown System Instrumentation and Controls

......Table 3.3.7.4-1 Remote Shutdown SystemInstrumentation and Controls

.....Table 4.3.7.4-1 (Deleted)

........................

Accident Monitoring Instrumentation

......................

Table 3.3.7.5-1 Accident Monitoring Instrumen-tation ...........................

Table 4.3.7.5-1 Accident Monitoring Instrumenta-tion Surveillance Requirements...

Source Range Monitors

................................

The information from page 3/4 3-89has been intentionally omitted.Refer to note on page ................................

Chlorine Detection System ................................

Toxic Gas Detection System ...............................

DELETED; Refer to note on page ...........................

... 3/4 3-66... 3/4 3-68... 3/4 3-73... 3/4 3-76... 3/4 3-77... 3/4 3-83... 3/4 3-84... 3/4 3-85... 3/4 3-87... 3/4 3-883/43/43/43/43-893-903-913-92LIMERICK

-UNIT 2i xAmendment No. 414, -36, 48, 9, 147 INDEXLIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGEINSTRUMENTATION (Continued)

(Deleted

) ........................................................

3/4 3-97The information from pages 3/4 3-98through 3/4 3-101 has been intentionally omitted.

Refer to note on page 3/4 3-98 .........................

3/4 3-98Offgas Monitoring Instrumentation

................................

3/4 3-103Table 3.3.7.12-1 OffgasMonitoring Instrumentation

.................................

3/4 3-104Table 4.3.7.12-1 OffgasMonitoring Instrumentation Surveillance Requirements

..................................

3/4 3-1073/4.3.8 (Deleted)

The information on pages 3/4 3-110 and3/4 3-111 has been intentionally omitted.Refer to note on page 3/4 3-110 ......................

3/4 3-1103/4.3.9 FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION

...........................................

3/4 3-112Table 3.3.9-1 Feedwater/Main Turbine TripSystem Actuation Instrumentation

...............

3/4 3-113Table 3.3.9-2 Feedwater/Main Turbine TripSystem Actuation Instrumen-tation Setpoints

...............................

3/4 3-114Table 4.3.9.1-1 (Deleted)

......................................

3/4 3-1153/4.4 REACTOR COOLANT SYSTEM3/4.4.1 RECIRCULATION SYSTEMRecirculation Loops ..............................................

3/4 4-1LIMERICK

-UNIT 2Ux Amendment No. 44, 42, 4-14, 147 INDEXLIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGEREACTOR COOLANT SYSTEM (Continued)

Figure 3.4.1.1-1 Deleted ...............................

3/4 4-3Jet Pumps .....................................................

3/4 4-4Recirculation Pumps ...........................................

3/4 4-5Idle Recirculation Loop Startup ...............................

3/4 4-63/4.4.2 SAFETY/RELIEF VALVES ..........................................

3/4 4-73/4.4.3 REACTOR COOLANT SYSTEM LEAKAGELeakage Detection Systems .....................................

3/4 4-8Operational Leakage ...........................................

3/4 4-9Table 3.4.3.2-1 Deleted ................................

3/4 4-113/4.4.4 (Deleted)

The information from pages 3/4 4-12through 3/4 4-14 has been intentionally omitted.Refer to note on page 3/4 4-12 ...............................

3/4 4-123/4.4.5 SPECIFIC ACTIVITY

.............................................

3/4 4-15Table 4.4.5-1 Primary Coolant Specific ActivitySample and Analysis Program ..............

3/4 4-173/4.4.6 PRESSURE/TEMPERATURE LIMITSReactor Coolant System ........................................

3/4 4-18Figure 3.4.6.1-1 Minimum Reactor Pressure VesselMetal Temperature Vs. ReactorVessel Pressure

.......................

3/4 4-20Table 4.4.6.1.3-1 Deleted ..............................

3/4 4-21Reactor Steam Dome ............................................

3/4 4-223/4.4.7 MAIN STEAM LINE ISOLATION VALVES ..............................

3/4 4-233/4.4.8 (DELETED)

.....................................................

3/4 4-24LIMERICK

-UNIT 2xi Amendment No. 4349,4-3-6,4-39,4-44, 160 INDEXLIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGEREACTOR COOLANT SYSTEM (Continued) 3/4.4.9 RESIDUAL HEAT REMOVALHot Shutdown

...............................................

3/4 4-25Cold Shutdown

..............................................

3/4 4-263/4.5 EMERGENCY CORE COOLING SYSTEMS3/4.5.1 ECCS -OPERATING

...........................................

3/4 5-13/4.5.2 ECCS -SHUTDOWN

............................................

3/4 5-63/4.5.3 SUPPRESSION CHAMBER ........................................

3/4 5-83/4.6 CONTAINMENT SYSTEMS3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity

..............................

3/4 6-1Primary Containment Leakage ................................

3/4 6-2Primary Containment Air Lock ...............................

3/4 6-5MSIV Leakage Alternate Drain Pathway .......................

3/4 6-7Primary Containment Structural Integrity

...................

3/4 6-8Drywell and Suppression Chamber Internal Pressure

..........

3/4 6-9Drywell Average Air Temperature

............................

3/4 6-10Drywell and Suppression Chamber Purge System ...............

3/4 6-113/4.6.2 DEPRESSURIZATION SYSTEMSSuppression Chamber ........................................

3/4 6-12Suppression Pool Spray .....................................

3/4 6-15Suppression Pool Cooling ...................................

3/4 6-163/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES .......................

3/4 6-17LIMERICK

-UNIT 2xiiAmendment No. -53, 107 INDEXLIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGECONTAINMENT SYSTEMS (Continued) 3/4.6.4 VACUUM RELIEFSuppression Chamber -Drywell Vacuum Breakers

.....................

3/4 6-443/4.6.5 SECONDARY CONTAINMENT Reactor Enclosure Secondary Containment Integrity

.................

3/4 6-46Refueling Area Secondary Containment Integrity

....................

3/4 6-47Reactor Enclosure Secondary Containment Automatic Isolation Valves ..................................................

3/4 6-48Refueling Area Secondary Containment Automatic Isolation Valves ..................................................

3/4 6-50Standby Gas Treatment System -Common System ......................

3/4 6-52Reactor Enclosure Recirculation System ............................

3/4 6-553/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROLDeleted ...........................................................

3/4 6-57Drywell Hydrogen Mixing System ....................................

3/4 6-58Drywell and Suppression Chamber Oxygen Concentration

..............

3/4 6-593/4.7 PLANT SYSTEMS3/4.7.1 SERVICE WATER SYSTEMSResidual Heat Removal Service Water System -CommonSystem ............................................................

3/4 7-1Emergency Service Water System -Common System ....................

3/4 7-3Ultimate Heat Sink ................................................

3/4 7-5LIMERICK

-UNIT 2xiiiAmendment No. 49, 135 INDEXLIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGEPLANT SYSTEMS (Continued) 3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM -COMMONSYSTEM ........................................................

3/4 7-63/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM .........................

3/4 7-93/4 .7.4 SNUBBERS

......................................................

3/4 7-11Figure 4.7.4-1 Sample Plan 2) For SnubberFunctional Test ...........................

3/4 7-163/4.7.5 SEALED SOURCE CONTAMINATION

...................................

3/4 7-173/4.7.6 DELETED; Refer to note on page ................................

3/4 7-193/4.7.7 DELETED; Refer to note. on page ................................

3/4 7-193/4.7.8 MAIN TURBINE BYPASS SYSTEM ....................................

3/4 7-333/4.8 ELECTRICAL POWER SYSTEMS3/4.8.1 A.C. SOURCESA.C. Sources -Operating

...................................

3/4 8-1Table 4.8.1.1.2-1 DELETED .................................

3/4 8-8A.C. Sources -Shutdown

....................................

3/4 8-93/4.8.2 D.C. SOURCESD.C. Sources -Operating

...................................

3/4 8-1010LIMERICK

-UNIT 2xivAmendment No. 6, 48, 150 INDEXLIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGEELECTRICAL POWER SYSTEMS (Continued)

Table 4.8.2.1-1 Battery Surveillance Requirements

.......................

3/4 8-13D.C. Sources -Shutdown

....................................

3/4 8-143/4.8.3 ONSITE POWER DISTRIBUTION SYSTEMSDistribution

-Operating

...................................

3/4 8-15Distribution

-Shutdown

....................................

3/4 8-183/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES(Deleted

) ..................................................

3/4 8-21(Deleted

) ..................................................

3/4 8-27Reactor Protection System Electric Power Monitoring

........

3/4 8-283/4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCH ........................................

3/4 9-13/4.9.2 INSTRUMENTATION

............................................

3/4 9-33/4.9.3 CONTROL ROD POSITION

.......................................

3/4 9-53/4.9.4 DECAY TIME .................................................

3/4 9-63/4.9.5 COMMUNICATIONS

.............................................

3/4 9-73/4.9.6 REFUELING PLATFORM

.........................................

3/4 9-83/4.9.7 CRANE TRAVEL -SPENT FUEL STORAGE POOL .....................

3/4 9-103/4.9.8 WATER LEVEL -REACTOR VESSEL ...............................

3/4 9-113/4.9.9 WATER LEVEL -SPENT FUEL STORAGE POOL ......................

3/4 9-12LIMERICK

-UNIT 2XVAmendment No. 1-ý, 170 INDEXLIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGEREFUELING OPERATIONS (Continued) 3/4.9.10 CONTROL ROD REMOVALSingle Control Rod Removal .................................

3/4 9-13Multiple Control Rod Removal ...............................

3/4 9-153/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level ...........................................

3/4 9-17Low Water Level ............................................

3/4 9-183/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY

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3/4 10-13/4.10.2 ROD WORTH MINIMIZER

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3/4 10-23/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS

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3/4 10-33/4.10.4 RECIRCULATION LOOPS ........................................

3/4 10-43/4.10.5 OXYGEN CONCENTRATION

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3/4 10-53/4.10.6 TRAINING STARTUPS

..........................................

3/4 10-63/4.10.7 SPECIAL INSTRUMENTATION

-INITIAL CORE LOADING .............

3/4 10-73/4.10.8 INSERVICE LEAK AND HYDROSTATIC TESTING .....................

3/4 10-93/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS The information from pages 3/4 11-1through 3/4 11-6 has been intentionally omitted.

Refer to note on page 3/4 11-1 ...................

3/4 11-1Liquid Holdup Tanks ........................................

3/4 11-73/4.11.2 GASEOUS EFFLUENTS The information from pages 3/4 11-8through 3/4 11-14 has been intentionally omitted.

Refer to note on page 3/4 11-8 ...................

3/4 11-8LIMERICK

-UNIT 2xviAmendment No. -14, 95 INDEXLIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGERADIOACTIVE EFFLUENTS (Continued)

Explosive Gas Mixture ........................................

3/4 11-15Main Condenser

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3/4 11-16The information on page 3/4 11-17 has beenintentionally omitted.

Refer to note on thispage ...................................................

3/4 11-173/4.11.3 (Deleted)

The information on pages 3/4 11-18 through3/4 11-20 has been intentionally omitted.Refer to note on page 3/4 11-18.3/4 .11.4 (Deleted)

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3/4 11-183/4.12 (Deleted)

The information on pages 3/4 12-1 through3/4 12-14 has been intentionally omitted.Refer to note on page 3/4 12-1 .....................

3/4 12-1LIMERICK

-UNIT 2xvi iAmendment No. 11 INDEXBASESSECTION PAGE3/4.0 APPLICABILITY

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B 3/4 0-13/4.1 REACTIVITY CONTROL SYSTEMS3/4.1.1 SHUTDOWN MARGIN ..............................................

B 3/4 1-13/4.1.2 REACTIVITY ANOMALIES

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B 3/4 1-13/4.1.3 CONTROL RODS .................................................

B 3/4 1-23/4.1.4 CONTROL ROD PROGRAM CONTROLS

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B 3/4 1-33/4.1.5 STANDBY LIQUID CONTROL SYSTEM ................................

B 3/4 1-43/4.2 POWER DISTRIBUTION LIMITS3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE .........................................................

B 3/4 2 -13/4.2.2 (DELETED)

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B 3/4 2-2LEFT INTENTIONALLY BLANK ................................................

B 3/4 2-33/4.2.3 MINIMUM CRITICAL POWER RATIO .................................

B 3/4 2-43/4.2.4 LINEAR HEAT GENERATION RATE ..................................

B 3/4 2-53/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION

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B 3/4 3-13/4.3.2 ISOLATION ACTUATION INSTRUMENTATION

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B 3/4 3-23/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION

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B 3/4 3-23/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION

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B 3/4 3-33/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION

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B 3/4 3-43/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION

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B 3/4 3-43/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation

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B 3/4 3-50LIMERICK

-UNIT 2xviiiAmendment No. 4, 321, 48 INDEXBASESSECTION PAGEINSTRUMENTATION (Continued)

(Deleted

) ..................................................

B 3/4 3-5(Deleted)

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

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B 3/4 3-5Accident Monitoring Instrumentation

.B 3/4 3-5Source Range Monitors

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(Deleted

) .........................................

Chlorine and Toxic Gas Detection Systems ..........

(Deleted

) .........................................

(Deleted

) .........................................

(Deleted

) .........................................

Offgas Monitoring Instrumentation

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(Deleted

) .........................................

FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION

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Bases Figure B 3/4.3-1 Reactor Vessel WaterLevel ...............

3/43/43/43/43/43/43/43/43-53-63-63-63-73-73-73-73/4.3.83/4.3.9.........

B 3/4 3-7B 3/4 3-83/4.4 REACTOR COOLANT SYSTEM3/4.4.1 RECIRCULATION SYSTEM ..............

3/4.4.2 SAFETY/RELIEF VALVES ..............

3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGELeakage Detection Systems .........

Operational Leakage ...............

3/4.4.4 CHEMISTRY

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

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BBB3/43/43/44-34-34-3aLIMERICK

-UNIT 2xixAmendment No. -14, 42, 44, 3,ý&, 69, -9, 42, 117 INDEXBASESSECTION PAGEREACTOR COOLANT SYSTEM (Continued) 3/4.4.5 SPECIFIC ACTIVITY

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B 3/4 4-43/4.4.6 PRESSURE/TEMPERATURE LIMITS ...................................

B 3/4 4-4Bases Table B 3/4.4.6-1 Reactor VesselToughness

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B 3/4 4-7Bases Figure B 3/4.4.6-1 Fast Neutron Fluence(E>1 MeV) At 1/4 T As AFunction of ServiceLife ...........................

B 3/4 4-83/4.4.7 MAIN STEAM LINE ISOLATION VALVES ..............................

B 3/4 4-63/4.4.8 (DELETED)

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B 3/4 4-63/4.4.9 RESIDUAL HEAT REMOVAL .........................................

B 3/4 4-63/4.5 EMERGENCY CORE COOLING SYSTEMS3/4.5.1 and 3/4.5.2 ECCS -OPERATING and SHUTDOWN

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B 3/4 5-13/4.5.3 SUPPRESSION CHAMBER .....................................

B 3/4 5-23/4.6 CONTAINMENT SYSTEMS3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity

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B 3/4 6-1Primary Containment Leakage .............................

B 3/4 6-1Primary Containment Air Lock ............................

B 3/4 6-1MSIV Leakage Control System .............................

B 3/4 6-1Primary Containment Structural Integrity

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B 3/4 6-2Drywell and Suppression Chamber InternalPressure

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B 3/4 6-2Drywell Average Air Temperature

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B 3/4 6-2Drywell and Suppression Chamber Purge System ............

B 3/4 6-23/4.6.2 DEPRESSURIZATION SYSTEMS ................................

B 3/4 6-3LIMERICK

-UNIT 2XXAmendment No. 160 INDEXBASESSECTION PAGECONTAINMENT SYSTEMS (Continued) 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES .................

B 3/4 6-43/4.6.4 VACUUM RELIEF ........................................

B 3/4 6-43/4.6.5 SECONDARY CONTAINMENT

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B 3/4 6-53/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL ...............

B 3/4 6-63/4.7 PLANT SYSTEMS3/4.7.1 SERVICE WATER SYSTEMS -COMMON SYSTEMS ...............

3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM -COMMON SYSTEM ........................................

3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM ................

3/4.7.4 SNUBBERS

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3/4.7.5 SEALED SOURCE CONTAMINATION

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3/4.7.6 (Deleted

) ............................................

3/4.7.7 (Deleted)

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3/4.7.8 MAIN TURBINE BYPASS SYSTEM ...........................

3/4.8 ELECTRICAL POWER SYSTEM3/4.8.1, 3/4.8.2, and3/4.8.3 A.C. SOURCES, D.C. SOURCES, AND ONSITE POWERDISTRIBUTION SYSTEMS .....................................

3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES ..............

3/4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCH ..................................

3/4.9.2 INSTRUMENTATION

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3/4.9.3 CONTROL ROD POSITION

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3/4.9.4 DECAY TIME ...........................................

3/4.9.5 COMMUNICATIONS

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B 3/4 7-1BBBBBBB3/43/43/43/43/43/43/47-17-ic7-27-37-47-47-5B 3/4 8-1B 3/4 8-3BBBBB3/43/43/43/43/49-19-19-19-19-1LIMERICK

-UNIT 2xxiAmendment No. 6, 49, 149 INDEXBASESSECTION PAGEREFUELING OPERATIONS (Continued) 3/4.9.6 REFUELING PLATFORM

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B 3/4 9-23/4.9.7 CRANE TRAVEL -SPENT FUEL STORAGE POOL .....................

B 3/4 9-23/4.9.8 and 3/4.9.9 WATER LEVEL -REACTOR VESSELAND WATER LEVEL -SPENT FUEL STORAGE POOL ..................

B 3/4 9-23/4.9.10 CONTROL ROD REMOVAL ........................................

B 3/4 9-23/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION

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B 3/4 9-23/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY

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B 3/4 10-13/4.10.2 ROD WORTH MINIMIZER

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B 3/4 10-13/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS

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B 3/4 10-13/4.10.4 RECIRCULATION LOOPS ........................................

B 3/4 10-13/4.10.5 OXYGEN CONCENTRATION

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B 3/4 10-13/4.10.6 TRAINING STARTUPS

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B 3/4 10-13/4.10.7 SPECIAL INSTRUMENTATION

-INITIAL CORE LOADING .............

B 3/4 10-13/4.10.8 INSERVICE LEAK AND HYDROSTATIC TESTING .....................

B 3/4 10-23/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS The information on page B 3/4 11-1 has beenintentionally omitted.

Refer to note on thispage .......................................................

B 3/4 11-1(Deleted

) ..................................................

B 3/4 11-2Liquid Holdup Tanks ........................................

B 3/4 11-23/4.11.2 GASEOUS EFFLUENTS (Deleted

) ..................................................

B 3/4 11-2The information on page B 3/4 11-3 has beenintentionally omitted.

Refer to note on thispage ......................................................

B 3/4 11-3(Deleted)

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B 3/4 11-4LIMERICK

-UNIT 2 xxii Amendment No. -14, 95 INDEXBASESSECTION PAGERADIOACTIVE EFFLUENTS (Continued)

Explosive Gas Mixture ...................................

B 3/4 11-4Main Condenser

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B 3/4 11-5(Deleted)

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B 3/4 11-53/4.11.3 (Deleted)

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B 3/4 11-53/4.11.4 (Deleted)

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B 3/4 11-53/4.12 (Deleted)

The information from pages B 3/4 12-1through B 3/4 12-2 has been intentionally omitted.

Refer to note on pageB 3/4 12-1 ...................................

B 3/4 12-1LIMERICK

-UNIT 2xxiiiAmendment No. 11 INDEXDESIGN FEATURESSECTION PAGE5.1 SITEExclusion Area ......................................................

5-1Figure 5.1.1-1 Exclusion Area ...........................

5-2Low Population Zone ....... .........................................

5-1Figure 5.1.2-1 Low Population Zone ......................

5-3Maps Defining UNRESTRICTED AREAS and SITE BOUNDARYfor Radioactive Gaseous and Liquid Effluents

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5-1Figure 5.1.3-1a Map Defining UNRESTRICTED AREAS forRadioactive Gaseous and Liquid Effluents

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5-4Figure 5.1.3-1b Map Defining UNRESTRICTED AREAS forRadioactive Gaseous and Liquid Effluents

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5-5(De leted ) ...........................................................

5 -1The figure on page 5-6 has been intentionally omitted.

Refer to note on page 5-6 ......................

5-65.2 CONTAINMENT Configuration

................................

Design Temperature and Pressure

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Secondary Containment

........................

5.3 REACTOR COREFuel Assemblies

..............................

Control Rod Assemblies

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5.4 REACTOR COOLANT SYSTEMDesign Pressure and Temperature

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Volume .......................................

5.5 FUEL STORAGE...................

...................

5-15-15-75-75-75-75-8Criticality

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5-8LIMERICK

-UNIT 2xxi vAmendment No. 11 1 INDEXDESIGN FEATURESSECTION PAGEFUEL STORAGE (Continued)

Drainage

..........................................................

5-8Capacity

..........................................................

5-85.6 COMPONENT CYCLIC OR TRANSIENT LIMIT ................................

5-8Table 5.6.1-1 Component Cyclic or Transient Limits ..........

5-9LIMERICK

-UNIT 2XXV INDEXADMINISTRATIVE CONTROLSSECTION PAGE6.1 RESPONSIBILITY

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6-16.2 ORGANIZATION

.......................................................

6-16.2.1 OFFSITE AND ONSITE ORGANIZATION

...................................

6-1Figure 6.2.1-1 DELETED ............................

6-36 .2 .2 UNIT STAFF ........................................................

6-2Figure 6.2.2-1 DELETED ............................

6-4Table 6.2.2-1 Minimum Shift CrewComposition

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6-56.2.3 DELETED; Refer to note on page ....................................

6-66.2.4 SHIFT TECHNICAL ADVISOR ...........................................

6-66.3 UNIT STAFF QUALIFICATIONS

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6-66 .4 DELETED ............................................................

6-76.5 DELETED6.5.1 Deleted .Deleted ....................................................

6-7Deleted ....................................................

6 -7Deleted ....................................................

6-7Deleted ....................................................

6 -7Deleted ....................................................

6-8Deleted ....................................................

6 -9LIMERICK

-UNIT 2xxviAmendment No. -, 60, 4-22, 138 INDEXADMINISTRATIVE CONTROLSSECTION6.5.2 DeletedDeleted .........

Deleted .........

6.5.3 Deleted .........

6.6 REPORTABLE EVENT ACTION.6.7 SAFETY LIMIT VIOLATION..

6.8 PROCEDURES AND PROGRAMS.

6.9 REPORTING REQUIREMENTS

..................

...... I ...........

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PAGE6-96-96-106-106-106-106-106-126-126-12a6-12a6-136.9.1ROUTINE REPORTS.... ... ... .. ... ... ... ... ..... .. ... ... .... ..6 -15Startup Report .......................................

Annual Reports .......................................

Monthly Operating Reports ............................

Annual Radiological Environmental Operating Report ...Annual Radioactive Effluent Release Report ...........

CORE OPERATING LIMITS REPORTS ........................

6.9.2 SPECIAL REPORTS ......................................

6.10 DELETED .....................................................

6.11 RADIATION PROTECTION PROGRAM ................................

6.12 HIGH RADIATION AREA .........................................

6-156-156-166-166-176-18a6-18a6-196-206-20LIMERICK

-UNIT 2xxvi iAmendment No. 4, 49, 498, 153 INDEXADMINISTRATIVE CONTROLSSECTION PAGE6.13 PROCESS CONTROL PROGRAM (PCP) .....................................

6-216.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) ............................

6-226.15 (Deleted

) .........................................................

6-226.16 CONTROL ROOM ENVELOPE HABITABILITY PROGRAM ........................

6-22LIMERICK

-UNIT 2xxvii iAmendment No. -14, 149 SECTION 1.0DEFINITIONS THIS PAGE INTENTIONALLY LEFT BLANK

1.0 DEFINITIONS

The following terms are defined so that uniform interpretation of thesespecifications may be achieved.

The defined terms appear in capitalized typeand shall be applicable throughout these Technical Specifications.

ACTION1.1 ACTION shall be that part of a Specification which prescribes remedialmeasures required under designated conditions.

AVERAGE PLANAR EXPOSURE1.2 The AVERAGE PLANAR EXPOSURE shall be applicable to a specific planar heightand is equal to the sum of the exposure of all the fuel rods in thespecified bundle at the specified height divided by the number of fuelrods in the fuel bundle.AVERAGE PLANAR LINEAR HEAT GENERATION RATE1.3 The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicable to a specific planar height and is equal to the sum of the LINEAR HEATGENERATION RATES for all the fuel rods in the specified bundle at thespecified height divided by the number of fuel rods in the fuel bundle.CHANNEL CALIBRATION 1.4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channeloutput such that it responds with the necessary range and accuracy to knownvalues of the parameter which the channel monitors.

The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/ortrip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNELCALIBRATION may be performed by any series of sequential, overlapping ortotal channel steps such that the entire channel is calibrated.

CHANNEL CHECK1.5 A CHANNEL CHECK shall be the qualitative assessment of channel behaviorduring operation by observation.

This determination shall include, wherepossible, comparison of the channel indication and/or status with otherindications and/or status derived from independent instrument channelsmeasuring the same parameter.

CHANNEL FUNCTIONAL TEST1.6 A CHANNEL FUNCTIONAL TEST shall be:a. Analog channels

-the injection of a simulated signal into the channelas close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips.b. Bistable channels

-the injection of a simulated signal into the sensorto verify OPERABILITY including alarm and/or trip functions.

The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is tested.LIMERICK

-UNIT 21-1 DEFINITIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removedand fuel in the vessel. The following exceptions are not considered to beCORE ALTERATIONS:

a) Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or specialmoveable detectors (including undervessel replacement);

andb) Control rod movement, provided there are no fuel assemblies in theassociated core cell.Suspension of CORE ALTERATIONS shall not preclude completion of movementof a component to a safe position.

CORE OPERATING LIMITS REPORT1.7a The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific documentthat provides the core operating limits for the current operating reloadcycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specifications 6.9.1.9 thru6.9.12. Plant operation within these limits is addressed in individual specifications.

CRITICAL POWER RATIO1.8 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in theassembly which is calculated by application of the (GEXL) correlation tocause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.DOSE EQUIVALENT 1-1311.9 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131, microcuries pergram, which alone would produce the same inhalation committed effective dose equivalent (CEDE) as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present.

The inhalation committed effective dose equivalent (CEDE) conversion factors used for thiscalculation shall be those listed in Table 2.1 of Federal Guidelines Report11, "Limiting Values of Radionuclide Intake and Air Concentration and DoseConversion Factors for Inhalation, Submersion, and Ingestion,"

ORNL, 1989,as described in Regulatory Guide 1.183. The factors in the column headed"effective" yield doses corresponding to the CEDE.DOWNSCALE TRIP SETPOINT (DTSP)1.9a The downscale trip setpoint associated with the Rod Block Monitor (RBM)rod block trip setting.1.10 (Deleted)

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME1.11 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that timeinterval from when the monitored parameter exceeds its ECCS actuation set-point at the channel sensor until the ECCS equipment is capable ofperforming its safety function, i.e., the valves travel to their requiredpositions, pump discharge pressures reach their required values, etc.Times shall include diesel generator starting and sequence loading delayswhere applicable.

The response time may be measured by any series ofsequential, overlapping or total steps such that the entire response timeis measured.

LIMERICK

-UNIT 21-2 Amendment No. 4, 4-8, 4-9, 4-46, 146 DEFINITIONS END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME1.12 The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall bethat time interval to complete suppression of the electric arc betweenthe fully open contacts of the recirculation pump circuit breaker frominitial movement of the associated:

a. Turbine stop valves, andb. Turbine control valves.This total system response time consists of two components, t.he instrumen-tation response time and the breaker arc suppression time. These timesmay be measured by any series of sequential, overlapping or total stepssuch that the entire response time is measured.

1.13 (Deleted) 1.14 (Deleted)

FREQUENCY NOTATION1.15 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.HIGH (POWER) TRIP SETPOINT (HTSP)1.15a The high power trip setpoint associated with the Rod Block Monitor (RBM)rod block trip setting applicable above 85% reactor thermal power.IDENTIFIED LEAKAGE1.16 IDENTIFIED LEAKAGE shall be:a. Leakage into collection

systems, such as pump seal or valve packingleaks, that is captured and conducted to a sump or collecting tank, orb. Leakage into the containment atmosphere from sources that are bothspecifically located and known either not to interfere with the operation of the leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE.INTERMEDIATE (POWER) TRIP SETPOINT (ITSP)1.16a The intermediate power trip setpoint associated with the Rod Block Monitor (RBM)rod block trip setting applicable between 65% and 85% reactor thermal power.ISOLATION SYSTEM RESPONSE TIME1.17 The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from whenthe monitored parameter exceeds its isolation actuation setpoint at thechannel sensor until the isolation valves travel to their required positions.

Times shall include diesel generator starting and sequence loading delayswhere applicable.

The response time may be measured by any series ofsequential, overlapping or total steps such that the entire response timeis measured.

LIMITING CONTROL ROD PATTERN1.18 A LIMITING CONTROL ROD PATTERN shall be a pattern which results in thecore being on a thermal hydraulic limit, i.e., operating on a limitingvalue for APLHGR, LHGR, or MCPR.LINEAR HEAT GENERATION RATE1.19 LINEAR HEAT GENERATION RATE (LHGR) shall be the heat generation per unitlength of fuel rod. It is the integral of the heat flux over the heattransfer area associated with the unit length.LIMERICK

-UNIT 21-3Amendment No. 48 DEFINITIONS LOGIC SYSTEM FUNCTIONAL TEST1.20 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, i.e., all relays and contacts, all trip units, solid state logic elements, etc, of a logic circuit, from sensor through and including the actuateddevice, to verify OPERABILITY.

The LOGIC SYSTEM FUNCTIONAL TEST may beperformed by any series of sequential, overlapping or total system stepssuch that the entire logic system is tested.LOW (POWER) TRIP SETPOINT (LTSP)1.20a The low power trip setpoint associated with the Rod Block Monitor (RBM)rod block trip setting applicable between 30% and 65% reactor thermal power.1.21 (Deleted)

MEMBER(S)

OF THE PUBLIC1.22 MEMBER OF THE PUBLIC means any individual except when that individual isreceiving an occupational dose.MAPFAC(F)-(MAPLHGR FLOW FACTOR)1.22a A core flow dependent multiplication factor used to flow bias the standardMaximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit.MAPFAC(P)-(POWER DEPENDENT MAPLHGR MULTIPLIER) 1.22b A core power dependent multiplication factor used to power bias the standardMaximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit.MINIMUM CRITICAL POWER RATIO (MCPR)1.23 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR whichexists in the core (for each class of fuel). Associated with the minimumcritical power ratio is a core flow dependent (MCPR(F))

and core powerdependent (MCPR(P))

minimum critical power ratio.OFFSITE DOSE CALCULATION MANUAL1.24 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting fromradioactive gaseous and liquid effluents, in the calculation of gaseousand liquid effluent monitoring alarm/trip setpoints, and in the conductof the Radiological Environmental Monitoring Program.

The ODCM shallalso contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2)descriptions of the information that should be included in the AnnualRadiological Environmental Operating and Annual Radioactive EffluentRelease Reports required by Specifications 6.9.1.7 and 6.9.1.8.OPERABLE

-OPERABILITY 1.25 A system, subsystem, train, component or device shall be OPERABLE or haveOPERABILITY when it is capable of performing its specified function(s) andwhen all necessary attendant instrumentation,

controls, electrical power,cooling or seal water, lubrication or other auxiliary equipment that arerequired for the system, subsystem, train, component, or device to perform itsfunction(s) are also capable of performing their related support function(s).

LIMERICK

-UNIT 21-4Amendment No. 4,45,49,448, 153 DEFINITIONS OPERATIONAL CONDITION

-CONDITION 1.26 An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive combination of mode switch position and average reactor coolant tempera-ture as specified in Table 1.2.PHYSICS TESTS1.27 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and (1) described in Chapter 14 of the FSAR, (2) authorized under theprovisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE1.28 PRESSURE BOUNDARY LEAKAGE shall be leakage through a nonisolable faultin a reactor coolant system component body, pipe wall or vessel wall.PRIMARY CONTAINMENT INTEGRITY 1.29 PRIMARY CONTAINMENT INTEGRITY shall exist when:a. All primary containment penetrations required to be closed duringaccident conditions are either:1. Capable of being closed by an OPERABLE primary containment automatic isolation system, or2. Closed by at least one manual valve, blind flange, ordeactivated automatic valve secured in. its closed position, except for valves that are opened under administrative controlas permitted by Specification 3.6.3.b. All primary containment equipment hatches are closed and sealed.c. The primary containment air lock is in compliance with therequirements of Specification 3.6.1.3.d. The primary containment leakage rates are within the limits ofSpecification 3.6.1.2.e. The suppression chamber is in compliance with the requirements of Specification 3.6.2.1.f. The sealing mechanism associated with each primary containment penetration; e.g., welds, bellows, or O-rings, is OPERABLE.

PROCESS CONTROL PROGRAM1.30 The PROCESS CONTROL PROGRAM (PCP) shall contain the provisions to assurethat the SOLIDIFICATION or dewatering and packaging of radioactive wastesresults in a waste package with properties that meet the minimum andstability requirements of 10 CFR Part 61 and other requirements for trans-portation to the disposal site and receipt at the disposal site. WithSOLIDIFICATION or dewatering, the PCP shall identify the processparameters influencing SOLIDIFICATION or dewatering based on laboratory scale and full scale testing or experience.

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-UNIT 21-5Amendment No. -14, 48, 107 DEFINITIONS PURGE -PURGING1.31 PURGE or PURGING shall be the controlled process of discharging air orgas from a confinement to maintain temperature,

pressure, humidity, concentration or other operating condition, in such a manner thatreplacement air or gas is required to purify the confinement.

RATED THERMAL POWER1.32 RATED THERMAL POWER shall be a total reactor core heat transfer rate tothe reactor coolant of 3515 MWt.REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall exist when:a. All reactor enclosure secondary containment penetrations required tobe closed during accident conditions are either:1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or2. Closed by at least one manual valve, blind flange, slide gatedamper or deactivated automatic valve secured in its closedposition, except as provided by Specification 3.6.5.2.1.

b. All reactor enclosure secondary containment hatches and blowout panelsare closed and sealed.c. The standby gas treatment system is in compliance with the requirements of Specification 3.6.5.3.d. The reactor enclosure recirculation system is in compliance with therequirements of Specification 3.6.5.4.e. At least one door in each access to the reactor enclosure secondary containment is closed.f. The sealing mechanism associated with each reactor enclosure secondary containment penetration, e.g., welds, bellows, or O-rings, is OPERABLE.

g. The pressure within the reactor enclosure secondary containment isless than or equal to the value required by Specification 4.6.5.1.1a.

REACTOR PROTECTION SYSTEM RESPONSE TIME1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval fromwhen the monitored parameter exceeds its trip setpoint at the channelsensor until de-energization of the scram pilot valve solenoids.

Theresponse time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

RECENTLY IRRADIATED FUEL1.35 RECENTLY IRRADIATED FUEL is fuel that has occupied part of a critical reactor corewithin the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY 1.36 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY shall exist when:a. All refueling floor secondary containment penetrations required tobe closed during accident conditions are either:LIMERICK

-UNIT 21-6Amendment No. 49,-51,f64,4-14, 163 DEFINITIONS REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY (Continued)

1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or2. Closed by at least one manual valve, blind flange, slide gatedamper or deactivated automatic valve secured in its closedposition, except as provided by Specification 3.6.5.2.2.

b. All refueling floor secondary containment hatches and blowout panels areclosed and sealed.c. The standby gas treatment system is in compliance with the requirements of Specification 3.6.5.3.d. At least one door in each access to the refueling floor secondary containment is closed.e. The sealing mechanism associated with each refueling floor secondary containment penetration, e.g., welds, bellows, or 0-rings, is OPERABLE.

f. The pressure within the refueling floor secondary containment is lessthan or equal to the value required by Specification 4.6.5.1.2a.

REPORTABLE EVENT1.37 A REPORTABLE EVENT shall be any of those conditions specified in Section50.73 to 10 CFR Part 50.RESTRICTED AREA1.37a RESTRICTED AREA means an area, access to which is limited by the licensee forthe purpose of protecting individuals against undue risks from exposure toradiation and radioactive materials.

RESTRICTED AREA does not include areasused as residential

quarters, but separate rooms in a residential building maybe set apart as a RESTRICTED AREA.1.38 (Deleted)

SHUTDOWN MARGIN1.39 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor issubcritical or would be subcritical assuming all control rods are fullyinserted except for the single control rod of highest reactivity worth whichis assumed to be fully withdrawn and the reactor is in the shutdowncondition; cold, i.e. 68°F; and xenon free.SITE BOUNDARY1.40 The SITE BOUNDARY shall be that line as defined in Figure 5.1.3-1a.

SOURCE CHECK1.41 A SOURCE CHECK shall be the qualitative assessment of channel response whenthe channel sensor is exposed to a radioactive source.LIMERICK

-UNIT 21-7Amendment No. 14,48,49,-,--8, 168 DEFINITIONS STAGGERED TEST BASIS1.42 A STAGGERED TEST BASIS shall consist of:a. A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into nequal subintervals.

b. The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

THERMAL POWER1.43 THERMAL POWER shall be the total reactor core heat transfer rate to thereactor coolant.TURBINE BYPASS SYSTEM RESPONSE TIME1.43A The TURBINE BYPASS SYSTEM RESPONSE TIME shall be that time interval fromwhen the turbine bypass control unit generates a turbine bypass valve flowsignal until the turbine bypass valves travel to their required position.

The response time may be measured by any series of sequential, overlapping, or total steps such that the entire response time is measured.

UNIDENTIFIED LEAKAGE1.44 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.UNRESTRICTED AREA1.45 UNRESTRICTED AREA means an area, access to which is neither limited norcontrolled by the licensee.

VENTILATION EXHAUST TREATMENT SYSTEM1.46 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed andinstalled to reduce gaseous radioiodine or radioactive material inparticulate form in effluents by passing ventilation or vent exhaust gasesthrough charcoal adsorbers and/or HEPA filters for the purpose of removingiodines or particulates from the gaseous exhaust stream prior to therelease to the environment (such a system is not considered to have anyeffect on noble gas effluents).

Engineered Safety Feature (ESF)atmospheric cleanup systems are not considered to be VENTILATION EXHAUSTTREATMENT SYSTEM components.

VENTING1.47 VENTING shall be the controlled process of discharging air or gas from aconfinement to maintain temperature,

pressure, humidity, concentration orother operating condition, in such a manner that replacement air or gas isnot provided or required during VENTING.

Vent, used in system names, doesnot imply a VENTING process.LIMERICK

-UNIT 21-8Amendment No. 4-6, 148 DEFINITIONS TABLE 1.1SURVEILLANCE FREQUENCY NOTATIONNOTATIONSDwMQSAAER (Refueling Interval)

S/UPN.A.FREQUENCY At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .At least once per 7 days.At least once per 31 days.At least once per 92 days.At least once per 184 days.At least once per 366 days.At least once per 18 months (550 days).At least once per 24 months (731 days).Prior to each reactor startup.Prior to each radioactive release.Not applicable.

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-UNIT 21-9Amendment No. 34 DEFINITIONS TABLE 1.2OPERATIONAL CONDITIONS CONDITION MODE SWITCHPOSITIONAVERAGE REACTORCOOLANT TEMPERATURE Any temperature

1. POWER OPERATION Run2. STARTUP3. HOT SHUTDOWN4. COLD SHUTDOWN5. REFUELING*

Startup/Hot StandbyShutdown#

Shutdown###*

Shutdown or Refuel**

Any temperature

> 200°F 200°F ****NA#The reactor mode switch mayposition to test the switchrods are verified to remainother technically qualified be placed in the Run or Startup/Hot Standbyinterlock functions provided that the controlfully inserted by a second licensed operator ormember of the unit technical staff.##The reactor mode switch may be placed in the Refuel position while a singlecontrol rod drive is being removed from the reactor pressure vessel perSpecification 3.9.10.1.

Fuel in the reactor vessel with the vessel head closure bolts less thanfully tensioned or with the head removed.**See Special Test Exceptions 3.10.1 and 3.10.3.***The reactor mode switch may be placed in the Refuel position while a singlecontrol rod is being moved provided that the one-rod-out interlock isOPERABLE.

- - - See Special Test Exception 3.10.8.LIMERICK

-UNIT 21-10Amendment No. 5O, 4-, 112 SECTION 2.0SAFETY LIMITSANDLIMITING SAFETY SYSTEM SETTINGS THIS PAGE INTENTIONALLY LEFT BLANK 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS2.1 SAFETY LIMITSTHERMAL POWER. Low Pressure or Low Flow2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactorvessel steam dome pressure less than 785 psig or core flow less than 10% of ratedflow.APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.ACTION:With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vesselsteam dome pressure less than 785 psig or core flow less than 10% of rated flow,be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements ofSpecification 6.7.1.THERMAL POWER. Hiah Pressure and Hiqh Flow2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.09 for tworecirculation loop operation and shall not be less than 1.12 for singlerecirculation loop operation with the reactor vessel steam dome pressure greaterthan 785 psig and core flow greater than 10% of rated flow.APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.ACTION:With MCPR less than 1.09 for two recirculation loop operation or less than 1.12for single recirculation loop operation and the reactor vessel steam dome pressuregreater than 785 psig and core flow greater than 10% of rated flow, be in at leastHOT SHUTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.7.1.REACTOR COOLANT SYSTEM PRESSURE2.1.3 The reactor coolant system pressure, as measured in the reactor vesselsteam dome, shall not exceed 1325 psig.APPLICABILITY:

OPERATION CONDITIONS 1, 2, 3, and 4.ACTION:With the reactor coolant system pressure, as measured in the reactor vessel steamdome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant systempressure less than or equal to 1325 psig within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with tilerequirements of Specification 6.7.1.LIMERICK

-UNIT 22-1Amendment No. 4-4, 8-, 9-, 9-, 44-4,12-7, 162 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGSSAFETY LIMITS (Continued)

REACTOR VESSEL WATER LEVEL2.1.4 The reactor vessel water level shall be above the top of theactive irradiated fuel.APPLICABILITY:

OPERATIONAL CONDITIONS 3, 4, and 5.ACTION:With the reactor vessel water level at or below the top of the activeirradiated fuel, manually initiate the ECCS to restore the water level,after depressurizing the reactor vessel, if required.

Comply with therequirements of Specification 6.7.1.LIMERICK

-UNIT 22-2 S-AFETY L IMITS AND LIMITING SAFETY SYSTEM SETTINGS2.2 ITMTTING SAFETY SYSTEM SETTINGSREACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor protection system instrumentation setpoints shall be setconsistent with the Trip Setpoint values shown in Table 2.2.1-1.APPLICABILITY:

As shown in Table 3.3.1-1.ACTION:With a reactor protection system instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2.1-1, declarethe channel inoperable*

and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status withits setpoint adjusted consistent with the Trip Setpoint value.*The APRM Simulated Thermal Power -Upscale Functional Unit need not be declaredinoperable upon entering single reactor recirculation loop operation provided thatthe flow-biased setpoints are adjusted within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months per Specification 3.4.1.1.LIMERICK

-UNIT 22-3Amendment No. 109 TABLE 2.2.1-1REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS FUNCTIONAL UNITTRIP SETPOINT1. Intermediate Range Monitor, Neutron Flux-High

2. Average Power Range Monitor:a. Neutron Flux-Upscale (Setdown)

b. Simulated Thermal Power -Upscale:-Two Recirculation Loop Operation

-Single Recirculation Loop Operation***

c. Neutron Flux -Upscale120/125 divisions of full scale15.0% of RATED THERMALPOWER0.65 W + 61.7% and116.6% of RATEDTHERMAL POWER< 0.65 (W-7.6%)

+ 61.5% and 116.6% of RATEDTHERMAL POWER118.3% of RATEDTHERMAL POWERALLOWABLE VALUES122/125 divisions of full scale< 20.0% of RATEDTHERMAL POWER0.65 W + 62.2% and117.0% of RATEDTHERMAL POWER0.65 (W-7.6%)

+ 62.0% and117.0% of RATEDTHERMAL POWER118.7% of RATEDTHERMAL POWERd. Inoperative

e. 2-Out-Of-4 Voterf. OPRM UpscaleN.A.N.A.N.A.N.A.N.A.3.4.5.6.7.8.Reactor Vessel Steam Dome Pressure

-HighReactor Vessel Water Level -Low, Level 3Main Steam Line Isolation Valve -ClosureDELETEDDrywell Pressure

-HighScram Discharge Volume Water Level -Higha. Level Transmitter

b. Float Switch1096 psig12.5 inches above instrument zero* 8% closedDELETED 1.68 psig 261' 1 1/4" elevation**

261' 1 1/4" elevation**

1103 psig11.0 inches aboveinstrument zero12% closedDELETED< 1.88 psig< 261' 9 1/4" elevation

< 261' 9 1/4" elevation

See Bases Figure B 3/4.3-1.** Equivalent to 25.58 gallons/scram discharge volume.*** The 7.6% flow "offset" for Single Loop Operation (SLO) is applied for W 7.6%. For flows W < 7.6%, the(W-7.6%)

term is set equal to zero.**** See COLR for OPRM period based detection algorithm trip setpoints.

OPRM Upscale trip output auto-enable (not bypassed) setpoints shall be APRM Simulated Thermal Power 29.5% and recirculation drive flow < 60%.LIMERICK

-UNIT 22-4Amendment No. 498,-5-1,5-2-9,-49, 163 0TABLE 2.2.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS FUNCTIONAL UNIT9. Turbine Stop Valve -Closure10. Turbine Control Valve Fast Closure,Trip Oil Pressure

-Low11. Reactor Mode Switch Shutdown Position12. Manual ScramTRIP SETPOINT 5% closedALLOWABLE VALUES< 7% closed> 465 psigN.A.N.A. 500 psigN.A.N.A.LIMERICK

-UNIT 22-4a THIS PAGE INTENTIONALLY LEFT BLANK BASESFORSECTION 2.0SAFETY LIMITSANDLIMITING SAFETY SYSTEM SETTINGS NOTEThe BASES contained in succeeding pages summarize the reasons for the Specifications in Section 2.0,but in accordance with 10 CFR 50.36 are not part ofthese Technical Specifications.

2.1 SAFETY LIMITSBASES

2.0 INTRODUCTION

The fuel cladding, reactor pressure vessel and primary system piping are theprinciple barriers to the release of radioactive materials to the environs.

Safety Limits are established to protect the integrity of these barriers duringnormal plant operations and anticipated transients.

The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limitis not violated.

Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that more than 99.9% of thefuel rods avoid transition boiling.

Meeting the Safety Limit can bedemonstrated by analysis that confirms less than 0.1% of fuel rods in the coreare susceptible to transition boiling or by demonstrating that the MCPR is notless than the values specified in Specification 2.1.2 for two recirculation loopoperation and for single recirculation loop operation.

Less than 0.1% of fuelrods in transition boiling and MCPR greater than the values specified for tworecirculation loop operation and for single recirculation loop operation represents a conservative margin relative to the conditions required to maintainfuel cladding integrity.

The fuel cladding is one of the physical barriers whichseparate the radioactive materials from the environs.

The integrity of thiscladding barrier is related to its relative freedom from perforations or cracking.

Although some corrosion or use related cracking may occur during the life of thecladding, fission product migration from this source is incrementally cumulative and continuously measurable.

Fuel cladding perforations,

however, can result fromthermal stresses which occur from reactor operation significantly above designconditions and the Limiting Safety System Settings.

While fission productmigration from cladding perforation is just as measurable as that from use relatedcracking, the thermally caused cladding perforations signal a threshold beyondwhich still greater thermal stresses may cause gross rather than incremental cladding deterioration.

Therefore, the fuel cladding Safety Limit is defined witha margin to the conditions which would produce onset of transition

boiling, MCPRof 1.0. These conditions represent a significant departure from the condition intended by design for planned operation.

2.1.1 THERMAL POWER, Low Pressure or Low FlowThe use of the (GEXL) correlation is not valid for all critical powercalculations at pressures below 785 psig or core flows less than 10% of ratedflow. Therefore, the fuel cladding integrity Safety Limit is established by othermeans. This is done by establishing a limiting condition on core THERMAL POWERwith the following basis. Since the pressure drop in the bypass region isessentially all elevation head, the core pressure drop at low power and flows willalways be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x10i lb/hr, bundle pressure drop is nearly independent of bundle power and hasa value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will begreater than 28 x 103 lb/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at thisflow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWERlimit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig isconservative.

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-UNIT 2 B 2-1 Amendment No. 4-4, 8-3, 4-, 9-7-, 4-1-4,-2-7-, 4-4, ECR LG 12-00035 SAFETY LIMITSBASES2.1.2 THERMAL POWER. Hiah Pressure and Hiah FlowThe fuel cladding integrity Safety Limit is set such that no fuel damageis calculated to occur if the limit is not violated.

Since the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleateboiling have been used to mark the beginning of the region where fuel damagecould occur. Although it is recognized that a departure from nucleate boilingwould not necessarily result in damage to BWR fuel rods, the critical power atwhich boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state andin the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for which morethan 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.

The analyses that demonstrate less than 0.1% of fuel rods enter transition boiling and determine the Safety Limit MCPR are performed using a statistical approach that combines all of the uncertainties in operating parameters and theprocedures used to calculate critical power. The analysis methods use.d to performthese calculations are described in Reference 1.

Reference:

1. "General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A (latest approved revision).

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-UNIT 2B 2-2ECR LG 12-00035 INTENTIONALLY LEFT BLANKLIMERICK

-UNIT 2B 2-3 INTENTIONALLY LEFT BLANKLIMERICK

-UNIT 2B 2-4 SAFETY LIMITSBASES2.1.3 REACTOR COOLANT SYSTEM PRESSUREThe Safety Limit for the reactor coolant system pressure has beenselected such that it is at a pressure below which it can be shown that theintegrity of the system is not endangered.

The reactor pressure vessel isdesigned to Section III of the ASME Boiler and Pressure Vessel Code 1968Edition, including Addenda through Summer 1969, which permits a maximum pres-sure transient of 110%, 1375 psig, of design pressure 1250 psig. The SafetyLimit of 1325 psig, as measured by the reactor vessel steam dome pressureindicator, is equivalent to 1375 psig at the lowest elevation of the reactorcoolant system. The reactor coolant system is designed to the ASME Boilerand Pressure Vessel Code, 1977 Edition, including Addenda through Summer 1977for the reactor recirculation piping, which permits a maximum pressure transient of 110%, 1375 psig of design pressure, 1250 psig for suction piping and 1500psig for discharge piping. The pressure Safety Limit is selected to be thelowest transient overpressure allowed by the ASME Boiler and Pressure Vessel CodeSection III, Class I.2.1.4 REACTOR VESSEL WATER LEVELWith fuel in the reactor vessel during periods when the reactor isshutdown, consideration must be given to water level requirements due to theeffect of decay heat. If the water level should drop below the top of theactive irradiated fuel during this period, the ability to remove decay heat isreduced.

This reduction in cooling capability could lead to elevated claddingtemperatures and clad perforation in the event that the water level became lessthan two-thirds of the core height. The Safety Limit has been established atthe top of the active irradiated fuel to provide a point which can be monitored and also provide adequate margin for effective action.LIMERICK

-UNIT 2B 2-5 2.2 LIMITING SAFETY SYSTEM SETTINGSBASES2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The Reactor Protection System instrumentation setpoints specified inTable 2.2.1-1 are the values at which the reactor trips are set for each para-meter. The Trip Setpoints have been selected to ensure that the reactor coreand reactor coolant system are prevented from exceeding their Safety Limitsduring normal operation and design basis anticipated operational occurrences and to assist in mitigating the consequences of accidents.

Operation with atrip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between eachTrip Setpoint and the Allowable Value is equal to or less than the driftallowance assumed for each trip in the safety analyses.

1. Intermediate Range Monitor, Neutron Flux -HighThe IRM system consists of 8 chambers, 4 in each of the reactor tripsystems.

The IRM is a 5 decade 10 range instrument.

The trip setpoint of 120divisions of scale is active in each of the 10 ranges. Thus as the IRM isranged up to accommodate the increase in power level, the trip setpoint isalso ranged up. The IRM instruments provide for overlap with both the APRMand SRM systems.The most significant source of reactivity changes during the powerincrease is due to control rod withdrawal.

In order to ensure that the IRMprovides the required protection, a range of rod withdrawal accidents havebeen analyzed.

The results of these analyses are in Section 15.4 of theFSAR. The most severe case involves an initial condition in which THERMALPOWER is at approximately 1% of RATED THERMAL POWER. Additional conservatism was taken in this analysis by assuming the IRM channel closest to the controlrod being withdrawn is bypassed.

The results of this analysis show that thereactor is shutdown and peak power is limited to 21% of RATED THERMAL POWERwith the peak fuel enthalpy well below the fuel failure threshold of 170 cal/gm.Based on this analysis, the IRM provides protection against local control roderrors and continuous withdrawal of control rods in sequence and provides backupprotection for the APRM.2. Average Power Range MonitorThe APRM system is divided into four APRM channels and four 2-Out-Of-4 Voterchannels.

The four voter channels are divided into two groups of two each, witheach group of two providing inputs to one RPS trip system. All four voters willtrip (full scram) when any two unbypassed APRM channels exceed their tripsetpoints.

APRM trip Functions 2.a, 2.b, 2.c, and 2.d are voted independently fromOPRM Upscale Function 2.f. Therefore, any Function 2.a, 2.b, 2.c, or 2.d tripfrom any two unbypassed APRM channels will result in a full trip in each of thefour voter channels.

Similarly, a Function 2.f trip from any two unbypassed APRMchannels will result in a full trip from each of the four voter channels.

For operation at low pressure and low flow during STARTUP, the APRM NeutronFlux-Upscale (Setdown) scram setting of 15% of RATED THERMAL POWER provides adequatethermal margin between the setpoint and the Safety Limits. The margin accommodates the anticipated maneuvers associated with power plant startup.

Effects of increasing pressure at zero or low void content are minor and cold water from sources available during startup is not much colder than that already in the system. Tempera-ture coefficients are small and control rod patterns are constrained by theRWM. Of all the possible sources of reactivity input, uniform control rodwithdrawal is the most probable cause of significant power increase.

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-UNIT 2B 2-6Amendment No. 4-09, 139 IMITING SAFFTY SYSTFM S[TTINGSBASESREACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Average Power Range Monitor (Continued)

Because the flux distribution associated with uniform rod withdrawals does notinvolve high local peaks and because several rods must be moved to change powerby a significant amount, the rate of power rise is very slow. Generally theheat flux is in near equilibrium with the fission rate. In an assumed uniformrod withdrawal approach to the trip level, the rate of power rise is not morethan 5% of RATED THERMAL POWER per minute and the APRM system would be morethan adequate to assure shutdown before the power could exceed the Safety Limit.The 15% Neutron Flux -Upscale (Setdown) trip remains active until the modeswitch is placed in the Run position.

The APRM trip system is calibrated using heat balance data taken duringsteady state conditions.

Fission chambers provide the basic input to thesystem and therefore the monitors respond directly and quickly to changes dueto transient operation for the case of the Neutron Flux -Upscale setpoint; i.e.,for a power increase, the THERMAL POWER of the fuel will be less than thatindicated by the neutron flux due to the time constants of the heat transferassociated with the fuel. For the Simulated Thermal Power -Upscale setpoint, a time constant of 6 +/- 0.6 seconds is introduced into the flow-biased APRM inorder to simulate the fuel thermal transient characteristics.

A more conservative maximum value is used for the flow-biased setpoint as shown in Table 2.2.1-1.A reduced Trip Setpoint and Allowable Value is provided for the Simulated Thermal Power -Upscale Function, applicable when the plant is operating in SingleLoop Operation (SLO) per LCO 3.4.1.1.

In SLO, the drive flow values (W) used inthe Trip Setpoint and Allowable Value equations is reduced by 7.6%. The 7.6% valueis established to conservatively bound the inaccuracy created in the coreflow/drive flow correlation due to back flow in the jet pumps associated with theinactive recirculation loop. The Trip Setpoint and Allowable Value thus maintainthermal margins essentially unchanged from those for two-loop operation.

The TripSetpoint and Allowable Value equations for single loop operation are only valid forflows down to W = 7.6%. The Trip Setpoint and Allowable Value do not go below61.5% and 62.0% RATED THERMAL POWER, respectively.

This is acceptable because back Iflow in the inactive recirculation loop is only an issue with drive flows ofapproximately 40% or greater (Reference 1).The APRM setpoints were selected to provide adequate margin for the SafetyLimits and yet allow operating margin that reduces the possibility of unneces-sary shutdown.

The APRM channels also include an Oscillation Power Range Monitor (OPRM)Upscale Function.

The OPRM Upscale Function provides compliance with GDC 10 andGDC 12, thereby providing protection from exceeding the fuel MCPR Safety Limit dueto anticipated thermal-hydraulic power oscillations.

The OPRM Upscale Functionreceives input signals from the local power range monitors (LPRMs) within thereactor core, which are combined into "cells" for evaluation by the OPRMalgorithms.

References 2, 3 and 4 describe three algorithms for detecting thermal-hydraulic instability related neutron flux oscillations:

the period baseddetection algorithm, the amplitude based algorithm, and the growth rate algorithm.

All three are implemented in the OPRM Upscale Function, but the safety analysistakes credit only for the period based detection algorithm.

The remaining algorithms provide defense in depth and additional protection against unanticipated oscillations.

OPRM Upscale Function OPERABILITY for Technical Specification purposes is based only on the period based detection algorithm.

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-UNIT 2 B 2-7 Amendment 4-.49.-14--3-9 Associated with Amendment 163 LIMITING SAFETY SYSTEM SETTINGSBASESREACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Average Power Range Monitor (Continued)

The OPRM Upscale trip output shall be automatically enabled (not bypassed) when APRM Simulated Thermal Power is 29.5% and recirculation drive flow is< 60% as indicated by APRM measured recirculation drive flow. (NOTE: 60%recirculation drive flow is the recirculation drive flow that corresponds to 60%of rated core flow. Refer to TS Bases 3/4.3.1 for further discussion concerning the recirculation drive flow/core flow relationship.)

This is the operating region where actual thermal-hydraulic instability and related neutron fluxoscillations may occur. See Reference 5 for additional discussion of OPRMUpscale trip enable region limits. These setpoints, which are sometimes referredto as the "auto-bypass" setpoints, establish the boundaries of the OPRM Upscaletrip enabled region. The APRM Simulated Thermal Power auto-enable setpoint has1% deadband while the drive flow setpoint has a 2% deadband.

The deadband forthese setpoints is established so that it increases the enabled region.An OPRM Upscale trip is issued from an APRM channel when the period baseddetection algorithm in that channel detects oscillatory changes in the neutronflux, indicated by the combined signals of the LPRM detectors in a cell, withperiod confirmations and relative cell amplitude exceeding specified setpoints.

One or more cells in a channel exceeding the trip conditions will result in achannel trip. An OPRM Upscale trip is also issued from the channel if either thegrowth rate or amplitude based algorithms detect oscillatory changes in theneutron flux for one or more cells in that channel.There are four "sets" of OPRM related setpoints or adjustment parameters:

a) OPRM trip auto-enable setpoints for APRM Simulated Thermal Power (29.5%) andrecirculation drive flow (60%); b) period based detection algorithm (PBDA)confirmation count and amplitude setpoints; c) period based detection algorithm tuning parameters; and d) growth rate algorithm (GRA) and amplitude basedalgorithm (ABA) setpoints.

The first set, the OPRM auto-enable region setpoints, are treated asnominal setpoints with no additional margins added as discussed in Reference 5.The settings, 29.5% APRM Simulated Thermal Power and 60% recirculation drive flow,are defined (limit values) in a note to Table 2.2.1-1.

The second set, the OPRMPBDA trip setpoints, are established in accordance with methodologies defined inReference 4, and are documented in the COLR. There are no allowable values forthese setpoints.

The third set, the OPRM PBDA "tuning" parameters, areestablished or adjusted in accordance with and controlled by station procedures.

The fourth set, the GRA and ABA setpoints, in accordance with References 2 and 3,are established as nominal values only, and controlled by station procedures.

3. Reactor Vessel Steam Dome Pressure-High High pressure in the nuclear system could cause a rupture to the nuclearsystem process barrier resulting in the release of fission products.

A pressureincrease while operating will also tend to increase the power of the reactor bycompressing voids thus adding reactivity.

The trip will quickly reduce theneutron flux, counteracting the pressure increase.

The trip setting is slightlyhigher than the operating pressure to permit normal operation without spurioustrips. The setting provides for a wide margin to the maximum allowable designpressure and takes into account the location of the pressure measurement comparedto the highest pressure that occurs in the system during a transient.

This tripsetpoint is effective at low power/flow conditions when the turbine stop valveand control fast closure trips are bypassed.

For a turbine trip or load rejection under these conditions, the transient analysis indicated an adequate margin tothe thermal hydraulic limit.LIMERICK

-UNIT 2 B 2-7a Amendment 4., 09, ,4-3,Associated with Amendment 163 INTENTIONALLY LEFT BLANK LIMITING SAFETY SYSTEM SETTINGSBASESREACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

4. Reactor Vessel Water Level-Low The reactor vessel water level trip setpoint has been used in transient analyses dealing with coolant inventory decrease.

The scram setting was chosenfar enough below the normal operating level to avoid spurious trips but highenough above the fuel to assure that there is adequate protection for the fueland pressure limits.5. Main Steam Line Isolation Valve-Closure The main steam line isolation valve closure trip was provided to limitthe amount of fission product release for certain postulated events. The MSIVsare closed automatically from measured parameters such as high steam flow, lowreactor water level, high steam tunnel temperature, and low steam line pressure.

The MSIVs closure scram anticipates the pressure and flux transients whichcould follow MSIV closure and thereby protects reactor vessel pressureand fuel thermal/hydraulic Safety Limits.6. DELETED7. Drywell Pressure-High High pressure in the drywell could indicate a break in the primary pressureboundary systems or a loss of drywell cooling.

The reactor is tripped in orderto minimize the possibility of fuel damage and reduce the amount of energy beingadded to the coolant and to the primary containment.

The trip setting wasselected as low as possible without causing spurious trips.LIMERICK

-UNIT 2B 2-8Amendment No. 52 LIMITING SAFETY SYSTEM SETTINGBASESREACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

8. Scram Discharge Volume Water Level-High The scram discharge volume receives the water displaced by the motion ofthe control rod drive pistons during a reactor scram. Should this volume fillup to a point where there is insufficient volume to accept the displaced waterat pressures below 65 psig, control rod insertion would be hindered.

The reactoris therefore tripped when the water level has reached a point high enough toindicate that it is indeed filling up, but the volume is still great enough toaccommodate the water from the movement of the rods at pressures below 65 psigwhen they are tripped.

The trip setpoint for each scram discharge volume isequivalent to a contained volume of 25.58 gallons of water.9. Turbine Stop Valve-Closure The turbine stop valve closure trip anticipates the pressure, neutronflux, and heat flux increases that would result from closure of the stopvalves. With a trip setting of 5% of valve closure from full open, theresultant increase in heat flux is such that adequate thermal margins aremaintained during the worst design basis transient.

10. Turbine Control Valve Fast Closure, Trip Oil Pressure-Low The turbine control valve fast closure trip anticipates the pressure, neutronflux, and heat flux increase that could result from fast closure of the turbinecontrol valves due to load rejection with or without coincident failure of theturbine bypass valves. The Reactor Protection System initiates a trip when fastclosure of the control valves is initiated by the fast acting solenoid valves andin less than 30 milliseconds after the start of control valve fast closure.

Thisis achieved by the action of the fast acting solenoid valves in rapidly reducinghydraulic trip oil pressure at the main turbine control valve actuator disc dumpvalves. This loss of pressure is sensed by pressure switches whose contacts formthe one-out-of-two-twice logic input to the Reactor Protection System. This tripsetting, a faster closure time, and a different valve characteristic from that ofthe turbine stop valve, combine to produce transients which are very similar tothat for the stop valve. Relevant transient analyses are discussed in Section15.2.2 of the Final Safety Analysis Report.11. Reactor Mode Switch Shutdown PositionThe reactor mode switch Shutdown position is a redundant channel to theautomatic protective instrumentation channels and provides additional manualreactor trip capability.

12. Manual ScramThe Manual Scram is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

LIMERICK

-UNIT 2B 2-9 LIMITING SAFETY SYSTEM SETTINGBASESREACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

REFERENCES:

1. NEDC-31300, "Single-Loop Operation Analysis for Limerick Generating

Station, Unit 1," August 1986.2. NEDO-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology,"

November 1995.3. NEDO-31960-A, Supplement 1, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology,"

November 1995.4. NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications,"

August 1996.5. BWROG Letter 96113, K. P. Donovan (BWROG) to L. E. Phillips (NRC),"Guidelines for Stability Option III 'Enable Region' (TAC M92882),"

September 17, 1996.LIMERICK

-UNIT 2B 2-10Amendment No. 139 I SECTIONS 3.0 and 4.0LIMITING CONDITIONS FOR OPERATION ANDSURVEILLANCE REQUIREMENTS THIS PAGE INTENTIONALLY LEFT BLANK 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in thesucceeding Specifications is required during the OPERATIONAL CONDITIONS or otherconditions specified therein; except that upon failure to meet the LimitingConditions for Operation, the associated ACTION requirements shall be met.3.0.2 Noncompliance with a Specification shall exist when the requirements ofthe Limiting Condition for Operation and associated ACTION requirements arenot met within the specified time intervals.

If the Limiting Condition forOperation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.

3.0.3 When a Limiting Condition for Operation is not met, except as providedin the associated ACTION requirements, within one hour action shall be initiated to place the unit in an OPERATIONAL CONDITION in which the Specification doesnot apply by placing it, as applicable, in:a. At least STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,b. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , andc. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .Where corrective measures are completed that permit operation under the ACTIONrequirements, the ACTION may be taken in accordance with the specified timelimits as measured from the time of failure to meet the Limiting Condition forOperation.

Exceptions to these requirements are stated in the individual Specifications.

This Specification is not applicable in OPERATIONAL CONDITION 4 or 5.3.0.4 When a Limiting Condition for Operation is not met, entry into anOPERATIONAL CONDITION or other specified condition in the Applicability shall onlybe made:a. When the associated ACTION requirements to be entered permit continued operation in the OPERATIONAL CONDITION or other specified condition inthe Applicability for an unlimited period of time; orb. After performance of a risk assessment addressing inoperable systemsand components, consideration of the results, determination of theacceptability of entering the OPERATIONAL CONDITION or other specified condition in the Applicability, and establishment of risk management

actions, if appropriate; exceptions to this Specification are statedin the individual Specifications; orc. When an allowance is stated in the individual value, parameter, orother Specification.

This Specification shall not prevent changes in OPERATIONAL CONDITIONS or otherspecified conditions in the Applicability that are required to comply with ACTIONrequirements or that are part of a shutdown of the unit.LIMERICK

-UNIT 23/4 0-1Amendment No. 132 APPLICABILITY SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL CONDITIONS or other specified conditions in the Applicability for individual LimitingConditions for Operation, unless otherwise stated in the Surveillance Requirement.

Failure to meet a Surveillance, whether such failure isexperienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the Limiting Condition forOperation.

Failure to perform a Surveillance within the specified Surveillance time interval and allowed extension per Specification 4.0.2, shall be failure tomeet the Limiting Condition for Operation except as provided inSpecification 4.0.3. Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.4.0.2 Each Surveillance Requirement shall be performed within the specified surveillance time interval with a maximum allowable extension not to exceed 25% ofthe surveillance interval.

4.0.3 If it is discovered that a Surveillance was not performed within itsspecified Surveillance time interval and allowed extension per Specification 4.0.2, then compliance with the requirement to declare the Limiting Condition for Operation not met may be delayed, from the time of discovery, up to 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months sor up to the limit of the specified Surveillance time interval, whichever isgreater.

This delay period is permitted to allow performance of theSurveillance.

A risk evaluation shall be performed for any Surveillance delayedgreater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and the risk impact shall be managed.If the Surveillance is not performed within the delay period, the LimitingCondition for Operation must immediately be declared not met, and theapplicable ACTION requirements must be entered.When the Surveillance is performed within the delay period and the Surveillance is not met, the Limiting Condition for Operation must immediately be declarednot met, and the applicable ACTION requirements must be entered.4.0.4 Entry into an OPERATIONAL CONDITION or other specified condition in theApplicability of a Limiting Condition for Operation shall only be made when theLimiting Condition for Operation's Surveillance Requirements have been met withintheir Surveillance time interval, except as provided in Specification 4.0.3.When a Limiting Condition for Operation is not met due to its Surveillance Requirements not having been met, entry into an OPERATIONAL CONDITION or otherspecified condition in the Applicability shall only be made in accordance withSpecification 3.0.4.This provision shall not prevent entry into OPERATIONAL CONDITIONS or otherspecified conditions in the Applicability that are required to comply withACTION requirements or that are part of a shutdown of the unit.4.0.5 Surveillance Requirements for inservice inspection and testing of ASMECode Class 1, 2, & 3 components shall be applicable as follows:a. Inservice inspection of ASME Code Class 1, 2, and 3 components shall beperformed in accordance with Section XI of the ASME Boiler and PressureVessel Code and applicable Addenda as required by 10 CFR Part 50, Section50.55a. Inservice testing of ASME Code Class 1, 2, and 3 pumps and valvesshall be performed in accordance with the ASME Code for Operation andMaintenance of Nuclear Power Plants (ASME OM Code) and applicable Addendaas required by 10 CFR Part 50, Section 50.55a.LIMERICK

-UNIT 23/4 0-2 Amendment No. 89, 4-2-4, 4-32, 155 APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued)

b. Surveillance intervals specified in Section XI of the ASME Boilerand Pressure Vessel Code and applicable Addenda for the inservice inspection activities, and the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda forinservice testing activities, shall be applicable as follows in theseTechnical Specifications:

Required frequencies ASME Code and applicable Addenda for performing inservice terminology for inservice inspection and testinginspection and testing activities activities Weekly At least once per 7 daysMonthly At least once per 31 daysQuarterly or every 3 months At least once per 92 daysSemiannually or every 6 months At least once per 184 daysEvery 9 months At least once per 276 daysYearly or annually At least once per 366 daysBiennially or every 2 years At least once per 731 daysc. The provisions of Specification 4.0.2 are applicable to the aboverequired frequencies for performing inservice inspection and testingactivities.

In addition, the provision of Specification 4.0.2 areapplicable to other normal and accelerated frequencies specified as 2years or less in the Inservice Testing Program for performing inservice testing activities.

d. Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements.

e. Nothing in the ASME Code shall be construed to supersede the requirements of any Technical Specification.

f. The Inservice Inspection (ISI) Program for piping identified in NRCGeneric Letter 88-01 shall be performed in accordance with the staffpositions on schedule, methods and personnel, and sample expansion included in the Generic Letter, or in accordance with alternate measuresapproved by the NRC staff. Details for implementation of theserequirements are included as augmented inspection requirements in the ISIProgram.LIMERICK

-UNIT 23/4 0-3Amendment No. 2, 84, 4-3-3, 155 INTENTIONALLY LEFT BLANK 3/4.1 REACTIVITY CONTROL SYSTEMS3/4.1.1 SHUTDOWN MARGINLIMITING CONDITION FOR OPERATION 3.1.1 The SHUTDOWN MARGIN shall be equal to or greater than:a. 0.38% Ak/k with the highest worth rod analytically determined, orb. 0.28% Ak/k with the highest worth rod determined by test.APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, 4, and 5.ACTION:With the SHUTDOWN MARGIN less than specified:

a. In OPERATIONAL CONDITION 1 or 2, reestablish the required SHUTDOWNMARGIN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTDOWN within the next12 hours.b. In OPERATIONAL CONDITION 3 or 4, immediately verify all insertable control rods to be inserted and suspend all activities that couldreduce the SHUTDOWN MARGIN. In OPERATIONAL CONDITION 4, establish SECONDARY CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .c. In OPERATIONAL CONDITION 5, suspend CORE ALTERATIONS and otheractivities that could reduce the SHUTDOWN MARGIN and insert allinsertable control rods within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Establish SECONDARY CONTAIN-MENT INTEGRITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .SURVEILLANCE REQUIREMENTS 4.1.1 The SHUTDOWN MARGIN shall be determined to be equal to or greater thanspecified at any time during the fuel cycle:a. By measurement, prior to or during the first startup after eachrefueling.

b. By measurement, within 500 MWD/T prior to the core average exposureat which the predicted SHUTDOWN MARGIN, including uncertainties andcalculation biases, is equal to the specified limit.c. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after detection of a withdrawn control rod that isimmovable, as a result of excessive friction or mechanical inter-ference, or is untrippable, except that the above required SHUTDOWNMARGIN shall be verified acceptable with an increased allowance for thewithdrawn worth of the immovable or untrippable control rod.LIMERICK

-UNIT 23/4 1-1 REACTIVITY CONTROL SYSTEMS3/4.1.2 REACTIVITY ANOMALIES LIMITING CONDITION FOR OPERATION 3.1.2 The reactivity difference between the actual core keff and the predicted core keff shall not exceed 1% Ak/k.APPLICABILITY:

OPERATIONAL CONDITION 1 and 2.ACTION:With the reactivity difference exceeding 1% Ak/k:a. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months perform an analysis to determine and explain the causeof the reactivity difference; operation may continue if the difference is explained and corrected.

b. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .SURVEILLANCE REQUIREMENTS 4.1.2 The reactivity difference between the actual core keff and the predicted core keff shall be verified to be less than or equal to 1% Ak/k:a. During the first startup following CORE ALTERATIONS, andb. At least once per 31 effective full power days during POWER OPERATION.

c. The provisions of Specification 4.0.4 are not applicable.

0LIMERICK

-UNIT 23/4 1-2Amendment No. 168 REACTIVITY CONTROL SYSTEMS3/4.1.3 CONTROL RODSCONTROL ROD OPERABILITY LIMITING CONDITION FOR OPERATION 3.1.3.1 All control rods and scram discharge volume vent and drain valves shall beOPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3***ACTION:a. With one withdrawn control rod inoperable due to being immovable, as aresult of excessive friction or mechanical interference, or known to beuntri ppabl e:1. Within 1 hour:a) Verify that the inoperable withdrawn control rod is separated from all other inoperable withdrawn control rods by at leasttwo control cells in all directions.

b) Disarm the associated directional control valves**

either:1) Electrically, or2) Hydraulically by closing the drive water and exhaustwater isolation valves.Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .2. Restore the inoperable withdrawn control rod to OPERABLE statuswithin 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT SHUTDOWN within the next 12hours.b. With one or more control rods trippable but inoperable for causesother than addressed in ACTION a, above:1. If the inoperable control rod(s) is withdrawn, within 1 hour:a) Verify that the inoperable withdrawn control rod(s) isseparated from all other inoperable withdrawn control rodsby at least two control cells in all directions, andb) Demonstrate the insertion capability of the inoperable with-drawn control rod(s) by inserting the control rod(s) atleast one notch by drive water pressure within the normaloperating range*.*The inoperable control rod may then be withdrawn to a position no furtherwithdrawn than its position when found to be inoperable.

- May be rearmed intermittently, under administrative

control, to permittesting associated with restoring the control rod to OPERABLE status.***OPERATIONAL CONDITION 3 is only applicable to the scram discharge volume vent anddrain valves.LIMERICK

-UNIT 23/4 1-3Amendment No. 431, 140 REACTIVITY CONTROL SYSTEMSLIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

Otherwise, insert the inoperable withdrawn control rod(s) anddisarm the associated directional control valves**

either:a) Electrically, orb) Hydraulically by closing the drive water and exhaust waterisolation valves.2. If the inoperable control rod(s) is inserted, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months disarmthe associated directional control valves**

either:a) Electrically, orb) Hydraulically by closing the drive water and exhaust waterisolation valves.Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .c. With more than 8 control rods inoperable, be in at least HOT SHUTDOWNwithin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .d. With one or more scram discharge volume (SDV) vent or drain lines with onevalve inoperable, restore the inoperable valve(s) to OPERABLE status within7 days or be in at least HOT SHUTDOWN within the next 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s***

and inCOLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .e. With one or more SDV vent or drain lines with both valves inoperable, isolate the associated line within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months **** or be in at least HOTSHUTDOWN within the next 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s***

and in COLD SHUTDOWN within thefollowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .SURVEILLANCE REOUIREMENTS 4.1.3.1.1 The scram discharge volume drain and vent valves shall bedemonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by:a. Verifying each valve to be open,* andb. Cycling each valve through at least one complete cycle of full travel.* These valves may be closed intermittently for testing under administrative controls.

- May be rearmed intermittently, under administrative

control, to permittesting associated with restoring the control rod to OPERABLE status.***Separate Action entry is allowed for each SDV vent and drain line.****An isolated line may be unisolated under administrative control to allow drainingand venting of the SDV.LIMERICK

-UNIT 23/4 1-4Amendment No. 4-3-], 4-3-, 147 REACTIVITY CONTROL SYSTEMSSURVEILLANCE REQUIREMENTS (Continued) 4.1.3.1.2 When above the preset power level of the RWM, all withdrawn control rods not required to have their directional control valves disarmedelectrically or hydraulically shall be demonstrated OPERABLE by moving eachcontrol rod at least one notch:a. In accordance with the Surveillance Frequency Control Program, andb. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from discovery that a control rod is immovable as aresult of excessive friction or mechanical interference.

4.1.3.1.3 All control rods shall be demonstrated OPERABLE by performance ofSurveillance Requirements 4.1.3.2, 4.1.3.4, 4.1.3.5, 4.1.3.6, and 4.1.3.7.4.1.3.1.4 The scram discharge volume shall be determined OPERABLE bydemonstrating:

a. The scram discharge volume drain and vent valves OPERABLE inaccordance with the Surveillance Frequency Control Program, byverifying that the drain and vent valves:1. Close within 30 seconds after receipt of a signal for controlrods to scram, and2. Open when the scram signal is reset.b. Proper level sensor response by performance of a CHANNEL FUNCTIONAL TEST of the scram discharge volume scram and control rod block levelinstrumentation in accordance with the Surveillance Frequency ControlProgram.LIMERICK

-UNIT 23/4 1-5Amendment No. 3-3,44,49,4-3-1,4A4-,

147 REACTIVITY CONTROL SYSTEMSCONTROL ROD MAXIMUM SCRAM INSERTION TIMESLIMITING CONDITION FOR OPERATION 3.1.3.2 The maximum scram insertion time of each control rod from the fullywithdrawn position to notch position 5, based on deenergization of thescram pilot valve solenoids as time zero, shall not exceed 7.0 seconds.APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.ACTION:With the maximum scram insertion time of one or more control rods exceeding 7 seconds:a. Declare the control rod(s) with the slow insertion time inoperable, andb. Perform the Surveillance Requirements of Specification 4.1.3.2c.

at leastonce per 60 days when operation is continued with three or more controlrods with maximum scram insertion times in excess of 7.0 seconds.Otherwise, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .SURVEILLANCE REQUIREMENTS 4.1.3.2 The maximum scram insertion time of the control rods shall be demon-strated through measurement and, during single control rod scram time tests,the control rod drive pumps shall be isolated from the accumulators:

a. For all control rods prior to THERMAL POWER exceeding 40% of RATEDTHERMAL POWER with reactor coolant pressure greater than or equal to950 psig, following CORE ALTERATIONS or after a reactor shutdown thatis greater than 120 days.b. For specifically affected individual control rods following maintenance on or modification to the control rod or control roddrive system which could affect the scram insertion time of thosespecific control rods in accordance with either "1" or "2" asfollows:l.a Specifically affected individual control rods shall be scramtime tested at zero reactor coolant pressure and the scraminsertion time from the fully withdrawn position to notchposition 05 shall not exceed 2.0 seconds, and1.b Specifically affected individual control rods shall be scramtime tested at greater than or equal to 950 psig reactorcoolant pressure prior to exceeding 40% of RATED THERMAL POWER.2. Specifically affected individual control rods shall bescram time tested at greater than or equal to 950 psigreactor coolant pressure.

c. For at least 10% of the control rods, with reactor coolant pressuregreater than or equal to 950 psig, on a rotating basis, and inaccordance with the Surveillance Frequency Control Program.LIMERICK

-UNIT 23/4 1-6Amendment No. 63, 4-32-, 147 REACTIVITY CONTROL SYSTEMSCONTROL ROD AVERAGE SCRAM INSERTION TIMESLIMITING CONDITION FOR OPERATION 3.1.3.3 The average scram insertion time of all OPERABLE control rods fromthe fully withdrawn

position, based on deenergization of the scram pilotvalve solenoids as time zero, shall not exceed any of the following:

Position Inserted FromFully Withdrawn 45392505Average Scram Inser-tion Time (Seconds) 0.430.861.933.49APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.ACTION:With the average scram insertion time exceeding any of the above limits, be inat least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .SURVEILLANCE REQUIREMENTS 4.1.3.3 All control rods shall be demonstrated OPERABLE by scram timetesting from the fully withdrawn position as required by Surveillance Requirement 4.1.3.2.LIMERICK

-UNIT 23/4 1-7 RFACTTVTTY CONTROl SYUTFMSFOUR CONTROL ROD GROUP SCRAM INSERTION TIMESLIMITING CONDITION FOR OPERATION 3.1.3.4 The average scram insertion time, from the fully withdrawn

position, for the three fastest control rods in each group of four control rods arrangedin a two-by-two array, based on deenergization of the scram pilot valve sole-noids as time zero, shall not exceed any of the following:

Position Inserted FromFully Withdrawn 4539255Average Scram Inser-tion Time (Seconds) 0.450.922.053.70APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.ACTION:With the average scram insertion times of control rods exceeding the above limits:a. Declare the control rods with the slower than average scram insertion times inoperable until an analysis is performed to determine thatrequired scram reactivity remains for the slow four control rod group,andb. Perform the Surveillance Requirements of Specification 4.1.3.2c.

at leastonce per 60 days when operation is continued with an average scraminsertion time(s) in excess of the average scram insertion time limit.Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .SURVEILLANCE REOUIREMENTS 4.1.3.4 All control rods shall be demonstrated from the fully withdrawn position as required byOPERABLE by scram time testingSurveillance Requirement 4.1.3.2.LIMERICK

-UNIT 23/4 1-8Amendment No. 132 REACTIVITY CONTROL SYSTEMSCONTROL ROD SCRAM ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.1.3.5 All control rod scram accumulators shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 5*.ACTION:a. In OPERATIONAL CONDITION 1 or 2:1. With one control rod scram accumulator inoperable, within 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s:a) Restore the inoperable accumulator to OPERABLE status, orb) Declare the control rod associated with the inoperable accumulator inoperable.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .2. With more than one control rod scram accumulator inoperable, declare the associated control rods inoperable and:a) If the control rod associated with any inoperable scramaccumulator is withdrawn, immediately verify that at leastone control rod drive pump is operating by verifying thatcontrol rod charging water header pressure is 14000 psig orby inserting at least one withdrawn control rod at least onenotch. If no control rod drive pump is operating and:1) If reactor pressure is g900 psig, then restart atleast one control rod drive pump within 20 minutesor place the reactor mode switch in the shutdownposition, or2) If reactor pressure is <900 psig, then place thereactor mode switch in the Shutdown position.

b) Insert the inoperable control rods and disarm the associated control valves either:1) Electrically, or2) Hydraulically by closing the drive water and exhaustwater isolation valves.Otherwise, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .b. In OPERATIONAL CONDITION 5*:1. With one withdrawn control rod with its associated scramaccumulator inoperable, insert the affected control rod anddisarm the associated directional control valves within onehour, either:a) Electrically, orb) Hydraulically by closing the drive water and exhaust waterisolation valves.*At least the accumulator associated with each withdrawn control rod. Notapplicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

LIMERICK

-UNIT 23/4 1-9Amendment No. 4, 105 REACTIVITY CONTROL SYSTEMSSURVEILLANCE REQUIREMENTS

2. With more than one withdrawn control rod with the associated scram accumulator inoperable or no control rod drive pump oper-ating, immediately place the reactor mode switch in the Shutdownposition.

4.1.3.5 Each control rod scram accumulator shall be determined OPERABLE:

a. In accordance with the Surveillance Frequency Control Program byverifying that the indicated pressure is greater than or equal to 955psig unless the control rod is inserted and disarmed or scrammed.

LIMERICK

-UNIT 23/4 1-10 Amendment No. 6,34,445,44-,

147 REACTIVITY CONTROL SYSTEMSCONTROL ROD DRIVE COUPLINGLIMITING CONDITION FOR OPERATION 3.1.3.6 All control rods shall be coupled to their drive mechanisms.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 5*.ACTION:a. In OPERATIONAL CONDITIONS 1 and 2 with one control rod not COuDledto1.its associated drive mechanism, within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s:If permitted by the RWM, insert the control rod drive mechanism to accomplish recoupling and verify recoupling by withdrawing the control rod, and:a) Observing any indicated response of the nuclear instrumenta-tion, andb) Demonstrating that the control rod will not go to the over-travel position.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .2. If recoupling is not accomplished on the first attempt or, ifnot permitted by the RWM, then until permitted by the RWM,declare the control rod inoperable, insert the control rod anddisarm the associated directional control valves**

either:a) Electrically, orb) Hydraulically by closing the drive water and exhaust waterisolation valves.Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .b. In OPERATIONAL CONDITION 5* with a withdrawn control rod not coupledto its associated drive mechanism, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:1. Insert the control rod to accomplish recoupling and verify recoup-ling by withdrawing the control rod and demonstrating that thecontrol rod will not go to the overtravel

position, or2. If recoupling is not accomplished, insert the control rod anddisarm the associated directional control valves**

either:a) Electrically, orb) Hydraulically by closing the drive water and exhaust waterisolation valves.*At least each withdrawn control rod. Not applicable to control rods removedper Specification 3.9.10.1 or 3.9.10.2.

- May be rearmed intermittently, under administrative

control, to permittesting associated with restoring the control rod to OPERABLE status.LIMERICK

-UNIT 23/4 1-11Amendment No. 132 REACTIVITY CONTROL SYSTEMSSURVEILLANCE REQUIREMENTS 4.1.3.6 Each affected control rod shall be demonstrated to be coupled to itsdrive mechanism by observing any indicated response of the nuclear instrumen-tation while withdrawing the control rod to the fully withdrawn position andthen verifying that the control rod drive does not go to the overtravel position:

a. Prior to reactor criticality after completing CORE ALTERATIONS thatcould have affected the control rod drive coupling integrity,

b. Anytime the control rod is withdrawn to the "Full out" position insubsequent operation, andc. Following maintenance on or modification to the control rod orcontrol rod drive system which could have affected the control roddrive coupling integrity.

4LIMERICK

-UNIT 23/4 1-12 REACTIVITY CONTROL SYSTEMSCONTROL ROD POSITION INDICATION LIMITING CONDITION FOR OPERATION 3.1.3.7 The control rod position indication system shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 5*.ACTION:a. In OPERATIONAL CONDITION 1 or 2 with one or more control rod positionindicators inoperable, within 1 hour:1. Determine the position of the control rod by using an alternate method, or:a) Moving the control rod, by single notch movement, to aposition with an OPERABLE position indicator, b) Returning the control rod, by single notch movement, toits original

position, andc) Verifying no control rod drift alarm at least once per12 hours, or2. Move the control rod to a position with an OPERABLE positionindicator, or3. When THERMAL POWER is:a) Within the preset power level of the RWM, declare thecontrol rod inoperable.

b) Greater than the preset power level of the RWM, declarethe control rod inoperable, insert the control rod and disarmthe associated directional control valves**

either:1) Electrically, or2) Hydraulically by closing the drive water and exhaustwater isolation valves.Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .b. In OPERATIONAL CONDITION 5* with a withdrawn control rod positionindicator inoperable, move the control rod to a position with anOPERABLE position indicator or insert the control rod.*At least each withdrawn control rod. Not applicable to control rods removedper Specification 3.9.10.1 or 3.9.10.2.

- May be rearmed intermittently, under administrative

control, to permittesting associated with restoring the control rod to OPERABLE status.LIMERICK

-UNIT 23/4 1-13Amendment No. 132 REACTIVITY CONTROL SYSTEMSSURVEILLANCE REQUIREMENTS 4.1.3.7 The control rod position indication system shall be determined OPERABLE by verifying:

a. In accordance with the Surveillance Frequency Control Program thatthe position of each control rod is indicated,

b. That the indicated control rod position changes during the movementof the control rod drive when performing Surveillance Requirement 4.1.3.1.2, andc. That the control rod position indicator corresponds to the controlrod position indicated by the "Full out" position indicator whenperforming Surveillance Requirement 4.1.3.6b.

LIMERICK

-UNIT 23/4 1-14Amendment No. 147 REACTIVITY CONTROL SYSTEMSCONTROL ROD DRIVE HOUSING SUPPORTLIMITING CONDITION FOR OPERATION 3.1.3.8 The control rod drive housing support shall be in place.APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2 and 3.ACTION:With the control rod drive housing support not in place, be in at least HOTSHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .SURVEILLANCE REOUIREMENTS 4.1.3.8 The control rod drive housing support shall be verified to be in placeby a visual inspection prior to startup any time it has been disassembled orwhen maintenance has been performed in the control rod drive housing supportarea.LIMERICK

-UNIT 23/4 1-15 RFACTTVTTY CONTROl SYSTFMS3/4.1.4 CONTROL ROD PROGRAM CONTROLSROD WORTH MINIMIZER LIMITING CONDITION FOR OPERATION 3.1.4.1 The rod worth minimizer (RWM) shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2*, **, when THERMAL POWER is lessthan or equal to 10% of RATED THERMAL POWER.ACTION:a. With the RWM inoperable after the first 12 control rods are fullywithdrawn, operation may continue provided that control rod movementand compliance with the prescribed control rod pattern are verifiedby a second licensed operator or technically qualified member of theunit technical staff.b. With the RWM inoperable before the first 12 control rods are fullywithdrawn, one startup per calendar year may be performed providedthat control rod movement and compliance with the prescribed controlrod pattern are verified by a second licensed operator ortechnically qualified member of the unit technical staff.c. Otherwise, with the RWM inoperable, control rod movement shall notbe permitted except by full scram.***

See Special Test Exception 3.10.2.**Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods ispermitted for the purpose of determining the OPERABILITY of the RWM prior towithdrawal of control rods for the purpose of bringing the reactor tocriticality.

- - Control rods may be moved, under administrative

control, to permit testingassociated with demonstrating OPERABILITY of the RWM.LIMERICK

-UNIT 23/4 1-16 REACTIVITY CONTROL SYSTEMSSURVEILLANCE REQUIREMENTS 4.1.4.1 The RWM shall be demonstrated OPERABLE:

a. In OPERATIONAL CONDITION 2 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to withdrawal ofcontrol rods for the purpose of making the reactor critical, and inOPERATIONAL CONDITION 1 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after RWM automatic initia-tion when reducing THERMAL POWER, by verifying proper indication ofthe selection error of at least one out-of-sequence control rod.b. In OPERATIONAL CONDITION 2 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to withdrawal ofcontrol rods for the purpose of making the reactor critical, byverifying the rod block function by demonstrating inability towithdraw an out-of-sequence control rod.c. In OPERATIONAL CONDITION 1 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after RWM automatic initiation when reducing THERMAL POWER, by verifying the rod blockfunction by demonstrating inability to withdraw an out-of-sequence control rod..d. By verifying that the control rod patterns and sequence input tothe RWM computer are correctly loaded following any loading of theprogram into the computer.

3.1.4.2 Deleted.4.1.4.2 Deleted.LIMERICK

-UNIT 23/4 1-17 REACTINVTY CONTROL FYORTOERM ROD BLOCK MONITORLIMITING CONDITION FOR OPERATION 3.1.4.3Both rod block monitor (RBM) channels shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than orequal to 30% of RATED THERMAL POWER and less than 90% of RATED THERMAL POWER withMCPR less than 1.70, or THERMAL POWER greater than or equal to 90% of rated withMCPR less than 1.40.ACTION:a. With one RBM channel inoperable:

1. Verify that the reactor is not operating on a LIMITING CONTROLROD PATTERN, and2. Restore the inoperable RBM channel to OPERABLE status within24 hours.Otherwise, place the inoperable rod block monitor channel in thetripped condition within the next hour.b. With both RBM channels inoperable, place at least one inoperable rodblock monitor channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .SURVEILLANCE REOUIREMENTS 4.1.4.3 Each of the above required RBM channels shall be demonstrated OPERABLEby performance of a:a. CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION for the OPERATIONAL CONDITIONS specified in Table 4.3.6-1 and at the frequencies specified in the Surveillance Frequency Control Program unlessotherwise noted in Table 4.3.6-1.b. CHANNEL FUNCTIONAL TEST prior to control rod withdrawal when thereactor is operating on a LIMITING CONTROL ROD PATTERN.LIMERICK

-UNIT 23/4 1-18Amendment No. 4A-, 147 REACTIVITY CONTROL SYSTEMS3/4.1.5 STANDBY LIQUID CONTROL SYSTEMLIMITING CONDITION FOR OPERATION 3.1.5 The standby liquid control system shall be OPERABLE and consist of thefollowing:

a. In OPERATIONAL CONDITIONS 1 and 2, two pumps and corresponding flow paths,b. In OPERATIONAL CONDITION 3, a minimum of one pump and corresponding flow path.APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2 and 3ACTION:a. With only one pump and corresponding explosive valve OPERABLE, inOPERATIONAL CONDITION 1 or 2, restore one inoperable pump andcorresponding explosive valve to OPERABLE status within 7 days or be inat least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .b. With standby liquid control system otherwise inoperable, in OPERATIONAL CONDITION 1, 2, or 3, restore the system to OPERABLE status within 8hours 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 COLDSHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .SURVEILLANCE REQUIREMENTS 4.1.5 The standby liquid control system shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying that:1. The temperature of the sodium pentaborate solution is within thelimits of Figure 3.1.5-1.2. The available volume of sodium pentaborate solution is at least 3160gallons.3. The temperature of the pump suction piping is within the limitsof Figure 3.1.5-1 for the most recent concentration analysis.

LIMERICK

-UNIT 23/4 1-19 Amendment No.

163 REACTIVITY CONTROL SYSTEMSSURVEILLANCE REQUIREMENTS (Continued)

b. In accordance with the Surveillance Frequency Control Program by:1. Verifying the continuity of the explosive charge.2. Determining by chemical analysis and calculation*

that theavailable weight of Boron-lO is greater than or equalto 185 lbs; the concentration of sodium pentaborate in solutionis less than or equal to 13.8% and within the limits ofFigure 3.1.5-1 and; the following equation is satisfied:

C x E x Q 113% wt. 29 atom % 86 gpmwhereC = Sodium pentaborate solution

(% by weight)Q = Two pump flowrate, as determined per surveillance requirement 4.1.5.c.E = Boron 10 enrichment (atom % Boron 10)3. Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise securedin position, is in its correct position.

4. Verifying that no more than two pumps are aligned for automatic operation.

c. Demonstrating that, when tested pursuant to Specification 4.0.5, theminimum flow requirement of 41.2 gpm per pump at a pressure of greaterthan or equal to 1230+/-25 psig is met.d. In accordance with the Surveillance Frequency Control Program by:1. Initiating at least one of the standby liquid control systemloops, including an explosive valve, and verifying that a flowpath from the pumps to the reactor pressure vessel is available by pumping demineralized water into the reactor vessel. Thereplacement charge for the explosive valve shall be from thesame manufactured batch as the one fired or from another batchwhich has been certified by having one of the batch success-fully fired. All injection loops shall be tested in 3operating cycles.2. Verify all heat-treated piping between storage tank and pumpsuction is unblocked.**

e. Prior to addition of Boron to storage tank verify sodium pentaborate enrichment to be added is 29 atom % Boron 10.This test shall also be performed anytime water or boron is added to the solu-tion or when the solution temperature drops below the limits of Figure 3.1.5-1 forthe most recent concentration

analysis, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after water or boronaddition or solution temperature is restored.

- This test shall also be performed whenever suction piping temperature drops belowthe limits of Figure 3.1.5-1 for the most recent concentration

analysis, within24 hours after solution temperature is restored.

LIMERICK

-UNIT 23/4 1-20 Amendment No.

163

". 8877J________

.2 ~ -* ja._______ I _______ ______ --i--.__________

.-J- I~A-. ~-,..i ----- -.. .-70C&aJ4.--..~~~ .......* ...... .OPERATI NG RANGE* ...........; ....._. _: ....". .'._. ....._. ..._. ._........: .... ........ ...., , .v ' .: .i........: .............." ... .......'.: : ..~~ ." " .: i ..... ............:.. ...... .. ..-. ........'6 .L "1 X- 1 i.o r10* r 1 .. .i "".I12J,24CoNc-rNTrAAT-ON, 0o YW-. -GFSODIUM PENTABORATE SOLUTIONTEMPERATURE/CONCENTRATION REQUIREMENTS FIGURE 3.1.5-13/4 1-21LIMERICK

-UNIT 2 (THIS PAGE LEFT BLANK INTENTIONALLY)

LIMERICK

-UNIT 23/4 1-22 3/4.2 POWER DISTRIBUTION LIMITS3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATELIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each typeof fuel as a function of axial location and AVERAGE PLANAR EXPOSURE shall bewithin limits based on applicable APLHGR limit values which have been determined by approved methodology for the respective fuel and lattice types. When handcalculations are required, the APLHGR for each type of fuel as a function of AVERAGEPLANAR EXPOSURE shall not exceed the limiting value for the most limiting lattice(excluding natural uranium) as shown in the CORE OPERATING LIMITS REPORT (COLR). Duringoperation, the APLHGR for each fuel type shall not exceed the above values multiplied bythe appropriate reduction factors for power and flow as defined in the COLR.APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than orequal to 25% of RATED THERMAL POWER.ACTION:With an APLHGR exceeding the limiting value, initiate corrective action within15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or,reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4hours.SURVEILLANCE REOUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the limitingvalue:a. In accordance with the Surveillance Frequency Control Program,b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least15% of RATED THERMAL POWER, andc. Initially and in accordance with the Surveillance Frequency Control Programwhen the reactor is operating with a LIMITING CONTROL ROD PATTERN forAPLHGR.d. The provisions of Specification 4.0.4 are not applicable.

LIMERICK

-UNIT 23/4 2-1Amendment No. 4, --4, 489, 147 Figures on pages3/4 2-2 thru 3/4 2-6have been removed from Technical Specifications, and relocated to theCORE OPERATING LIMITS REPORT.Technical Specifications pages3/4 2-3 thru 3/4 2-6ahave been INTENTIONALLY OMITTED.LIMERICK

-UNIT 23/4 2-2Amendment No. 4 1 POWER DISTRIBUTION LIMITSSection 3/4.2.2 (DELETED)

INFORMATION CONTAINED ONTHIS PAGE HAS BEENDELETEDLIMERICK

-UNIT 23/4 2-7Amendment No. 48 POWER DISTRIBUTION LIMITS3/4.2.3 MINIMUM CRITICAL POWER RATIOLIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater thanthe rated MCPR limit adjusted by the MCPR(P) and MCPR(F) factors as shown in theCORE OPERATING LIMITS REPORT, provided that the end-of-cycle recirculation pumptrip (EOC-RPT) system is OPERABLE per Specification 3.3.4.2 and the main turbinebypass system is OPERABLE per Specification 3.7.8, with:T (cave -cB)TA -TBwhere:TA = 0.86 seconds, control rod average scram insertion time limit to notch 39 per Specification 3.1.3.3,T = 0.672 + 1.65 N1 11/2(0.016),

nnI " Ini=il1Tave = =nY- Nii=1n =number of surveillance tests performed to date in cycle,Ni =number of active control rods measured in the ith surveillance test,Ti = average scram time to notch 39 of all rods measuredin the it" surveillance test, andN1 =total number of active rods measured in Specification 4.1.3.2.a.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to25% of RATED THERMAL POWER.LIMERICK

-UNIT 23/4 2-8Amendment No. 4, 4-4, 4-6, 48 POWER DISTRIBUTION LIMITSLIMITING CONDITION FOR OPERATION (Continued)

ACTIONa. With the end-of-cycle recirculation pump trip system inoperable perSpecification 3.3.4.2, operation may continue provided that, within 1hour, MCPR is determined to be greater than or equal to the rated MCPRlimit as a function of the average scram time (shown in the CORE OPERATING LIMITS REPORT) EOC-RPT inoperable curve, adjusted by the MCPR(P) and MCPR(F)factors as shown in the CORE OPERATING LIMITS REPORT.b. With MCPR less than the applicable MCPR limit adjusted by the MCPR(P) andMCPR(F) factors as shown in the CORE OPERATING LIMITS REPORT, initiatecorrective action within 15 minutes and restore MCPR to within the requiredlimit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMALPOWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .c. With the main turbine bypass system inoperable per Specification 3.7.8,operation may continue provided that, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , MCPR is determined to be greater than or equal to the rated MCPR limit as a function of theaverage scram time (shown in the CORE OPERATING LIMITS REPORT) mainturbine bypass valve inoperable curve, adjusted by the MCPR(P) and MCPR(F)factors as shown in the CORE OPERATING LIMITS REPORT.SURVEILLANCE REQUIREMENTS 4.2.3 MCPR, with:a. T = 1.0 prior to performance of the initial scram time measurements for the cycle in accordance with Specification 4.1.3.2a and duringreactor startups prior to control rod scram time tests in accordance with Specification 4.1.3.2.b.1.b, orb. T as defined in Specification 3.2.3 used to determine the limitwithin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the conclusion of each scram time surveillance testrequired by Specification 4.1.3.2,shall be determined to be equal to or greater than the applicable MCPR limitincluding application of the MCPR(P) and MCPR(F) factors as determined from theCORE OPERATING LIMITS REPORT.a. In accordance with the Surveillance Frequency Control Program,b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least15% of RATED THERMAL POWER, andc. Initially and in accordance with the Surveillance Frequency ControlProgram when the reactor is operating with a LIMITING CONTROL ROD PATTERNfor MCPR.d. The provisions of Specification 4.0.4 are not applicable.

LIMERICK

-UNIT 23/4 2-9Amendment No. 4, 4-6, 48, &3, 147 Figures on pages3/4 2-10 thru 3/4 2-11have been removed fromTechnical Specifications, andrelocated to the CORE OPERATING LIMITS REPORT.Technical Specifications pages3/4 2-10a thru 3/4 2-11have been INTENTIONALLY OMITTED.LIMERICK

-UNIT 23/4 2-10Amendment No. 4 1 THIS PAGE INTENTIONALLY LEFT BLANK POWER DISTRIBUTION LIMITS3/4.2.4 LINEAR HEAT GENERATION RATELIMITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed the value in theCORE OPERATING LIMITS REPORT.APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than orequal to 25% of RATED THERMAL POWER.ACTION:With the LHGR of any fuel rod exceeding the limit, initiate corrective actionwithin 15 minutes and restore the LHGR to within the limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months orreduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next4 hours.SURVEILLANCE REQUIREMENTS 4.2.4 LHGRs shall be determined to be equal to or less than the limit:a. In accordance with the Surveillance Frequency Control Program,b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of atleast 15% of RATED THERMAL POWER, andc. Initially and in accordance with the Surveillance Frequency ControlProgram when the reactor is operating on a LIMITING CONTROL ROD PATTERNfor LHGR.d. The provisions of Specification 4.0.4 are not applicable.

LIMERICK

-UNIT 23/4 2-12Amendment No. 4, 147 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LJIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels shownin Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM RESPONSETIME as shown in Table 3.3.1-2.APPLICABILITY:

As shown in Table 3.3.1-1.ACTION:Note: Separate condition entry is allowed for each channel.a. With the number of OPERABLE channels in either trip system for one or moreFunctional Units less than the Minimum OPERABLE Channels per Trip Systemrequired by Table 3.3.1-1, within one hour for each affected functional unit either verify that at least one* channel in each trip system isOPERABLE or tripped or that the trip system is tripped, or place eitherthe affected trip system or at least one inoperable channel in theaffected trip system in the tripped condition.

b. With the number of OPERABLE channels in either trip system less than theMinimum OPERABLE Channels per Trip System required by Table 3.3.1-1, placeeither the inoperable channel(s) or the affected trip system**

in thetripped condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .***

c. With the number of OPERABLE channels in both trip systems for one or moreFunctional Units less than the Minimum OPERABLE Channels per Trip Systemrequired by Table 3.3.1-1, place either the inoperable channel(s) in onetrip system or one trip system in the tripped condition within 6hours**.***

d. If within the allowable time allocated by Actions a, b or c, it is notdesired to place the inoperable channel or trip system in trip (e.g., fullscram would occur), Then no later than expiration of that allowable timeinitiate the action identified in Table 3.3.1-1 for the applicable Functional Unit.* For Functional Units 2.a, 2.b, 2.c, 2.d, and 2.f, at least two channels shall beOPERABLE or tripped.

For Functional Unit 5, both trip systems shall have eachchannel associated with the MSIVs in three main steam lines (not necessarily thesame main steam lines for both trip systems)

OPERABLE or tripped.

For Function9, at least three channels per trip system shall be OPERABLE or tripped.** For Functional Units 2.a, 2.b, 2.c, 2.d, and 2.f, inoperable channels shall beplaced in the tripped condition to comply with Action b. Action c does notapply for these Functional Units.*** A channel or trip system which has been placed in the tripped condition tosatisfy Action b. or c. may be returned to the untripped condition underadministrative control for up to two hours solely to perform testing required todemonstrate its operability or the operability of other equipment providedAction a. continues to be satisfied.

LIMERICK

-UNIT 23/4 3-1Amendment No. -1_, 34, 4-0-9, 139 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.1.1 Each reactor protection system instrumentation channel shall bedemonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS shown inTable 4.3.1.1-1 and at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3.1.1-1.

4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of allchannels shall be performed in accordance with the Surveillance Frequency ControlProgram, except Table 4.3.1.1-1 Functions 2.a, 2.b, 2.c, 2.d, 2.e, and 2.f.Functions 2.a, 2.b, 2.c, 2.d, and 2.f do not require separate LOGIC SYSTEMFUNCTIONAL TESTS. For Function 2.e, tests shall be performed in accordance with theSurveillance Frequency Control Program.

LOGIC SYSTEM FUNCTIONAL TEST for Function2.e includes simulating APRM and OPRM trip conditions at the APRM channel inputs tothe voter channel to check all combinations of two tripped inputs to the 2-Out-Of-4 voter logic in the voter channels.

4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor tripfunctional unit shown in Table 3.3.1-2 shall be demonstrated to be within itslimit in accordance with the Surveillance Frequency Control Program.

Each test shallinclude at least one channel per trip system such that all channels are tested atleast once every N times the frequency specified in the Surveillance Frequency ControlProgram where N is the total number of redundant channels in a specific reactortrip system.LIMERICK

-UNIT 23/4 3-1laAmendment No. 4--9, 1-34, 147 TABLE 3.3.1-1REACTOR PROTECTION SYSTEM INSTRUMENTATION APPLICABLE MINIMUMOPERATIONAL OPERABLE CHANNELSCONDITIONS PER TRIP SYSTEM (a)FUNCTIONAL UNIT1. Intermediate Range Monitors(b):

a. Neutron Flux -HighACTION23(i ,5(i)23(i ,5(i4(i)4(i)b. Inoperative

2. Average Power Range Monitor(e):

a. Neutron Flux -Upscale (Setdown)

b. Simulated Thermal Power -Upscalec. Neutron Flux -Upscaled. Inoperative

e. 2-Out-Of-4 Voterf. OPRM Upscale3. Reactor Vessel Steam DomePressure

-High4. Reactor Vessel Water Level -Low,Level 35. Main Steam Line Isolation Valve-Closure2111, 21, 21(o)(p)333(d)333(d)3(m)3(m)3(m)3(m)23(m)12312314411101, 2(f)221, 241(g)I/valveLIMERICK

-UNIT 23/4 3-2Amendment No. ;, 04 , 104 , 139 INTENTIONALLY LEFT BLANK TABLE 3.3.1-1 (Continued)

RFACTOR PROTECTION SYSTEM INSTRUMENTATION FUNCTIONAL UNIT6. DELETEDAPPLICABLE OPERATIONAL CONDITIONS DELETEDMINIMUMOPERABLE CHANNELSPER TRIP SYSTEM (a)DELETED2ACTIONDELETED17. DrywellPressure

-High1, 2(h)8. Scram Discharge Volume WaterLevel -Higha. Level Transmitter

b. Float Switch9. Turbine Stop Valve -Closure10. Turbine Control Valve Fast Closure,Trip Oil Pressure

-Low11. Reactor Mode Switch ShutdownPosition1, 25(i1, 25(i1(j)1(j)22224(k)2(k)1313661 ,3,1,3,24524522222217312. Manual Scram189LIMERICK

-UNIT 23/4 3-3Amendment No. 52 TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION STATEMENTS ACTION 1 -Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .ACTION 2 -Verify all insertable control rods to be inserted in the coreand lock the reactor mode switch in the Shutdown positionwithin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .ACTION 3 -Suspend all operations involving CORE ALTERATIONS and insertall insertable control rods within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .ACTION 4 -Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .ACTION 5 -Be in STARTUP with the main steam line isolation valves closedwithin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .ACTION 6 -Initiate a reduction in THERMAL POWER within 15 minutes andreduce turbine first stage pressure until the function isautomatically

bypassed, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .ACTION 7 -Verify all insertable control rods to be inserted within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .ACTION 8 -Lock the reactor mode switch in the Shutdown position within1 hour.ACTION 9 -Suspend all operations involving CORE ALTERATIONS, andinsert all insertable control rods and lock the reactormode switch in the Shutdown position within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .ACTION 10 -a.. If the condition exists due to a common-mode OPRM deficiency*,

then initiate alternate method to detect and suppress thermal-hydraulic instability oscillations within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND restorerequired channels to OPERABLE status within 120 days,ORb. Reduce THERMAL POWER to < 25% RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .* Unanticipated characteristic of the instability detection algorithm or equipment that renders all OPRM channelsinoperable at once.LIMERICK

-UNIT 23/4 3-4Amendment No. 4-9,-1-1-2-,4--3, 161 TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a) 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 forrequired surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

(b) This function shall automatically be bypassed when the reactor mode switch isin the Run position.

(c) DELETED(d) The noncoincident NMS reactor trip function logic is such that all channels goto both trip systems.

Therefore, when the "shorting links" are removed, theMinimum OPERABLE Channels Per Trip System is 6 IRMs.(e) An APRM channel is inoperable if there are less than 3 LPRM inputs per levelor less than 20 LPRM inputs to an APRM channel, or if more than 9 LPRM inputsto the APRM channel have been bypassed since the last APRM calibration (weeklygain calibration).

(f) This function is not required to be OPERABLE when the reactor pressurevessel head is removed per Specification 3.10.1.(g) This function shall be automatically bypassed when the reactor mode switchis not in the Run position.

(h) This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

(i) With any control rod withdrawn.

Not applicable to control rods removed perSpecification 3.9.10.1 or 3.9.10.2.

(j) This function shall be automatically bypassed when turbine first stagepressure is equivalent to a THERMAL POWER of less than 29.5% of RATED THERMALPOWER.(k) Also actuates the EOC-RPT system.(l) DELETED(m) Each APRM channel provides inputs to both trip systems.(n) DELETED(o) With THERMAL POWER 25% RATED THERMAL POWER. The OPRM Upscale trip outputshall be automatically enabled (not bypassed) when APRM Simulated Thermal Poweris 29.5% and recirculation drive flow is < 60%. The OPRM trip output may beautomatically bypassed when APRM Simulated Thermal Power is < 29.5% orrecirculation drive flow is 60%.(p) A minimum of 23 cells, each with a minimum of 2 OPERABLE LPRMs, must beOPERABLE for an OPRM channel to be OPERABLE.

LIMERICK

-UNIT 23/4 3-5Amendment No.

163 TABLE 3.3.1-2REACTOR PROTECTION SYSTEM RESPONSE TIMESRESPONSE TIMEFUNCTIONAL UNIT (Seconds)

1. Intermediate Range Monitors:

a. Neutron Flux -High N.A.b. Inoperative N.A.2. Average Power Range Monitor*:

a. Neutron Flux -Upscale (Setdown)

N.A.b. Simulated Thermal Power -Upscale N.A.c. Neutron Flux -Upscale N.A.d. Inoperative N.A.e. 2-Out-Of-4 Voter 0.05*f. OPRM Upscale N.A.3. Reactor Vessel Steam Dome Pressure

-High 0.554. Reactor Vessel Water Level -Low, Level 3 1.05#5. Main Steam Line Isolation Valve -Closure 0.066. DELETED DELETED7. Drywell Pressure

-High N.A.8. Scram Discharge Volume Water Level -Higha. Level Transmitter N.A.b. Float Switch N.A.9. Turbine Stop Valve -Closure 0.0610. Turbine Control Valve Fast Closure,Trip Oil Pressure

-Low O.08**11. Reactor Mode Switch Shutdown Position N.A.12. Manual Scram N.A.* Neutron detectors, APRM channel and 2-Out-Of-4 Voter channel digital electronics are exempt fromresponse time testing.

Response time shall be measured from activation of the 2-Out-Of-4 Voteroutput relay. For application of Specification 4.3.1.3, the redundant outputs from each 2-Out-Of-4 Voter channel are considered part of the same channel, but the OPRM and APRM outputs are considered to be separate

channels, so N = 8. Testing of OPRM and APRM outputs shall alternate.

- Measured from start of turbine control valve fast closure.# Sensor is eliminated from response time testing for the RPS circuits.

Response time testingand conformance to the administrative limits for the remaining channel including trip unitand relay logic are required.

LIMERICK

-UNIT 23/4 3-6Amendment No. -5,9, -3, 49, 139 TAO.3.1.1-1 r:;YSTFM INSTRIIMFNTATION SIIRVFTIIANCF RFnIITRFMFNTS PFACTflR PpnTFCTTnfN CHANNELCHANNEL FUNCTIONAL FUNCTIONAL UNIT CHECK (n) TEST (n)CHANNELCALIBRATION(a)(n)

OPERATIONAL CONDITIONS FOR WHICHSURVEILLANCE REQUIRED1. Intermediate Range Monitors:

a. Neutron Flux -Highb. Inoperative

2. Average Power Range Monitor(f):

a. Neutron Flux -Upscale (Setdown)

b. Simulated Thermal Power -Upscalec. Neutron Flux -Upscale(b)N.A.(b)Qj)(l)(e)N.A.(d), (g),(0), (p)(d)N.A.23(i), 4(i), 5(i)2, 3(i), 4(i), 5(i)2111, 21, 21(m)1, 2(h)1, 2d. Inoperative

e. 2-Out-Of-4 Voterf. OPRM UpscaleN.A.N.A.(e)(c)(g)3. Reactor Vessel Steam DomePressure

-High4. Reactor Vessel Water Level-Low, Level 35. Main Steam Line Isolation Valve -ClosureN.A.I6. DELETED7. Drywell Pressure

-High8. Scram Discharge Volume WaterLevel -Higha. Level Transmitter

b. Float Switch1, 21, 2, 5(i)N.A. 1, 2, 5(i)LIMERICK

-UNIT 23/4 3-7Amendment No. 2,.-5, 163 TABLE 4.3.1.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICHFUNCTIONAL UNIT CHECK (n) TEST (n) CALIBRATION(a)(n)

SURVEILLANCE REQUIRED9. Turbine Stop Valve -Closure N.A. 110. Turbine Control Valve FastClosure, Trip OilPressure

-Low N.A. 111. Reactor Mode SwitchShutdown Position N.A. N.A. 1, 2, 3, 4, 512. Manual Scram N.A. N.A. 1, 2, 3, 4, 5(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 afterentering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for a least 1/2decades during each controlled

shutdown, if not performed within the previous 7 days.(c) Calibration includes verification that the OPRM Upscale trip auto-enable (not-bypass) setpoint for APRMSimulated Thermal Power is 29.5% and for recirculation drive flow is < 60%.(d) The more frequent calibration shall consist of the adjustment of the APRM channel to conform to the power valuescalculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER 25% of RATED THERMAL POWER.Adjust the APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER.(e) CHANNEL FUNCTIONAL TEST shall include the flow input function, excluding the flow transmitter.

(f) The LPRMs shall be calibrated at least once per 2000 effective full power hours (EFPH).(g) The less frequent calibration includes the flow input function.

(h) This function is not required to be OPERABLE when the reactor pressure vessel head is removed perSpecification 3.10.1.Ci) With any control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

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

During this time, CORE ALTERATIONS shall be suspended, and no control rod shallbe moved from its existing position.

(k) DELETED(1) Not required to be performed when entering OPERATIONAL CONDITION 2 from OPERATIONAL CONDITION 1 until 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months safter entering OPERATIONAL CONDITION 2.(m) With THERMAL POWER 25% of RATED THERMAL POWER.(n) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.LIMERICK

-UNIT 23/4 3-8Amendment No.

163 TABLE 4.3.1.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS (o) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated toverify that it is functioning as required before returning the channel to service.(p) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Trip Setpointat the completion of the surveillance; otherwise, the channel shall be declared inoperable.

Setpoints more conservative than the Trip Setpoint are acceptable provided that the as-found and as-left tolerances apply to the actual setpointimplemented in the surveillance procedures (field setting) to confirm channel performance.

The methodologies used todetermine the as-found and the as-left tolerances are specified in the associated Technical Specifications Bases.LIMERICK

-UNIT 23/4 3-8aAmendment No. 163 I INTENTIONALLY LEFT BLANK INSTRUMENTATION 3/4.3.2.

ISOLATION ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The isolation actuation instrumentation channels shown in Table 3.3.2-1 shall beOPERABLE with their trip setpoints set consistent with the values shown in the TripSetpoint column of Table 3.3.2.-2 and with ISOLATION SYSTEM RESPONSE TIME as shown in Table3.3.2-3.APPLICABILITY:

As shown in Table 3.3.2-1.ACTION:a) With an isolation actuation instrumentation channel trip setpoint lessconservative than the value shown in the Allowable Values column of Table 3.3.2-2, declare the channel inoperable until the channel is restored to OPERABLEstatus with its trip setpoint adjusted consistent with the Trip Setpoint value.b) With the number of OPERABLE channels less than required by the Minimum OPERABLEChannels per Trip System requirements for one trip system:1. If placing the inoperable channel(s) in the tripped condition would causean isolation, the inoperable channel(s) shall be restored to OPERABLE statuswithin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . If this cannot be accomplished, the ACTION required by Table3.3.2-1 for the affected trip function shall be taken, or the channel shallbe placed in the tripped condition.

or2. If placing the inoperable channel(s) in the tripped condition would not causean isolation, the inoperable channel(s) and/or that trip system shallbe placed in the tripped condition within:a) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for trip functions common* to RPS Instrumentation, b) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for trip functions not common* to RPS Instrumentation.

Trip functions common to RPS Actuation Instrumentation are shownin Table 4.3.2.1-1.

LIMERICK

-UNIT 23/4 3-9Amendment No. 7-, 3, 132 INSTRUMENTATION LIMITING CONDIQION FOR OPERATION (Continued)

ACTION: (Continued)

C. With the number of OPERABLE channels less than required by the MinimumOPERABLE Channels per Trip System requirement for both trip systems,place at least one trip system**

in the tripped condition within 1 hourand take the ACTION required by Table 3.3.2-1.SURVEILLANCE REQUIREMENTS 4.3.2.1 Each isolation actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, andCHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS shown in Table4.3.2.1-1 and at the frequencies specified in the Surveillance Frequency ControlProgram unless otherwise noted in Table 4.3.2.1-1.

4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operations of allchannels shall be performed in accordance with the Surveillance Frequency ControlProgram.4.3.2.3 The ISOLATION SYSTEM RESPONSE TIME of each isolation trip function shownin Table 3.3.2-3 shall be demonstrated to be within its limit in accordance with theSurveillance Frequency Control Program.

Each test shall include at least one channelper trip system such that all channels are tested at least once every N times thefrequency specified in accordance with the Surveillance Frequency Control Program,where N is the total number of redundant channels in a specific isolation tripsystem.** The trip system need not be placed in the tripped condition if this wouldcause the Trip Function to occur. When a trip system can be placed in thetripped condition without causing the Trip Function to occur, place the tripsystem with the most inoperable channels in the tripped condition; if bothsystems have the same number of inoperable

channels, place either trip systemin the tripped condition.

LIMERICK

-UNIT 23/4 3-10Amendment No. 4-4, -4, 147 TABLE 3.3.2-1ISOLATION ACTUATION INSTRUMENTATION MINIMUM APPLICABLE ISOLATION OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION SIGNAL (a) PER TRIP SYSTEM (b) CONDITION ACTION1. MAIN STEAM LINE ISOLATION

a. Reactor Vessel Water Level1) Low, Low-Level 2 B 2 1, 2, 3 212) Low, Low, Low-Level 1 C 2 1, 2, 3 21b. DELETED DELETED DELETED DELETED DELETEDc. Main Steam LinePressure

-Low P 2 1 22d. Main Steam LineFlow -High E 2/line 1, 2, 3 20e. Condenser Vacuum -Low Q 2 1, 2**, 3** 21f. Outboard MSIV RoomTemperature

-High F(f) 2 1, 2, 3 21g. Turbine Enclosure

-Main SteamLine Tunnel Temperature

-High F(f) 14 1, 2, 3 21h. Manual Initiation NA 2 1, 2, 3 242. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION

a. Reactor Vessel Water LevelLow -Level 3 A 2 1, 2, 3 23b. Reactor Vessel (RHR Cut-InPermissive)

Pressure

-High V 2 1, 2, 3 23c. Manual Initiation NA 1 1, 2, 3 24LIMERICK

-UNIT 23/4 3-11Amendment No. 52 TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION MINIMUM APPLICABLE ISOLATION OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION SIGNAL (a) PER TRIP SYSTEM (b) CONDITION ACTION3 REACTOR WATER CLEANUP SYSTEM ISOLATION

a. RWCS A Flow -High J 1 1, 2, 3 23b. RWCS Area Temperature

-High J 6 1, 2, 3 23c. RWCS Area Ventilation A Temperature

-High J 6 1, 2, 3 23d. SLCS Initiation y(d) NA 1, 2, 3 23e. Reactor Vessel Water Level -Low, Low -Level 2 B 2 1, 2, 3 23f. Manual Initiation NA 1 1, 2, 3 244. HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION

a. HPCI Steam LineA Pressure

-High L 1 1, 2, 3 23b. HPCI Steam SupplyPressure

-Low LA 2 1, 2, 3 23c. HPCI Turbine Exhaust Diaphragm Pressure

-High L 2 1, 2, 3 23d. HPCI Equipment RoomTemperature

-High L 1 1, 2, 3 23e. HPCI Equipment RoomA Temperature

-High L 1 1, 2, 3 23LIMERICK

-UNIT 23/4 3-12 0 _ _TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION MINIMUMISOLATION OPERABLE CHANNELSTRIP FUNCTION SIGNAL (a) PER TRIP SYSTEM (b)4. HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION (Continued)

f. HPCI Pipe Routing AreaTemperature

-High L 4g. Manual Initiation NA(e) 1/systemh. HPCI Steam Line A Press Timer NA 15. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION

a. RCIC Steam LineA Pressure

-High K 1b. RCIC Steam Supply Pressure

-Low KA 2c. RCIC Turbine Exhaust Diaphragm Pressure

-High K 2d. RCIC Equipment RoomTemperature

-High K 1e. RCIC Equipment RoomA Temperature

-High K 1f. RCIC Pipe Routing AreaTemperature

-High K 4g. Manual Initiation NA(e) 1/systemh. RCIC Steam LineA Pressure Timer NA 1APPLICABLE OPERATIONAL CONDITION ACTION1,1,1,1,1,2,2,2,2,2,333332324231, 2, 31, 2, 31, 2, 323232323232324231,2,2,331, 2, 3LIMERICK

-UNIT 23/4 3-13 TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION MINIMUM APPLICABLE ISOLATION OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION SIGNAL'a)

PER TRIP SYSTEM CONDITION ACTION6. PRIMARY CONTAINMENT ISOLATION

a. Reactor Vessel Water Level1) Low, Low -Level 2 B 2 1, 2, 3 202) Low, Low, Low -Level 1 C 2 1, 2, 3 20b. Drywell Pressure

-High H 2 1, 2, 3 20c. North Stack EffluentRadiation

-High (g) W 1 1, 2, 3 23d. Deletede. Reactor Enclosure Ventilation Exhaust Duct-Radiation

-High S 2 1, 2, 3 23f. Deletedg. Deletedh. Drywell Pressure

-High/Reactor Pressure

-Low G 2/2 1, 2, 3 26i. Primary Containment Instrument M 1 1, 2, 3 26Gas Line to Drywell APressure

-Lowj. Manual Initiation NA 1 1, 2, 3 24LIMERICK

-UNIT 23/4 3-14Amendment No. 74 0TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION MINIMUMISOLATION TRIP FUNCTION

7. SECONDARY CONTAINMENT ISOLATION

a. Reactor Vessel Water LevelLow, Low -Level 2 Bb. Drywell Pressure

-High Hc.1. Refueling Area Unit 1 Ventilation Exhaust Duct Radiation

-High R2. Refueling Area Unit 2 Ventilation Exhaust Duct Radiation

-High Rd. Reactor Enclosure Ventilation ExhaustDuct Radiation

-High Se. Deletedf. Deletedg. Reactor Enclosure Manual Initiation NAh. Refueling Area Manual Initiation NAOPERABLE CHANNELSPER TRIP SYSTEM (b)APPLICABLE OPERATIONAL CONDITION ACTION2525222221,1,2, 32, 31, 2, 31, 2, 32525252425LIMERICK

-UNIT 23/4 3-15Amendment No. 74 TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION ACTION STATEMENTS ACTION 20 -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 thenext 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .ACTION 21 -Be in at least STARTUP with the associated isolation valves closed within 6hours 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 SHUTDOWNwithin the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .ACTION 22 Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .ACTION 23 -In OPERATIONAL CONDITION 1 or 2, verify the affected system isolation valvesare closed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and declare the affected system inoperable.

InOPERATIONAL CONDITION 3, be in at least COLD SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .ACTION 24 -Restore the manual initiation function to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months orclose the affected system isolation valves within the next hour and declarethe affected system inoperable or be in at least HOT SHUTDOWN within the next12 hours and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .ACTION 25 -Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment systemoperating within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .ACTION 26 -Close the affected system isolation valves within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .TABLE NOTATIONS Required when (1) handling RECENTLY IRRADIATED FUEL in the secondary containment, or (2) during operations with a potential for draining the reactorvessel with the vessel head removed and fuel in the vessel.** May be bypassed under administrative

control, with all turbine stop valves closed.# During operation of the associated Unit 1 or Unit 2 ventilation exhaust system.(a) DELETED(b) A channel may be placed in an inoperable status for up to 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months sfor required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring thatparameter.

Trip functions common to RPS Actuation Instrumentation areshown in Table 4.3.2.1-1.

In addition, for the HPCI system and RCIC systemisolation, provided that the redundant isolation valve, inboard or outboard, asapplicable, in each line is OPERABLE and all required actuation instrumentation forthat valve is OPERABLE, one channel may be placed in an inoperable status for up to 8hours for required surveillance without placing the channel or trip system in thetripped condition.

LIMERICK

-UNIT 23/4 3-16Amendment No. 7, -, 4--, 146 TABLE 3.3.2-1 (Continued)

TABLE NOTATIONS (c) Actuates secondary containment isolation valves. Signal B, H, S, and R alsostart the standby gas treatment system.(d) RWCU system inlet outboard isolation valve closes on SLCS "B" initiation.

RWCU system inlet inboard isolation valve closes on SLCS "A" or SLCS "C"initiation.

(e) Manual initiation isolates the steam supply line outboard isolation valve andonly following manual or automatic initiation of the system.(f) In the event of a loss of ventilation the temperature

-high setpoint may beraised by 50°F for a period not to exceed 30 minutes to permitrestoration of the ventilation flow without a spurious trip. During the 30minute period, an operator, or other qualified member of the technical staff,shall observe the temperature indications continuously, so that, in the eventof rapid increases in temperature, the main steam lines shall be manuallyisolated.

(g) Wide range accident monitor per Specification 3.3.7.5.LIMERICK

-UNIT 23/4 3-17Amendment No. --7, -4, 107 TABLE 3.3.2-2ISOLATION ACTUATION INSTR[JMFNTATTON SETPOINTS ALLOWABLE VALUETRIP FUNCTION1. MAIN STEAM LINE ISOLATION

a. Reactor Vessel Water Level1) Low, Low -Level 22) Low, Low, Low -Level 1b. DELETEDc. Main Steam LinePressure

-Lowd. Main Steam LineFlow -Highe. Condenser Vacuum -Lowf. Outboard MSIV RoomTemperature

-Highg. Turbine Enclosure

-Main SteamLine Tunnel Temperature

-Highh. Manual Initiation

2. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION

a. Reactor Vessel Water LevelLow -Level 3b. Reactor Vessel (RHR Cut-inPermissive)

Pressure

-Highc. Manual Initiation TRIP SETPOINT> -38 inches*> -129 inches*DELETED> 756 psig< 122.1 psid10.5 psia> -45 inches> -136 inchesDELETED> 736 psig< 123 psid>10.1 psia/ 10.9 psia< 1920F< 1650FN.A.< 200°F< 175°FN.A.> 12.5 inches*> 11.0 inches< 75 psigN.A.< 95 psigN.A.LIMERICK

-UNIT 23/4 3-18Amendment No. -54, 52 0TABLE 3.3.2-2 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SETPOINTS TRIP FUNCTION3. REACTOR WATER CLEANUP SYSTEM ISOLATION

a. RWCS A Flow -Highb. RWCS Area Temperature

-Highc. RWCS Area Ventilation A Temperature

-Highd. SLCS Initiation

e. Reactor Vessel Water Level -Low, Low, -Level 2f. Manual Initiation

4. HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION

a. HPCI Steam Line A Pressure

-Highb. HPCI Steam Supply Pressure

-Lowc. HPCI Turbine Exhaust Diaphragm Pressure

-Highd. HPCI Equipment RoomTemperature

-Highe. HPCI Equipment RoomA Temperature

-Highf. HPCI Pipe Routing AreaTemperature

-Highg. Manual Initiation

h. HPCI Steam Line A Pressure

-TimerTRIP SETPOINT< 54.9 gpm< 155°F or _ 1320F**< 520F or _ 320F**N.A.> -38 inches *N.A.< 974" H20> 100 psig< 10 psig2250FALLOWABLE VALUE< 65.2 gpm< 160°F or _ 137°F**< 60°F or < 400F**N.A.> -45 inchesN.A.< 984" H20> 90 psig< 20 psig> 2180F, _< 247°F< 108.50F> 1650F, < 200°FN.A.2.5 < T < 13 seconds< 1040F1750FN.A.3 < r < 12.5 secondsLIMERICK

-UNIT 23/4 3-19Amendment No. 44,-5-4,4-2-3, 164 TABLE 3.3.2-2 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE VALUETRIP FUNCTION5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION

a. RCIC Steam Line APressure

-Highb. RCIC Steam Supply Pressure

-Lowc. RCIC Turbine Exhaust Diaphragm Pressure

-Highd. RCIC Equipment RoomTemperature

-Highe. RCIC Equipment RoomA Temperature

-Highf. RCIC Pipe Routing AreaTemperature

-HighTRIP SETPOINT< 373" H20> 64.5 psig< 10.0 psig< 381" H20> 56.5 psig< 20.0 psig205°F> 1980F, < 2270F< 1090F< 113.50Fg.h.Manual Initiation RCIC Steam Line A Pressure Timer175°FN.A.3 < T < 12.5 seconds> 1650F, < 200°FN.A.2.5 < T < 13 secondsLIMERICK

-UNIT 23/4 3-20Amendment No.4-6, 51 0TABLE 3.3.2-2 (Continued)

IS(wATMIN ACTIIATIMN INSTRUMENTATION SETPOINTS ALLOWABLE VALUETRIP FUNCTION6. PRIMARY CONTAINMENT ISOLATION

a. Reactor Vessel Water Level1. Low, Low -Level 22. Low, Low, Low, Level 1b. Drywell Pressure

-Highc. North Stack EffluentRadiation

-Highd. Deletede. Reactor Enclosure Ventilation ExhaustDuct -Radiation

-Highf. Deletedg. Deletedh. Drywell Pressure

-High/Reactor Pressure

-Lowi. Primary Containment Instrument Gas to Drywell A Pressure

-Lowj. Manual Initiation TRIM SETPOINT> -38 inches*> -129 inches*< 1.68 psig< 2.1 pCi/cc> -45 inches> -136 inches* 1.88 psig* 4.0 pCi/cc< 1.35 mR/h< 1.5 mR/h< 1.68 psig/> 455 psig (decreasing)

> 2.0 psi1.88 psig/435 psig (decreasing) 1.9 psiN.A.N.A.LIMERICK

-UNIT 23/4 3-21Amendment No. 74 TABLE 3.3.2-2 (Continued)

ISOLATTON ACTHATTON TNSTR1JMFNTATTON SFTPOINTS TRIP FUNCTION7. SECONDARY CONTAINMENT ISOLATION

a. Reactor Vessel Water Level -Low, Low -Level 2b. Drywell Pressure

-Highc.1. Refueling Area Unit 1 Ventilation Exhaust Duct Radiation

-High2. Refueling Area Unit 2 Ventilation Exhaust Duct Radiation

-Highd. Reactor Enclosure Ventilation ExhaustDuct Radiation

-Highe. Deletedf. Deletedg. Reactor Enclosure ManualInitiation

h. Refueling Area Manual Initiation TRIP SETPOINTALLOWABLE VALUE> -38 inches*< 1.68 psig< 2.0 mR/h< 2.0 mR/h< 1.35 mR/h-45 inches1.88 psig< 2.2 mR/h< 2.2 mR/h< 1.5 mR/hN.A.N.A.N.A.N.A.* See Bases Figure B 3/4 3-1.** The low setpoints are for the RWCU Heat Exchanger Rooms; the high setpoints are for the pump rooms.LIMERICK

-UNIT 23/4 3-22Amendment No. 74 TABLE 3.3.2-3ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIMETRIP FUNCTION RESPONSE TIME (Seconds)#

1. MAIN STEAM LINE ISOLATION

a. Reactor Vessel Water Level1) Low, Low -Level 2 N.A.2) Low, Low, Low -Level 1 <1.0H*b. DELETED DELETEDc. Main Steam LinePressure

-Low <I.0###*d. Main Steam LineFlow -High <0.5#*e. Condenser Vacuum -Low N.A.f. Outboard MSIV RoomTemperature

-High N.A.g. Turbine Enclosure

-Main SteamLine Tunnel Temperature

-High N.A.h. Manual Initiation N.A.2. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION

a. Reactor Vessel Water LevelLow -Level 3 N.A.b. Reactor Vessel (RHR Cut-InPermissive)

Pressure

-High N.A.c. Manual Initiation N.A.3. REACTOR WATER CLEANUP SYSTEM ISOLATION

a. RWCS A Flow -High N.A.#b. RWCS Area Temperature

-High N.A.c. RWCS Area Ventilation A Temperature

-High N.A.d. SLCS Initiation N.A.e. Reactor Vessel Water Level -Low, Low -Level 2 N.A.f. Manual Initiation N.A.LIMERICK

-UNIT 23/4 3-23Amendment No. , 93 TABLE 3.3.2-3 (Continued)

ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIMETRIP FUNCTION RESPONSE TIME (Seconds)#

4. HIGH PRESSURE COOLANT INJECTION SYSTEMISOLATION

a. HPCI Steam LineA Pressure

-High N.A.b. HPCI Steam SupplyPressure

-Low N.A.c. HPCI Turbine Exhaust Diaphragm Pressure

-High N.A.d. HPCI Equipment RoomTemperature

-High N.A.e. HPCI Equipment RoomA Temperature

-High N.A.f. HPCI Pipe Routing AreaTemperature

-High N.A.g. Manual Initiation N.A.5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION

a. RCIC Steam LineA Pressure

-High N.A.b. RCIC Steam Supply Pressure

-Low N.A.c. RCIC Turbine Exhaust Diaphragm Pressure

-High N.A.d. RCIC Equipment RoomTemperature

-High N.A.e. RCIC Equipment RoomA Temperature

-High N.A.f. RCIC Pipe Routing AreaTemperature

-High N.A.g. Manual Initiation N.A.I.LIMERICK

-UNIT 23/4 3-24Amendment No. 93 TABLE 3.3.2-3 (Continued)

ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIMETRIP FUNCTION RESPONSE TIME (Seconds)#

6. PRIMARY CONTAINMENT ISOLATION

a. Reactor Vessel Water Level1) Low, Low -Level 2 N.A.2) Low, Low, Low -Level 1 N.A.b. Drywell Pressure

-High N.A.c. North Stack EffluentRadiation

-High N.A.d. Deletede. Reactor Enclosure Ventilation ExhaustDuct -Radiation High N.A.f. Deletedg. Deletedh. Drywell Pressure High/Reactor Pressure

-Low N.A.i. Primary Containment Instrument Gas toDrywell A Pressure

-Low N.A.j. Manual Initiation N.A.7. SECONDARY CONTAINMENT ISOLATION

a. Reactor Vessel Water LevelLow, Low -Level 2 N.A.b. Drywell Pressure

-High N.A.c.1. Refueling Area Unit 1 Ventilation Exhaust Duct Radiation

-High N.A.2. Refueling Area Unit 2 Ventilation Exhaust Duct Radiation

-High N.A.d. Reactor Enclosure Ventilation ExhaustDuct Radiation

-High N.A.e. DeletedLIMERICK

-UNIT 23/4 3-25Amendment No. 1-4, 93 TABLE 3.3.2-3 (Continued)

ISOLATION SYSTEM INSTR[JMENTATION RFSPONSF TTMFTRIP FUNCTIONRESPONSE TIME (Seconds)#

f. Deletedg. Reactor Enclosure ManualInitiation N.A.N.A.h. Refueling Area Manual Initiation TABLE NOTATIONS (a) DELETED(b) DELETED* Isolation system instrumentation response timegenerator delays assumed for MSIVs.for MSIV only.No diesel** DELETED# Isolation system instrumentation response time specified for the TripFunction actuating each valve group shall be added to isolation timefor valves in each valve group to obtain ISOLATION SYSTEM RESPONSETIME for each valve.#f With 45 second time delay.f Sensor is eliminated from response time testing forcircuits.

Response time testing and conformance tothe remaining channel including trip unit and relaythe MSIV actuation logicthe administrative limits forlogic are required.

LIMERICK

-UNIT 23/4 3-26Amendment No. 2, 7-4, 43, 107 TABLE 4.3.2.1-1 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICHCHECK (a) TEST (a) CALIBRATION(a)

SURVEILLANCE REQUIREDTRIP FUNCTION1. MAIN STEAM LINE ISOLATION

a. Reactor Vessel Water Level1) Low, Low, Level 22) Low, Low, Low -Level 1b. DELETEDc. Main Steam LinePressure

-Lowd. Main Steam LineFlow -Highe. Condenser Vacuum -Lowf. Outboard MSIV RoomTemperature

-Highg. Turbine Enclosure

-Main SteamLine Tunnel Temperature

-Highh. Manual Initiation

2. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION

a. Reactor Vessel Water Level##Low -Level 31, 2, 31, 2, 3DELETED11, 2, 31, 2**, 3**1, 2, 31.2,2,33N.A.N.A.b. Reactor Vessel (RHR Cut-InPermissive)

Pressure

-Highc. Manual Initiation 1, 2, 31, 2, 31, 2, 3N.A.N.A.LIMERICK

-UNIT 23/4 3-27Amendment No. 44, 32, , 147 TABLE 4.3.2.1-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNELCHANNEL FUNCTIONAL CHANICHECK (a) TEST (a) CALIBRTRIP FUNCTIONNELAT ION (a)OPERATIONAL CONDITIONS FOR WHICHSURVEILLANCE REQUIRED3. REACTOR WATER CLEANUP SYSTEM ISOLATION

a. RWCS A Flow -Highb. RWCS Area Temperature

-Highc. RWCS Area Ventilation A Temperature

-Highd. SLCS Initiation

e. Reactor Vessel Water LevelLow, Low, -Level 2f. Manual Initiation

4. HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION

a. HPCI Steam LineA Pressure

-Highb. HPCI Steam SupplyPressure, Lowc. HPCI Turbine Exhaust Diaphragm Pressure

-Highd. HPCI Equipment RoomTemperature

-Highe. HPCI Equipment RoomA Temperature

-Highf. HPCI Pipe Routing AreaTemperature

-Highg. Manual Initiation

h. HPCI Steam LineA Pressure TimerN.A.N.A.N.A.N.A.1, 2, 31, 2, 31, 2, 31, 2, 31, 2, 31, 2, 31 2, 31 2, 31 2, 31 2, 31 2, 31 2, 31, 2, 31, 2, 3N.A.N.A.N.A.LIMERICK

-UNIT 23/4 3-28Amendment No. -14, 32?, 147 TABLE 4.3.2.1-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNELCHANNEL FUNCTIONAL TRIP FUNCTION CHECK (a) TEST (a)CHANNELCALIBRATION(a)

OPERATIONAL CONDITIONS FOR WHICHSURVEILLANCE REQUIRED5. REACTOR CORE ISOLATION COOLTNG SYUTFM TrSlATTnN

a. RCIC Steam LineA Pressure

-Highb. RCIC Steam SupplyPressure

-Lowc. RCIC Turbine Exhaust Diaphragm Pressure

-Highd. RCIC Equipment RoomTemperature

-Highe. RCIC Equipment RoomA Temperature

-Highf. RCIC Pipe Routing AreaTemperature

-Highg. Manual Initiation

h. RCIC Steam LineA Pressure Timer1, 2, 31 2, 31 2, 31 2, 31 2, 31 2, 31, 2, 31, 2, 3N.A.N.A.N.A.LIMERICK

-UNIT 23/4 3-29Amendment No. -321, 147 TABLE 4.3.2.1-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICHTRIP FUNCTION CHECK (a) TEST (a) CALIBRATION(a)

SURVEILLANCE REQUIRED6. PRIMARY CONTAINMENT ISOLATION

a. Reactor Vessel Water Level1) Low, Low -Level 2 1, 2, 32) Low, Low, Low -Level 1 1, 2, 3b. Drywell Pressure

1. -High 1, 2, 3c. North Stack EffluentRadiation

-High 1, 2, 3d. Deletede. Reactor Enclosure Ventilation Exhaust Duct -Radiation

-High 1, 2, 3f. Deletedg. Deletedh. Drywell Pressure

-High/Reactor Pressure

-Low 1, 2, 3i. Primary Containment Instrument Gas to Drywell A Pressure

-Low N.A. 1, 2, 3j. Manual Initiation N.A. N.A. 1, 2, 3LIMERICK

-UNIT 23/4 3-30Amendment No. , B?, 4, 147 0TABLE 4.3.2.1-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNELCHANNEL FUNCTIONAL CHANNELOPERATIONAL CONDITIONS FOR WHICHSURVEILLANCE REQUIREDTRIP FUNCTION7. SECONDARY CONTAINMENT ISOLATION

a. Reactor Vessel Water LevelLow, Low -Level 2b. Drywell Pressure##

-Highc.1. Refueling Area Unit 1 Ventilation Exhaust Duct Radiation

-High2. Refueling Area Unit 2 Ventilation Exhaust Duct Radiation

-Highd. Reactor Enclosure Ventilation Exhaust Duct Radiation

-Highe. Deletedf. Deletedg. Reactor Enclosure Manual Initiation

h. Refueling AreaManual Initiation CHECK(a)

TEST(a) CALIBRATION(a 1,2, 32, 31, 2, 3N.A.N.A.N.A.N.A.1, 2, 3(a) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted*Required when (1) handling RECENTLY IRRADIATED FUEL in the secondary containment, or (2) duringpotential for draining the reactor vessel with the vessel head removed and fuel in the vessel.**When not administratively bypassed and/or when any turbine stop valve is open.#During operation of the associated Unit 1 or Unit 2 ventilation exhaust system.##These trip functions (2a, 6b, and 7b) are common to the RPS actuation trip function.

in the table.operations with aLIMERICK

-UNIT 23/4 3-31Amendment No. 4-3, -5,2, --4, 4r, 147 INSTRUMENTATION 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3 The emergency core cooling system (ECCS) actuation instrumentation channels shown in Table 3.3.3-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.3-2and with EMERGENCY CORE COOLING SYSTEM RESPONSE TIME as shown in Table 3.3.3-3.APPLICABILITY:

As shown in Table 3.3.3-1ACTION:a. With an ECCS actuation instrumentation channel trip setpoint lessconservative than the value shown in the Allowable Values column ofTable 3.3.3-2, declare the channel inoperable until the channel isrestored to Operable status with its trip setpoint adjusted consistent with the Trip Setpoint value.b. With one or more ECCS actuation instrumentation channels inoperable, take the ACTION required by Table 3.3.3-1.c. With either ADS trip system subsystem inoperable, restore theinoperable trip system to OPERABLE status within:1. 7 days, provided that the HPCI and RCIC systems are OPERABLE.

2. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .Otherwise, be in at least HOT SHUTDOWN within the next 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months sand reduce reactor steam dome pressure to less than or equal to100 psig within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .SURVEILLANCE REOUIREMENTS 4.3.3.1 Each ECCS actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST andCHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS shown in Table4.3.3.1-1 and at the frequencies specified in the Surveillance Frequency ControlProgram unless otherwise noted in Table 4.3.3.1-1.

4.3.3.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation ofall channels shall be performed in accordance with the Surveillance Frequency Control Program.4.3.3.3 The ECCS RESPONSE TIME of each ECCS trip function shown in Table 3.3.3-3shall be demonstrated to be within the limit in accordance with the Surveillance Frequency Control Program.

Each test shall include at least one channel per trip.system such that all channels are tested at least once every N times the frequency specified in the Surveillance Frequency Control Program where N is the totalnumber of redundant channels in a specific ECCS trip system.LIMERICK

-UNIT 23/4 3-32Amendment No. -4, 147 0TABLE 3.3.3-1EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION MINIMUM OPERABLECHANNELS PERTRIPTRIP FUNCTION FUNCTION' APPLICABLE OPERATIONAL CONDITIONS ACTION1. CORE SPRAY SYSTEM***

a. Reactor Vessel Water Level -Low Low Low, Level 1b. Drywell Pressure

-Highc. Reactor Vessel Pressure

-Low (Permissive)

d. Manual Initiation

2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM***

2/pumpbl) 2/pumplbl 1, 2, 3, 4*, 5*1, 2, 3,1, 2, 34*, 5*1, 2, 3, 4*, 5*6(b)2(e)a.b.C.d.Reactor Vessel Water Level -Low Low Low, Level 1Drywell Pressure

-HighReactor Vessel Pressure

-Low (Permissive)

Injection Valve Differential Pressure-Low (Permissive) 2221/valve1,1,1,1,2,2,2,2,3, 4*, 5*333, 4*, 5*30303132333030313133343435353133e. Manual Initiation 11, 2, 3, 4*, 5*3. HIGH PRESSURE COOLANT INJECTION SYSTEMwa .b.C.d.e.f.Reactor Vessel Water Level -Low Low, Level 2Drywell Pressure

-HighCondensate Storage Tank Level LowSuppression Pool Water Level HighReactor Vessel Water Level -High, Level 8Manual Initiation 442 (c)24(d)1/system1 2, 31 2, 31 2, 31 2, 31 2, 31 2, 3LIMERICK

-UNIT 23/4 3-33 TABLE 3.3.3-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION MINIMUM OPERABLECHANNELS PERTRIPFUNCTION'd' APPLICABLE OPERATIONAL CONDITIONS TRIP FUNCTIONACTION4. AUTOMATIC DEPRESSURIZATION SYSTEM#***

a.b.C.d.e.f.g.h.Reactor Vessel Water Level -Low Low Low, Level 1Drywell Pressure

-HighADS TimerCore Spray Pump Discharge Pressure

-High (Permissive)

RHR LPCI Mode Pump Discharge Pressure High(Permissive)

Reactor Vessel Water Level -Low, Level 3 (Permissive)

Manual Initiation ADS Drywell Pressure Bypass Timer221241221,1,1,1,1,1,1,2,2,2,2,2,2,2,2,333333333030313131313331TOTAL NO.OF CHANNELS(f)

MINIMUMCHANNELS CHANNELSTO TRIP OPERABLE5. LOSS OF POWER1. 4.16 kV Emergency Bus Under-voltage (Loss of Voltage)2. 4.16 kV Emergency Bus Under-voltage (Degraded Voltage)APPLICABLE OPERATIONAL CONDITIONS 1, 2, 3, 4**, 5**1, 2, 3, 4**, 5**ACTION1/bus1/source/

bus1/bus1/bus3637I/source/

I/source/

bus bus***The Minimum OPERABLE Channels Per Trip Function is per subsystem.

LIMERICK

-UNIT 23/4 3-340 TABLE 3.3.3-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION TABLE NOTATIONS (a) 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 forrequired surveillance without placing the trip system in the trippedcondition provided at least one OPERABLE channel in the same tripsystem is monitoring that parameter.

(b) Also provides input to actuation logic for the associated emergency diesel generators.

(c) One trip system. Provides signal to HPCI pump suction valves only.(d) On 1 out of 2 taken twice logic, provides a signal to trip the HPCIpump turbine only.(e) The manual initiation push buttons start the respective core spray pumpand diesel generator.

The "A" and "B" logic manual push buttons alsoactuate an initiation permissive in the injection valve opening logic.(f) A channel as used here is defined as the 127 bus relay for Item 1. andthe 127, 127Y, and 127Z feeder relays with their associated time delayrelays taken together for Item 2.* When the system is required to be OPERABLE per Specification 3.5.2.# Not required to be OPERABLE when reactor steam dome pressure is lessthan or equal to 100 psig.** Required when ESF equipment is required to be OPERABLE.

S Not required to be OPERABLE when reactor steam dome pressure is lessthan or equal to 200 psig.LIMERICK

-UNIT 23/4 3-35Amendment No. 17 ACTION 30 -ACTION 31 -ACTION 32 -ACTION 33 -ACTION 34 -TABLE 3.3.3-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ACTION STATEMENTS With the number of OPERABLE channels less than required by theMinimum OPERABLE Channels per Trip Function requirement:

a. With one channel inoperable, place the inoperable channel inthe tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the associated system inoperable.

b. With more than one channel inoperable, declare the associated system inoperable.

With the number of OPERABLE channels less than required by theMinimum OPERABLE Channels per Trip Function requirement, declare theassociated ECCS inoperable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .With the number of OPERABLE channels less than required by theMinimum OPERABLE Channels per Trip Function requirement, place theinoperable channel in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .With the number of OPERABLE channels less than required by theMinimum OPERABLE Channels per Trip Function requirement, restore theinoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare theassociated ECCS inoperable.

With the number of OPERABLE channels less than required by theMinimum OPERABLE Channels per Trip Function requirement:

a. For one channel inoperable, place the inoperable channel in thetripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the HPCI systeminoperable.

b. With more than one channel inoperable, declare the HPCI systeminoperable.

With the number of OPERABLE channels less than required by theMinimum OPERABLE Channels per Trip Function requirement, place atleast one inoperable channel in the tripped condition within 24hours or declare the HPCI system inoperable.

With the number of OPERABLE channels less than the Total Number ofChannels, declare the associated emergency diesel generator and theassociated offsite source breaker that is not supplying the businoperable and take the ACTION required by Specification 3.8.1.1 or3.8.1.2, as appropriate.

ACTION 35 -ACTION 36 -LIMERICK

-UNIT 23/4 3-36Amendment No. 4-7, 120 TABLE 3.3.3-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ACTION STATEMENTS With the number of OPERABLE channels one less than the Total Numberof Channels, place the inoperable device in the bypassed condition subject to the following conditions:

ACTION 37Inoperable DeviceCondition 127-1IXOX 127Y-11XOX 127Z-IIXOX 127Y-11XOX and 127Z-11XOX operable127-11XOX and 127Z-11XOX operable127-11XOX and 127Y-11XOX operable.

127Z-11YOY operable for the other 3 breakersmonitoring that source, offsite source gridvoltage for that source is maintained at orabove 230kV (for the 101 Safeguard Bus Source)or 525kV (for the 201 Safeguard Bus Source),Load Tap Changer for that source is in serviceand in automatic operation, and the electrical buses and breaker alignments are maintained within bounds of approved plant procedures.

or, place the inoperable channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months andtake the Action required by Specification 3.8.1.1 or 3.8.1.2, as appropriate.

Operation may then continue until performance of the next required CHANNELFUNCTIONAL TEST.LIMERICK

-UNIT 2 3/4 3-36a Amendment No. 120 INTENTIONALLY LEFT BLANK 0TABLE 3.3.3-2FMFRGFNCY CORF Ol1 ITNG SYSTFM ACT[]ATION INSTRUMENTATION SETPOINTS ALLOWABLE VALUETRIP FUNCTIONTRIP SETPOINT1. CORE SPRAY SYSTEMa.b.C.d.Reactor Vessel Water Level -Low Low Low, Level 1Drywell Pressure

-HighReactor Vessel Pressure

-LowManual Initiation

> -129 inches** 1.68 psig* 455 psig,(decreasing)

N.A.> -136 inches< 1.88 psig> 435 psig, (decreasing)

N.A.2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEMa.b.c.d.e.Reactor Vessel Water Level -Low Low Low, Level 1Drywell Pressure

-HighReactor Vessel Pressure

-LowInjection Valve Differential Pressure

-LowManual Initiation

3. HIGH PRESSURE COOLANT INJECTION SYSTEMabcdefReactor Vessel Water Level -Low Low, Level 2Drywell Pressure

-HighCondensate Storage Tank Level LowSuppression Pool Water Level HighReactor Vessel Water Level -High, Level 8Manual Initiation

> -129 inches** 1.68 psig> 455 psig,(decreasing)

> 74 psid, (decreasing)

N.A.> -38 inches*< 1.68 psig> 167.8 inches**< 24 feet 1.5 inches< 54 inchesN.A.> -129 inches*< 1.68 psig< 105 seconds> 145 psig,(increasing)

> 125 psig,(increasing)

> 12.5 inchesN.A.< 420 seconds> -45 inches< 1.88 psig> 164.3 inches< 24 feet 3 inches< 60 inches -N.A.> -136 inches< 1.88 psig> 435 psig, (decreasing)

> 64 psid and < 84 psidN.A.4. AUTOMATIC DEPRESSURIZATION SYSTEMabcdefg.h.Reactor Vessel Water Level -Low Low Low, Level 1Drywell Pressure

-HighADS TimerCore Spray Pump Discharge Pressure

-HighRHR LPCI Mode Pump Discharge Pressure-High Reactor Vessel Water Level-Low, Level 3Manual Initiation ADS Drywell Pressure Bypass Timer> -136 inches< 1.88 psig< 117 seconds> 125 psig, (increasing),

> 115 psig, (increasing)

> 11.0 inchesN.A.< 450 seconds*See Bases Figure B 3/4.3-1.**Corresponds to 2.3 feet indicated.

LIMERICK

-UNIT 23/4 3-37 TABLE 3.3.3-2 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE VALUETRIP FUNCTION5. LOSS OF POWERa. 4.16 kV Emergency Bus(Loss of Voltage)b. 4.16 kV Emergency Bus(Degraded Voltage)TRIP SETPOINTRELAYUndervoltage Undervol tageNANA127-11XRELAY127-11XOX and102-11XOX 127Y-11XOX**

and127Y-1-11XOX 127Z-11XOX and162Y-11XOX

a. 4.16 kV Basis2905 +/- 115 voltsb. 120 V Basis83 +/- 3 voltsc. < 1 second timedelaya. 4.16 kV Basis3640 +/- 91 voltsb. 120 V Basis104 +/- 3 voltsc. < 52 second timedelaya. 4.16 kV Basis3910 +/- 11 voltsb. 120 V Basis111.7 +/- 0.3 voltsc. < 10 second timedelaya. 4.16 kV Basis3910 +/- 11 voltsb. 120 V Basis111.7 +/- 0.3 voltsc. < 61 second timedelay2905 +/- 145 volts83 +/- 4 volts< 1.5 second timedelay3640 +/- 182 volts104 +/- 5.2 volts< 60 second timedelay3910 +/- 19 volts111.7 +/- 0.5 volts< 11 second timedelay3910 +/- 19 volts111.7 +/- 0.5 volts< 64 second timedelay127Z-11XOX and162Z-11XOX

- This is an inverse time delay voltage relay. The voltages shown area trip. Some voltage conditions will result in decreased trip timesthe maximum that will not result inLIMERICK

-UNIT 23/4 3-38 I.2.3.4.5.TABLE 3.3.3-3EMERGENCY CORE COOLING SYSTEM RESPONSE TIMESRESPONSE TIME (Seconds)

CORE SPRAY SYSTEM < 27#LOW PRESSURE COOLANT INJECTION MODEOF RHR SYSTEM < 40#-AUTOMATIC DEPRESSURIZATION SYSTEM N.A.HIGH PRESSURE COOLANT INJECTION SYSTEM < 60#LOSS OF POWER N.A.# ECCS actuation instrumentation is eliminated from response time testing.LIMERICK

-UNIT 23/4 3-39Amendment No. &&, 93 TABLE 4.3.3.1-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNELCHANNEL FUNCTIONAL TRIP FUNCTION CHECK(a)

TEST (a)CHANNELCALIBRATION(a)

OPERATIONAL CONDITIONS FOR WHICHSURVEILLANCE REQUIRED1. CORE SPRAY SYSTEMa. Reactor Vessel Water Level -Low Low Low, Level 1b. Drywell Pressure

-Highc. Reactor Vessel Pressure

-Lowd. Manual Initiation

2. LOW PRESSURE COOLANT INJECTION MODE1,1,1,1,2,2,2,2,3, 4*, 5*33, 4*, 5*3, 4*, 5*N.A.N.A.(OF RHR %YSTFMOF RHR SYSTEMa. Reactor Vessel Water Level -Low Low Low, Level 1b. Drywell Pressure

-Highc. Reactor Vessel Pressure

-Lowd. Injection Valve Differential Pressure

-Low (Permissive)

e. Manual Initiation 1,1,2,2,2,3, 4*, 5*331, 2, 3, 4*, 5*1, 2, 3, 4*, 5*N.A.N.A.3. HIGH PRESSURE COOLANT INJECTION SYSTEM***

a. Reactor Vessel Water Level -Low Low, Level 2b. Drywell Pressure

-Highc. Condensate Storage Tank LevelLowd. Suppression Pool Water LevelHighe. Reactor Vessel Water Level -High, Level 8f. Manual Initiation 1, 2, 31, 2, 31, 2, 31, 2, 31, 2, 31, 2, 3N.A.N.A.LIMERICK

-UNIT 23/4 3-40Amendment No. 4-7-, -34, 147 0TABLE 4.3.3.1-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNELCHANNEL FUNCTIONAL TRIP FUNCTION CHECK (a) TEST (a)CHANNELCALIBRATION(a)

OPERATIONAL CONDITIONS FOR WHICHSURVEILLANCE REQUIRED4. AUTOMATIC DEPRESSURIZATION SYSTEM#a. Reactor Vessel Water Level -Low Low Low, Level 1b. Drywell Pressure

-Highc. ADS Timerd. Core Spray Pump Discharge Pressure

-Highe. RHR LPCI Mode Pump Discharge Pressure

-Highf. Reactor Vessel Water Level -Low,Level 3g. Manual Initiation

h. ADS Drywell Pressure Bypass Timer1,1,1,2, 32, 32, 3N.A.1, 2, 31, 2, 3N.A.N.A.N.A.1,1,2, 32, 32, 35. LOSS OF POWERa. 4.16 kV Emergency Bus Undervoltage (Loss of Voltage)##

N.A.N.A.b. 4.16 kV Emergency Bus Under-voltage (Degraded Voltage)(a) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in* When the system is required to be OPERABLE per Specification 3.5.2.** Required OPERABLE when ESF equipment is required to be OPERABLE.

Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 200 psig.# Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 100 psig.## Loss of Voltage Relay 127-11X is not field setable.1, 2, 3, 4**, 5**1, 2, 3, 4**, 5**the table.LIMERICK

-UNIT 23/4 3-41Amendment No. -1, 147 INSTRUMENTATION 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.4.1 The anticipated transient without scram recirculation pump trip(ATWS-RPT) system instrumentation channels shown in Table 3.3.4.1-1 shall beOPERABLE with their trip setpoints set consistent with values shown in the TripSetpoint column of Table 3.3.4.1-2.

APPLICABILITY:

OPERATIONAL CONDITION 1.ACTION:a. With an ATWS recirculation pump trip system instrumentation channeltrip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4.1-2, declare the channel inoperable untilthe channel is restored to OPERABLE status with the channel tripsetpoint adjusted consistent with the Trip Setpoint value.b. With the number of OPERABLE channels one less than required by theMinimum OPERABLE Channels per Trip System requirement for one orboth trip systems, place the inoperable channel(s) in the trippedcondition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .c. With the number of OPERABLE channels two or more less than requiredby the Minimum OPERABLE Channels per Trip System requirement for onetrip system and:1. If the inoperable channels consist of one reactor vessel waterlevel channel and one reactor vessel pressure

channel, place bothinoperable channels in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or,if this action will initiate a pump trip, declare the trip systeminoperable.

2. If the inoperable channels include two reactor vessel water levelchannels or two reactor vessel pressure

channels, declare thetrip system inoperable.

d. With one trip system inoperable, restore the inoperable trip systemto OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least STARTUP withinthe next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .e. With both trip systems inoperable, restore at least one trip systemto OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least STARTUP withinthe next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .SURVEILLANCE REQUIREMENTS 4.3.4.1.1 Each of the required ATWS recirculation pump trip system instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK,CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies specified in the Surveillance Frequency Control Program.4.3.4.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation ofall channels shall be performed in accordance with the Surveillance Frequency Control Program.LIMERICK

-UNIT 23/4 3-42Amendment No. 3-3, 34, 147 TABLE 3.3.4.1-1 ATWS RFCTRCtJLATION PIIMP TRIP SYSTEM INSTRUMENTATION TRIP FUNCTIONMINIMUM OPERABLE CHANNELS PERTRIP SYSTEM *1. Reactor Vessel Water Level -Low Low, Level 22. Reactor Vessel Pressure

-High22* One 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 forrequired surveillance provided the other channel is OPERABLE.

LIMERICK

-UNIT 23/4 3-43Amendment No. 33 TABLE 3.3.4.1-2 ATWS RECIRCULATION PHMP TRIP SYSTFM TNVTRIJMFNTATTIN SýFTPnTNTS TRIP FUNCTION1. Reactor Vessel, Water Level -Low Low, Level 22. Reactor Vessel Pressure

-HighTRIPSETPOINT> -38 inches** 1149 psigALLOWABLE VALUE> -45 inches< 1156 psig* See Bases Figure B3/4.3-1.

LIMERICK

-UNIT 23/4 3-44Amendment No. 51 INFORMATION ON THIS PAGE HAS BEEN DELETEDLIMERICK

-UNIT 23/4 3-45Amendment No. 3-3, 147 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.4.2 The end-of-cycle recirculation pump trip (EOC-RPT) systeminstrumentation channels shown in Table 3.3.4.2-1 shall be OPERABLE withtheir trip setpoints set consistent with the values shown in the Trip Setpointcolumn of Table 3.3.4.2-2 and with the END-OF-CYCLE RECIRCULATION PUMP TRIPSYSTEM RESPONSE TIME as shown in Table 3.3.4.2-3.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than orequal to 29.5% of RATED THERMAL POWER.ACTION:a. With an end-of-cycle recirculation pump trip system instrumentation channel trip setpoint less conservative than the value shown in theAllowable Values column of Table 3.3.4.2-2, declare the channelinoperable until the channel is restored to OPERABLE status with thechannel setpoint adjusted consistent with the Trip Setpoint value.b. With the number of OPERABLE channels one less than required by theMinimum OPERABLE Channels per Trip System requirement for one or bothtrip systems, place the inoperable channel(s) in the tripped condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .c. With the number of OPERABLE channels two or more less than requiredby the Minimum OPERABLE Channels per Trip System requirement for onetrip system and:1. If the inoperable channels consist of one turbine control valvechannel and one turbine stop valve channel, place both inoperable channels in the tripped condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .2. If the inoperable channels include two turbine control valvechannels or two turbine stop valve channels, declare the tripsystem inoperable.

d. With one trip system inoperable, restore the inoperable trip systemto OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or take the ACTION required bySpecification 3.2.3.e. With both trip systems inoperable, restore at least one trip systemto OPERABLE status within one hour or take the ACTION required bySpecification 3.2.3.0LIMERICK

-UNIT 23/4 3-46Amendment No. -4, 163 INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.4.2.1 Each of the required end-of-cycle recirculation pump trip systeminstrumentation channels shall be demonstrated OPERABLE by the performance ofthe CHANNEL FUNCTIONAL TEST, including trip system logic testing, and CHANNELCALIBRATION operations at the frequencies specified in the Surveillance Frequency Control Program.4.3.4.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation ofall channels shall be performed in accordance with the Surveillance Frequency Control Program.4.3.4.2.3 The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME ofeach trip function shown in Table 3.3.4.2-3 shall be demonstrated to be withinits limit in accordance with the Surveillance Frequency Control Program.

Eachtest shall include at least the logic of one type of channel input, turbinecontrol valve fast closure or turbine stop valve closure, such that both types ofchannel inputs are tested in accordance with the Surveillance Frequency ControlProgram.

The measured time shall be added to the most recent breaker arcsuppression time and the resulting END-OF-CYCLE-RECIRCULATION PUMP TRIP SYSTEMRESPONSE TIME shall be verified to be within its limit.4.3.4.2.4 The time interval necessary for breaker arc suppression from energi-zation of the recirculation pump circuit breaker trip coil shall be measured inaccordance with the Surveillance Frequency Control Program.LIMERICK

-UNIT 23/4 3-47Amendment No. -4, 147 TABLE 3.3.4.2-1 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTR[JMENTATION 0MINIMUMOPERABLE CHANNELSPER TRIP SYSTEM*TRIP FUNCTION1. Turbine Stop Valve -Closure 2**2. Turbine Control Valve-Fast Closure 2*** A trip system 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 forrequired surveillance provided that the other trip system is OPERABLE.

- This function shall be automatically bypassed when turbine first stagepressure is equivalent to THERMAL POWER LESS than 29.5% of RATED THERMAL POWER.LIMERICK

-UNIT 23/4 3-48Amendment No. 3-3, 163 TABLE 3.3.4.2-2 END-OF-CYCLE RECIRCULATION PUMP TRIP SETPOINTS TRIP FUNCTION1. Turbine Stop Valve-Closure

2. Turbine Control Valve-Fast ClosureTRIP SETPOINT< 5% closed> 500 psigALLOWABLE VALUE< 7% closed* 465 psigLIMERICK

-UNIT 23/4 3-49 TABLE 3.3.4.2-3 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIMETRIP FUNCTION1. Turbine Stop Valve-Closure

2. Turbine Control Valve-Fast ClosureRESPONSE TIME (Milliseconds)

< 175< 175LIMERICK

-UNIT 23/4 3-50 INFORMATION ON THIS PAGE HAS BEEN DELETEDLIMERICK

-UNIT 23/4 3-51Amendment No. 3-3, 147 INSTRUMENTATION 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.5 The reactor core isolation cooling (RCIC) system actuation instrumentation channels shown in Table 3.3.5-1 shall be OPERABLE with theirtrip setpoints set consistent with the values shown in the Trip Setpointcolumn of Table 3.3.5-2.APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3 with reactor steamdome pressure greater than 150 psig.ACTION:a. With a RCIC system actuation instrumentation channel trip setpointless conservative than the value shown in the Allowable Valuescolumn of Table 3.3.5-2, declare the channel inoperable until thechannel is restored to OPERABLE status with its trip setpointadjusted consistent with the Trip Setpoint value.b. With one or more RCIC system actuation instrumentation channelsinoperable, take the ACTION required by Table 3.3.5-1.SURVEILLANCE REQUIREMENTS 4.3.5.1 Each of the required RCIC system actuation instrumentation channelsshall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNELFUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies specified in the Surveillance Frequency Control Program.

CHANNEL CHECK and CHANNELCALIBRATION are not required for manual initiation.

4.3.5.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation ofall channels shall be performed in accordance with the Surveillance Frequency Control Program.LIMERICK

-UNIT 23/4 3-52Amendment No. 34, 147 TABLE 3.3.5-1REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION FUNCTIONAL UNITSa. Reactor Vessel Water Level -Low Low, Level 2b. Reactor Vessel Water Level -High, Level 8c. Condensate Storage Tank WaterLevel -Lowd. Manual Initiation MINIMUMOPERABLE CHANNELSPER TRIP FUNCTION*

4#4#2"*I/system***

ACTION50515253*A channel may be placed in an inoperable status for up to 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months sfor required surveillance without placing the trip system in thetripped condition provided all other channels monitoring thatparameter are OPERABLE.

- One trip system with one-out-of-two logic.***One trip system with one channel.#One trip system with one-out-of-two twice logic.Amendment No. 17 1LIMERICK

-UNIT 23/4 3-53 ACTION 50TABLE 3.3.5-1 (Continued)

REACTOR CORE ISOLATION COOLING SYSTEMACTION STATEMENTS With the number of OPERABLE channels less than required by theMinimum OPERABLE Channels per Trip Function requirement:

a. With one channel inoperable, place the inoperable channel inthe tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the RCICsystem inoperable.

b. With more than one channel inoperable, declare the RCIC systeminoperable.

With the number of OPERABLE channels less than required by theminimum OPERABLE channels per Trip System requirement, declare theRCIC system inoperable within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .With the number of OPERABLE channels less than required by theMinimum OPERABLE Channels per Trip System requirement, place atleast one inoperable channel in the tripped condition within 24hours or declare the RCIC system inoperable.

With the number of OPERABLE channels one less than required by theMinimum OPERABLE channels per Trip System requirement, restore theinoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare theRCIC system inoperable.

ACTION 51ACTION 52ACTION 53 -LIMERICK

-UNIT 23/4 3- 54Amendment No. 17 1 0 TABLE 3.3.5-2REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE FUNCTIONAL UNITS TRIP SETPOINT VALUEa. Reactor Vessel Water Level -Low Low, Level 2 >-38 inches* >-45 inchesb. Reactor Vessel Water Level -High, Level 8 < 54 inches < 60 inchesc. Condensate Storage Tank Level -Low > 135.8"* inches > 132.3 inchesd. Manual Initiation N.A. N.A.*See Bases Figure B 3/4 3-1.**Corresponds to 2.3 feet indicated.

LIMERICK

-UNIT 23/4 3-55 INFORMATION ON THIS PAGE HAS BEEN DELETEDLIMERICK

-UNIT 23/4 3-56Amendment No. -1, 147 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 3.3.6-1shall be OPERABLE with their trip setpoints set consistent with the valuesshown in the Trip Setpoint column of Table 3.3.6-2.APPLICABILITY:

As shown in Table 3.3.6-1.ACTION:a. With a control rod block instrumentation channel trip setpoint**

lessconservative than the value shown in the Allowable Values column ofTable 3.3.6-2, declare the channel inoperable until the channel isrestored 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 MinimumOPERABLE Channels per Trip Function requirement, take the ACTIONrequired by Table 3.3.6-1.SURVEILLANCE REQUIREMENTS 4.3.6 Each of the above required control rod block trip systems andinstrumentation channels shall be demonstrated OPERABLE*

by the performance ofthe CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS shown in Table 4.3.6-1 and at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted inTable 4.3.6-1.*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 requiredsurveillance without placing the trip system in the tripped condition, providedat least one other operable channel in the same trip system is monitoring thatparameter.

- The APRM Simulated Thermal Power -Upscale Functional Unit need not be declaredinoperable upon entering single reactor recirculation loop operation provided that theflow-biased setpoints are adjusted within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months per Specification 3.4.1.1.LIMERICK

-UNIT 23/4 3-57Amendment No. 3, Th 147 TABLE 3.3.6-1CONTROL ROD BLOCK INSTRUMENTATION MINIMUM APPLICABLE OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION PER TRIP FUNCTION CONDITIONS ACTION1. ROD BLOCK MONITOR(a)

a. Upscale 2 1* 60b. Inoperative 2 1* 60c. Downscale 2 1* 602. APRMa. Simulated Thermal Power -Upscale 3 1 61b. Inoperative 3 1, 2 61c. Neutron Flux -Downscale 3 1 61d. Simulated Thermal Power -Upscale (Setdown) 3 2 61e. Recirculation Flow -Upscale 3 1 61f. LPRM Low Count 3 1, 2 613. SOURCE RANGE MONITORS

a. Detector not full in (b) 3 2 612 5 61b. Upscale(c) 3 2 612 5 61c. Inoperative(c) 3 2 612 5 61d. Downscale d) 3 2 612 5 614. INTERMEDIATE RANGE MONITORSa. Detector not full in 6 2, 5** 61b. Upscale 6 2, 5** 61c. Inoperative 6 2, 5** 61d. Downscale'e) 6 2, 5** 615. SCRAM DISCHARGE VOLUMEa. Water Level-High 2 1, 2, 5** 626. DELETED DELETED DELETED DELETED7. REACTOR MODE SWITCH SHUTDOWN POSITION 2 3, 4 63LIMERICK

-UNIT 23/4 3-58Amendment No. -7, 4-0-9, 139 TABLE 3.3.6-1 (Continued)

CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION ACTION STATEMENTS ACTION 60 -Declare the affected RBM channel inoperable and take the ACTIONrequired by Specification 3.1.4.3.ACTION 61 -With the number of OPERABLE Channels:

a. One less than required by the Minimum OPERABLE Channels per TripFunction requirement, restore the inoperable channel to OPERABLEstatus within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or place the inoperable channel in thetripped condition.

b. Two or more less than required by the Minimum OPERABLE Channelsper Trip Function requirement, place at least one inoperable channel in the tripped condition within one hour.ACTION 62 With the number of OPERABLE channels less than required by theMinimum OPERABLE Channels per Trip Function requirement, placethe inoperable channel in the tripped condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .ACTION 63 With the number of OPERABLE channels less than required by theMinimum OPERABLE Channels per Trip Function requirement, initiatea rod block.NOTES* For OPERATIONAL CONDITION of Specification 3.1.4.3.** With more than one control rod withdrawn.

Not applicable to control rodsremoved per Specification 3.9.10.1 or 3.9.10.2.

These channels are not required when sixteen or fewer fuel assemblies, adjacent to the SRMs, are in the core.(a) The RBM shall be automatically bypassed when a peripheral control rod isselected or the reference APRM channel indicates less than 30% ofRATED THERMAL POWER.(b) This function shall be automatically bypassed if detector count rate is> 100 cps or the IRM channels are on range 3 or higher.(c) This function is automatically bypassed when the associated IRM channelsare on range 8 or higher.(d) This function is automatically bypassed when the IRM channels are onrange 3 or higher.Ce) This function is automatically bypassed when the IRM channels are onrange 1.(f) DELETEDLIMERICK

-UNIT 23/4 3-59Amendment No. 3, 3-3, 49, 109 TABLE 3.3.6-2CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS TRIP SETPOINTTRIP FUNCTIONALLOWABLE VALUE1. ROD BLOCK MONITORa. Upscale(a)

1) Low Trip Setpoint (LTSP)2) Intermediate Trip Setpoint (ITSP)3) High Trip Setpoint (HTSP)******b.C.d.Inoperative Downscale (DTSP)Power Range Setpoint'b)

1) Low Power Setpoint (LPSP)2) Intermediate Power Setpoint (IPSP)3) High Power Setpoint (HPSP)N/AN/A2. APRMa. Simulated Thermal Power -Upscale:-Two Recirculation Loop Operation

-Single Recirculation Loop Operation****

28.1% RATED THERMAL POWER63.1% RATED THERMAL POWER83.1% RATED THERMAL POWER0.65 W + 54.3% and108.0% of RATEDTHERMAL POWER 0.65 (W-7.6%)

+ 54.1% and 108.0% of RATEDTHERMAL POWER28.4% RATED63.4% RATED83.4% RATEDTHERMAL POWERTHERMAL POWERTHERMAL POWER0.65 W + 54.7% and108.4% of RATEDTHERMAL POWER0.65 (W-7.6%)

+ 54.5% and108.4% of RATEDTHERMAL POWERb. Inoperative N.A.N.A.c. Neutron Flux -Downscale POWERd. Simulated Thermal Power -Upscale(Setdown)

e. Recirculation Flow -Upscale 3.2% of RATED THERMALPOWER 12.0% of RATED THERMALPOWER2.8% of RATED THERMAL13.0% of RATED THERMALPOWER**f. LPRM Low Count< 20 per channel< 3 per axial level3. SOURCE RANGE MONITORSa. Detector not full inb. Upscalec. Inoperative

d. Downscale N.A._< 1 x 105 cpsN.A._ 3 cps**< 20 per channel< 3 per axial levelN.A._< 1.6 x 10' cpsN.A.> 1.8 cps**LIMERICK

-UNIT 23/4 3-60Amendment No. 48,154,44-1-49, 163 0TABLE 3.3.6-2 (Continued)

CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS TRIP FUNCTIONTRIP SETPOINT4. INTERMEDIATE RANGE MONITORSa. Detector not full inb. Upscalec. Inoperative

d. Downscale

5. SCRAM DISCHARGE VOLUMEa. Water Level-High

a. Float SwitchN.A._< 108/125 divisions offull scaleN.A.> 5/125 divisions of fullscale_< 257' 7 3/8" elevation***

ALLOWABLE VALUEN.A._< 110/125 divisions offull scaleN.A.> 3/125 divisions of fullscale< 257' 9 3/8" elevation

6. DELETED7. REACTOR MODE SWITCH SHUTDOWNPOSITION* Refer to the COLR for these setpoints.

DELETEDDELETEDN.A.N.A.** May be reduced, provided the Source Range Monitor has an observedor above the curve shown in Figure 3.3.6-1.Equivalent to 13.56 gallons/scram discharge volume.** The 7.6% flow "offset" for Single Loop Operation (SLO) is applied(W-7.6%)

term is set equal to zero.count rate and signal-to-noise ratio onfor W _> 7.6%. For flows W < 7.6%, the(a) There are three upscale trip levels. Each is applicable only over its specified operating core thermalpower range. All RBM trips are automatically bypassed below the low power setpoint (LPSP). The upscaleLTSP is applied between the low power setpoint (LPSP) and the intermediate power setpoint (IPSP). Theupscale ITSP is applied between the intermediate power setpoint and the high power setpoint (HPSP).The HTSP is applied above the high power setpoint.

(b) Power range setpoints control enforcement of appropriate upscale trips over the proper core thermalpower ranges. The power signal to the RBM is provided by the APRM.LIMERICK

-UNIT 23/4 3-60aAmendment No. 4,4,49,449, 14-0, 139

[A3 .0 f 1 I I I I I I I I H2.02.8...........

f ill! IJ 11112.6llll I ffIiS2.21.8 I I I I I I I ____________

-1.6oOOI ~I I I .i-r-i ,__,___0.6 .....~~ -. ---...2 4 6 8 10 12 14 16 18 20 22 24 26 28 30SIGNAL-TO-NOISE RATIOSRM COUNT RATE VERSUS SIGNAL-TO-NOISE RATIOFigure 3.3.6-10LIMERICK

-UNIT 23/4 3-60bAmendment No. 3 0 0 0TABLE 4.3.6-1CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICHTRIP FUNCTION CHECK (h) TEST (h) CALIBRATION(a)(h)

SURVEILLANCE REQUIRED1. ROD BLOCK MONITORa. Upscale N.A. (c) 1*b. Inoperative N.A. (c) N.A. 1*c. Downscale N.A. (c) 1*2. APRMa. Simulated Thermal Power -Upscale N.A. 1b. Inoperative N.A. N.A. 1, 2c. Neutron Flux -Downscale N.A. 1d. Simulated Thermal Power -Upscale (Setdown)

N.A. 2e. Recirculation Flow -Upscale N.A. If. LPRM Low Count N.A. 1, 23. SOURCE RANGE MONITORSa. Detector not full in N.A. (e) N.A. 2, 5b. Upscale N.A. (e) 2, 5c. Inoperative N.A. (e) N.A. 2, 5d. Downscale N.A. (e) 2, 54. INTERMEDIATE RANGE MONITORSa. Detector not full in N.A. N.A. 2, 5**b. Upscale N.A. 2, 5**c. Inoperative N.A. N.A. 2, 5**d. Downscale N.A. 2, 5**5. SCRAM DISCHARGE VOLUMEa. Water Level -High N.A. 1, 2, 5**6. DELETED7. REACTOR MODE SWITCH SHUTDOWNPOSITION N.A. (g) N.A. 3, 4LIMERICK

-UNIT 2 3/4 3-61 Amendment No. -, , 4-8, -3, -9, 4-3-9, 147Crr.eted by letter- dated May 28, 2002 TABLE 4.3.6-1 (Continued)

CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATIONS (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b) Deleted.(c) Includes reactor manual control multiplexing system input.* For OPERATIONAL CONDITION of Specification 3.1.4.3.** With more than one control rod withdrawn.

Not applicable to controlrods removed per Specification 3.9.10.1 or 3.9.10.2.

- - Deleted.(d) Deleted(e) The provisions of Specification 4.0.4 are not applicable provided that thesurveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the IRMs are on Range 2 or belowduring a shutdown.

(f) Deleted(g) The provisions of Specification 4.0.4 are not applicable provided that thesurveillance is performed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after the Reactor Mode Switch has beenplaced in the shutdown position.

(h) Frequencies are specified in the Surveillance Frequency Control Program unlessotherwise noted in the table.LIMERICK

-UNIT 2 3/4 3-62 Amendment No. 5, 48, 4-3, 4-9-9, 147 INSTRUMENTATION 3/4.3.7 MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.7.1 The radiation monitoring instrumentation channels shown in Table3.3.7.1-1 shall be OPERABLE with their alarm/trip setpoints within the specified limits.APPLICABILITY:

As shown in Table 3.3.7.1-1.

ACTION:a. With a radiation monitoring instrumentation channel alarm/trip setpoint exceeding the value shown in Table 3.3.7.1-1, adjust thesetpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channelinoperable.

b. With one or more radiation monitoring channels inoperable, take theACTION required by Table 3.3.7.1-1.

c. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.7.1 Each of the above required radiation monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNELCHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for theconditions shown in Table 4.3.7.1-1 and at the frequencies specified in theSurveillance Frequency Control Program unless otherwise noted in Table4.3.7.1-1.

LIMERICK

-UNIT 23/4 3-63Amendment No. 147 TABLE 3.3.7.1-1 RADIATION MONITORING INSTRUMENTATION INSTRUMENTATION MINIMUM CHANNELSOPERABLEAPPLICABLE CONDITIONS ALARM/TRIP SETPOINTACTION1. Main Control Room NormalFresh Air Supply Radiation Monitor2. Area Monitorsa. Criticality Monitors41,2,3,and *1 x 10-, pCi/cc701) Spent FuelStorage Pool21(a)> 5 mR/h and 20mR/h(b)

b. Control Room DirectRadiation Monitor3. Reactor Enclosure CoolingWater Radiation MonitorAt All TimesAt All TimesN.A. (b)7173721< 3 x Background`b)

LIMERICK

-UNIT 23/4 3- 64Amendment No. 146 TABLE 3.3.7.1-1 (Continued)

RADIATION MONITORING INSTRUMENTATION TABLE NOTATIONS

When RECENTLY IRRADIATED FUEL is being handled in the secondary containment or duringoperations with a potential for draining the reactor vessel with the vessel headremoved and fuel in the vessel.(a) With fuel in the spent fuel storage pool.(b) Alarm only.ACTION STATEMENTS ACTION 70 With one monitor inoperable, restore the inoperable monitor tothe OPERABLE status within 7 days or, within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,initiate and maintain operation of the control room emergency filtration system in the radiation isolation mode of operation.

With two or more of the monitors inoperable, within one hour,initiate and maintain operation of the control room emergency filtration system in the radiation mode of operation.

ACTION 71 With one of the required monitor inoperable, assure a portablecontinuous monitor with the same alarm setpoint is OPERABLE inthe vicinity of the installed monitor during any fuel movement.

If no fuel movement is being made, perform area surveys of themonitored area with portable monitoring instrumentation at leastonce per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .ACTION 72 -With the required monitor inoperable, obtain and analyze atleast one grab sample of the monitored parameter at least onceper 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .ACTION 73 -With the required monitor inoperable, assure a portable alarmingmonitor is OPERABLE in the vicinity of the installed monitor orperform area surveys of the monitored area with portable monitor-ing instrumentation at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .LIMERICK

-UNIT 23/4 3-65Amendment No. 146 TABLE 4.3.7.1-1 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNELCHANNEL FUNCTIONAL INSTRUMENTATION CHECK(c)

TEST(c)CHANNELCALIBRATION(c)

OPERATIONAL CONDITIONS FORWHICH SURVEILLANCE IS REQUIRED1. Main Control Room NormalFresh Air Supply Radiation Monitor2. Area Monitorsa. Criticality Monitors1) Spent Fuel StoragePool1, 2, 3, and *(a)b. Control Room DirectRadiation MonitorAt All Times3. Reactor Enclosure CoolingWater Radiation Monitor (b)At All TimesLIMERICK UNIT 23/4 3-66Amendment No. --456, 147 TABLE 4.3.7.1-1 (Continued)

RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATIONS

When RECENTLY IRRADIATED FUEL is being handled in the secondary containment or duringoperations with a potential for draining the reactor vessel with the vessel headremoved and fuel in the vessel.(a) With fuel in the spent fuel storage pool.(b) The initial CHANNEL CALIBRATION shall be performed using one or moreof the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers thatparticipate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy andmeasurement range. For subsequent CHANNEL CALIBRATION, sources that havebeen related to the initial calibration shall be used.(c) Frequencies are specified in the Surveillance Frequency Control Programunless otherwise noted in the table.LIMERICK

-UNIT 23/4 3-67Amendment No. 4-46, 147 Section 3.3.7.2 (Deleted) 0THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE TRM. TECHNICAL SPECIFICATIONS PAGES 3/4 3-69 THROUGH 3/4 3-72 OF THISSECTION HAVE BEEN INTENTIONALLY OMITTED.LIMERICK

-UNIT 23/4 3-68Amendment No. -3, 153 Section 3.3.7.3 (Deleted)

THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEENRELOCATED TO THE ODCM. TECHNICAL SPECIFICATIONS PAGES 3/4 3-74 THROUGH3/4 3-75 OF THIS SECTION HAVEBEEN INTENTIONALLY OMITTED.LIMERICK

-UNIT 23/4 3-73Amendment No. 11 I INSTRUMENTATION REMOTE SHUTDOWN SYSTEM INSTRUMENTATION AND CONTROLSLIMITING CONDITION FOR OPERATION 3.3.7.4 The remote shutdown system instrumentation and controls shown inTable 3.3.7.4-1 shall be OPERABLE.

0APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.ACTION:a. With the number of OPERABLE remote shutdown system instrumentation channels less than required by Table 3.3.7.4-1, restore the inoperable channel(s) to OPERABLE status within 7 days or be in at least HOTSHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .b. With therequiredOPERABLEthe nextnumber of OPERABLE remote shutdown system controls less thanin Table 3.3.7.4-1, restore the inoperable control(s) tostatus within 7 days or be in at least HOT SHUTDOWN within12 hours.SURVEILLANCE REQUIREMENTS 4.3.7.4.1 Each of the above required remote shutdown monitoring instrumentation channels shall be demonstrated OPERABLE by performance of the CHANNEL CHECK*and CHANNEL CALIBRATION operations at the frequencies specified in theSurveillance Frequency Control Program.4.3.7.4.2 Each of the above remote shutdown control switch(es) and controlcircuits shall be demonstrated OPERABLE by verifying its capability to performits intended function(s) in accordance with the Surveillance Frequency ControlProgram.* Control is not required to be transferred to perform the CHANNEL CHECK.LIMERICK

-UNIT 23/4 3-76Amendment No. 34, 4-34, 147 TABLE 3.3.7.4-1 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION AND CONTROLSMINIMUMINSTRUMENTS INSTRUMENT OPERABLE1. Reactor Vessel Pressure

12. Reactor Vessel Water Level 13. Safety/Relief Valve Position, 3 valves 1/valve4. Suppression Chamber Water Level 15. Suppression Chamber Water Temperature

16. Drywell Pressure

17. Drywell Temperature

18. RHR System Flow 19. RHR Service Water Pump Discharge Pressure 110. RHR Heat Exchanger Service Water Outlet Pressure 111. RCIC System Flow 112. RCIC Turbine Speed 113. Emergency Service Water Pump Discharge Pressure 114. Condensate Storage Tank Level 115. RHR Heat Exchanger Bypass Valve (HV-C-51-2FO48A)

Position Indication (0 -100%) 116. RCIC Turbine Tripped Indication 117. RCIC Turbine Bearing Oil Pressure Low Indication 118. RCIC Bearing Oil Temperature High Indication 119. RHR Heat Exchanger Discharge Line High Radiation Indication LIMERICK

-UNIT 23/4 3- 77 TABLE 3.3.7.4-1 (Continued)

REMOTE SHUTDOWN SYSTEM CONTROLSRCIC SYSTEMHSS-49-291 HSS-49-292 HSS-49-293 HSS-49-295 HSS-49-296 HV-49-2F076 HV-49-2F060 HV-50-212 HV-50-2F045 HV-49-2FO08 HV-49-2FO07 HV-49-2F031 HV-49-2F029 HV-49-2F010 HV-49-2F019 HV-49-2F022 HV-50-2F046 HV-49-2F012 HV-49-2F013 20P22020P219HV-49-2FO02 Control-Transfer SwitchControl-Transfer SwitchControl-Transfer SwitchControl-Transfer SwitchControl-Transfer SwitchControl-Steam Line warmup bypass valveControl-RCIC turb exhaust to suppression poolisolation Control-Turb trip throttle valveControl-Turbine steam supply valveControl-Turbine steam line outboard isolation valveControl-Turbine steam line inboard isolation valveControl-RCIC pump suction from suppression poolControl-RCIC pump suction from suppression poolControl-RCIC pump suction from condensate storagetankControl-Minimum flow bypass to suppression poolControl-Test return to condensate storage tankControl-RCIC turbine cooling water valveControl-RCIC pump disch valveControl-RCIC pump disch valveControl-Vacuum tank condensate pumpControl-Barometric condenser vacuum pumpControl-Barometric condenser vacuum pump dischLIMERICK

-UNIT 23/4 3-78 Table 3.3.7.4-1 (Continued)

RCIC SYSTEM (Continued)

HV-49-2F080 Control-Vacuum breaker outboard isolation valveHV-49-2F084 Control-Vacuum breaker inboard isolation valveFIC-49-2RO01 Controller-RCIC discharge flow controlE51-S45 RCIC Turbine Trip BypassNUCLEAR BOILER SYSTEMHSS-41-291 Control-Transfer switchPSV-41-2FO13A PSV-41-2F013C PSV-41-2FO13N RHR SYSTEMHSS-51-195 HSS-51-196 HSS-51-292 HSS-51-293 HSS-51-294 HSS-51-295 HSS-51-296 HSS-51-297 HSS-51-298 HV-51-2FO09 HV-51-2FO08 HV-51-2FO06A HV-51-2FO06B HV-51-2FO04A 2AP202Control-Main steam lineControl-Main steam lineControl-Main steam linesafety/relief valvesafety/relief valvesafety/relief valveControl-Transfer switchControl-Transfer switchControl-Transfer switchControl-Transfer switchControl-Transfer switchControl-Transfer switchControl-Transfer switchControl-Transfer switchControl-Transfer switchControl-RHR pump shutdown cooling suction inboardisolation Control-RHR shutdown cooling suction outboardisolation Control-2A RHR loop shutdown cooling suctionControl-2B RHR loop shutdown cooling suctionControl-2A RHR pump suctionControl-2A RHR pumpLIMERICK

-UNIT 23/4 3-79 Table 3.3.7.4-1 (Continued)

RHR SYSTEM (Continued)

HV-43-2FO23A HSS-43-291 HV-51-2FO07A HV-51-2FO48A HV-51-2FO15A HV-51-2FO16A HV-51-2FO17A HV-51-2FO24A HV-51-2FO27A HV-51-2FO47A HV-51-2FO03A HV-51-2F049 HV-51-225A Control-Recirculation pump A suction valveControl-Transfer switchControl-2A RHR pump minimum flow bypass valveControl-2A heat exchanger shell side bypassControl-2A shutdown cooling injection valveControl-Reactor containment sprayControl-2A RHR loop injection valveControl-2A RHR loop test returnControl-Suppression pool sparger isolation Control-2A Heat exchanger shell side inletControl-2A Heat exchanger shell side outletControl-RHR Discharge to radwaste outboard isolation Control-2A/2C test return line to suppression poolRHR SERVICE W)HSS-12-015A-2 HSS-12-015C-2 HSS-12-016A-2 HSS-12-016C-2 ATER SYSTEMControl-Spray pond/cooling tower selectControl-Spray pond/cooling tower selectControl-Spray/bypass selectControl-Spray/bypass selectLIMERICK

-UNIT 23/4 3-80Amendment No. 47 Table 3.3.7.4-1 (Continued)

RHR SERVICE WATER SYSTEM (Continued)

HSS-12-094 Control-Transfer switchHSS-12-093 Control-Transfer switchHV-51-2FO14A Control-2A RHR heat exchanger tube side inletOCPSO6 Control-RHR Service Water pumpHV-51-2FO68A Control-2A RHR Heat exchanger tube side outletEMERGENCY SERVICEOAP548HV-11-011A HSS-11-091 HSS-11-092 HSS-11-093 WATER SYSTEMControl-A emergency serControl-A emergency ser\service waterControl-Transfer switchControl-Transfer switchControl-Transfer switchvice water pumpvice water disch to RHRThe following valves of the ESW and RHRSW systems are actuated by signals fromthe transfer switches:

HV-12-005 ESW and RHRSW pumps wetwell intertie gateHV-11-O15A ESW loop A discharge to RHRSW loop BHV-12-017A ESW and RHRSW cooling tower return cross-tie STANDBY AC POWER SUPPLY152-11509/CSR 101-D21 Safeguard 152-11609/CSR 101-D22 Safeguard 152-11709/CSR 101-D23 Safeguard 152-11502/CSR 201-021 Safeguard 152-11602/CSR 201-D22 Safeguard 152-11702/CSR 201-D23 Safeguard 152-11505/CSR D214 Safeguard LCSWGRSWGRSWGRSWGRSWGRSWGRXFMRfeeder bkr.feeder bkr.feeder bkr.feeder bkr.feeder bkr.feeder bkr.breakerLIMERICK

-UNIT 23/4 3-81 Table 3.3.7.4-1 (Continued)

STANDBY AC POWER SUPPLY (Continued) 152-11605/CSR D224 Safeguard LC XFMR breaker152-11705/CSR D234 Safeguard LC XFMR breaker143-115/CS Transfer switch143-116/CS Transfer switch143-117/CS Transfer switchLIMERICK

-UNIT 23/4 3-82 INFORMATION ON THIS PAGE HAS BEEN DELETEDLIMERICK

-UNIT 23/4 3-83Amendment No. 147 INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.7.5 The accident monitoring instrumentation channels shown in Table 3.3.7.5-1 shall be OPERABLE.

APPLICABILITY:

As shown in Table 3.3.7.5-1.

ACTION:With one or more accident monitoring instrumentation channels inoperable, takethe ACTION required by Table 3.3.7.5-1.

SURVEILLANCE REQUIREMENTS 4.3.7.5 Each of the above required accident monitoring instrumentation channelsshall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNELCALIBRATION operations at the frequencies specified in the Surveillance Frequency Control Program unless otherwise noted in Table 4.3.7.5-1.

LIMERICK

-UNIT 23/4 3-84Amendment No. 147 0TABLE 3.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION INSTRUMENT

1. Reactor Vessel Pressure2. Reactor Vessel Water Level3. Suppression Chamber Water Level4. Suppression Chamber Water Temperature REQUIRED NUMBEROF CHANNELS2228, 6 locations MINIMUMCHANNELSOPERABLE1116,1/location APPLICABLE OPERATIONAL CONDITIONS 1,21,21,21,2ACTION808080805.6.7.8.9.10.11.12.13.Del etedDrywell PressureDeletedDeletedDeletedDeletedPrimary Containment Post-LOCA Radiation MonitorsNorth Stack Wide Range Accident Monitor**

Neutron Flux211,28043*23*1,2,31,2,31,281818021LIMERICK

-UNIT 23/4 3-85Amendment No. 4-1--, 4--3-5, 4-4-], 152 Table 3.3.7.5-1 (Continued)

ACCIDENT MONITORING INSTRUMENTATION TABLE NOTATIONS

Three noble gas detectors with overlapping ranges (10. to 10', 10. to102, 10I to 10i pCi/cc).**High range noble gas monitor.ACTION STATEMENTS ACTION 80 -a. With the number of OPERABLE accident monitoring instrumentation channels less than the Required Number of Channels shown in Table3.3.7.5-1, restore the inoperable channel(s) to OPERABLE statuswithin 7 days or be in at least HOT SHUTDOWN within the next12 hours.b. With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum Channels OPERABLE requirements ofTable 3.3.7.5-1, restore the inoperable channel(s) to OPERABLEstatus within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT SHUTDOWN within thenext 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .ACTION 81 -With the number of OPERABLE accident monitoring instrumentation channels less than required by the Minimum Channels OPERABLErequirement, initiate the preplanned alternate method of monitor-ing the appropriate parameters within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , anda. Either restore the inoperable channel(s) to OPERABLE status within7 days of the event, orb. Prepare and submit a Special Report to the Commission pursuant toSpecification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans andschedule for restoring the system to OPERABLE status.ACTION 82 -DELETED0LIMERICK

-UNIT 23/4 3-86Amendment No. 4-14, 135 0TABLE 4.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL0CHANNELLIBRATION (a)INSTRUMENT

1. Reactor Vessel Pressure2. Reactor Vessel Water Level3. Suppression Chamber Water Level4. Suppression Chamber Water Temperature

5. Deleted6. Primary Containment Pressure7. Deleted8. Deleted9. Deleted10. Deleted11. Primary Containment Post LOCA Radiation Monitors12. North Stack Wide Range Accident Monitor***

13. Neutron FluxCHECK (a) CA**(a) Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.**CHANNEL CALIBRATION shall consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/h and a one point calibration check of the detector below 10 R/h with aninstalled or portable gamma source.***High range noble gas monitors.

LIMERICK

-UNIT 23/4 3-87Amendment No. 8, --3-5, 4-4-i, 7-, 152 INSTRUMENTATION SOURCE RANGE MONITORSLIMITING CONDITION FOR OPERATION 3.3.7.6 At least the following source range monitor channels shall be OPERABLE:

a. In OPERATIONAL CONDITION 2*, three.b. In OPERATIONAL CONDITION 3 and 4, two.APPLICABILITY:

OPERATIONAL CONDITIONS 2*#, 3, and 4.ACTION:a. In OPERATIONAL CONDITION 2* with one of the above required sourcerange monitor channels inoperable, restore at least three source rangemonitor channels to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at leastHOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .b. In OPERATIONAL CONDITION 3 or 4 with one or more of the above requiredsource range monitor channels inoperable, verify all insertable controlrods to be inserted in the core and lock the reactor mode switch inthe Shutdown position within I hour.SURVEILLANCE REOUIREMENTS 4.3.7.6 Each of the above required source range monitor channels shall bedemonstrated OPERABLE by:a. Performance of a:1. CHANNEL CHECK in accordance with the Surveillance Frequency Control Program:a) in CONDITION 2*, andb) in CONDITION 3 or 4.2. CHANNEL CALIBRATION**

in accordance with the Surveillance Frequency Control Program.b. Performance of a CHANNEL FUNCTIONAL TEST in accordance with theSurveillance Frequency Control Program.c. Verifying, prior to withdrawal of control rods, that the SRM countrate is at least 3.0 cps*** with the detector fully inserted.#

With IRM's on range 2 or below in CONDITION 2.**Neutron detectors may be excluded from CHANNEL CALIBRATION.

- - May be reduced, provided the source range monitor has an observed count rateand signal-to-noise ratio on or above the curve shown in Figure 3.3.6-1.#During initial startup test program, SRM detectors may be partially withdrawn prior to IRM on-scale indication provided that the SRM channelsremain on scale above 100 cps and respond to changes in the neutron flux.LIMERICK

-UNIT 23/4 3-88Amendment No. -, 34, 63, 147 INSTRUMENTATION Section 3/4.3.7.7 THE INFORMATION FROM THIS TECHNICAL SPECIFICATION HAS BEEN RELOCATED TO THE TECHNICAL REQUIREMENTS MANUAL (TRM)LIMERICK

-UNIT 23/4 3-89Amendment No. 79 INSTRUMENTATION CHLORINE DETECTION SYSTEMLIMITING CONDITION FOR OPERATION 3.3.7.8.1 Two independent chlorine detection system subsystems shall beOPERABLE with their alarm and trip setpoints adjusted to actuate at achlorine concentration of less than or equal to 0.5 ppmAPPLICABILITY:

All OPERATIONAL CONDITIONS.

ACTION:a. With one chlorine detection subsystem inoperable, restore theinoperable detection system to OPERABLE status within 7 days or,within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , initiate and maintain operation of at leastone control room emergency filtration system subsystem in thechlorine isolation mode of operation.

b. With both chlorine detection subsystem inoperable, within 1 hourinitiate and maintain operation of at least one control room emer-gency filtration system subsystem in the chlorine isolation mode ofoperation.

SURVEILLANCE REQUIREMENTS 4.3.7.8.1 Each of the above required chlorine detection system subsystems shallbe demonstrated OPERABLE by performance of a CHANNEL CHECK, CHANNEL FUNCTIONAL TEST,and CHANNEL CALIBRATION in accordance with the Surveillance Frequency Control Program.LIMERICK

-UNIT 23/4 3-90Amendment No. B-3, 44, 147 INSTRUMENTATION TOXIC GAS DETECTION SYSTEMLIMITING CONDITION FOR OPERATION 3.3.7.8.2 Three independent toxic gas detection system subsystems shall beOPERABLE with their alarm setpoints adjusted to actuate at a toxic gas concen-tration of less than or equal to:MONITORSET POINTCHEMICAL (Ppm)Ammonia 25Ethylene Oxide 50Formaldehyde 5Vinyl Chloride 10Phosgene 0.4APPLICABILITY:

All OPERATIONAL CONDITIONS.

ACTION:a. With one toxic gas detection subsystem inoperable, place the inoperable subsystem in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .b. With two toxic gas detection system subsystems inoperable, place oneinoperable subsystem in the tripped condition within I hour, restore oneinoperable detection subsystem to OPERABLE status within 7 days, orinitiate and maintain operation of at least one control room emergency filtration subsystem in the chlorine isolation mode of operation.

c. With three toxic gas detection subsystems inoperable, within 1 hourinitiate and maintain operation of at least one control room emergency filtration subsystem in the chlorine isolation mode of operation.

SURVEILLANCE REQUIREMENTS 4.3.7.8.2 Each of the above required toxic gas detection system subsystems shall be demonstrated OPERABLE by performance of a CHANNEL CHECK, CHANNELFUNCTIONAL TEST, and CHANNEL CALIBRATION in accordance with the Surveillance Frequency Control Program.LIMERICK

-UNIT 23/4 3-91Amendment No. 44, 4-5, 147 INSTRUMENTATION Section 3/4.7.9 (Deleted)

THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTIONHAS BEEN RELOCATED TO THE TECHNICAL REQUIREMENTS MANUAL (TRM) FIREPROTECTION SECTION.

TECHNICAL SPECIFICATIONS PAGES 3/4 3-92 THROUGH3/4 3-96 OF THIS SECTION HAVE BEEN INTENTIONALLY OMITTED.LIMERICK

-UNIT 23/4 3-92Amendment No. 2-5, 68 INFORMATION CONTAINED ON THIS PAGE HAS BEEN DELETEDLIMERICK

-UNIT 23/4 3-97Amendment No. -34, 117 Section 3.3.7.11 (Deleted)

THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEENRELOCATED TO THE ODCM. TECHNICAL SPECIFICATIONS PAGES 3/4 3-99 THROUGH3/4 3-102 OF THIS SECTION HAVEBEEN INTENTIONALLY OMITTED.Amendment No. 11 IfeLIMERICK

-UNIT 23/4 3-98 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 limitsof 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:a. With an offgas monitoring instrumentation channel alarm/trip setpointless conservative than required by the above Specification, declarethe 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.

Restorethe inoperable instrumentation to OPERABLE status within the timespecified in the ACTION or explain why this inoperability was notcorrected in a timely manner in the next Annual Radioactive Effluent Release Report.c. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 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 specified in theSurveillance Frequency Control Program unless otherwise noted in Table 4.3.7.12-1.

LIMERICK

-UNIT 23/4 3-103Amendment No. -14, ýý, 449, 147 TABLE 3.3.7.12-1 OFFGAS MONITORING TNNSTRIIMENTATION MINIMUM CHANNELSOPERABLEINSTRUMENT APPLICABILITY ACTION1. MAIN CONDENSER OFFGAS TREATMENT SYSTEMEXPLOSIVE GAS MONITORING SYSTEMa. Hydrogen Monitor1**1102. (Deleted)

3. (Deleted)

4. MAIN CONDENSER OFFGAS PRE-TREATMENT RADIOACTIVITY MONITORa. Noble Gas Activity Monitor1**1155. (Deleted)

LIMERICK

-UNIT 23/4 3-104Amendment No.ill TABLE 3.3.7.12-1 (Continued)

TABLE NOTATIONS

(Deleted)

- During operation of the main condenser steam jet air ejector and offgastreatment system.*** (Deleted)

ACTION STATEMENTS ACTION 110 -With the number of channels OPERABLE less than required by theMinimum Channels OPERABLE requirement, operation of maincondenser offgas treatment system may continue for up to30 days provided grab samples are collected at least onceper 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .ACTION 111-114 (Deleted)

ACTION 115 -With the number of channels OPERABLE less than required bythe Minimum Channels OPERABLE requirement, releases to theenvironment may continue for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided thatthe North Stack Effluent Noble Gas Activity Monitor isOPERABLE; otherwise, be in at least HOT SHUTDOWN within12 hours.LIMERICK

-UNIT 23/4 3-105Amendment No. 11 1 INTENTIONALLY LEFT BLANKAmendment No. 11 ILIMERICK

-UNIT 23/4 3-106 0TABLE 4.3.7.12-1 OFFGAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNELCHANNEL SOURCE CHANNEL FUNCTIONAL CHECK (5) CHECK (5) CALIBRATION(5)

TEST (5)MODES IN WHICHSURVEILLANCE IS REQUIREDINSTRUMENT

1. MAIN CONDENSER OFFGAS TREATMENT SYSTEM EXPLOSIVE GAS MONITORING SYSTEMa. Hydrogen MonitorN.A.(3)**2. (Deleted)

3. (Deleted)

4. MAIN CONDENSER OFFGASPRE-TREATMENT RADIOACTIVITY MONITOR (STEAM JET AIREJECTOR)a. Noble gas activity monitor(2)(1)**5. (Deleted)

LIMERICK

-UNIT 23/4 3-107Amendment No. 14, 147 TABLE 4.3.7.12-1 (Continued)

TABLE NOTATIONS

(Deleted)

- During operation of the main condenser steam jet air ejector and offgastreatment system.*** (Deleted)

(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarmannunciation occurs if any of the following conditions exists:1. Instrument indicates measured levels above the alarm/trip setpoint.

2. Circuit failure.3. Instrument indicates a downscale failure.4. Instrument controls not set in operate mode.(2) The initial CHANNEL CALIBRATION shall be performed using one or more of thereference standards certified by the National Institute of Standards and Technology (NIST), previously National Bureau of Standards, or usingstandards that have been obtained from suppliers that participate inmeasurement assurance activities with NIST. These standards shall permitcalibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been relatedto the initial calibration shall be used.(3) The CHANNEL CALIBRATION shall include the use of standard gas samplescontaining a nominal:1. 0.0 volume percent hydrogen, balance nitrogen, and2. 4 volume percent hydrogen, balance nitrogen.

(4) (Deleted)

(5) Frequencies are specified in the Surveillance Frequency Control Program unlessotherwise noted in the table.LIMERICK

-UNIT 23/4 3-108Amendment No. -14, 147 INTENTIONALLY LEFT BLANKLIMERICK

-UNIT 23/4 3-109Amendment No. 11 1 Section 3/4.3.8 (Deleted)

THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEENRELOCATED TO THE TRM.TECHNICAL SPECIFICATIONS PAGE 3/4 3-111HAS BEEN INTENTIONALLY OMITTED.LIMERICK

-UNIT 23/4 3-110Amendment No. 64, 153 PAGE INTENTIONALLY LEFT BLANK INSTRUMENTATION 3/4.3.9 FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.9 The feedwater/main turbine trip system actuation instrumentation channelsshown in the Table 3.3.9-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.9-2.APPLICABILITY:

As shown in Table 3.3.9-1.ACTION:a. With a feedwater/main turbine trip system actuation instrumentation channel trip setpoint less conservative than the value shown in theAllowable Values column of Table 3.3.9-2, declare the channel inoper-able and either place the inoperable channel in the tripped condition until the channel is restored to OPERABLE status with its trip set-point adjusted consistent with the Trip Setpoint value, or declarethe associated system inoperable.

b. With the number of OPERABLE channels one less thanMinimum OPERABLE Channels requirement, restore theto OPERABLE status within 7 days or be in at leastthe next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .required by theinoperable channelSTARTUP withinc. With the number of OPERABLE channels two less than required by theMinimum OPERABLE Channels requirement, restore at least one of theinoperable channels to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in atleast STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .SURVEILLANCE REQUIREMENTS 4.3.9.1 Each of the required feedwater/main turbine trip system actuation instrumentation channels shall be demonstrated OPERABLE*

by the performance of theCHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations at thefrequencies specified in the Surveillance Frequency Control Program.4.3.9.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation ofall channels shall be performed in accordance with the Surveillance Frequency Control Program.* 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 requiredsurveillance without placing the trip system in the tripped condition.

LIMERICK

-UNIT 23/4 3-112Amendment No. -3, 3-4, 147 TABLE 3.3.9-1FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION MINIMUM OPERABLECHANNELS PERTRIP SYSTEMTRIP FUNCTIONAPPLICABLE OPERATIONAL CONDITIONS 1*1. Reactor Vessel WaterLevel-High, Level 84* With Thermal Power greater than or equal to 25% of Rated Thermal Power.LIMERICK

-UNIT 23/4 3-113Amendment No. 55 TABLE 3.3.9-2FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SETPOINTS TRIP FUNCTIONTRIP SETPOINT 54 inches*ALLOWABLE VALUE 55.5 inches1. Reactor Vessel Water Level-High, Level 8*See Bases Figure B 3/4.3-1LIMERICK

-UNIT 23/4 3-114 INFORMATION ON THIS PAGE HAS BEEN DELETEDLIMERICK

-UNIT 23/4 3-115Amendment No. --3, 5, 147 PAGE INTENTIONALLY LEFT BLANK 3/4.4 REACTOR COOLANT SYSTEM3/4.4.1 RECIRCULATION SYSTEMRECIRCULATION LOOPSLIMITING CONDITION FOR OPERATION 3.4.1.1Two reactor coolant system recirculation loops shall be in operation.

APPLICABILITY:

OPERATIONAL CONDITIONS 1* and 2*.ACTION:a. With one reactor coolant system recirculation loop not in operation:

1. Within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s:a. Place the recirculation flow control system in the LocalManual mode, andb. Reduce THERMAL POWER to < 74.9% of RATED THERMAL POWER, and,c. Limit the speed of the operating recirculation pump to lessthan or equal to 90% of rated pump speed, andd. Verify that the differential temperature requirements ofSurveillance Requirement 4.4.1.1.5 are met if THERMALPOWER is 30% of RATED THERMAL POWER or the recirculation loop flow in the operating loop is < 50% of rated loopflow, or suspend the THERMAL POWER or recirculation loopflow increase.

See Special Test Exception 3.10.4.LIMERICK

-UNIT 23/4 4-1Amendment No. 48, , 4-3-9, 163 REACTOR COOLANT SYSTEMLIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

2. Within 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s:Reduce the Average Power Range Monitor (APRM) Simulated Thermal Power-Upscale Scram and Rod Block Trip Setpoints and Allowable Values, tothose applicable for single recirculation loop operation perSpecifications 2.2.1 and 3.3.6, or declare the associated channel(s) inoperable and take the actions required by the referenced specifications.

3. Otherwise be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .b. With no reactor coolant system recirculation loops in operation, initiatemeasures to place the unit in at least HOT SHUTDOWN within the next 12hours.LIMERICK

-UNIT 23/4 4-1aAmendment No. 4-, 4-0-9, 4-31-, 139 REACTOR COOLANT SYSTEMcýIIPVFTI I fnHIITPPMfNTRl 4.4.1.1.1 DELETED4.4.1.1.2 DELETED4.4.1.1.3 DELETED4.4.1.1.4 operation, that:With one reactor coolant system recirculation loop not inin accordance with the Surveillance Frequency Control Program, verifya. Reactor THERMAL POWER is < 74.9% of RATED THERMAL POWER,b. The recirculation flow control system is in the Local Manual mode,andc. The speed of the operating recirculation pump is < 90% of rated pumpspeed.4.4.1.1.5 With one reactor coolant system recirculation loop not inoperation, within 15 minutes prior to either THERMAL POWER increase orrecirculation loop flow increase, verify that the following differential temperature requirements are met if THERMAL POWER is < 30% of RATED THERMALPOWER or the recirculation loop flow in the operating recirculation loop is 50% of rated loop flow.a. 1450F between reactor vessel steam space coolant anddrain line coolant,b. 50°F between the reactor coolant within the loop notand the coolant in the reactor pressure vessel, andbottom headin operation

c. 50°F between the reactor coolant within the loop not in operation and the operating loop.The differential temperature requirements of Specification 4.4.1.1.5b.

and c.do not apply when the loop not in operation is isolated from the reactorpressure vessel.LIMERICK

-UNIT 23/4 4-2 Amendment No. -7,3-.,4,-55-,4-3,9,-1-4-ý,

163 INTENTIONALLY LEFT BLANK CONTENTS OF THIS PAGE HAVE BEEN DELETEDLIMERICK

-UNIT 23/4 4- 3Amendment No. 139 REACTOR COOLANT SYSTEMJET PUMPSLIMITING CONDITION FOR OPERATION 3.4.1.2 All jet pumps shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.ACTION:With one or more jet pumps inoperable, be in at least HOT SHUTDOWN within12 hours.SURVEILLANCE REOUIREMENTS 4.4.1.2 All jet pumps shall be demonstrated OPERABLE as follows:a. During two recirculation loop operation, each of the above requiredjet pumps shall be demonstrated OPERABLE prior to THERMAL POWERexceeding 25% of RATED THERMAL POWER and in accordance with theSurveillance Frequency Control Program while greater than 25% ofRATED THERMAL POWER by determining recirculation loop flow, totalcore flow and diffuser-to-lower plenum differential pressure foreach jet pump and verifying that no two of the following conditions occur when both recirculation loop indicated flows are in compliance with Specification 3.4.1.3.1. The indicated recirculation loop flow differs by more than 10%from the established pump speed-loop flow characteristics.

2. The indicated total core flow differs by more than 10% from theestablished total core flow value derived from recirculation loop flow measurements.

3. The indicated diffuser-to-lower plenum differential pressure ofany individual jet pump differs from the established patternsby more than 10%.LIMERICK

-UNIT 23/4 4-4Amendment No. 4-4-7-, 157 REACTOR COOLANT SYSTEMSURVEILLANCE REOUIREMENTS (Continued)

b. During single recirculation loop operation, each of the aboverequired jet pumps shall be demonstrated OPERABLE in accordance withthe Surveillance Frequency Control Program by verifying that no twoof the following conditions occur:1. The indicated recirculation loop flow in the operating loopdiffers by more than 10% from the established pump speed-loop flow characteristics.

2. The indicated total core flow differs by more than 10% from theestablished total core flow value derived from singlerecirculation loop flow measurements.

3. The indicated diffuser-to-lower plenum differential pressure ofany individual jet pump differs from established singlerecirculation loop patters by more than 10%.c. The provisions of Specification 4.0.4 are not applicable providedthat this surveillance is performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding 25% of RATED THERMAL POWER and upon entering single recirculation loop operation.

LIMERICK

-UNIT 23/4 4-4aAmendment No. 4-4-4, 157 THIS PAGE INTENTIONALLY LEFT BLANK REACTOR COOLANT SYSTEMRECIRCULATION PUMPSLIMITING CONDITION FOR OPERATION 3.4.1.3 Recirculation loop flow mismatch shall be maintained within:a. 5% of each other with core flow greater than or equal to 70% ofrated core flow.b. 10% of each other with core flow less than 70% of rated core flow.APPLICABILITY:

OPERATIONAL CONDITIONS 1* and 2* during two recirculation loopoperation.

ACTION:With the recirculation loop flows different by more than the specified limits, either:a. Restore the recirculation loop flows to within the specified limitwithin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , orb. Shutdown one of the recirculation loops within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months andtake the ACTION required by Specification 3.4.1.1.SURVEILLANCE REOUIREMENTS 4.4.1.3 Recirculation loop flow mismatch shall be verified to be within thelimits in accordance with the Surveillance Frequency Control Program.*See Special Test Exception 3.10.4.LIMERICK

-UNIT 23/4 4- 5Amendment No. 147 REACTOR COOLANT SYSTEMIDLE RECIRCULATION LOOP STARTUPLIMITING CONDITION FOR OPERATION 3.4.1.4 An idle recirculation loop shall not be started unless the temperature differential between the reactor pressure vessel steam space coolant and thebottom head drain line coolant is less than or equal to 1450F, and:a. When both loops have been idle, unless the temperature differential between the reactor coolant within the idle loop to be started upand the coolant in the reactor pressure vessel is less than or equalto 500F, orb. When only one loop has been idle, unless the temperature differential between the reactor coolant within the idle and operating recircula-tion loops is less than or equal to 50OF and the operating loopflow rate is less than or equal to 50% of rated loop flow.APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, and 4.ACTION:With temperature differences and/or flow rates exceeding the above limits,suspend startup of any idle recirculation loop.SURVEILLANCE REQUIREMENTS 4.4.1.4 The temperature differentials and flow rate shall be determined to bewithin the limits within 15 minutes prior to startup of an idle recirculation loop.LIMERICK

-UNIT 23/4 4-6 REACTOR COOLANT SYSTEM3/4.4.2 SAFETY/RELIEF VALVESLIMITING CONDITION FOR OPERATION 3.4.2 The safety valve function of at least 12 of the following reactor coolant systemsafety/relief valves shall be OPERABLE with the specified code safety valve function liftsettings:*#

455safety/relief valves @ 1170 psig +/-3%safety/relief valves @ 1180 psig +/-3%safety/relief valves @ 1190 psig +/-3%APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.ACTION:a. With the safety valve function of one or more of the above requiredsafety/relief valves inoperable, be in at least HOT SHUTDOWN within 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months sand in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .b. DELETEDc. DELETEDSURVEILLANCE REOUIREMENTS 4.4.2.1 DELETED4.4.2.2 At least 1/2 of the safety relief valves shall be removed, set pressure testedand reinstalled or replaced with spares that have been previously set pressure tested andstored in accordance with manufacturer's recommendations in accordance with theSurveillance Frequency Control Program, and they shall be rotated such that all 14 safetyrelief valves are removed, set pressure tested and reinstalled or replaced with sparesthat have been previously set pressure tested and stored in accordance with manufacturer's recommendations in accordance with the Surveillance Frequency Control Program.

All safetyvalves will be recertification tested to meet a +/-1% tolerance prior to returning thevalves to service.The lift setting pressure shall correspond to ambient conditions of thevalves at nominal operating temperatures and pressures.

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

LIMERICK

-UNIT 23/4 4-7 Amendment No. 2-1, 3-3, 34, -4, 48, 4-1-9,4-4-, 147 REACTOR COOLANT SYSTEM3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGELEAKAGE DETECTION SYSTEMSLIMITING CONDITION FOR OPERATION 3.4.3.1 The following reactor coolant leakage detection systems shallbe OPERABLE:

a. The primary containment atmosphere gaseous radioactivity monitoring system,b. The drywell sump monitoring system,c. The drywell unit coolers condensate flow rate monitoring system, andd. The primary containment pressure and temperature monitoring system.APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.**- The primary containment gaseous radioactivity monitor is not required to beoperable until Operational Condition 2.ACTIONS:A. With the primary containment atmosphere gaseous radioactivity monitoring systeminoperable, analyze grab samples of primary containment atmosphere at least onceper 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND restore primary containment atmosphere gaseous radioactivity monitoring system to OPERABLE status within 30 days.B. With the drywell sump monitoring system inoperable, restore the drywell sumpmonitoring system to OPERABLE status within 30 days AND increase monitoring ifrequency of drywell unit cooler condensate flow rate (SR 4.4.3.2.1.c) to onceevery 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .C. With the drywell unit coolers condensate flow rate monitoring system inoperable, AND the primary containment atmosphere gaseous radioactivity monitoring systemOPERABLE, perform a channel check of the primary containment atmosphere gaseousradioactivity monitoring system (SR 4.4.3.1.a) once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .D. With the primary containment pressure and temperature monitoring systeminoperable, restore the primary containment pressure and temperature monitoring system to OPERABLE status within 30 days. Note: All other Tech Spec LimitingConditions For Operation and Surveillance Requirements associated with theprimary containment pressure/temperature monitoring system still apply. AffectedTech Spec Sections include:

3/4.3.7.5, 4.4.3.2.1.

3/4.6.1.6.

and 3/4.6.1.7.

E. With the primary containment atmosphere gaseous radioactivity monitoring systeminoperable AND the drywell unit coolers condensate flow rate monitoring systeminoperable, restore the primary containment atmosphere gaseous radioactivity monitoring system to OPERABLE status within 30 days OR restore the drywell unitcoolers condensate flow rate monitoring system to OPERABLE status within 30 days.With the primary containment atmosphere gaseous radioactivity monitoring systeminoperable, analyze grab samples of primary containment atmosphere at least onceper 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .LIMERICK

-UNIT 23/4 4-8Amendment No. -4,4-943,4-32-,-1-9-3,169 REACTOR COOLANT SYSTEMACTIONS (Continued)

F. With the drywell floor drain sump monitoring system inoperable AND the drywellunit coolers condensate flow rate monitoring system inoperable analyze grabsamples of the primary containment atmosphere once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , AND monitorReactor Coolant System leakage by administrative means once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ANDrestore either the drywell floor drain sump monitoring system to OPERABLE statuswithin 7 days OR restore the drywell unit coolers condensate flow rate monitoring system to OPERABLE status within 7 days.G. With any other two or more leak detection systems inoperable other than ACTIONS Eand F above OR with required Actions and associated Completion Time of ACTIONS A,B, C, D, E or F not met, be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND in COLD SHUTDOWNwithin the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .SURVEILLANCE REQUIREMENTS 4.4.3.1 The reactor coolant system leakage detection systems shall be demonstrated operable by:a. Perform a CHANNEL CHECK of the primary containment atmosphere gaseousradioactivity monitoring system in accordance with the Surveillance Frequency Control Program.b. Perform a CHANNEL FUNCTIONAL TEST of required leakage detection instrumentation in accordance with the Surveillance Frequency ControlProgram.

This does not apply to containment pressure and temperature monitoring system.c. Perform a CHANNEL CALIBRATION of required leakage detection instrumentation in accordance with the Surveillance Frequency Control Program.

This doesnot apply to containment pressure and temperature monitoring system.d. Monitor primary containment pressure AND primary containment temperature inaccordance with the Surveillance Frequency Control Program.LIMERICK

-UNIT 23/4 4-8aAmendment No. 4-443,4-4-7-,4--3, 167 INTENTIONALLY LEFT BLANK REACTOR COOLANT SYSTEMOPERATIONAL LEAKAGELIMITING CONDITION FOR OPERATION 3.4.3.2 Reactor coolant system leakage shall be limited to:a. No PRESSURE BOUNDARY LEAKAGE.b. 5 gpm UNIDENTIFIED LEAKAGE.c. 30 gpm total leakage.d. 25 gpm total leakage averaged over any 24-hour period.e. 1 gpm leakage at a reactor coolant system pressure of 950 +/-10 psig from anyreactor coolant system pressure isolation valve.**f. 2 gpm increase in UNIDENTIFIED LEAKAGE over a 24-hour period.APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.ACTION:a. With any PRESSURE BOUNDARY

LEAKAGE, be in at least HOT SHUTDOWN within 12hours and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .b. With any reactor coolant system leakage greater than the limits in b, cand/or d above, reduce the leakage rate to within the limits within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months sor 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 SHUTDOWNwithin the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .c. With any reactor coolant system pressure isolation valve leakage greaterthan the above limit, isolate the high pressure portion of the affectedsystem from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least oneother closed manual, deactivated automatic, or check* valves, or be in atleast 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 thefollowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .d. With one or more of the high/low pressure interface valve leakage pressuremonitors inoperable, restore the inoperable monitor(s) to OPERABLE statuswithin 7 days or verify the pressure to be less than the alarm setpoint atleast once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; restore the inoperable monitor(s) to OPERABLEstatus within 30 days or be in at least HOT SHUTDOWN within the next 12hours and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .e. With any reactor coolant system leakage greater than the limit in f above,identify the source of leakage within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHUTDOWNwithin 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 .Which have been verified not to exceed the allowable leakage limit at the lastrefueling outage or after the last time the valve was disturbed, whichever is morerecent.** Pressure isolation valve leakage is not included in any other allowable operational leakage specified in Section 3.4.3.2.LIMERICK

-UNIT 23/4 4-9Amendment No. 9, 4-34, 144 REACTOR COOLANT SYSTEMSURVEILLANCE REOUIREMENTS 4.4.3.2.1 The reactor coolant system leakage shall be demonstrated to be withineach of the above limits by:a. Monitoring the primary containment atmospheric gaseous radioactivity inaccordance with the Surveillance Frequency Control Program (not a meansof quantifying leakage),

b. Monitoring the drywell floor drain sump and drywell equipment draintank flow rate in accordance with the Surveillance Frequency ControlProgram,c. Monitoring the drywell unit coolers condensate flow rate in accordance with the Surveillance Frequency Control Program,d. Monitoring the primary containment pressure in accordance with theSurveillance Frequency Control Program (not a means of quantifying leakage),

e. Monitoring the reactor vessel head flange leak detection system inaccordance with the Surveillance Frequency Control Program, andf. Monitoring the primary containment temperature in accordance with theSurveillance Frequency Control Program (not a means of quantifying leakage).

4.4.3.2.2 Each reactor coolant system pressure isolation valve shall be demonstrated OPERABLE by leak testing pursuant to Specification 4.0.5 and verifying the leakage ofeach valve to be within the specified limit:a. In accordance with the Surveillance Frequency Control Program, andb. Prior to returning the valve to service following maintenance, repairor replacement work on the valve which could affect its leakage rate.The provisions of Specification 4.0.4 are not applicable for entry intoOPERATIONAL CONDITION 3.4.4.3.2.3 The high/low pressure interface valve leakage pressure monitors shallbe demonstrated OPERABLE with alarm setpoints set less than the specified allowable values by performance of a CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION at thefrequencies specified in the Surveillance Frequency Control Program.LIMERICK

-UNIT 2 3/4 4-10 Amendment No. --2, -34, -1A4, 147 TABLE 3.4.3.2-1 (Deleted)

THE INFORMATION FROM THIS TECHNICAL SPECIFICATION SECTION HASBEEN RELOCATED TO THE TECHNICL REQUIREMENTS MANUAL (TRM).LIMERICK

-UNIT 23/4 4-11Amendment No. 144 REACTOR COOLANT SYSTEM3/4.4.4 (Deleted)

THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE TECHNICAL REQUIREMENTS MANUAL (TRM). TECHNICAL SPECIFICATIONS PAGES 3/4 4-13AND 3/4 4-14 HAVE BEEN INTENTIONALLY OMITTED.LIMERICK

-UNIT 23/4 4-12Amendment No. 4-32, 136 REACTOR COOLANT SYSTEM3/4.4.5 SPECIFIC ACTIVITYLIMITING CONDITION FOR OPERATION 3.4.5 The specific activity of the primary coolant shall be limited to:a. Less than or equal to 0.2 microcurie per gram DOSE EQUIVALENT 1-131.b. (Deleted)

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, and 4.ACTION:a. In OPERATIONAL CONDITION 1, 2, or 3 with the specific activity ofthe primary coolant;1. Greater than 0.2 microcurie per gram DOSE EQUIVALENT 1-131 butless than or equal to 4 microcuries per gram, DOSE EQUIVALENT 1-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time intervalor greater than 4.0 microcuries per gram DOSE EQUIVALENT 1-131,be in at least HOT SHUTDOWN with the main steam line isolation valves closed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The provisions of Specification 3.0.4.c are applicable.

2. (Deleted)

b. In OPERATIONAL CONDITION 1, 2, 3, or 4, with the specific activityof the primary coolant greater than 0.2 microcuries per gram DOSEEQUIVALENT 1-131, perform the sampling and analysis requirements ofItem 4.a of Table 4.4.5-1 until the specific activity of the primarycoolant is restored to within its limit.c. In OPERATIONAL CONDITION 1 or 2, with:1. THERMAL POWER changed by more than 15% of RATED THERMAL POWERin 1 hour*, or2. The off-gas level, at the SJAE, increased by more than 10,000microcuries per second in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months during steady-state operation at release rates less than 75,000 microcuries per second, or3. The off-gas level, at the SJAE, increased by more than 15% inI hour during steady-state operation at release rates greaterthan 75,000 microcuries per second,perform the sampling and analysis requirements of Item 4.b ofTable 4.4.5-1 until the specific activity of the primary coolantis restored to within its limit.*Not applicable during the startup test program.LIMERICK

-UNIT 23/4 4-15Amendment No. 4-321, 136 REACTOR COOLANT SYSTEMSURVEILLANCE REOUIREMENTS 4.4.5 The specific activity of the reactor coolant shall be demonstrated tobe within the limits by performance of the sampling and analysis program ofTable 4.4.5-1.LIMERICK

-UNIT 23/4 4-16 TABLE 4.4.5-1PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAMTYPE OF MEASUREMENT AND ANALYSISSAMPLE AND ANALYSISFREQUENCY OPERATIONAL CONDITIONS IN WHICH SAMPLEAND ANALYSIS IS REOUIRED1. (Deleted)

2. Isotopic Analysis for DOSEEQUIVALENT 1-131 Concentration In accordance with theSurveillance Frequency Control Program13. (Deleted)

4. Isotopic Analysis for Iodinea) At least once per 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months swhenever the specificactivity exceeds a limit,as required by ACTION b.b) At least one sample, between2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following thechange in THERMAL POWER oroff-gas level, as requiredby ACTION c.1**, 2**, 3**, 4**1, 25. Isotopic Analysis of an Off-gas Sample Including Quantitative Measurements for at least Xe-133,Xe-135, and Kr-88In accordance with theSurveillance Frequency Control Program1**Until the specific activity of the primary coolant system is restored to within its limits.LIMERICK

-UNIT 23/4 4-17Amendment No. 4436, 147 REACTOR COOLANT SYSTEM3/4.4.6 PRESSURE/TEMPERATURE LIMITSREACTOR COOLANT SYSTEMLIMITING CONDITION FOR OPERATION 3.4.6.1 The reactor coolant system temperature and pressure shall be limitedin accordance with the limit lines shown on Figure 3.4.6.1-1 (1) curve Afor hydrostatic or leak testing; (2) curve B for heatup by non-nuclear means, cooldown following a nuclear shutdown and low power PHYSICS TESTS; and(3) curve C for operations with a critical core other than low power PHYSICS TESTS,with:a. A maximum heatup of 100°F in any 1-hour period,b. A maximum cooldown of 100OF in any 1-hour period,c. A maximum temperature change of less than or equal to 20'F in any1-hour period during inservice hydrostatic and leak testing opera-tions above the heatup and cooldown limit curves, andd. The reactor vessel flange and head flange temperature greater thanor equal to 70°F when reactor vessel head bolting studs are undertension.APPLICABILITY:

At all times.ACTION:With any of the above limits exceeded, restore the temperature and/or pressureto within the limits within 30 minutes; perform an engineering evaluation todetermine the effects of the out-of-limit condition on the structural integrity of the reactor coolant system; determine that the reactor coolant system remainsacceptable for continued operations or be in at least HOT SHUTDOWN within 12hours and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .SURVEILLANCE REQUIREMENTS 4.4.6.1.1 During system heatup, cooldown and inservice leak and hydrostatic testing operations, the reactor coolant system temperature and pressure shallbe determined to be within the above required heatup and cooldown limits and tothe right of the limit lines of Figure 3.4.6.1-1 curves A, B or C as applicable, in accordance with the Surveillance Frequency Control Program.LIMERICK

-UNIT 2 3/4 4-18 Amendment No. 4-14, 1470 REACTOR COOLANT SYSTEMSURVEILLANCE REQUIREMENTS (Continued) 4.4.6.1.2 The reactor coolant system temperature and pressure shall bedetermined to be to the right of the criticality limit line of Figure 3.4.6.1-1 curve C within 15 minutes prior to the withdrawal of control rods tobring the reactor to criticality and in accordance with the Surveillance Frequency Control Program during system heatup.4.4.6.1.3 DELETED4.4.6.1.4 DELETED4.4.6.1.5 The reactor vessel flange and head flange temperature shall beverified to be greater than or equal to 70°F:a. In OPERATIONAL CONDITION 4 when reactor coolant system temperature is:1. 1000F, in accordance with the Surveillance Frequency ControlProgram.2. 900F, in accordance with the Surveillance Frequency ControlProgram.b. Within 30 minutes prior to and in accordance with the Surveillance Frequency Control Program during tensioning of the reactor vesselhead bolting studs.LIMERICK

-UNIT 23/4 4-19Amendment No. -4, 444, 4-3-9, 147 1400130012001100a 100049000wnc)8000 700600wo 5004003u0:60400A22 A B CA, B, C -CORE BELTLINEAFTER ASSUMED 82°F SHIFTFROM AN INITIAL PLATE RTNDTF-- OF40°FA -SYSTEM HYDROTEST WITH FUEL IN THE VESSELB -NON-NUCLEAR HEATUP/COOLDOWN LIMIT312 PSIG C -NUCLEAR (CORE CRITICAL)

A )FiLIMITCURVES A, B, C ARE VALID UPBTU /- TO 32 EFPY OF OPERATION 70°F jCURVE A22 IS VALID UP TO22 EFPY OF OPERATION 0 25 50 75 100 125 150 175 200 225 250 275 300 325 350 375 4002001000MINIMUM REACTOR VESSEL METAL TEMPERATURE (fF)MINIMUM REACTOR VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSUREFIGURE 3.4.6.1-1 LIMERICK

-UNIT 23/4 4-20Amendment No. -4, 80, ;--11, 1250 INFORMATION CONTAINED ON THIS PAGE HAS BEEN DELETEDLIMERICK

-UNIT 23/4 4-21Amendment No. -94, 130 REACTOR COOLANT SYSTEMREACTOR STEAM DOMELIMITING CONDITION FOR OPERATION 3.4.6.2The pressure in the reactor steam dome shall be less than 1053 psig.APPLICABILITY:

OPERATIONAL CONDITIONS 1* and 2*.ACTION:With the reactor steam dome pressure exceeding 1053 psig, reduce the pressureto less than 1053 psig within 15 minutes or be in at least HOT SHUTDOWN within12 hours.SURVEILLANCE REOUIREMENTS 4.4.6.2than 1053Program.The reactor steam dome pressure shall be verified to be lesspsig in accordance with the Surveillance Frequency Control*Not applicable during anticipated transients.

Amendment No. .5-1, 1470LIMERICK

-UNIT 23/4 4-22 REACTOR COOLANT SYSTEM3/4.4.7 MAIN STEAM LINE ISOLATION VALVESULMITING CONDITION FOR OPERATION 3.4.7 Two main steam line isolation valves (MSIVs) per main steam line shallbe OPERABLE with closing times greater than or equal to 3 and less than orequal to 5 seconds.APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.ACTION:With one or more MSIVs inoperable:

a. Maintain at least one MSIV OPERABLE in each affected main steamline that is open and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , either:1. Restore the inoperable valve(s) to OPERABLE status, or2. Isolate the affected main steam line by use of a deactivated MSIV in the closed position.

b. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and inCOLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .SURVEILLANCE REQUIREMENTS 4.4.7 Each of the above required MSIVs shall be demonstrated OPERABLE byverifying full closure between 3 and 5 seconds when tested pursuant toSpecification 4.0.5.LIMERICK

-UNIT 23/4 4-23Amendment No. 132 REACTOR COOLANT SYSTEM3/4.4.8 (DELETED)

PAGE INTENTIONALLY LEFT BLANKLIMERICK

-UNIT 23/4 4-24Amendment No. 160 REACTOR COOLANT SYSTEM3/4.4.9 RESIDUAL HEAT REMOVAL.HOT SHUTDOWNLIMITING CONDITION FOR OPERATION 3.4.9.1 Two (2) independent RHR shutdown cooling subsystems shall be OPERABLE, and,with no recirculation pump in operation, at least one (1) RHR shutdown coolingsubsystem shall be in operation.

** ***Each independent RHR shutdown cooling subsystem shall consist of at least:a. One OPERABLE RHR pump, andb. One OPERABLE RHR heat exchanger, not common to the two (2) independent subsystems.

APPLICABILITY:

OPERATIONAL CONDITION 3, with reactor vessel pressure less thanthe RHR cut-in permissive setpoint.

ACTION:a. With less than the above required independent RHR shutdown cooling subsystems

OPERABLE, immediately initiate corrective action to return therequired independent subsystems to OPERABLE status as soon as possible.

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, verify theavailability of at least one alternate method capable of decay heat removalfor each inoperable independent RHR shutdown cooling subsystem.

Be in atleast COLD SHUTDOWN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .****

b. With no independent RHR shutdown cooling subsystem in operation, immediately initiate corrective action to return at least one (1) independent subsystem to operation as soon as possible.

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish reactorcoolant circulation by an alternate method and monitor reactor coolanttemperature and pressure at least once per hour.SURVEILLANCE REQUIREMENTS 4.4.9.1 At least one independent RHR shutdown cooling subsystem or alternate methodshall be determined to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program.*One independent RHR shutdown cooling subsystem may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months forsurveillance testing provided the other independent subsystem is OPERABLE and inoperation.

- The shutdown cooling pump may be removed from operation for up to 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months sper 8-hour period provided the other independent subsystem is OPERABLE.

- - The independent RHR shutdown cooling subsystem may be removed from operation duringhydrostatic testing.****Whenever two or more RHR subsystems are inoperable, if unable to attain COLDSHUTDOWN as required by this ACTION, maintain reactor coolant temperature aslow as practical by use of alternate heat removal methods.LIMERICK

-UNIT 2 3/4 4-25 Amendment No. 6-1,8, 147 REACTOR COOLANT SYSTEMCOLD SHUTDOWNLIMITING CONDITION FOR OPERATION 3.4.9.2 Two (2) RHR shutdown cooling subsystems shall be OPERABLE, and with norecirculation pump in operation, at least one (1) RHR shutdown coolingsubsystem shall be in operation.

** ***APPLICABILITY:

OPERATIONAL CONDITION 4.ACTION: #a. With one (1) or two (2) RHR shutdown cooling subsystems inoperable:

1. Within one (1) hour, and once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, verify analternate method of decay heat removal is available for eachinoperable RHR shutdown cooling subsystem.

b. With no RHR shutdown cooling subsystems in operation and no recirculation pump in operation:

1. Within one (1) hour from discovery of no reactor coolantcirculation, and once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter, verify reactorcoolant circulating by an alternate method; and2. Once per hour monitor reactor coolant temperature and pressure.

SURVEILLANCE REQUIREMENTS 4.4.9.2 At least one (1) RHR shutdown cooling subsystem or recirculation pump isoperating or an alternate method shall be determined to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program.Both RHR shutdown cooling subsystems and recirculation pumps may be removed fromoperation for up two (2) hours per eight hour (8) period.*** One (1) RHR shutdown cooling subsystem may be inoperable for up to two (2) hoursfor the performance of Surveillances.

- - - The shutdown cooling subsystem may be removed from operation during hydrostatic testing.# Separate Action entry is allowed for each shutdown cooling subsystem.

LIMERICK

-UNIT 23/4 4-26Amendment No. 6-1, 82, 147 3/4.5 EMERGENCY CORE COOLING SYSTEMS3/4.5.1 ECCS -OPERATING LIMITING CONDITION FOR OPERATION 3.5.1 The emergency core cooling systems shall be OPERABLE with:a. The core spray system (CSS) consisting of two subsystems with eachsubsystem comprised of:1. Two OPERABLE CSS pumps, and2. An OPERABLE flow path capable of taking suction from thesuppression chamber and transferring the water through the spraysparger to the reactor vessel.b. The low pressure coolant injection (LPCI) system of the residual heatremoval system consisting of four subsystems with each subsystem comprised of:1. One OPERABLE LPCI pump, and2. An OPERABLE flow path capable of taking suction from thesuppression chamber and transferring the water to the reactorvessel.c. The high pressure coolant injection (HPCI) system consisting of:1. One OPERABLE HPCI pump, and2. An OPERABLE flow path capable of taking suction from thesuppression chamber and transferring the water to the reactorvessel.d. The automatic depressurization system (ADS) with at least fiveOPERABLE ADS valves.APPLICABILITY:

OPERATIONAL CONDITION 1, 2* ** #, and 3* ***The HPCI system is not required to be OPERABLE when reactor steam domepressure is less than or equal to 200 psig.**The ADS is not required to be OPERABLE when the reactor steam dome pressure isless that or equal to 100 psig.#See Special Test Exception 3.10.6.##Two LPCI subsystems of the RHR system may be inoperable in that they are alignedin the shutdown cooling mode when reactor vessel pressure is less than theRHR Shutdown cooling permissive setpoint.

LIMERICK

-UNIT 23/4 5-1Amendment No. 0, 49, 153 EMERGENCY CORE COOLING SYSTEMSLIMITING CONDITION FOR OPERATION (Continued)

ACTION:a. For the core spray system:1. With one CSS subsystem inoperable, provided that at least two LPCIsubsystems are OPERABLE, restore the inoperable CSS subsystem toOPERABLE status within 7 days or be in at least HOT SHUTDOWN withinthe 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 subsystems inoperable, be in at least HOT SHUTDOWNwithin 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 CSSsubsystem is OPERABLE, restore the inoperable LPCI pump to OPERABLEstatus within 30 days or be in at least HOT SHUTDOWN within thenext 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 notremoved from one closed RHR cross-tie valve operator, close theopen valve and/or remove power from the closed valves operatorwithin 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 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months sand 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 powernot removed from both closed RHR cross-tie valve operators, orwith one RHR cross-tie valve open and power not removed fromthe other RHR cross-tie valve operator, be in at least HOTSHUTDOWN within 12.hours and in COLD SHUTDOWN within the next24 hours.4. With two LPCI subsystems inoperable, provided that at least one CSSsubsystem is OPERABLE, restore at least three LPCI subsystems toOPERABLE status within 7 days or be in at least HOT SHUTDOWN withinthe 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. With three LPCI subsystems inoperable, provided that both CSSsubsystems are OPERABLE, restore at least two LPCI subsystems toOPERABLE 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 withinthe next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .6. With all four LPCI subsystems inoperable, be in at least HOTSHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next24 hours.**Whenever both shutdown cooling subsystems are inoperable, if unable to attainCOLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.LIMERICK

-UNIT 23/4 5-2Amendment No. -58, 9, 92 EMERGFNCY CfRF COOITNG SYSTFMSLIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

c. For the HPCI system:1. With the HPCI system inoperable, provided the CSS, the LPCIsystem, the ADS and the RCIC system are OPERABLE, restore theHPCI system to OPERABLE status within 14 days or be in at leastHOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor, steamdome pressure to 200 psig within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .2. With the HPCI system inoperable, and one CSS subsystem, and/orLPCI subsystem inoperable, and provided at least one CSS subsystem, three LPCI subsystems, and ADS are operable, restore the HPCI toOPERABLE within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or be in HOT SHUTDOWN in the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,and in COLD SHUTDOWN in the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .3. Specification 3.0.4.b is not applicable to HPCI.d. For the ADS:1. With one of the above required ADS valves inoperable, providedthe HPCI system, the CSS and the LPCI system are OPERABLE, restore the inoperable ADS valve to OPERABLE status within14 days or be in at least HOT SHUTDOWN within the next 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months sand reduce reactor steam dome pressure to 100 psig withinthe next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .2. With two or more of the above required ADS valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactorsteam dome pressure to 100 psig within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .e. With a CSS and/or LPCI header AP instrumentation channel inoperable, restore the inoperable channel to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ordetermine the ECCS header AP locally at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;otherwise, declare the associated CSS and/or LPCI, as applicable, inoperable.

f. In the event an ECCS system is actuated and injects water into thereactor coolant system, a Special Report shall be prepared and sub-mitted to the Commission pursuant to Specification 6.9.2 within90 days describing the circumstances of the actuation and the totalaccumulated actuation cycles to date. The current value of theusage factor for each affected safety injection nozzle shall beprovided in this Special Report whenever its value exceeds 0.70.LIMERICK

-UNIT 23/4 5-3Amendment No. -8, 132 EMERGENCY CORE COOLING SYSTEMSSURVEILLANCE REOUIREMENTS 4.5.1 The emergency core cooling systems shall be demonstrated OPERABLE by:a. In accordance with the Surveillance Frequency Control Program:1. For the CSS, the LPCI system, and the HPCI system:a) Verifying by venting at the high point vents that thesystem piping from the pump discharge valve to the systemisolation valve is filled with water.b) Verifying that each valve (manual, power-operated, orautomatic) in the flow path that is not locked, sealed, orotherwise secured in position, is in its correct*

position.

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

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

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

b. Verifying that, when tested pursuant to Specification 4.0.5:1. Each CSS pump in each subsystem develops a flow of at least3175 gpm against a test line pressure corresponding to a reactorvessel to primary containment differential pressure of = 105 psidplus head and line losses.2. Each LPCI pump in each subsystem develops a flow of at least10,000 gpm against a test line pressure corresponding to areactor vessel to primary containment differential pressure of> 20 psid plus head and line losses.3. The HPCI pump develops a flow of at least 5600 gpm against atest line pressure which corresponds to a reactor vesselpressure of 1040 psig plus head and line losses when steam isbeing supplied to the turbine at 1040, +13, -120 psig.**c. In accordance with the Surveillance Frequency Control Program:1. For the CSS, the LPCI system, and the HPCI system, performing asystem functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flowpath actuates to its correct position.

Actual injection ofcoolant into the reactor vessel may be excluded from this test.Except that an automatic valve capable of automatic return to its ECCS positionwhen an ECCS signal is present may be in position for another mode of operation.

- The provisions of Specification 4.0.4 are not applicable provided thesurveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure isadequate to perform the test. If OPERABILITY is not successfully demonstrated within the 12-hour period, reduce reactor steam dome pressure to less than200 psig within the following 72-hours.

LIMERICK

-UNIT 23/4 5-4Amendment No. -4, 4, 147 EMERGENCY CORE COOLING SYSTEMSSURVEILLANCE REOUIREMENTS (Continued)

2. For the HPCI system, verifying that:a) The system develops a flow of at least 5600 gpm against atest line pressure corresponding to a reactor vessel pressureof 200 psig plus head and line losses, when steam is beingsupplied to the turbine at 200 + 15, -0 psig.**b) The suction is automatically transferred from the condensate storage tank to the suppression chamber on a condensate storage tank water level -low signal and on a suppression chamber water level -high signal.3. Performing a CHANNEL CALIBRATION of the CSS, LPCI, and HPCIsystem discharge line "keep filled" alarm instrumentation.

4. Performing a CHANNEL CALIBRATION of the CSS header AP instru-mentation and verifying the setpoint to be the allowable valueof 4.4 psid.5. Performing a CHANNEL CALIBRATION of the LPCI header AP instru-mentation and verifying the setpoint to be the allowable valueof 3.0 psid.d. For the ADS:1. In accordance with the Surveillance Frequency Control Program,verify ADS accumulator gas supply header pressure is 90 psig.2. In accordance with the Surveillance Frequency Control Program:a) Performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating

sequence, but excluding actual valve actuation.

b) Verify that when tested pursuant to Specification 4.0.5 thateach ADS valve is capable of being opened.c) DELETED** The provisions of Specification 4.0.4 are not applicable provided thesurveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure isadequate to perform the test. If HPCI OPERABILITY is not successfully demonstrated within the 12-hour period, reduce reactor steam dome pressureto less than 200 psig within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .LIMERICK

-UNIT 23/4 5-5Amendment No. %4, 84, 4-1-6, 147 EMERGENCY CORE COOLING SYSTEMS3/4 5.2 ECCS -SHUTDOWNLIMITING CONDITION FOR OPERATION 3.5.2 At least two of the following shall be OPERABLE:

a. Core spray system (CSS) subsystems with a subsystem comprised of:1. Two OPERABLE CSS pumps, and2. An OPERABLE flow path capable of taking suction from at leastone of the following water sources and transferring the waterthrough the spray sparger to the reactor vessel:a) From the suppression

chamber, orb) When the suppression chamber water level is less than thelimit or is drained, from the condensate storage tankcontaining at least 135,000 available gallons of water,equivalent to a level of 29 feet.b. Low pressure coolant injection (LPCI) system subsystems with asubsystem comprised of:1. One OPERABLE LPCI pump, and2. An OPERABLE flow path capable of taking suction from thesuppression chamber and transferring the water to the reactorvessel.**

APPLICABILITY:

OPERATIONAL CONDITIONS 4 and 5*.ACTION:a. With one of the above required subsystems inoperable, restore atleast two subsystems to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or suspendall operations with a potential for draining the reactor vessel.b. With both of the above required subsystems inoperable, suspend COREALTERATIONS and all operations with a potential for draining thereactor vessel. Restore at least one subsystem to OPERABLE statuswithin 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or establish SECONDARY CONTAINMENT INTEGRITY withinthe next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .*The ECCS is not required to be OPERABLE provided that the reactor vessel headis removed, the cavity is flooded, the spent fuel pool gates are removed, andwater level is maintained within the limits of Specifications 3.9.8 and 3.9.9.**One LPCI subsystem may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned and not otherwise inoperable.

LIMERICK

-UNIT 23/4 5-6Amendment No. 59 EMFRGFNCY cOnRF mnOITNG SýYSTFMSSURVEILLANCE REQUIREMENTS 4.5.2.1 At least the above required ECCS shall be demonstrated OPERABLE perSurveillance Requirement 4.5.1.*4.5.2.2 The core spray system shall be determined OPERABLE in accordance with theSurveillance Frequency Control Program by verifying the condensate storage tankrequired volume when the condensate storage tank is required to be OPERABLE perSpecification 3.5.2a.2.b).

One LPCI subsystem may be considered OPERABLE during alignment and operation for decay heat removal if capable of being manually realigned and not otherwise inoperable.

LIMERICK

-UNIT 23/4 5-7Amendment No. -59, 147 EMERGENCY CORE COOLING SYSTEMS3/4.5.3 SUPPRESSION CHAMBERLIMITING CONDITION FOR OPERATION 3.5.3 The suppression chamber shall be OPERABLE:

a. In OPERATIONAL CONDITIONS 1, 2, and 3 with a contained water volume ofat least 122,120 ft3, equivalent to a level of 22'0".b. In OPERATIONAL CONDITION 4 and 5* with a contained water volume of atleast 88,815 ft3, equivalent to a level of 16'0", except that thesuppression chamber level may be less than the limit or may be drainedprovided that:1. No operations are performed that have a potential for drainingthe reactor vessel,2. The reactor mode switch is locked in the Shutdown or Refuelposition,

3. The condensate storage tank contains at least 135,000 available gallons of water, equivalent to a level of 29 feet, and4. The core spray system is OPERABLE per Specification 3.5.2 withan OPERABLE flow path capable of taking suction from thecondensate storage tank and transferring the water through thespray sparger to the reactor vessel.APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, 4, and 5*.ACTION:a. In OPERATIONAL CONDITION 1, 2 or 3 with the suppression chamber waterlevel less than the above limit, restore the water level to withinthe limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT SHUTDOWN within thenext 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. In OPERATIONAL CONDITION 4 or 5* with the suppression chamber waterlevel less than the above limit or drained and the above requiredconditions not satisfied, suspend CORE ALTERATIONS and all operations that have a potential for draining the reactor vessel and lock thereactor mode switch in the Shutdown position.

Establish SECONDARY CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .*The suppression chamber is not required to be OPERABLE provided that thereactor vessel head is removed, the cavity is flooded or being flooded fromthe suppression pool, the spent fuel pool gates are removed when the cavityis flooded, and the water level is maintained within the limits ofSpecifications 3.9.8 and 3.9.9.LIMERICK

-UNIT 23/4 5-8 FMERGENCY CORE COOLING SYSTEMSSURVEILLANCE REOUIREMENTS 4.5.3.1 The suppression chamber shall be determined OPERABLE by verifying the water level to be greater than or equal to, as applicable:

a. 22'0" in accordance with the Surveillance Frequency Control Program.b. 16'0" in accordance with the Surveillance Frequency Control Program.4.5.3.2 With the suppression chamber level less than the above limit ordrained in OPERATIONAL CONDITION 4 or 5*, in accordance with the Surveillance Frequency Control Program:a. Verify the required conditions of Specification 3.5.3b. to besatisfied, orb. Verify footnote conditions

to be satisfied.

The suppression chamber is not required to be OPERABLE provided that thereactor vessel head is removed, the cavity is flooded or being flooded fromthe suppression pool, the spent fuel pool gates are removed when the cavityis flooded, and the water level is maintained within the limits ofSpecifications 3.9.8 and 3.9.9.LIMERICK

-UNIT 23/4 5-9Amendment No. 147 THIS PAGE INTENTIONALLY LEFT BLANK 3/4.6 CONTAINMENT SYSTEMS3/4.6.1 PRIMARY CONTAINMENT PRIMARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2*, and 3.ACTION:Without PRIMARY CONTAINMENT INTEGRITY, restore PRIMARY CONTAINMENT INTEGRITY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 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 inCOLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .SURVEILLANCE REQUIREMENTS 4.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be demonstrated:

a. After each closing of each penetration subject to Type B testing,except the primary containment air locks, if opened following Type Aor B test, by leak rate testing in accordance with the Primary Containment Leakage Rate Testing Program.b. In accordance with the Surveillance Frequency Control Program byverifying that all primary containment penetrations**

not capable ofbeing closed by OPERABLE containment automatic isolation valves andrequired to be closed during accident conditions are closed by valves,blind flanges, or deactivated automatic valves secured in position, except for valves that are opened under administrative control aspermitted by Specification 3.6.3.c. By verifying the primary containment air lock is in compliance withthe requirements of Specification 3.6.1.3.d. By verifying the suppression chamber is in compliance with therequirements of Specification 3.6.2.1.* See Special Test Exception 3.10.1**Except valves, blind flanges, and deactivated automatic valves which are locatedinside the containment, and are locked, sealed, or otherwise secured in theclosed position.

These penetrations shall be verified closed during each COLDSHUTDOWN except such verification need not be performed when the primarycontainment has not been deinerted since the last verification or more often thanonce per 92 days.LIMERICK

-UNIT 23/4 6-1Amendment No. 8-4, 4-0-ý, 147 CONTAINMENT SYSTEMSPRIMARY CONTAINMENT LEAKAGELIMITING CONDITION FOR OPERATION 3.6.1.2 Primary containment leakage rates shall be limited to:a. An overall integrated leakage rate (Type A Test) in accordance withthe Primary Containment Leakage Rate Testing Program.b. A combined leakage rate in accordance with the Primary Containment Leakage Rate Testing Program for all primary containment penetrations and all primary containment isolation valves that are subject to Type Band C tests, except for: main steam line isolation valves*,

valveswhich are hydrostatically tested, and those valves where an exemption toAppendix J of 10 CFR 50 has been granted.c. *Less than or equal to 100 scf per hour through any one main steamisolation valve not to exceed 200 scf per hour for all four main steamlines, when tested at Pt, 22.0 psig.d. A combined leakage rate of less than or equal to 1 gpm times thetotal number of containment isolation valves in hydrostatically tested lines which penetrate the primary containment, when tested at1.10 P, , 48.4 psig.APPLICABILITY:

When PRIMARY CONTAINMENT INTEGRITY is required perSpecification 3.6.1.1.ACTION:With:a. The measured overall integrated primary containment leakage rate(Type A Test) exceeding the leakage rate specified in the PrimaryContainment Leakage Rate Testing Program, orb. The measured combined leakage rate exceeding the leakage rate specified in the Primary Containment Leakage Rate Testing Program for all primarycontainment penetrations and all primary containment isolation valvesthat are subject to Type B and C tests, except for: main steam lineisolation valves*,

valves which are hydrostatically tested, and thosevalves where an exemption to Appendix J of 10 CFR 50 has been granted,orc. The measured leakage rate exceeding 100 scf per hour through any onemain steam isolation valve, or exceeding 200 scf per hour for all fourmain steam lines, ord. The measured combined leakage rate for all containment isolation valves in hydrostatically tested lines which penetrate the primarycontainment exceeding 1 gpm times the total number of such valves,restore:a. The overall integrated leakage rate(s) (Type A Test) to be in accordance with the Primary Containment Leakage Rate Testing Program, and*Exemption to Appendix J of 10 CFR Part 50.LIMERICK

-UNIT 23/4 6-2Amendment No. 3, 84-, 107 CONTAINMENT SYSTEMSLIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

b. The combined leakage rate to be in accordance with the Primary Containment Leakage Rate Testing Program for all primary containment penetrations and allprimary containment isolation valves that are subject to Type B and C tests,except for: main steam line isolation valves*,

valves which are hydrostatically tested, and those valves where an exemption to Appendix J of 10 CFR 50 has beengranted, andc. The leakage rate to <100 scf per hour for any main steam isolation valvethat exceeds 100 scf per hour, and restore the combined maximum pathwayleakage to <200 scf per hour, andd. The combined leakage rate for all containment isolation valves inhydrostatically tested lines which penetrate the primary containment to less than or equal to 1 gpm times the total number of such valves,prior to increasing reactor coolant system temperature above 200°F.SURVEILLANCE REQUIREMENTS 4.6.1.2 The primary containment leakage rates shall be demonstrated to be inaccordance with the Primary Containment Leakage Rate TestingProgram, or approved exemptions, for the following:

a. Type A Testb. Type B and C Tests (including air locks)c. Main Steam Line Isolation Valvesd. Hydrostatically tested Containment Isolation Valves*Exemption to Appendix "Y' to 10 CFR Part 50.LIMERICK

-UNIT 23/4 6-3Amendment No. 3, 7-4, 4-, 40-7-, 146 THIS PAGE ISINTENTIONALLY LEFT BLANKLIMERICK

-UNIT 23/4 6-4Amendment No. -34, 44, 81 CONTAINMENT SYSTEMSPRIMARY CONTAINMENT AIR LOCKLIMITING CONDITION FOR OPERATION 3.6.1.3 The primary containment air lock shall be OPERABLE with:a. Both doors closed except when the air lock is being used for normaltransit entry and exit through the containment, then at least oneair lock door shall be closed, andb. An overall air lock leakage rate in accordance with the Primary Containment Leakage Rate Testing Program.APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2*, and 3.ACTION:a. With one primary containment air lock door inoperable:

1.2.3.Maintain at least the OPERABLE air lock door closed and eitherrestore the inoperable air lock door to OPERABLE status within24 hours or lock the OPERABLE air lock door closed.Operation may then continue until performance of the nextrequired overall air lock leakage test provided that the OPERABLEair lock door is verified to be locked closed at least once per31 days.Otherwise, be in at least HOT SHUTDOWN within the next 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months sand in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .b. With the primary containment air lock inoperable, except as aresult of an inoperable air lock door, maintain at least one airlock door closed; restore the inoperable air lock to OPERABLEstatus within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT SHUTDOWN within thenext 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 .*See Special Test Exception 3.10.1.LIMERICK

-UNIT 23/4 6-5Amendment No. 8-4, 132

.CONTAINMENT SYSTEMSSURVEILLANCEROIEET 4.6.1.3 The primary containment air lock shall be demonstrated OPERABLE:

a. By verifying the seal leakage rate is in accordance with the Primary Containment Leakage Rate Testing Program.b. By conducting an overall air lock leakage test in accordance with thePrimary Containment Leakage Rate Testing Program.c. In accordance with the Surveillance Frequency Control Program byverifying that only one door in the air lock can be opened at atime.'***Except that the airlock doors need not be opened to verify interlock OPERA-BILITY when the primary containment is inerted, provided that the airlockdoors' interlock is tested within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after the primary containment hasbeen deinerted and provided the shield door to the airlock is maintained locked closed.LIMERICK

-UNIT 23/4 6-6Amendment No. 8-1, 147 CONTAINMENT SYSTEMSMSIV LEAKAGE ALTERNATE DRAIN PATHWAYLIMITING CONDITION FOR OPERATION 3.6.1.4 The MSIV Leakage Alternate Drain Pathway shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.ACTION:With the MSIV Leakage Alternate Drain Pathway inoperable, restorethe pathway to OPERABLE status within 30 days or be in at leastHOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWNwithin the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .SURVEILLANCE REQUIREMENTS 4.6.1.4 The MSIV Leakage Alternate Drain Pathway shall be demonstrated OPERABLE:

a. In accordance with 4.0.5, by cycling each motor operated valve,required to be repositioned, through at least one completecycle of full travel.LIMERICK

-UNIT 23/4 6-7Amendment No. -34, 53 CONTAINMENT SYSTEMSPRIMARY CONTAINMENT STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.5 The structural integrity of the primary containment shall bemaintained at a level consistent with the acceptance criteria in Specification 4.6.1.5.APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.ACTION:With the structural integrity of the primary containment not conforming to theabove requirements, restore the structural integrity to within the limitswithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-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 inCOLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .SURVEILLANCE REQUIREMENTS 4.6.1.5.1 The structural integrity of the exposed accessible interior andexterior surfaces of the primary containment, including the liner plate, shallbe determined by a visual inspection of those surfaces.

This inspection shallbe performed in accordance with the Primary Containment Leakage Rate TestingProgram.4.6.1.5.2 Reports Any abnormal degradation of the primary containment structure detected during the above required inspections shall be reported ina Special Report to the Commission pursuant to Specification 6.9.2 within30 days. This report shall include a description of the condition of theliner and concrete, the inspection procedure, the tolerances on cracking, andthe corrective actions taken.I TMFTICK -UNIT 2 3/4 6-R Amendment Nn R1 0 CONTAINMENT SYSTEMSDRYWELL AND SUPPRESSION CHAMBER INTERNAL PRESSURELIMITING CONDITION FOR OPERATION 3.6.1.6 Drywell and suppression chamber internal pressure shall be maintained between -1.0 and +2.0 psig.APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.ACTION:With the drywell and/or suppression chamber internal pressure outside of thespecified limits, restore the internal pressure to within the limit within1 hour 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 COLDSHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .SURVEILLANCE REQUIREMENTS 4.6.1.6 The drywell and suppression chamber internal pressure shall bedetermined to be within the limits in accordance with the Surveillance Frequency Control Program.LIMERICK

-UNIT 23/4 6-9Amendment No. 147 CONTAINMENT SYSTEMSDRYWELL AVERAGE AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.7 Drywell average air temperature shall not exceed 1450F.APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.ACTION:With the drywell average air temperature greater than 1450F, reduce theaverage air temperature to within the limit within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at leastHOT 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 thefollowinq 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .SURVEILLANCE REQUIREMENTS 4.6.1.7 The drywell average air temperature shall be the volumetric averageof the temperatures at the following locations and shall be determined to bewithin the limit in accordance with the Surveillance Frequency Control Program:Approximate Elevation

a. 330'b. 320'c. 260'd. 248'Number ofInstalled Sensors*3336* At least one reading from each elevation is required for a volumetric averagecalculation.

LIMERICK

-UNIT 23/4 6-10Amendment No. 4-2-1, 4-, 153 CONTAINMENT SYSTEMSDRYWELL AND SUPPRESSION CHAMBER PURGE SYSTEMLIMITING CONDITION FOR OPERATION 3.6.1.8 The drywell and suppression chamber purge system may be in operation with the supply and exhaust isolation valves in one supply line and one exhaustline open for inerting, deinerting, pressure

control, ALARA or air quality considerations for personnel entry, or Surveillances that require the valves to be open.APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.ACTION:a. With a drywell and/or suppression chamber purge supply and/or exhaustisolation valve open, except as permitted above, close the valve(s)within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHUTDOWN within the next 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months sand in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .SURVEILLANCE REDUIREMENTS 4.6.1.8 In accordance with the Surveillance Frequency Control Program, verify eachprimary containment purge valve [18" or 24"] is closed.*,

- - Only required to be met in OPERATIONAL CONDITIONS 1, 2, and 3.** Not required to be met when the primary containment purge valves are openfor inerting, deinerting, pressure

control, ALARA or air qualityconsiderations for personnel entry, or Surveillances that require these valvesto be open.LIMERICK

-UNIT 23/4 6-11Amendment No. L4-, @-1, 147 CONTAINMENT SYSTEMS3/4.6.2 DEPRESSURIZATION SYSTEMSSUPPRESSION CHAMBERLIMITING CONDITION FOR OPERATION 3.6.2.1 The suppression chamber shall be OPERABLE with:a. The pool water:1. Volume* between 122,120 ft3 and 134,600 ft3, equivalent to alevel between 22' 0" and 24' 3", and a2. Maximum average temperature of 95°F except that the maximumaverage temperature may be permitted to increase to:a) 105'F during testing which adds heat to the suppression chamber.b) 110°F with THERMAL POWER less than or equal to 1% of RATEDTHERMAL POWER.c) 120°F with the main steam line isolation valves closedfollowing a scram, one in each of the eight locations.

b. Drywell-to-suppression chamber bypass leakage less than or equal to10% of the acceptable A/NK design value of 0.0500 ft2.c. At least eight suppression pool water temperature instrumentation indicators, one in each of the eight locations.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.ACTION:a. With the suppression chamber water level outside the above limits,restore the water level to within the limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be inat 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 SHUTDOWNwithin the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .b. With the suppression chamber average water temperature greater than 950F,restore the average temperature to less than or equal to 95°F within24 hours 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 inCOLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , except, as permitted above:1. With the suppression chamber average water temperature greaterthan 1050F during testing which adds heat to the suppression

chamber, stop all testing which adds heat to the suppression chamber and restore the average temperature to less than 950Fwithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT SHUTDOWN within the next12 hours 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 the suppression chamber average water temperature greater than:a) 95°F for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and THERMAL POWER greater than1% of RATED THERMAL POWER, be in at least HOT SHUTDOWNwithin 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) 1100F, place the reactor mode switch in the Shutdownposition and operate at least one residual heat removal loopin the suppression pool cooling mode.*Includes the volume inside the pedestal.

LIMERICK

-UNIT 23/4 6-12 CONTAINMENT SYSTEMSLIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

3. With the suppression chamber average water temperature greaterthan 1200F, depressurize the reactor pressure vessel to lessthan 200 psig within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .c. With only one suppression chamber water level indicator OPERABLE and/orwith less than eight suppression pool water temperature indicators, one in each of the eight locations

OPERABLE, restore the inoperable indicator(s) to OPERABLE status within 7 days or verify suppression chamber water level and/or temperature to be within the limits at leastonce per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .d. With no suppression chamber water level indicators OPERABLE and/or withless than seven suppression pool water temperature indicators coveringat least seven locations

OPERABLE, restore at least one water levelindicator and at least seven water temperature indicators to OPERABLEstatus within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT SHUTDOWN within the next12 hours and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .e. With the drywell-to-suppression chamber bypass leakage in excess ofthe limit, restore the bypass leakage to within the limit prior toincreasing reactor coolant temperature above 200'F.SURVEILLANCE REDUIREMENTS 4.6.2.1 The suppression chamber shall be demonstrated OPERABLE:

a. By verifying the suppression chamber water volume to be within thelimits in accordance with the Surveillance Frequency Control Program.b. In accordance with the Surveillance Frequency Control Program byverifying the suppression chamber average water temperature to be lessthan or equal to 950F, except:1. At least once per 5 minutes during testing which adds heat tothe suppression

chamber, by verifying the suppression chamberaverage water temperature less than or equal to 1050F.2. At least once per hour when suppression chamber average watertemperature is greater than or equal to 950F, by verifying:

a) Suppression chamber average water temperature to be lessthan or equal to 1100F, andb) THERMAL POWER to be less than or equal to 1% of RATED THERMALPOWER 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after suppression chamber average watertemperature has exceeded 95°F for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .3. At least once per 30 minutes following a scram with suppression chamber average water temperature greater than or equal to 950F,by verifying suppression chamber average water temperature lessthan or equal to 120'F.LIMERICK

-UNIT 23/4 6-13Amendment No. 147 CONTAINMENT SYSTEMSSURVEILLANCE REOUIREMENTS (Continued)

c. By verifying at least 8 suppression pool water temperature indicators inat least 8 locations, OPERABLE by performance of a CHANNEL CHECK, CHANNELFUNCTIONAL TEST and CHANNEL CALIBRATION at the frequencies specified inthe Surveillance Frequency Control Program with the temperature alarmsetpoint for:1. High water temperature:

a) First setpoint

< 950Fb) Second setpoint

< 1050Fc) Third setpoint

< 1100Fd) Fourth setpoint

< 120°Fd. By verifying at least two suppression chamber water level indicators OPERABLE by performance of a CHANNEL CHECK, CHANNEL FUNCTIONAL TESTand CHANNEL CALIBRATION at the frequencies specified in theSurveillance Frequency Control Program with the water level alarmsetpoint for high water level < 24'1-1/2".

e. Drywell-to-suppression chamber bypass leak tests shall be conducted tocoincide with the Type A test at an initial differential pressure of 4 psiand verifying that the A/vrk' calculated from the measured leakage is withinthe specified limit. If any drywell-to-suppression chamber bypass leaktest 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 every24 months until two consecutive tests meet the specified limit, at which timethe test schedule may be resumed.f. By conducting a leakage test on the drywell-to-suppression chambervacuum breakers at a differential pressure of at least 4.0 psi andverifying that the total leakage area A/VF contributed by all vacuumbreakers is less than or equal to 24% of the specified limit and the leakagearea for an individual set of vacuum breakers is less than or equal to 12% ofthe specified limit. The vacuum breaker leakage test shall be conducted duringeach refueling outage for which the drywell-to-suppression chamber bypass leaktest in Specification 4.6.2.1.e is not conducted.

LIMERICK

-UNIT 23/4 6-14Amendment No. 4,3-3,34,#-,8-,,92, 147 CONTAINMENT SYSTEMSSUPPRESSION POOL SPRAYLIMITING 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:a. One OPERABLE RHR pump, andb. An OPERABLE flow path capable of recirculating water from thesuppression chamber through an RHR heat exchanger and thesuppression pool spray sparger(s).

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.ACTION:a. With one suppression pool spray loop inoperable, restore the inoperable loop to OPERABLE status within 7 days or be in at least HOT SHUTDOWNwithin 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 leastone 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 HOTSHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN*

within thefollowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .1R\/PFTI I ANC'FPRFnIIIPFMPNT 4.6.2.2 TheOPERABLE:

suppression pool spray mode of the RHR system shall be demonstrated

a. In accordance with the Surveillance Frequency Control Program byverifying that each valve (manual, power-operated, or automatic) in theflow 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 ofat least 500 gpm on recirculation flow through the RHR heat exchanger and the suppression pool spray sparger when tested pursuant to Speci-fication 4.0.5.* Whenever both RHR subsystems are inoperable, if unable to attain COLDSHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.LIMERICK

-UNIT 23/4 6-15Amendment No. W , 0, 4, 147 CONTAINMENT SYSTEMSSUPPRESSION POOL COOLINGLIMITING CONDITION FOR OPERATION 3.6.2.3 The suppression pool 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, andb. 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:a. With one suppression pool cooling loop inoperable, restore the inoperable loop to OPERABLE status within 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s** or be in at least HOT SHUTDOWNwithin 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 24hours.b. With both suppression pool cooling loops inoperable, be in at least HOTSHUTDOWN 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 REOUIREMENTS 4.6.2.3 The suppression pool cooling mode of the RHR system shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program byverifying that each valve (manual, power-operated, or automatic) in theflow 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 atleast 10,000 gpm on recirculation flow through the flow path including the RHR heat exchanger and its associated closed bypass valve, thesuppression pool and the full flow test line when tested pursuant toSpecification 4.0.5.*Whenever both RHR subsystems are inoperable, if unable to attain COLD SHUTDOWNas required by this ACTION, maintain reactor coolant temperature as low aspractical by use of alternate heat removal methods.**During the extended 7-day Allowed Outage Time (AOT) specified by TS LCO 3.7.1.1,Action a.3.a) or a.3.b) to allow for RHRSW subsystem piping repairs, the 72-hour AOTfor one inoperable suppression pool cooling loop may also be extended to 7 days forthe same 7-day period.LIMERICK

-UNIT 23/4 6-16Amendment No. 3,74,4,4,41-4, 165 CONTAINMENT SYSTEMS3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVESLIMITING CONDITION FOR OPERATION 3.6.3 Each primary containment isolation valve and each instrumentation line excessflow check valve shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.ACTION:a. With one or more of the primary containment isolation valves inoperable,**

maintain at least one isolation valve OPERABLE in each affectedpenetration that is open and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:1. Restore the inoperable valve(s) to OPERABLE status, or2. Isolate each affected penetration by use of at least one de-activated automatic valve secured in the isolated position,*

or3. Isolate each affected penetration by use of at least one closedmanual valve or blind flange.*Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months andin COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .b. With one or more of the instrumentation line excess flowcheck valves inoperable, operation may continue and the provisions ofSpecification 3.0.3 are not applicable provided that within 4hours either:1. The inoperable valve is returned to OPERABLE status, or2. The instrument line is isolated and the associated instrument is declared inoperable.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months andin COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .c. With one or more scram discharge volume vent or drain valves inoperable, perform the applicable actions specified in Specification 3.1.3.1.* Isolation valves closed to satisfy these requirements may be reopened on anintermittent basis under administrative control.** Except for the scram discharge volume vent and drain valves.LIMERICK

-UNIT 23/4 6-17Amendment No. 4-0-7, 41_, 4-321, 153 CONTAINMENT SYSTEMSSURVEILLANCE REQUIREMENTS 4.6.3.1 Each primary containment isolation valve shall be demonstrated OPERABLE prior toreturning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated

actuator, control or power circuit by cycling the valvethrough at least one complete cycle of full travel and verifying the specified isolation time.4.6.3.2 Each primary containment automatic isolation valve shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by verifying that on a containment isolation test signal each automatic isolation valve actuatesto its isolation position.

4.6.3.3 The isolation time of each primary containment power operated orautomatic valve shall be determined to be within its limit when tested pursuant toSpecification 4.0.5.4.6.3.4 A representative sample of instrumentation line excess flow check valves shallbe demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program,such that each valve is tested in accordance with the Surveillance Frequency ControlProgram, by verifying that the valve checks flow.*4.6.3.5 Each traversing in-core probe system explosive isolation valve shallbe demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program byverifying the continuity of the explosive charge.b. In accordance with the Surveillance Frequency Control Program byremoving the explosive squib from the explosive valve, such that each .explosive squib in each explosive valve will be tested in accordance with the Surveillance Frequency Control Program, and initiating theexplosive squib. The replacement charge for the exploded squib shallbe from the same manufactured batch as the one fired or from anotherbatch which has been certified by having at least one of that batchsuccessfully fired. No squib shall remain in use beyond the expiration of its shelf-life and/or operating life, as applicable.

The reactor vessel head seal leakage detection line (penetration 29A) excess flow checkvalve is not required to be tested pursuant to this requirement.

LIMERICK

-UNIT 2 3/4 6-18 Amendment No. 34,4-9-,-4-9, 147 0 TABLE 3.6.3-1 (Deleted)

THE INFORMATION FROM THISTECHNICAL SPECIFICATION SECTIONHAS BEEN RELOCAED TO THETECHNICAL REQUIREMENTS MANUAL (TRM), PCIV SECTION.TECHNICAL SPECIFICATION PAGES 3/4 6-19 THROUGH 3/4 6-43aHAVE BEEN INTENTIONALLY OMITTED.LIMERICK

-UNIT 23/4 6-19Amendment No. , -53, 107 PAGE INTENTIONALLY LEFT BLANK CONTAINMENT SYSTEMS3/4.6.4 VACUUM RELIEFSUPPRESSION CHAMBER -DRYWELL VACUUM BREAKERSLIMITING CONDITION FOR OPERATION 3.6.4.1 Three pairs of suppression chamber -drywell vacuum breakers shall beOPERABLE and all suppression chamber -drywell vacuum breakers shall be closed.APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.ACTION:a. With one or more vacuum breakers in one of the three required pairs ofsuppression chamber -drywell vacuum breaker pairs inoperable for openingbut known to be closed, restore at least one inoperable pair of vacuumbreakers 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 HOTSHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within thefollowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .b. With one suppression chamber -drywell vacuum breaker open, verifythe other vacuum breaker in the pair to be closed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ;restore the open vacuum breaker to the closed position within 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months sor 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 COLDSHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .c. With one position indicator of any suppression chamber -drywellvacuum breaker inoperable:

1. Verify the other vacuum breaker in the pair to be closed within2 hours and at least once per 15 days thereafter, or2. Verify the vacuum breaker(s) with the inoperable positionindicator to be closed by conducting a test which demonstrates that the AP is maintained at greater than or equal to 0.7 psifor one hour without makeup within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least onceper 15 days thereafter.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months andin COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .LIMERICK

-UNIT 23/4 6-44Amendment No. 9 0INTENTIONALLY LEFT BLANK CONTAINMENT SYSTEMSSURVEILLANCE REQUIREMENTS 4.6.4.1Each suppression chamber -drywell vacuum breaker shall be:a. Verified closed in accordance with the Surveillance Frequency ControlProgram.b. Demonstrated OPERABLE:

1. In accordance with the Surveillance Frequency Control Programand within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after any discharge of steam to thesuppression chamber from the safety/relief valves, by cyclingeach vacuum breaker through at least one complete cycle of fulltravel.2. In accordance with the Surveillance Frequency Control Programby verifying both position indicators OPERABLE by observing expected valve movement during the cycling test.3. In accordance with the Surveillance Frequency Control Program by:a) Verifying each valve's opening setpoint, from the closedposition, to be 0.5 psid +/- 5%, andb) Verifying both position indicators OPERABLE by performance of a CHANNEL CALIBRATION.

c) Verifying that each outboard valve's position indicator iscapable of detecting disk displacement 0.050",

and eachinboard valve's position indicator is capable of detecting disk displacement

>0.120".LIMERICK

-UNIT 23/4 6-45Amendment No. 34, 49, 147 CONTAINMENT SYSTEMS3/4.6.5 SECONDARY CONTAINMENT RFAC.TOR FNCIOSURE SECONDARY CONTATNMFNT TNTFURTTY LIMITING CONDITION FOR OPERATION 3.6.5.1.1 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.ACTION:Without REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY, restore REACTORENCLOSURE SECONDARY CONTAINMENT INTEGRITY within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOTSHUTDOWN 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 24hours.SURVEILLANCE REQUIREMENTS 4.6.5.1.1 strated by:REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall be demon-a. Verifying in accordance with the Surveillance Frequency ControlProgram that the pressure within the reactor enclosure secondary containment is greater than or equal to 0.25 inch of vacuum watergauge.b. Verifying in accordance with the Surveillance Frequency ControlProgram that:1. All reactor enclosure secondary containment equipment hatches andblowout panels are closed and sealed.2. At least one door in each access to the reactor enclosure secondary containment is closed.3. All reactor enclosure secondary containment penetrations notcapable of being closed by OPERABLE secondary containment auto-matic isolation dampers/valves and required to be closed duringaccident conditions are closed by valves, blind flanges, slidegate dampers or deactivated automatic dampers/valves secured inposition.

c. In accordance with the Surveillance Frequency Control Program:1. Verifying that one standby gas treatment subsystem will draw downthe reactor enclosure secondary containment to greater than orequal to 0.25 inch of vacuum water gauge in less than or equal to916 seconds with the reactor enclosure recirc system in operation, and2. Operating one standby gas treatment subsystem for one hour andmaintaining greater than or equal to 0.25 inch of vacuum watergauge in the reactor enclosure secondary containment at a flowrate not exceeding 2500 cfm with wind speeds of < 7.0 mph asmeasured on the wind instrument on Tower 1, elevation 30' or,if that instrument is unavailable, Tower 2, elevation 159'.LIMERICK

-UNIT 23/4 6-46Amendment No. -4, , 86, 147 CONTAINMENT SYSTEMS3/4.6.5 SECONDARY CONTAINMENT REFUELING AREA SECONDARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.5.1.2 REFUELING AREA SECONDARY CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY:

When RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining thereactor vessel, with the vessel head removed and fuel in the vessel.ACTION:Without REFUELING AREA SECONDARY CONTAINMENT INTEGRITY, suspend handling ofRECENTLY IRRADIATED FUEL in the secondary containment, and operations with a potential for draining the reactor vessel. The provisions of Specifica-tion 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.6.5.1.2 REFUELING AREA SECONDARY CONTAINMENT INTEGRITY shall be demonstrated by:a. Verifying in accordance with the Surveillance Frequency ControlProgram that the pressure within the refueling area secondary containment is greater than or equal to 0.25 inch of vacuum watergauge.b. Verifying in accordance with the Surveillance Frequency ControlProgram that:1. All refueling area secondary containment equipment hatches andblowout panels are closed and sealed.2. At least one door in each access to the refueling area secondary containment is closed.3. All refueling area secondary containment penetrations not capableof being closed by OPERABLE secondary containment automatic iso-lation dampers/valves and required to be closed during accidentconditions are closed by valves, blind flanges, slide gatedampers or deactivated automatic dampers/valves secured inposition.

c. In accordance with the Surveillance Frequency Control Program:Operating one standby gas treatment subsystem for one hour and main-taining greater than or equal to 0.25 inch of vacuum water gaugein the refueling area secondary containment at a flow rate not exceeding 764 cfm.LIMERICK

-UNIT 23/4 6-47Amendment No. -4, 444, 147 CONTAINMENT SYSTEMSREACTOR ENCLOSURE SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVESLIMITING CONDITION FOR OPERATION 3.6.5.2.1 The reactor enclosure secondary containment ventilation system auto-matic isolation valves shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.ACTION:With one or more of the reactor secondary containment ventilation systemautomatic isolation valves inoperable, maintain at least one isolation valveOPERABLE in each affected penetration that is open and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months either:a. Restore the inoperable valves to OPERABLE status, orb. Isolate each affected penetration by use of at least one deactivated valve secured in the isolation

position, orc. Isolate each affected penetration by use of at least one closed manualvalve, blind flange or slide gate damper.Otherwise, in OPERATIONAL CONDITION 1, 2, or 3, be in at least HOT SHUTDOWNwithin 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 REOUIREMENTS 4.6.5.2.1 Each reactor enclosure secondary containment ventilation systemautomatic isolation valve shall be demonstrated OPERABLE:

a. Prior to returning the valve to service after maintenance, repair orreplacement work is performed on the valve or its associated

actuator, control or power circuit by cycling the valve through at least one completecycle of full travel and verifying the specified isolation time.b. In accordance with the Surveillance Frequency Control Program byverifying that on a containment isolation test signal each isolation valve actuates to its isolation position.

c. By verifying the isolation time to be within its limit in accordance with the Surveillance Frequency Control Program.LIMERICK

-UNIT 23/4 6-48Amendment No. 34, 4-9, 147 THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE TRM.LIMERICK

-UNIT 23/4 6-49Amendment No. -6, 153 CONTAINMENT SYSTEMSREFUELING AREA SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVESLIMITING CONDITION FOR OPERATION 3.6.5.2.2 The refueling area secondary containment ventilation system automatic isolation valves shall be OPERABLE.

APPLICABILITY:

When RECENTLY IRRADIATED FUEL is being handled in the secondary containment, or during operations with a potential for draining thereactor vessel, with the vessel head removed and fuel in the vessel.ACTION:With one or more of the refueling area secondary containment ventilation systemautomatic isolation valves inoperable, maintain at least one isolation valve OPERABLEin each affected penetration that is open and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months either:a. Restore the inoperable valves to OPERABLE status, orb. Isolate each affected penetration by use of at least one deactivated valve secured in the isolation

position, orc. Isolate each affected penetration by use of at least one closed manualvalve, blind flange or slide gate damper.Otherwise, suspend handling of RECENTLY IRRADIATED FUEL in the refueling area secondary containment, and operations with a potential for draining thereactor vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.6.5.2.2 Each refueling area secondary containment ventilation system auto-matic isolation valve shall be demonstrated OPERABLE:

a. Prior to returning the valve to service after maintenance, repair orreplacement work is performed on the valve or its associated

actuator, control or power circuit by cycling the valve through at least onecomplete cycle of full travel arid verifying the specified isolation time.b. In accordance with the Surveillance Frequency Control Program byverifying that on a containment isolation test signal each isolation valve actuates to its isolation position.

c. By verifying the isolation time to be within its limit in accordance with the Surveillance Frequency Control Program.LIMERICK

-UNIT 23/4 6-50Amendment No. -34, 4ý, -446, 147 THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE TRM.LIMERICK

-UNIT 23/4 6-51Amendment No. 49, 153 THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE TRM.LIMERICK

-UNIT 23/4 6-51aAmendment No. 64, 153 CONTAINMENT SYSTEMSSTANDBY GAS TREATMENT SYSTEM -COMMON SYSTEMLIMITING CONDITION FOR OPERATION 3.6.5.3 Two independent standby gas treatment subsystems shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, and when (1) irradiated fuel isbeing handled in the refueling area secondary containment, or (2) during COREALTERATIONS, or (3) during operations with a potential for draining the reactorvessel with the vessel head removed and fuel in the vessel.ACTION:a. In OPERATIONAL CONDITION 1, 2, or 3:1. With the Unit 1 diesel generator for one standby gas treatment subsystem inoperable for more than 30 days, be in at least HOTSHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN withinthe following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .2. With one standby gas treatment subsystem inoperable, restorethe inoperable subsystem to OPERABLE status within 7 days, orbe 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 inCOLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .3. With one standby gas treatment subsystem inoperable and theother standby gas treatment subsystem with an inoperable Unit 1diesel generator, restore the inoperable subsystem to OPERABLEstatus or restore the inoperable Unit 1 diesel generator toOPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or be in at least HOT SHUTDOWNwithin the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within thefollowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .4. With the Unit 1 diesel generators for both standby gastreatment system subsystems inoperable for more than72 hours, be in at least HOT SHUTDOWN within the next 12hours and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .b. When (1) irradiated fuel is being handled in the refueling areasecondary containment, or (2) during CORE ALTERATIONS, or (3) duringoperations with a potential for draining the reactor vessel with thevessel head removed and fuel in the vessel.:1. With one standby gas treatment subsystem inoperable, restorethe inoperable subsystem to OPERABLE status within 7 days, orsuspend handling of irradiated fuel in the secondary containment, CORE ALTERATIONS, and operations with a potential for drainingthe reactor vessel. The provisions of Specification 3.0.3 arenot applicable.

2. With both standby gas treatment subsystems inoperable, if inprogress, suspend handling of irradiated fuel in the secondary containment, CORE ALTERATIONS, and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.

LIMERICK

-UNIT 23/4 6-52Amendment No. 43k, 146 CONTAINMENT SYSTEMSSURVEILLANCE REOUIREMENTS 4.6.5.3 Each standby gas treatment subsystem shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program byinitiating, from the control room, flow through the HEPA filters andcharcoal adsorbers and verifying that the subsystem operates with theheaters OPERABLE.

LIMERICK

-UNIT 23/4 6-52aAmendment No. 147 CONTAINMENT SYSTEMSSURVEILLANCE REQUIREMENTS (Continued)

b. In accordance with the Surveillance Frequency Control Program or (1)after any structural maintenance on the HEPA filter or charcoaladsorber

housings, or (2) following

painting, fire, or chemicalrelease in any ventilation zone communicating with the subsystem by:1. Verifying that the subsystem satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05%and uses the test procedure guidance in Regulatory Positions C.5.a,C.5.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978,and the system flow rate is 5764 cfm +/- 10%.2. Verifying within 31 days after removal that a laboratory analysisof a representative carbon sample obtained in accordance withRegulatory Position C.6.b of Regulatory Guide 1.52, Revision 2,March 1978, shows the methyl iodide penetration of less than 0.5%when tested in accordance with ASTM D3803-1989 at a temperature of300C (86°F), at a relative humidity of 70% and at a face velocityof 66 fpm.3. Verify that when the fan is running the subsystem flowrate is2800 cfm minimum from each reactor enclosure (Zones I and II)and 2200 cfm minimum from the refueling area (Zone III) whentested in accordance with ANSI N510-1980.

4. Verify that the pressure drop across the refueling area to SGTSprefilter is less than 0.25 inches water gage while operating at a flow rate of 2400 cfm +/- 10%.c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,shows the methyl iodide penetration of less than 0.5% when testedin accordance with ASTM D3803-1989 at a temperature of 30'C (86°F),at a relative humidity of 70% and at a face velocity of 66 fpm.d. In accordance with the Surveillance Frequency Control Program by:1. Verifying that the pressure drop across the combined HEPAfilters and charcoal adsorber banks is less than 9.1 incheswater gauge while operating the filter train at a flow rate of8400 cfm +/- 10%.LIMERICK

-UNIT 23/4 6-53Amendment No. 34, 89, 4-06, 147 CONTAINMENT SYSTEMSSURVEILLANCE REQUIREMENTS (Continued)

2. Verifying that the fan starts and isolation valves necessary todraw a suction from the refueling area or the reactor enclosure recirculation discharge open on each of the following test signals:a) Manual initiation from the control room, andb) Simulated automatic initiation signal.3. Verifying that the temperature differential across each heateris > 150F when tested in accordance with ANSI N510-1980.

e. After each complete or partial replacement of a HEPA filter bank byverifying that the HEPA filter bank satisfies the inplace penetration and leakage testing acceptance criteria of less than 0.05% inaccordance with ANSI N510-1980 while operating the system at a flowrate of 5764 cfm +/- 10%.f. After each complete or partial replacement of a charcoal adsorberbank by verifying that the charcoal adsorber bank satisfies the inplacepenetration and leakage testing acceptance criteria of less than 0.05%in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of5764 cfm +/- 10%.g. After any major system alteration:

1. Verify that when the SGTS fan is running the subsystem flowrateis 2800 cfm minimum from each reactor enclosure (Zones I andII) and 2200 cfm minimum from the refueling area (Zone III).2. Verify that one standby gas treatment subsystem will drawdownreactor enclosure Zone II secondary containment to greater thanor equal to 0.25 inch of vacuum water gauge in less than orequal to 916 seconds with the reactor enclosure recirculation system in operation and the adjacent reactor enclosure andrefueling area zones are in their isolation modes.LIMERICK

-UNIT 23/4 6-54Amendment No. 15-1, 86 CONTAINMENT SYSTEMSREACTOR ENCLOSURE RECIRCULATION SYSTEMLIMITING CONDITION FOR OPERATION 3.6.5.4 Two independent reactor enclosure recirculation subsystems shall beOPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.ACTION:a. With one reactor enclosure recirculation subsystem inoperable, restorethe inoperable subsystem to OPERABLE status within 7 days, or be inat 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 SHUTDOWNwithin the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .b. With both reactor enclosure recirculation subsystems inoperable, bein 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 COLDSHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .SURVEILLANCE REQUIREMENTS 4.6.5.4 Each reactor enclosure recirculation subsystem shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program byinitiating, from the control room, flow through the HEPA filtersand charcoal adsorbers and verifying that the subsystem operatesproperly.

b. In accordance with the Surveillance Frequency Control Program or(1) after any structural maintenance on the HEPA filter or charcoaladsorber

housings, or (2) following

painting, fire, or chemicalrelease in any ventilation zone communicating with the subsystem by:1. Verifying that the subsystem satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05%and uses the test procedure guidance in Regulatory Positions C.5.a,C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978,and the system flow rate is 60,000 cfm +/- 10%.2. Verifying within 31 days after removal that a laboratory analysisof a representative carbon sample obtained in accordance withRegulatory Position C.6.b of Regulatory Guide 1.52, Revision 2,March 1978, shows the methyl iodide penetration of less than 2.5%when tested in accordance with ASTM D3803-1989 at a temperature of30'C (86°F) and a relative humidity of 70%.3. Verifying a subsystem flow rate of 60,000 cfm +/- 10% during systemoperation when tested in accordance with ANSI N510-1980.

LIMERICK

-UNIT 23/4 6-55Amendment No. -34, 4-46, 147 CONTAINMENT SYSTEMSSURVEILLANCE REOUIREMENTS (Continued)

c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,shows the methyl iodide penetration of less than 2.5% when testedin accordance with ASTM D3803-1989 at a temperature of 30'C (86°F)an a relative humidity of 70%.d. In accordance with the Surveillance Frequency Control Program by:1. Verifying that the pressure drop across the combined prefilter, upstream and downstream HEPA filters, and charcoal adsorberbanks is less than 6 inches water gauge while operating thefilter train at a flow rate of 60,000 cfm +/- 10%, verifying thatthe prefilter pressure drop is less than 0.8 inch water gaugeand that the pressure drop across each HEPA is less than 2inches water gauge.2. Verifying that the filter train starts and the isolation valveswhich take suction on and return to the reactor enclosure openon each of the following test signals:a. Manual initiation from the control room, andb. Simulated automatic initiation signal.e. After each complete or partial replacement of a HEPA filter bank byverifying that the HEPA filter bank satisfies the inplace penetration and leakage testing acceptance criteria of less than 0.05% inaccordance with ANSI N510-1980 while operating the system at a flowrate of 60,000 cfm +/- 10%.f. After each complete or partial replacement of a charcoal adsorberbank by verifying that the charcoal adsorber bank satisfies theinplace penetration and leakage testing acceptance criteria of lessthan 0.05% in accordance with ANSI N510-1980 for a halogenated hydro-carbon refrigerant test gas while operating the system at a flowrate of 60,000 cfm +/- 10%.ITMFRICfK

-UNIT 2 3/4 6-56 Amendment No -34 4- 147 0 CONTAINMENT SYSTEMS3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROLPRIMARY CONTAINMENT HYDROGEN RECOMBINER SYSTEMSLIMITING CONDITION FOR OPERATION 3.6.6.1 DELETEDLIMERICK

-UNIT 23/4 6-57Amendment No. 34, 135 CONTAINMENT SYSTEMSDRYWELL HYDROGEN MIXING SYSTEMLIMITING CONDITION FOR OPERATION 3.6.6.2 Four independent drywell unit cooler hydrogen mixing subsystems (2AV212, 2BV212, 2GV212, 2HV212) shall be OPERABLE with each subsystem consist-ing of one unit cooler fan.APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.ACTION:With one drywell unit cooler hydrogen mixing subsystem inoperable, restorethe inoperable subsystem to OPERABLE status within 30 days or be in at leastHOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .SURVEILLANCE REQUIREMENTS 4.6.6.2 Each drywell unit cooler hydrogen mixing subsystem shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by:a. Starting the system from the control room, andb. Verifying that the system operates for at least 15 minutes.LIMERICK

-UNIT 23/4 6-58Amendment No. 147 CONTAINMENT SYSTEMSDRYWELL AND SUPPRESSION CHAMBER OXYGEN CONCENTRATION LIMITING CONDITION FOR OPERATION 3.6.6.3 The drywell and suppression chamber atmosphere oxygen concentration shall be less than 4% by volume.APPLICABILITY:

OPERATIONAL CONDITION 1*, during the time period:a. Within 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s** after THERMAL POWER is greater than 15% of RATEDTHERMAL POWER, following

startup, tob. Within 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s** prior to reducing THERMAL POWER to less than 15% ofRATED THERMAL POWER, preliminary to a scheduled reactor shutdown.

ACTION:With the drywell and/or suppression chamber oxygen concentration exceeding the limit, restore the oxygen concentration to within the limitwithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least STARTUP within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .SURVEILLANCE REOUIREMENTS 4.6.6.3 The drywell and suppression chamber oxygen concentration shallbe verified to be within the limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER isgreater than 15% of RATED THERMAL POWER and in accordance with the Surveillance Frequency Control Program thereafter.

See Special Test Exception 3.10.5.**Specification 3.6.1.8 is applicable during this 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period.LIMERICK

-UNIT 23/4 6-59Amendment No. 147 THIS PAGE INTENTIONALLY LEFT BLANK 3/4.7 PLANT SYSTEMS3/4.7.1 SERVICE WATER SYSTEMSRESIDUAL HEAT REMOVAL SERVICE WATER SYSTEM -COMMON SYSTEMLIMITING CONDITION FOR OPERATION 3.7.1.1 At least the following independent residual heat removal service water(RHRSW) system subsystems, with each subsystem comprised of:a. Two OPERABLE RHRSW pumps, andb. An OPERABLE flow path capable of taking suction from the RHR servicewater pumps wet pits which are supplied from the spray pond or thecooling tower basin and transferring the water through one Unit 2RHR heat exchanger, shall be OPERABLE:

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

b. In OPERATIONAL CONDITIONS 4 and 5, the subsystem(s) associated withsystems 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 toOPERABLE status within 30 days, or be in at least HOT SHUTDOWNwithin the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within thefollowing 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 atleast one of the inoperable RHRSW pumps to OPERABLE statuswithin 7 days or be in at least HOT SHUTDOWN within the next12 hours 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 theinoperable subsystem to OPERABLE status with at least oneOPERABLE RHRSW pump within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , unless otherwise specified in a) or b) below**,

or be in at least HOT SHUTDOWN within thenext 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24hours.a) When the 'A' RHRSW subsystem is inoperable to allow forrepairs of the 'A' RHRSW subsystem piping, with LimerickGenerating Station Unit I shutdown, reactor vessel headremoved and reactor cavity flooded, the 72-hour AllowedOutage Time may be extended to 7 days once every othercalendar year with the following compensatory measuresestablished:

- Only one of these two Actions, either a.3.a) or a.3.b), may be entered on Unit 2in a calendar year. However, if either Unit 1 TS LCO 3.7.1.1, Action a.3.a) ora.3.b) has previously been entered in the calendar year, then Unit 2 Actiona.3.a) or a.3.b) may not be entered during that same calendar year.LIMERICK

-UNIT 23/4 7-1Amendment No. 3,-0,49, 165 PLANT SYSTEMSLIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

1) The following systems and subsystems will be protected in accordance with applicable station procedures:

'B' RHRSW subsystem

'B' ESW loop* 'B' and 'D' RHR subsystems

D12, D22, and D24 4kV buses and emergency dieselgenerators

Division 2 and Division 4 Safeguard DC, and2) The 'A' and 'B' loop of ESW return flow shall bealigned to the operable

'B' RHRSW return header only.The ESW return valves to the 'B' RHRSW return header(i.e., HV-11-015A and HV-11-015B) will beadministratively controlled in the open position andde-energized prior to entering the extended AOT. TheESW return valves to the 'A' RHRSW return header (i.e.,HV-11-O11A and HV-11-011B) will be administratively controlled in the closed position and de-energized aspart of the work boundary.

b) When the 'B' RHRSW subsystem is inoperable to allow forrepairs of the 'B' RHRSW subsystem piping, with LimerickGenerating Station Unit 1 shutdown, reactor vessel headremoved and reactor cavity flooded, the 72-hour AllowedOutage Time may be extended to 7 days once every othercalendar year with the following compensatory measuresestablished:

1) The following systems and subsystems will be protected in accordance with applicable station procedures:

" 'A' RHRSW subsystem

'A' ESW loop" 'A' and 'C' RHR subsystems

" Dll, D21, and D23 4kV buses and emergency dieselgenerators

" Division 1 and Division 3 Safeguard DC, and2) The 'A' and 'B' loop of ESW return flow shall bealigned to the operable

'A' RHRSW return header only.The ESW return valves to the 'A' RHRSW return header(i.e., HV-11-O11A and HV-11-O11B) will beadministratively controlled in the open position andde-energized prior to entering the extended AOT. TheESW return valves to the 'B' RHRSW return header (i.e.,HV-11-015A and HV-11-015B) will be administratively controlled in the closed position and de-energized aspart of the work boundary.

4. With both RHRSW subsystems otherwise inoperable, restore atleast one subsystem to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be inat least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLDSHUTDOWN*

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 SHUTDOWNas required by this ACTION, maintain reactor coolant temperature as low aspractical by use of alternate heat removal methods.LIMERICK

-UNIT 2 3/4 7-1a Amendment No. 4O,7-9,49, 165 THIS PAGE INTENTIONALLY LEFT BLANK PLANT SYSTEMSLIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

5. With two RHRSW pump/diesel generator pairs* inoperable, restoreat least one inoperable RHRSW pump/diesel generator pair* toOPERABLE status within 30 days or be in at least HOT SHUTDOWNwithin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .6. With three RHRSW pump/diesel generator pairs* inoperable, restore at least one inoperable RHRSW pump/diesel generator pair* to OPERABLE status within 7 days or be in at least HOTSHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within thefollowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .7. With four RHRSW pump/diesel generator pairs* inoperable, restore at least one inoperable RHRSW pump/diesel generator pair* 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 HOTSHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within thefollowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .b. In OPERATIONAL CONDITION 3 or 4 with the RHRSW subsystem(s),

which isassociated with an RHR loop required OPERABLE by Specification 3.4.9.1or 3.4.9.2, inoperable, declare the associated RHR loop inoperable andtake the ACTION required by Specification 3.4.9.1 or 3.4.9.2, asapplicable.

c. In OPERATIONAL CONDITION 5 with the RHRSW subsystem(s),

which isassociated with an RHR loop required OPERABLE by Specification 3.9.11.1 or 3.9.11.2, inoperable, declare the associated RHR systeminoperable and take the ACTION required by Specification 3.9.11.1or 3.9.11.2, as applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.1 At least the above required residual heat removal service water systemsubsystem(s) shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program byverifying that each valve in the flow path that is not locked, sealed,or otherwise secured in position, is in its correct position.

A RHRSW pump/diesel generator pair consists of a RHRSW pump and its associated diesel generator.

If either a RHRSW pump or its associated diesel generator becomes inoperable, then the RHRSW pump/diesel generator pair is inoperable.

LIMERICK

-UNIT 23/4 7-2Amendment No. 4-3-21, 147 PLANT SYSTEMSEMERGENCY SERVICE WATER SYSTEM -COMMON SYSTEMLIMITING CONDITION FOR OPERATION 3.7.1.2 At least the following independent emergency service water system loops,with each loop comprised of:a. Two OPERABLE emergency service water pumps, andb. An OPERABLE flow path capable of taking suction from the emergency service water pumps wet pits which are supplied from the spray pond orthe cooling tower basin and transferring the water to the associated Unit2 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:a. In OPERATION CONDITION 1, 2, or 3:1. With one emergency service water pump inoperable, restore the inoperable pump to OPERABLE status within 45 days or be in at least HOT SHUTDOWN withinthe 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 pump in each loop inoperable, restore atleast one inoperable pump to OPERABLE status within 30 days or be in atleast 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 thefollowing 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, declareall equipment aligned to the inoperable loop inoperable**,

restore theinoperable loop to OPERABLE status with at least one OPERABLE pump within 72hours" 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 COLDSHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .*When handling irradiated fuel in the secondary containment.

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

Those diesel generators not aligned to the OPERABLE emergency service water system loop shall be declaredinoperable and the actions of 3.8.1.1 taken.'During the extended 7-day Allowed Outage Time (AOT) specified by TS LCO 3.7.1.1, Actiona.3.a) or a.3.b) to allow for RHRSW subsystem piping repairs, the 72-hour AOT for oneinoperable emergency service water system loop may also be extended to 7 days for the same7-day period.LIMERICK

-UNIT 23/4 7-3Amendment No. 8,.-Q,2 165 PLANT SYSTEMSLIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

4. With three ESW pump/diesel generator pairs** inoperable, restoreat least one inoperable ESW pump/diesel generator pair** toOPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOWNwithin the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within thefollowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .5. With four ESW pump/diesel generator pairs** inoperable, restoreat least one inoperable ESW pump/diesel generator pair** toOPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWNwithin 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. In OPERATIONAL CONDITION 4 or 5:1. With only one emergency service water pump and its associated flow path OPERABLE, restore at least two pumps with at leastone flow path to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or declarethe associated safety related equipment inoperable and take theACTION required by Specifications 3.5.2 and 3.8.1.2.c. In OPERATIONAL CONDITION

1. With only one emergency service water pump and its associated flow path OPERABLE, restore at least two pumps with at leastone flow path to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or verifyadequate cooling remains available for the diesel generators required to be OPERABLE or declare the associated diesel genera-tor(s) inoperable and take the ACTION required by Specifica-tion 3.8.1.2.

The provisions of Specification 3.0.3 are notapplicable.

SURVEILLANCE REOUIREMENT 4.7.1.2 At least the above required emergency service water system loop(s)shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying that each valve (manual, power-operated, or automatic) that is not locked,sealed, or otherwise secured in position, is in its correct position.

b. In accordance with the Surveillance Frequency Control Program by verifying that:1. Each automatic valve actuates to its correct position on itsappropriate ESW pump start signal.2. Each pump starts automatically when its associated dieselgenerator starts.* When handling irradiated fuel in the secondary containment.

- - An ESW pump/diesel generator pair consists of an ESW pump and its associated diesel generator.

If either an ESW pump or its associated diesel generator becomes inoperable, than the ESW pump/diesel generator pair is inoperable.

LIMERICK

-UNIT 23/4 7-4Amendment No. -4, 147 PLANT SYSTEMSULTIMATE HEAT SINKLIMITING CONDITION FOR OPERATION 3.7.1.3 The spray pond shall be OPERABLE with:a. A minimum pond water level at or above elevation 250'-10" Mean SeaLevel, andb. A pond water temperature of less than or equal to 88°F.APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, 4, 5, and *ACTION:With the requirements of the above specification not satisfied:

a. In OPERATIONAL CONDITION 1, 2, or 3, be in at least HOT SHUTDOWNwithin 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. In OPERATIONAL CONDITION 4 or 5, declare the RHRSW system and theemergency service water system inoperable and take the ACTIONrequired by Specifications 3.7.1.1 and 3.7.1.2.c. In OPERATIONAL CONDITION

, declare the emergency service water systeminoperable and take the ACTION required by Specification 3.7.1.2.The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.3 The spray pond shall be determined OPERABLE:

a. By verifying the pond water level to be greater than its limitin accordance with the Surveillance Frequency Control Program.b. By verifying the water surface temperature (within the upper two feetof the surface) to be less than or equal to 880F:1. in accordance with the Surveillance Frequency Control Programwhen the spray pond temperature is greater than or equal to80°F; and2. in accordance with the Surveillance Frequency Control Programwhen the spray pond temperature is greater than or equal to850F; and3. in accordance with the Surveillance Frequency Control Programwhen the spray pond temperature is greater than 32°F.c. By verifying all piping above the frost line is drained:1. within one (1) hour after being used when ambient airtemperature is below 40'F; or2. when ambient air temperature falls below 40°F if the pipinghas not been previously drained.*When handling irradiated fuel in the secondary containment.

LIMERICK

-UNIT 23/4 7-5Amendment No. -54, 147 PLANT SYSTEMS3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM -COMMON SYSTEMLIMITING CONDITION FOR OPERATION 3.7.2 Two independent control room emergency fresh air supply system subsystems shall be OPERABLE.

NOTE: The main control room envelope (CRE) boundary may be opened intermittently underadministrative control.APPLICABILITY:

All OPERATIONAL CONDITIONS and when RECENTLY IRRADIATED FUEL is beinghandled in the secondary containment, or during operations with apotential for draining the reactor vessel.ACTION:a. In OPERATIONAL CONDITION 1, 2, or 3:1. With the Unit 1 diesel generator for one control room emergency freshair supply subsystem inoperable for more than 30 days, be in at leastHOT 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 thefollowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .2. With one control room emergency fresh air supply subsystem inoperable for reasons other than Condition a.5, restore the inoperable subsystem to OPERABLE status within 7 days, or be in at least HOTSHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within thefollowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .3. With one control room emergency fresh air supply subsystem inoperable for reasons other than Condition a.5, and the other control roomemergency fresh air supply subsystem with an inoperable Unit 1 dieselgenerator, restore the inoperable subsystem to OPERABLE status orrestore the Unit 1 diesel generator to OPERABLE status within 72hours, or be in at least HOT SHUTDOWN within the next 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months sand inCOLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .4. With the Unit 1 diesel generators for both control room emergency fresh air supply subsystems inoperable for more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ,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 inCOLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .5. With one or more control room emergency fresh air supplysubsystems inoperable due to an inoperable CRE boundary,

a. Initiate action to implement mitigating actions immediately orbe in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN withinthe following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; andb. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , verify mitigating actions ensure CREoccupant exposures to radiological and chemical hazardswill not exceed limits and actions to mitigate exposure tosmoke hazards are taken or be in HOT SHUTDOWN within 12hours and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ;andLIMERICK

-UNIT 23/4 7-6Amendment No. 4-321, 1-46, 149 PLANT SYSTEMSLIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

c. Restore CRE boundary to operable status within 90 days or be in HOTSHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .b. In OPERATIONAL CONDITION 4, 5 or when RECENTLY IRRADIATED FUEL is beinghandled in the secondary containment, or during operations with apotential for draining the reactor vessel:1. With one control room emergency fresh air supply subsystem inoperable for reasons other than Condition b.3, restore theinoperable subsystem to OPERABLE status within 7 days, or initiateand maintain operation of the OPERABLE subsystem in the radiation isolation mode of operation.

2. With both control room emergency fresh air supply subsystem inoperable for reasons other than Condition b.3, immediately suspend handling of RECENTLY IRRADIATED FUEL in the secondary containment and operations with a potential for draining the reactorvessel. The provisions of Specification 3.0.3 are not applicable.

3. With one or more control room emergency fresh air subsystems inoperable due to an inoperable CRE boundary, immediately suspend handling of RECENTLY IRRADIATED FUEL in the secondary containment and operations with a potential for draining thereactor vessel. The provisions of Specification 3.0.3 are notapplicable.

SURVEILLANCE REOUIREMENTS 4.7.2.1 Each control room emergency fresh air supply subsystem shall bedemonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program byverifying the control room air temperature to be less than or equal to850F effective temperature.

b. In accordance with the Surveillance Frequency Control Program on aSTAGGERED TEST BASIS by initiating, from the control room, flow throughthe HEPA filters and charcoal adsorbers and verifying that the subsystem operates with the heaters OPERABLE.

c. In accordance with the Surveillance Frequency Control Program or (1)after any structural maintenance on the HEPA filter or charcoal adsorberhousings, or (2) following

painting, fire, or chemical release in anyventilation zone communicating with the subsystem by:1. Verifying that the subsystem satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than0.05% and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2,March 1978, and the system flow rate is 3000 cfm +/- 10%.LIMERICK

-UNIT 23/4 7-6aAmendment No. -4, 4-9, 153 INTENTIONALLY LEFT BLANK PLANT SYSTEMSSURVEILLANCE REOUIREMENTS (Continued)

2. Verifying within 31 days after removal that a laboratory analysisof a representative carbon sample obtained in accordance withRegulatory Position C.6.b of Regulatory Guide 1.52, Revision 2,March 1978, shows the methyl iodide penetration of less than 2.5%when tested in accordance with ASTM D3803-1989 at a temperature of 300C (860F) and a relative humidity of 70%.3. Verifying a subsystem flow rate of 3000 cfm +/- 10% during subsystem operation when tested in accordance with ANSI N510-1980.

d. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,shows the methyl iodide penetration of less than 2.5% when testedin accordance with ASTM D3803-1989 at a temperature of 30'C (860F)and a relative humidity of 70%.e. In accordance with the Surveillance Frequency Control Program by:1. Verifying that the pressure drop across the combined prefilter, upstream and downstream HEPA filters, and charcoal adsorber banksis less than 6 inches water gauge while operating the subsystem at a flow rate of 3000 cfm +/- 10%; verifying that the prefilter pressure drop is less than 0.8 inch water gauge and that thepressure drop across each HEPA is less than 2 inches water gauge.2. Verifying that on each of the below chlorine isolation modeactuation test signals, the subsystem automatically switchesto the chlorine isolation mode of operation and the isolation valves close within 5 seconds:a) Outside air intake high chlorine, andb) Manual initiation from the control room.3. Verifying that on each of the below radiation isolation modeactuation test signals, the subsystem automatically switches tothe radiation isolation mode of operation:

a) Outside air intake high radiation, andb) Manual initiation from control room.LIMERICK

-UNIT 23/4 7-7Amendment No. 34, 4-"6, 4, 149 PLANT SYSTEMSSURVEILLANCE REQUIREMENTS (Continued)

f. After each complete or partial replacement of a HEPA filter bank byverifying that the HEPA filter bank satisfies the inplace penetra-tion and bypass leakage testing acceptance criteria of less than 0.05%in accordance with ANSI N510-1980 while operating the system at aflow rate of 3000 cfm +/- 10%.g. After each complete or partial replacement of a charcoal adsorberbank by verifying that the charcoal adsorber bank satisfies theinplace penetration and bypass leakage testing acceptance criteriaof less than 0.05% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at aflow rate of 3000 cfm +/- 10%.4.7.2.2 The control room envelope boundary shall be demonstrated OPERABLE:

a. At a frequency in accordance with the Control Room Envelope Habitability Program by performance of control room envelope unfiltered air inleakage testing in accordance with the Control Room Envelope Habitability Program.LIMERICK

-UNIT 23/4 7-8Amendment No. 149 PLANT SYSTEMS3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEMLIMITING CONDITION FOR OPERATION 3.7.3 The reactor core isolation cooling (RCIC) system shall be OPERABLE withan OPERABLE flow path capable of automatically taking suction from thesuppression pool and transferring the water to the reactor pressure vessel.APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3 with reactor steam domepressure greater than 150 psig.ACTION:a. With the RCIC system inoperable, operation may continue provided theHPCI system is OPERABLE; restore the RCIC system to OPERABLE status within14 days. Otherwise, be in at least HOT SHUTDOWN within the next 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months sand reduce reactor steam dome pressure to less than or equal to 150 psigwithin the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .b. In the event the RCIC system is actuated and injects water into thereactor coolant system, a Special Report shall be prepared and sub-mitted to the Commission pursuant to Specification 6.9.2 within90 days describing the circumstances of the actuation and the totalaccumulated actuation cycles to date.c. Specification 3.0.4.b is not applicable to RCIC.SUIRVEILLANCE REOUIREMENTS 4.7.3 The RCIC system shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by:1. Verifying by venting at the high point vents that the systempiping from the pump discharge valve to the system isolation valve is filled with water.2. Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise securedin position, is in its correct position.

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

b. In accordance with the Surveillance Frequency Control Program byverifying that the RCIC pump develops a flow of greater than or equalto 600 gpm in the test flow path with a system head corresponding toreactor vessel operating pressure when steam is being supplied to theturbine at 1040 + 13, -120 psig.*The provisions of Specification 4.0.4 are not applicable provided thesurveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure isadequate to perform the test. If OPERABILITY is not successfully demonstrated within the 12-hour period, reduce reactor steam dome pressure to less than150 psig within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .LIMERICK

-UNIT 23/4 7-9Amendment No. S-i, 44-, 147

.PLANT SYSTEMSSURVEILLANCE REQUIREMENTS (Continued) 10c. In accordance with the Surveillance Frequency Control Program by:1. Performing a system functional test which includes simulated automatic actuation and restart and verifying that eachautomatic valve in the flow path actuates to its correctposition.

Actual injection of coolant into the reactorvessel may be excluded.

2. Verifying that the system will develop a flow of greater thanor equal to 600 gpm in the test flow path when steam issupplied to the turbine at a pressure of 150 + 15, -0 psig.*3. Verifying that the suction for the RCIC system is automatically transferred from the condensate storage tank to the suppression pool on a condensate storage tank water level-low signal.4. Performing a CHANNEL CALIBRATION of the RCIC system discharge line "keep filled" level alarm instrumentation.

The provisions of Specification 4.0.4 are not applicable provided thesurveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure isadequate to perform the tests. If OPERABILITY is not successfully demonstrated within the 12-hour period, reduce reactor steam dome pressure to less than150 psig within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .LIMERICK

-UNIT 23/4 7-10Amendment No. -34, 147 PLANT SYSTEMS3/4.7.4 SNUBBERSLIMITING CONDITION FOR OPERATION 3.7.4 All snubbers shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3. OPERATIONAL CONDITIONS 4 and5 for snubbers located on systems required OPERABLE in those OPERATIONAL CONDITIONS.

ACTION:With one or more snubbers inoperable on any system, within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months replace orrestore the inoperable snubber(s) to OPERABLE status and perform an engineering evaluation per Specification 4.7.4g on the attached component or declare theattached system inoperable and follow the appropriate ACTION statement for thatsystem.SURVEILLANCE REOUIREMENTS 4.7.4 Each snubber shall be demonstrated OPERABLE by performance of thefollowing augmented inservice inspection program and the requirements ofSpecification 4.0.5.a. Inspection TypesAs used in this specification, type of snubber shall mean snubbers of thesame design and manufacturer, irrespective of capacity.

b. Visual Inspections Snubbers are categorized as inaccessible or accessible during reactoroperation.

Each of these categories (inaccessible and accessible) may beinspected independently according to the schedule determined by Table4.7.4-1.

The visual inspection interval for each type of snubber shallbe determined based upon the criteria provided in Table 4.7.4-1 and thefirst inspection interval determined using this criteria shall be basedupon the previous inspection interval as established by the requirements in effect before amendment no. 15.LIMERICK

-UNIT 23/4 7-11Amendment No. 1, 15 TABLE 4.7.4-1SNUBBER VISUAL INSPECTION INTERVALNUMBER OF UNACCEPTABLE SN[BBERSPopulation or Category(Notes 1 and 2)Column AExtend Interval(Notes 3 and 6)Column BRepeat Interval(Notes 4 and 6)Column CReduce Interval(Notes 5 and 6)1 0 0 180 0 0 2100 0 1 4150 0 3 8200 2 5 13300 5 12 25400 8 18 36500 12 24 48750 20 40 781000 or greater 29 56 109Note 1: The next visual inspection interval for a snubber population orcategory size shall be determined based upon the previous inspection interval and the number of unacceptable snubbers found during thatinterval.

Snubbers may be categorized, based upon theiraccessibility during power operation, as accessible or inaccessible.

These catergories may be examined separately or jointly.

However,the licensee must make and document that decision before anyinspection and shall use that decision as the basis upon which todetermine the next inspection interval for that category.

Note 2: Interpolation between population or category sizes and the number ofunacceptable snubbers is permissible.

Use next lower integer for thevalue of the limit for Columns A, B, or C if that integer includes afractional value of unacceptable snubbers as determined byinterpolation.

Note 3: If the number of unacceptable snubbers is equal to or less than thenumber in Column A, the next inspection interval may be twice theprevious interval but not greater than 48 months.Note 4: If the number of unacceptable snubbers is equal to or less than thenumber in Column B but greater than the number in Column A, the nextinspection interval shall be the same as the previous interval.

LIMERICK

-UNIT 23/4 7-11aAmendment No. 15 TABLE 4.7.4-1 (continued)

SNUBBER VISUAL INSPECTION INTERVALNote 5: If the number of unacceptable snubbers is equal to or greater thanthe number in Column C, the next inspection interval shall be two-thirds of the previous interval.

However, if the number ofunacceptable snubbers is less than the number in Column C but greaterthan the number in Column B, the next interval shall be reducedproportionally by interpolation, that is, the previous interval shallbe reduced by a factor that is one-third of the ratio of thedifference between the number of unacceptable snubbers found duringthe previous interval and the number in Column B to the difference inthe numbers in Columns B and C.Note 6: The provisions of Specification 4.0.2 are applicable for allinspection intervals up to and including 48 months.LIMERICK

-UNIT 23/4 7-11bAmendment No. 15 PLANT SYSTEMSSURVEILLANCE REOUIREMENTS (Continued)

c. Visual Inspection Acceptance CriteriaVisual inspections shall verify (1) that there are no visible indications of damage or impaired OPERABILITY, (2) attachments to the foundation orsupporting structure are secure, and (3) fasteners for attachment of thesnubber to the component and to the snubber anchorage are secure.Snubbers which appear inoperable as a result of visual inspections shallbe classified as unacceptable and may be reclassified acceptable for thepurpose of establishing the next visual inspection

interval, providing that: (1) the cause of the rejection is clearly established and remediedfor that particular snubber and for other snubbers irrespective of type onthat system that may be generically susceptible; and/or (2) the affectedsnubber is functionally tested in the as found condition and determined OPERABLE per Specifications 4.7.4f. For those snubbers common to morethan one system, the OPERABILITY of such snubbers shall be considered inassessing the surveillance schedule for each of the related systems.

Areview and evaluation shall be performed and documented to justifycontinued operation with an unacceptable snubber.

If continued operation cannot be justified, the snubber shall be declared inoperable and theACTIONS requirements shall be met.d. Transient Event Inspection An inspection shall be performed of all snubbers attached to sections ofsystems that have experienced unexpected, potentially damaging transients, as determined from a review of operational data or a visual inspection ofthe systems, within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for accessible systems and 6 months forinaccessible systems following this determination.

In addition tosatisfying the visual inspection acceptance

criteria, freedom-of-motion ofmechanical snubbers shall be verified using at least one of the following:

(1) manually induced snubber movement; or (2) evaluation of in-placesnubber piston setting; or (3) stroking the mechanical snubber through itsfull range of travel.LIMERICK

-UNIT 23/4 7-12Amendment No. 15 PLANT SYSTEMSSURVEILLANCE REOUIREMENTS (Continued)

e. Functional TestsAt least once per 24 months a representative sample of each type ofsnubber shall be tested using the following sample plans. The sample plan(s) shallbe selected for each type prior to the test period and cannot be changedduring the test period. The NRC Regional Administrator shall be notified inwriting of the sample plan(s) selected for each type prior to the testperiod or the sample plan(s) used in the prior test period shall beimplemented:

1) At least 13.3% of the total population of a snubber type shall befunctionally tested. For each snubber of that type that does notmeet the functional test acceptance criteria of Specification 4.7.4f.,

an additional sample of at least 1/2 the size of theinitial sample shall be tested until the total number tested is equalto the initial sample multiplied by the factor, 1+C/2, where C is thetotal number of unacceptable snubbers or until all the snubbers ofthat type have been tested; or2) A representative sample of 37 snubbers of a snubber type shall befunctionally tested in accordance with Figure 4.7.4-1.

"C" is the totalnumber of snubbers of that type found not meeting theacceptance requirements of Specification 4.7.4f. The cumulative numberof snubbers of the type tested is denoted by "N". If at any time thepoint plotted falls in the "Accept" region, testing of snubbers of thattype may be terminated.

When the point plotted lies in the "Continue Testing" region, additional snubbers of that type shall be tested untilthe point falls in the "Accept" region, or all the snubbers of thattype have been tested.LIMERICK

-UNIT 23/4 7-13Amendment No. 4-9, 42 PLANT SYSTEMSSURVEILLANCE REQUIREMENTS (Continued)

The representative sample selected for the function test sample plansshall be randomly selected from the snubbers of each type and reviewedbefore beginning the testing.

The review shall ensure as far as practical that they are representative of the various configurations, operating environments, range of size, and capacity of snubbers of that type.Snubbers placed in the same locations as snubbers which failed in theprevious functional test period shall be retested at the time of the nextfunctional test period but shall not be included in the sample plan, andfailure of this functional test shall not be the sole cause for increasing the sample size under the sample plan. Testing equipment failure duringfunctional testing may invalidate the day's testing and allow that day'stesting to resume anew at a later time provided all snubbers tested withthe failed equipment during the day of equipment failure are retested.

If during the functional

testing, additional testing is required due tofailure of snubbers, the unacceptable snubbers may be catergorized intofailure mode group(s).

A failure mode group shall include allunacceptable snubbers that have a given failure mode and all othersnubbers subject to the same failure mode. Once a failure mode group hasbeen established, it can be separated for continued testing apart from thegeneral population of snubbers.

However, all unacceptable snubbers in thefailure mode group shall be counted as one unacceptable snubberfor additional testing in the general population.

Testing in the failuremode group shall be based on the number of unacceptable snubbers and shallcontinue in accordance with the sample plan selected for the type or untilall snubbers in the failure mode group have been tested. Any additional unacceptable snubbers found in the failure mode group shall be counted forcontinue testing only for that test failure mode group. In the event thata snubber(s) becomes included in more than one test failure mode group, itshall be counted in each failure mode group and shall be subject to thecorrective action of each test failure mode group.f. Functional Test Acceptance CriteriaThe snubber functional test shall verify that:1) Activation (restraining action) is achieved within the specified range in both tension and compression;

2) Snubber bleed, or release rate where required, is present in bothtension and compression, within the specified range (hydraulic snubbers only);3) For mechanical

snubbers, the force required to initiate or maintainmotion of the snubber is within the specified range in bothdirections of travel; and4) For snubbers specifically required not to displace under continuous load, the ability of the snubber to withstand load withoutdisplacement.

LIMERICK

-UNIT 2 3/4 7-14 Amendment No. 191 PLANT SYSTEMSSURVEILLANCE REOUIREMENTS (Continued)

Testing methods may be used to measure parameters indirectly orparameters other than those specified if those results can be correlated to the specified parameters through established methods.g. Functional Test Failure AnalysisAn engineering evaluation shall be made of each failure to meet thefunctional test acceptance criteria to determine the cause of the failure.The results of this evaluation shall be used, if applicable, in selecting snubbers to be tested in an effort to determine the OPERABILITY of othersnubbers irrespective of type which may be subject to the same failuremode.For the snubbers found inoperable, an engineering evaluation shall beperformed on the components to which the inoperable snubbers are attached.

The purpose of this engineering evaluation shall be to determine if thecomponents to which the inoperable snubbers are attached were adversely affected by the inoperability of the snubbers in order to ensure that thecomponent remains capable of meeting the designed service.h. Functional Testing of Repaired and Replaced SnubbersSnubbers which fail the visual inspection or the functional testacceptance criteria shall be repaired or replaced.

Replacement snubbersand snubbers which have repairs which might affect the functional testresult shall be tested to meet the functional test criteria beforeinstallation in the unit. Mechanical snubbers shall have met theacceptance criteria subsequent to their most recent service, and thefreedom-of-motion test must have been performed within 12 months beforebeing installed in the unit.i. Snubber Service Life Replacement ProqramThe service life of all snubbers shall be monitored to ensure that theservice life is not exceeded between surveillance inspections.

Themaximum expected service life for various seals, springs, and othercritical parts shall be extended or shortened based on monitored testresults and failure history.

Critical parts shall be replaced so thatthe maximum service life will not be exceeded during a period when thesnubber is required to be OPERABLE.

The parts replacements shall bedocumented and the documentation shall be retained in accordance withSpecification 6.10.3.LIMERICK

-UNIT 23/4 7-15Amendment No. 19 1 109876054 CONTINUETESTING00 10 20 30 40 50 60 70 80 90 100NFigure 4.7.4-1SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TESTLIMERICK

-UNIT 23/4 7-16Amendment No. 19 PLANT SYSTEMS3/4.7.5 SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION 3.7.5 Each sealed source containing radioactive material either in excess of100 microcuries of beta and/or gamma emitting material or 5 microcuries of alphaemitting material shall be free of greater than or equal to 0.005 microcurie of removable contamination.

APPLICABILITY:

At all times.ACTION:a. With a sealed source having removable contamination in excess of theabove limit, withdraw the sealed source from use and either:1. Decontaminate and repair the sealed source, or2. Dispose of the sealed source in accordance with Commission Regulations.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.5.1 Test Requirements

-Each sealed source shall be tested for leakageand/or contamination by:a. The licensee, orb. Other persons specifically authorized by the Commission or anAgreement State.The test method shall have a detection sensitivity of at least 0.005 microcurie per test sample.4.7.5.2 Test Frequencies

-Each category of sealed sources, excluding startupsources and fission detectors previously subjected to core flux, shall be testedat the frequency described below:a. Sources in use -In accordance with the Surveillance Frequency ControlProgram for all sealed sources containing radioactive material:

1. With a half-life greater than 30 days, excluding Hydrogen 3, and2. In any form other than gas.LIMERICK

-UNIT 23/4 7-17Amendment No. 147 PLANT SYSTEMSSURVEILLANCE REOUIREMENTS (Continued)

b. Stored sources not in use -Each sealed source and fission detectorshall be tested prior to use or transfer to another licensee unlesstested within the previous 6 months. Sealed sources and fissiondetectors transferred without a certificate indicating the last testdate shall be tested prior to being placed into use.c. Startup sources and fission detectors

-Each sealed startup source* andfission detector shall be tested within 31 days prior to beingsubjected to core flux or installed in the core and following repair ormaintenance to the source.4.7.5.3 Reports -A report shall be prepared and submitted to the Commission onan annual basis if sealed source or fission detector leakage tests reveal thepresence of greater than or equal to 0.005 microcurie of removable contamination.

Except the Cf-252 startup sources which shall be tested within 6 months prior tobeing subjected to core flux or installed in the core and following repair ormaintenance to the source.LIMERICK

-UNIT 23/4 7-18 PLANT SYSTEMSSection 3/4.7.6 through 3/4.7.7 (Deleted)

THE INFORMATION FROM THESE TECHNICAL SPECIFICATIONS SECTIONSHAVE BEEN RELOCATED TO THE TECHNICAL REQUIREMENTS MANUAL (TRM) FIREPROTECTION SECTION.

TECHNICAL SPECIFICATIONS PAGES 3/4 7-19 THROUGH3/4 7-32 HAVE BEEN INTENTIONALLY OMITTED.LIMERICK

-UNIT 23/4 7-19Amendment No. 68 INTENTIONALLY LEFT BLANK PLANT SYSTEMS3/4.7.8 MAIN TURBINE BYPASS SYSTEMLIMITING CONDITION FOR OPERATION 3.7.8 The main turbine bypass system shall be OPERABLE as determined by thenumber of operable main turbine bypass valves being greater than or equal to thatspecified in the CORE OPERATING LIMITS REPORT.APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than orequal to 25% of RATED THERMAL POWER.ACTION: With the main turbine bypass system inoperable, restore the system toOPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or take the ACTION required by Specification 3.2.3.c.SURVEILLANCE REOUIREMENTS 4.7.8 The main turbine bypass system shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program:a. By cycling each turbine bypass valve through at least one completecycle of full travel,b. By performing a system functional test which includes simulated automatic actuation, and by verifying that each automatic valve actuates to itscorrect position, andc. By determining TURBINE BYPASS SYSTEM RESPONSE TIME to be less than or equalto the value specified in the CORE OPERATING LIMITS REPORT.LIMERICK

-UNIT 23/4 7-33Amendment No. -34, 147 PAGE INTENTIONALLY LEFT BLANK 3/4.8 ELECTRICAL POWER SYSTEMS3/4.8.1 A.C. SOURCESA.C. SOURCES -OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall beOPERABLE:

a. Two physically independent circuits between the offsite transmission network and the onsite Class 1E distribution system, andb. Four separate and independent diesel generators, each with:1. A separate day tank containing a minimum of 250 gallons of fuel,2. A separate fuel storage system containing a minimum of 33,500gallons of fuel, and3. A separate fuel transfer pump.APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.ACTION:a. With one diesel generator of the above required A.C. electrical powersources inoperable, demonstrate the OPERABILITY of the remaining A.C.sources by performing Surveillance Requirement 4.8.1.1.1.a within24 hours and at least once per 7 days thereafter.

If the dieselgenerator became inoperable due to any cause other than an inoperable support system, an independently testable component, or preplanned preventive maintenance or testing, demonstrate the OPERABILITY of theremaining operable diesel generators by performing Surveillance Requirement 4.8.1.1.2.a.4 for one diesel generator at a time, within24 hours, unless the absence of any potential common-mode failure forthe remaining diesel generators is determined.

Restore the inoperable diesel generator to OPERABLE status within 30 days or be in at leastHOT 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 thefollowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . See also ACTION e.b. With two diesel generators of the above required A.C. electrical powersources inoperable, demonstrate the OPERABILITY of the remaining A.C.sources by performing Surveillance Requirement 4.8.1.1.1.a within1 hour and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter.

If either of thediesel generators became inoperable due to any cause other than aninoperable support system, an independently testable component, orpreplanned preventive maintenance or testing, demonstrate theOPERABILITY of the remaining diesel generators by performing Surveillance Requirement 4.8.1.1.2.a.4 for one diesel generator at atime, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , unless the absence of any potential common-mode failure for the remaining diesel generators is determined.

Restore atleast one of the inoperable diesel generators to OPERABLE status within72 hours* 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 andin COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . See also ACTION e.*During the extended 7-day Allowed Outage Time (AOT) specified by TS LCO 3.7.1.1,Action a.3.a) or a.3.b) to allow for RHRSW subsystem piping repairs, the 72-hourAOT for two inoperable diesel generators may also be extended to 7 days for thesame 7-day period.LIMERICK

-UNIT 23/4 8-1Amendment No. 4--54, 4-5-4, 165 ELECTRICAL POWER SYSTEMSLIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

c. With three diesel generators of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirement 4.8.1.1.1.a within1 hour and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; and performSurveillance Requirement 4.8.1.1.2.a.4 for the remaining dieselgenerator, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . Restore at least one of the inoperable dieselgenerators to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at leastHOT 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 thefollowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . See also ACTION e.d. With one offsite circuit and one diesel generator of the above requiredA.C. electrical power sources inoperable, demonstrate the OPERABILITY ofthe remaining A.C. sources by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter.

If the diesel generator became inoperable due to any cause other than aninoperable support system, an independently testable component, orpreplanned preventive maintenance or testing, demonstrate theOPERABILITY of the remaining diesel generators by performing Surveillance Requirement 4.8.1.1.2.a.4 for one diesel generator at atime, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , unless the absence of any potential common-mode failure for the remaining diesel generators is determined.

Restore atleast two offsite circuits to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from timeof initial loss or be in at least HOT SHUTDOWN within the next 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months sand in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . See also ACTION e.LIMERICK

-UNIT 23/4 8-1aAmendment No. 150 ELECTRICAL POWER SYSTEMSLIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

e. In addition to the ACTIONS above:1. For two train systems, with one or more diesel generators ofthe above required A.C. electrical power sources inoperable, verify within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter that at least one of the required two train system subsystem, train, components, and devices is OPERABLE and its associated diesel generator is OPERABLE.

Otherwise, restore either theinoperable diesel generator or the inoperable system subsystem to an OPERABLE status within 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s* or be in at least HOTSHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN withinthe following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .2. For the LPCI systems, with two or more diesel generators of theabove required A.C. electrical power sources inoperable, verifywithin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter thatat least two of the required LPCI system subsystems, trains,components and devices are OPERABLE and its associated dieselgenerator is OPERABLE.

Otherwise, be in at least HOT SHUTDOWNwithin 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 .This ACTION does not apply for those systems covered inSpecifications 3.7.1.1 and 3.7.1.2.*During the extended 7-day Allowed OutageAction a.3.a) or a.3.b) to allow for RHRSIAOT may also be extended to 7 days for thiTime (AOT) specified by TS LCO 3.7.1.1,W subsystem piping repairs, the 72-houre same 7-day period.LIMERICK

-UNIT 23/4 8-2Amendment No. 165 ELECTRICAL POWER SYSTEMSLIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

f. With one offsite circuit of the above required A.C. electrical powersources inoperable, demonstrate the OPERABILITY of the remaining A.C.sources by performing Surveillance Requirement 4.8.1.1.1.a within1 hour and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter.

Restore at least twooffsite circuits to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at leastHOT 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 thefollowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .g. With two of the above required offsite circuits inoperable, restore atleast one of the inoperable offsite circuits to OPERABLE status within24 hours 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 .With only one offsite circuit restored to OPERABLE status, restore atleast two offsite circuits to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months fromtime of initial loss or be in at least HOT SHUTDOWN within the next 12hours and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .h. With one offsite circuit and two diesel generators of the above requiredA.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter.

Ifeither of the diesel generators became inoperable due to any cause otherthan an inoperable support system, an independently testable component, or preplanned preventive maintenance or testing, demonstrate theOPERABILITY of the remaining diesel generators by performing Surveillance Requirement 4.8.1.1.2.a.4 for one diesel generator at atime, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , unless the absence of any potential common-mode failure for the remaining diesel generators is determined.

Restore atleast one of the above required inoperable A.C. sources to OPERABLEstatus within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in at least HOT SHUTDOWN within the next 12hours and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Restore atleast two offsite circuits and at least three of the above requireddiesel generators to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from time ofinitial loss 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 andin COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . See also ACTION e.i. Specification 3.0.4.b is not applicable to diesel generators.

LIMERICK

-UNIT 23/4 8-2aAmendment No. 4-3-2, 150 ELECTRICAL POWER SYSTEMSSURVEILLANCE REQUIREMENTS 4.8.1.1.1 Each of the above required independent circuits between the offsitetransmission network and the onsite Class 1E distribution system shall be:a. Determined OPERABLE in accordance with the Surveillance Frequency Control Program by verifying correct breaker alignments and indicated power availability, andb. Demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by transferring, manually and automatically, unit powersupply from the normal circuit to the alternate circuit.4.8.1.1.2 Each of the above required diesel generators shall be demonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program on aSTAGGERED TEST BASIS by:1. Verifying the fuel level in the day fuel tank.2. Verifying the fuel level in the fuel storage tank.3. Verifying the fuel transfer pump starts and transfers fuel fromthe storage system to the day fuel tank.4. Verify that the diesel can start* and gradually accelerate tosynchronous speed with generator voltage and frequency at4280 +/- 120 volts and 60 +/- 1.2 Hz.5. Verify diesel is synchronized, gradually loaded* to anindicated 2700-2800 kW** and operates with this load for atleast 60 minutes.6. Verifying the diesel generator is aligned to provide standbypower to the associated emergency busses.7. Verifying the pressure in all diesel generator air startreceivers to be greater than or equal to 225 psig.*This test shall be conducted in accordance with the manufacturer's recommendations regarding engine prelube and warmup procedures, and asapplicable regarding loading and shutdown recommendations.

- This band is meant as guidance to avoid routine overloading of the engine.Loads in excess of this band for special testing under direct monitoring bythe manufacturer or momentary variations due to changing bus loads shall notinvalidate the test.LIMERICK

-UNIT 23/4 8-3Amendment No. 44, .6-, 4-4-i, 150 FIFCTRTCAI POWFR SY;TFM;SURVEILLANCE REOUIREMENTS (Continued)

b. By removing accumulated water:1) From the day tank in accordance with the Surveillance Frequency ControlProgram and after each occasion when the diesel is operated for greaterthan 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and2) From the storage tank in accordance with the Surveillance Frequency Control Program.c. By sampling new fuel oil in accordance with ASTM D4057-81 prior toaddition to the storage tanks and:1) By verifying in accordance with the tests specified in ASTMD975-81 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 specificgravity of within 0.0016 at 60/600F, when compared to thesupplier's certificate or an absolute specific gravityat 60/60°F of greater than or equal to 0.83 but less than orequal to 0.89 or an API gravity at 60°F of greater than orequal to 27 degrees but less than or equal to 39 degrees.b) A kinematic viscosity at 40°C of greater than or equal to1.9 centistokes, but less than or equal to 4.1 centistokes, if gravity was not determined by comparison with thesupplier's certification.

c) A flash point equal to or greater than 1250F, andd) A clear and bright appearance with proper color whentested in accordance with ASTM D4176-82.

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

d. In accordance with the Surveillance Frequency Control Program byobtaining 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. In accordance with the Surveillance Frequency Control Program by:1) Deleted2) Verifying each diesel generator's capability to reject a load ofgreater than or equal to that of its single largest post-accident load, and:a) Following load rejection, the frequency is 66.5 Hz;b) Within 1.8 seconds following the load rejection, voltage is 4285 +/-420 volts, and frequency is 60 +/- 1.2 Hz; andc) After steady-state conditions are reached, voltage is maintained at 4280 +/- 120 volts.LIMERICK

-UNIT 2 3/4 8-4 Amendment No. -.6-5.9-1--1,42-,4~1~,150 ELECTRICAL POWER SYSTEMSSURVEILLANCE REQUIREMENTS (Continued)

3. Verifying the diesel generator capability to reject a load of 2850kW without tripping.

The generator voltage shall not exceed 4784volts during and following the load rejection.

4. Simulating a loss-of-offsite power by itself, and:a) Verifying deenergization of the emergency buses and loadshedding from the emergency buses.b) Verifying the diesel generator starts* on the auto-start signal, energizes the emergency buses within 10 seconds,energizes the auto-connected loads through the individual load timers and operates for greater than or equal to 5 minuteswhile its generator is loaded with the shutdown loads. Afterenergization, the steady-state voltage and frequency of theemergency buses shall be maintained at 4280 +/- 120 volts and60 +/- 1.2 Hz during this test.5. Verifying that on an ECCS actuation test signal, without loss-of-offsite power, the diesel generator starts* on the auto-start signaland operates on standby for greater than or equal to 5 minutes.

Thegenerator voltage and frequency shall reach 4280 +/- 120 volts and 60 +1.2 Hz within 10 seconds after the auto-start signal; the steadystate generator voltage and frequency shall be maintained withinthese limits during this test.6. Simulating a loss-of-offsite power in conjunction with an ECCSactuation test signal, and:a) Verifying deenergization of the emergency buses and loadshedding from the emergency buses.b) Verifying the diesel generator starts* on the auto-start signal, energizes the emergency buses within 10 seconds,energizes the auto-connected shutdown loads through theindividual load timers and operates for greater than orequal to 5 minutes while its generator is loaded with theemergency loads. After energization, the steady-state voltage and frequency of the emergency buses shall bemaintained at 4280 +/- 120 volts and 60 +/- 1.2 Hz duringthis test.7. Verifying that all automatic diesel generator trips, except engineoverspeed and generator differential over-current are automatically bypassed upon an ECCS actuation signal.*This test shall be conducted in accordance with the manufacturer's recommendations regarding engine prelube and warm up procedures, and asapplicable regarding loading and shutdown recommendations.

LIMERICK

-UNIT 23/4 8-5Amendment No. -34, 4-, 147 ELECTRICAL POWER SYSTEMSSURVEILLANCE REQUIREMENTS (Continued)

8. a) Verifying the diesel generator operates*

for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .During the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of this test, the diesel generator shall beloaded to an indicated 2950-3050 kW** and during the remaining 22hours of this test, the diesel generator shall be loaded to anindicated 2700-2800 kW**.b) Verifying that, within 5 minutes of shutting down the dieselgenerator after the diesel generator has operated*

for at least 2hours at an indicated 2700-2800 kW**, the diesel generator starts*.The generator voltage and frequency shall reach 4280 +/- 120 voltsand 60 +/- 1.2 Hz within 10 seconds after the start signal.9. Verifying that the auto-connected loads to each diesel generator do notexceed the 2000-hour rating of 3100 kW.10. Verifying the diesel generator's capability to:a) Synchronize with the offsite power source while thegenerator is loaded with its emergency loads upon asimulated restoration of offsite power,b) Transfer its loads to the offsite power source, andc) Be restored to its standby status.11. Verifying that with the diesel generator operating in a test modeand connected to its bus, a simulated ECCS actuation signaloverrides the test mode by (1) returning the diesel generator tostandby operation, and (2) automatically energizes theemergency loads with offsite power.12. Verifying that the automatic load sequence timers are OPERABLE withthe interval between each load block within +/- 10% of its designinterval.

This test shall be conducted in accordance with the manufacturer's recommendations regarding engine prelube and warmup procedures, and asapplicable regarding loading and shutdown recommendations.

- This band is meant as guidance to avoid routine overloading of the engine.Loads in excess of this band for special testing under direct monitoring bythe manufacturer or momentary variations due to changing bus loads shall notinvalidate the test.LIMERICK

-UNIT 23/4 8- 6Amendment No. 34,4,45,4404, 147 ELECTRICAL POWER SYSTEMSSURVFT ILLANCFREntJIRFMFNTS (Cnntinued)

13. Verifying that the following diesel generator lockout featuresprevent diesel generator starting only when required:

a) Control Room Switch In Pull-To-Lock (With Local/Remote Switch in Remote)b) Local/Remote Switch in Local.c) Emergency Stopf. In accordance with the Surveillance Frequency Control Program orafter any modifications which could affect diesel generator interdependence by starting*

all four diesel generators simultaneously, during shutdown, and verifying that all four dieselgenerators accelerate to at least 882 rpm in less than or equal to10 seconds.g. In accordance with the Surveillance Frequency Control Program by:1. Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypochlorite orequivalent

solution, and2. Performing a pressure test of those portions of the diesel fueloil system designed to Section III, subsection ND of the ASMECode in accordance with ASME Code Section XI Article IWD-5000.

This test shall be conducted in accordance with the manufacturer's recommendations regarding engine prelube and warmup procedures, and asapplicable regarding loading and shutdown recommendations.

LIMERICK

-UNIT 23/4 8- 7Amendment No. -4, 147 ELECTRICAL POWER SYSTEMSSURVEILLANCE REQUIREMENTS (Continued)

h. In accordance with the Surveillance Frequency Control Program thediesel generator shall be started*

and verified to accelerate tosynchronous speed in less than or equal to 10 seconds.

The generator voltage and frequency shall reach 4280 + 120 volts and 60 +/- 1.2 Hzwithin 10 seconds after the start signal. The diesel generator shallbe started for this test by using one of the following signals:a) Manual***

b) Simulated loss-of-offsite power by itself.c) Simulated loss-of-offsite power in conjunction with an ECCSactuation test signal.d) An ECCS actuation test signal by itself.The generator shall be manually synchronized to its appropriate emergency bus, loaded to an indicated 2700-2800 KW** and operate forat least 60 minutes.

This test, if it is performed so it coincides with the testing required by Surveillance Requirement 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5, may also serve to concurrently meet thoserequirements as well.4.8.1.1.3 Deleted*This test shall be conducted in accordance with the manufacturer's recommendations regarding engine prelube and warmup procedures, and asapplicable regarding loading and shutdown recommendations.

- This band is meant as guidance to avoid routine overloading of the engine.Loads in excess of this band for special testing under direct monitoring bythe manufacturer or momentary variations due to changing bus loads shallnot invalidate the test.***If diesel generator started manually from the control room, 10 secondsafter the automatic prelube period.LIMERICK

-UNIT 23/4 8-7aAmendment No. -5, 4-4-7, 150 INFORMATION ON THIS PAGE HAS BEEN DELETED.LIMERICK

-UNIT 23/4 8-8Amendment No. --4-7, 150 THIS PAGE INTENTIONALLY LEFT BLANK FIFCTRTCAI POWER SYSTEMSA.C. SOURCES -SHUTDOWNLIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall beOPERABLE:

a. One circuit between the offsite transmission network and the onsiteClass 1E distribution system, andb. Two diesel generators each with:1. A day fuel tank containing a minimum of 250 gallons of fuel.2. A fuel storage system containing a minimum of 33,500 gallonsof fuel.3. A fuel transfer pump.APPLICABILITY:

OPERATIONAL CONDITIONS 4, 5, and *ACTION:a. With less than the above required A.C. electrical power sourcesOPERABLE, suspend CORE ALTERATIONS, handling of irradiated fuel inthe secondary containment, operations with a potential for drainingthe reactor vessel and crane operations over the spent fuel storagepool when fuel assemblies are stored therein.

In addition, when inOPERATIONAL CONDITION 5 with the water level less than 22 feet abovethe reactor pressure vessel flange, immediately initiate corrective action to restore the required power sources to OPERABLE status assoon as practical.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.8.1.2 At least the above required A.C. electrical power sources shall bedemonstrated OPERABLE per Surveillance Requirements 4.8.1.1.1 and 4.8.1.1.2.

When handling irradiated fuel in the secondary containment.

LIMERICK

-UNIT 23/4 8-9Amendment No. 4-54, 154 ELECTRICAL POWER SYSTEMS3/4.8.2 D.C. SOURCESD.C. SOURCES -OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 As a minimum, the following D.C. electrical power sources shall beOPERABLE:

a. Division 1, Consisting

1. 125-Volt Battery2. 125-Volt Battery3. 125-Volt Battery4. 125-Volt Batteryb. Division 2, Consisting

1. 125-Volt Battery2. 125-Volt Battery3. 125-Volt Battery4. 125-Volt Batteryc. Division 3, Consisting

1. 125-Volt Battery2. 125-Volt Batteryd. Division 4, Consisting

1. 125-Volt Battery2. 125-Volt Batteryof:2A1 (2AID101).

2A2 (2A2D101).

Charger 2BCA1 (2AID103).

Charger 2BCA2 (2A2D103).

of:2B1 (2BID101).

2B2 (2B2D101).

Charger 2BCB1 (2B1D103).

Charger 2BCB2 (2B2D103).

of:2C (2CD101).

Charger 2BCC (2CD103).

of:20 (2DD1O1).

Charger 2BCD (2DD103).

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.ACTION:a. With one or two battery chargers on one division inoperable:

1. Restore battery terminal voltage to greater than or equalestablished float voltage within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ,to the minimum2. Verify associated Division 1 or 2 float current < 2 amps, or Division 3or 4 float current < 1 amp within 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months and once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months sthereafter, and3. Restore battery charger(s) to OPERABLE status within 7 days.b. With one or more batteries inoperable due to:1. One or two batteries on one division with one or more battery cells floatvoltage < 2.07 volts, perform 4.8.2.1.a.1 and 4.8.2.1.a.2 within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months sfor affected battery(s) and restore affected cell(s) voltage > 2.07 voltswithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .2. Division 1 or 2 with float current > 2 amps, or with Division 3 or 4with float current > 1 amp, perform 4.8.2.1.a.2 within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months foraffected battery(s) and restore battery float current to within limitswithin 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months .LIMERICK

-UNIT 23/4 8-10Amendment No. 126 FIFfTRICAI PnWFR SYSTFMSLIMITING CONDITION FOR OPERATION ACTION: (Continued)

3. One or two batteries on one division with one or more cells electrolyte level less than minimum established design limits, if electrolyte levelwas below the top of the plates restore electrolyte level to above top ofplates within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and verify no evidence of leakage(*)

within 12hours. In all cases, restore electrolyte level to greater than or equalto minimum established design limits within 31 days.4. One or two batteries on one division with pilot cell electrolyte temperature less than minimum established design limits, rest6re batterypilot cell temperature to greater than or equal to minimum established design limits within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .5. Batteries in more than one division

affected, restore battery parameters for all batteries in all but one division to within limits within 2hours.6. Mi) Any battery having both (Action b.1) one or more battery cells floatvoltage < 2.07 volts and (Action b.2) float current not withinlimits, and/or(ii) Any battery not meeting any Action b.1 through b.5,Restore the battery parameters to within limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .c. With any battery(ies) on one division of the above required D.C. electrical power sources inoperable for reasons other than Action b., restore theinoperable division battery to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLDSHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .(*) Contrary to the provisions of Specification 3.0.2, if electrolyte level wasbelow the top of the plates, the verification that there is no evidence of leakageis required to be completed regardless of when electrolyte level is restored.

LIMERICK

-UNIT 23/4 8-10aAmendment No. 126 I THIS PAGE INTENTIONALLY LEFT BLANK ELECTRICAL POWER SYSTEMSSURVEILLANCE REOUIREMENTS 4.8.2.1 Each of the above required division batteries and chargers shall bedemonstrated OPERABLE:

a. In accordance with the Surveillance Frequency Control Program by verifying that:1. Each Division 1 and 2 battery float current is < 2 amps, andDivision 3 and 4 battery float current is < 1 amp when batteryterminal voltage is greater than or equal to the minimumestablished float voltage of 4.8.2.1.a.2, and2. Total battery terminal voltage for each 125-volt battery is greaterthan or equal to the minimum established float voltage.b. In accordance with the Surveillance Frequency Control Program by verifying that:1. Each battery pilot cell voltage is > 2.07 volts,2. Each battery connected cell electrolyte level is greater than orequal to minimum established design limits, and3. The electrolyte temperature of each pilot cell is greater than or equalto minimum established design limits.c. In accordance with the Surveillance Frequency Control Program by verifying that each battery connected cell voltage is > 2.07 volts.d. In accordance with the Surveillance Frequency Control Program by verifying that:1. The battery chargers will supply the currents listed below atgreater than or equal to the minimum established float voltagefor at least 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s:ChargerCurrent (Amperes) 2BCA12BCA22BCB12BCB22BCC2BCD30030030030075752. The battery capacity is adequate to supply and maintain in OPERABLEstatus the required emergency loads for the design duty cycle whensubjected to a battery service test.LIMERICK

-UNIT 23/4 8-11 Amendment No. 84, 5, 4-2-6, 147Corrected by letter- dated Jun~e 19, 1995 ELECTRICAL POWER SYSTEMSSURVEILLANCE REQUIREMENTS (Continued)

e. In accordance with the Surveillance Frequency Control Program byverifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test or modifiedperformance discharge test. The modified performance discharge test maybe performed in lieu of the battery service test (Specification 4.8.2.1.d.2).

f. Performance discharge tests or modified performance discharge tests of batterycapacity shall be given as follows:1. In accordance with the Surveillance Frequency Control Program when:(a) The battery shows degradation or(b) The battery has reached 85% of expected life with battery capacity< 100% of manufacturer's rating, and2. In accordance with the Surveillance Frequency Control Program whenbattery has reached 85% of expected life with battery capacity

>_ 100%of manufacturer's rating.LIMERICK

-UNIT 23/4 8-12Amendment No. -4, -54, 1-2-6, 147CP...r-lted by letterP dated- Ju.ne 19, 1995 TABLE 4.8.2.1-1 (DELETED)

THE INFORMATION ON THE PAGE HAS BEEN DELETEDLIMERICK

-UNIT 23/4 8-13Amendment No. 98, 126 ELECTRICAL POWER SYSTEMSD.C. SOUIRCES

-SHUTDOWNLIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, two of the following four divisions of the D.C.electrical power sources system shall be OPERABLE with:a. Division 1, Consisting

1. 125-Volt Battery2. 125-Volt Battery3. 125-Volt Battery4. 125-Volt Batteryb. Division 2, Consisting

1. 125-Volt Battery2. 125-Volt Battery3. 125-Volt Battery4. 125-Volt Batteryc. Division 3, Consisting

1. 125-Volt Battery2. 125-Volt Batteryd. Division 4, Consisting

1. 125-Volt Battery2. 125-Volt Batteryof:2A1 (2A10101).

2A2 (2A2D101).

Charger 2BCA1 (2A1D103).

Charger 2BCA2 (2A2D103).

of:2B1 (2BID101).

2B2 (2B2D101).

Charger 2BCB1 (2B1D103).

Charger 2BCB2 (2B2D103).

of:2C (2CD101).

Charger 2BCC (2CD103).

of:2D (2DD101).

Charger 2BCD (2DD103).

APPLICABILITY:

OPERATIONAL CONDITIONS 4, 5, and *ACTION:a. With one or two required battery chargers on one required divisioninoperable:

1. Restore battery terminal voltage to greater than or equal to theminimum established float voltage within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ,2. Verify associated Division 1 or 2 float current < 2 amps, or Division3 or 4 float current < 1 amp within 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months and once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months sthereafter, and3. Restore battery charger(s) to OPERABLE status within 7 days.b. With one or more required batteries inoperable due to:1. One or two batteries on one division with one or more battery cellsfloat voltage < 2.07 volts, perform 4.8.2.1.a.1 and 4.8.2.1.a.2 within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for affected battery(s) and restore affected cell(s)voltage > 2.07 volts within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .*When handling irradiated fuel in the secondary containment.

LIMERICK

-UNIT 23/4 8-14Amendment No. 126 ELECTRICAL POWER SYSTEMSLIMITING CONDITION FOR OPERATION ACTION: (Continued)

2. Division 1 or 2 with float current > 2 amps, or with Division 3 or 4with float current > 1 amp, perform 4.8.2.1.a.2 within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months foraffected battery(s) and restore battery float current to within limitswithin 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months .3. One or two batteries on one division with one or more cellselectrolyte level less than minimum established design limits, ifelectrolyte level was below the top of the plates restore electrolyte level to above top of plates within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and verify no evidence ofleakage(*)

within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . In all cases, restore electrolyte levelto greater than or equal to minimum established design limits within31 days.4. One or two batteries on one division with pilot cell electrolyte temperature less than minimum established design limits, restorebattery pilot cell temperature to greater than or equal to minimumestablished design limits within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .5. Batteries in more than one division

affected, restore batteryparameters for all batteries in one division to within limits within2 hours.6. (i) Any battery having both (Action b.1) one or more battery cellsfloat voltage < 2.07 volts and (Action b.2) float current notwithin limits, and/or(ii) Any battery not meeting any Action b.1 through b.5,Restore the battery parameters to within limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .c. 1. With the requirements of Action a. and/or Action b. not met, or2. With less than two divisions of the above required D.C. electrical powersources OPERABLE for reasons other than Actions a. and/or b.,Suspend CORE ALTERATIONS, handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.d. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.8.2.2 At least the above required batteries and chargers shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.1.(*) Contrary to the provisions of Specification 3.0.2, if electrolyte level was belowthe top of the plates, the verification that there is no evidence of leakage isrequired to be completed regardless of when electrolyte level is restored.

LIMERICK

-UNIT 23/4 8-14aAmendment No. 126 I THIS PAGE INTENTIONALLY LEFT BLANK ELECTRICAL POWER SYSTEMS3/4.8.3 ONSITE POWER DISTRIBUTION SYSTEMSDISTRIBUTION

-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 2 Division 1, Consisting of:a)b)c)4160-VAC Bus:480-VAC Load Center:480-VAC Motor Control Centers:d) 120-VAC Distribution Panels:2. Unit 2 Division 2, Consisting of:a) 4160-VAC Bus:b) 480-VAC Load Center:c) 480-VAC Motor Control Centers:d) 120-VAC Distribution Panels:3. Unit 2 Division 3, Consisting of:D21 (20A115)D214 (20B201)D214-R-C (20B213)D214-R-G (20B211)D214-R-G1 (20B215)D214-D-G (20B515)20Y10120Y206D22 (20A116)D224 (20B202)D224-R-C (20B214)D224-R-G (20B212)D224-R-G1 (20B216)D224-D-G (20B516)20Y10220Y207D23 (20A117)D234 (20B203)D234-R-H1 (20B221)D234-R-H (20B217)D234-R-E (20B223)D234-D-G (20B517)20Y10320Y163D24 (20A118)D244 (20B204)D244-R-H1 (20B222)D244-R-H (20B218)D244-R-E (20B224)D244-D-G (20B518)20Y10420Y164a)b)c)4160-VAC Bus:480-VAC Load Center:480-VAC Motor Control Centers:d) 120-VAC Distribution Panels:4. Unit 2 Division 4, Consisting of:a) 4160-VAC Bus:b) 480-VAC Load Center:c) 480-VAC Motor Control Centers:d) 120-VAC Distribution Panels:LIMERICK

-UNIT 23/4 8-15 ELECTRICAL POWER SYSTEMSLIMITING CONDITION FOR OPERATION (Continued)

5. Unit 1 and Common Division 1, Consisting of:abC4160-VAC Bus:480-VAC Load Center:480-VAC Motor Control Centers:d) 120-VAC Distribution Panels:6. Unit 1 and Common Division 2, Consisting of:a)b)c)4160-VAC Bus:480-VAC Load Center:480-VAC Motor Control Centers:d) 120-VAC Distribution Panels:7. Unit 1 and Common Division 3, Consisting of:DllD114D114-R-CD114-R-C1 D114-D-GD114-S-LlOY101lOY20601Y501D12D124D124-R-CD124-R-C1 D124-D-GD124-S-LlOY10210Y20702Y501D13D134D134-R-ED134-C-BD134-D-GD234-S-L1DY10310Y16303Y501D14D144D144-R-ED144-C-BD144-D-GD244-S-L10Y104lOY16404Y501(IOAl15)(10B201)(10B213)(IOB219)(10B515)(00B519)(10Al16)(10B202)10B214)10B220)10B516)(00B520)a)b)c)4160-VAC Bus:480-VAC Load Center:480-VAC Motor Control Centers:(10A117)(10B203)(10B223)(00B131)(10B517)(00B521)d) 120-VAC Distribution:

8. Unit 1 and Common Division 4, Consisting of:a)b)c)4160-VAC Bus:480-VAC Load Center:480-VAC Motor ControlCenters:10A118)10B204)10B224)(00B132)(1OB518)(00B522)d) 120-VAC Distribution:

LIMERICK

-UNIT 23/4 8-16 ELECTRICAL POWER SYSTEMSLIMITING CONDITION FOR OPERATION (Continued)

b. D.C. Power Distribution Panels1. Unit 2 Division 1, Consisting of:a) 250-V DC Fuse Box: 2FA (2AD105)b) 250-V DC Motor Control Center: 2DA (20D201)c) 125-V DC Distribution Panels: 2PPA1 (2AD102)2PPA2 (2AD501)2PPA3 (2AD162)2. Unit 2 Division 2, Consisting of:a) 250-V DC Fuse Box: 2FB (2BD105)b) 250-V DC Motor Control Centers:

2DB-1 (20D202)2DB-2 (20D203)c) 125-V DC Distribution Panels: 2PPB1 (2BD102)2PPB2 (2BD501)2PPB3 (2BD162)3. Unit 2 Division 3, Consisting of:a) 125-V DC Fuse Box: 2FC (2CD105)b) 125-V DC Distribution Panels: 2PPC1 (2CD102)2PPC2 (2CD501)2PPC3 (2CD162)4. Unit 2 Division 4, Consisting of:a) 125-V DC Fuse Box: 2FD (2DD105)b) 125-V DC Distribution Panels: 2PPD1 (2DD102)2PPD2 (2DD501)2PPD3 (2DD162)5. Unit 1 and Common Division 1, Consisting of:a) 250-V DC Fuse Box: 1FA (1ADlO5)b) 125-V DC Distribution Panels: 1PPA1 (1AD102)1PPA2 (1AD501)6. Unit 1 and Common Division 2, Consisting of:a) 250-V DC Fuse Box: 1FB (IBDI05)b) 125-V DC Distribution Panels: 1PPB1 (IBD102)1PPB2 (TBD501)7. Unit 1 and Common Division 3, Consisting of:a) 125-V DC Fuse Box: 1FC (ICD105)b) 125-V DC Distribution Panels: 1PPC1 lCD102)1PPC2 lCD501)8. Unit 1 and Common Division 4, Consisting of:a) 125-V DC Fuse Box: IFD (DD105)b) 125-V DC Distribution Panels: 1PPD1 (1DD102)1PPD2 (1DD501)LIMERICK

-UNIT 23/4 8-16aAmendment No. 102 THIS PAGE INTENTIONALLY LEFT BLANK ELECTRICAL POWER SYSTEMSLIMITING CONDITION FOR OPERATION (Continued)

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.ACTION:a. With one of the above required Unit 2 A.C. distribution systemdivisions not energized, reenergize the division within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months orbe 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 COLDSHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .b. With one of the above required Unit 2 D.C. distribution systemdivisions not energized, reenergize the division within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or bein 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 SHUTDOWNwithin the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .c. With any-of the above required Unit 1 and common AC and/or DCdistribution system divisions not energized, declare the associated common equipment inoperable, and take the appropriate ACTION forthat system.SURVEILLANCE REOUIREMENTS 4.8.3.1 Each of the above required power distribution system divisions shallbe determined energized in accordance with the Surveillance Frequency ControlProgram by verifying correct breaker alignment and voltage on thebusses/MCCs/panels.

LIMERICK

-UNIT 23/4 8-17Amendment No. 147 ELECTRICAL POWER SYSTEMSDTLTRINTCION

-SHUTDOWNLIMITING CONDITION FOR OPERATION 3.8.3.2 As a minimum, 2 of the 4 divisions of theshall be energized with:a. A.C. power distribution:

1. Unit 2 Division 1, Consisting of:power distribution systema)b)c)4160-VAC Bus:480-VAC Load Center:480-VAC Motor Control Centers:d) 120-VAC Distribution Panels:2. Unit 2 Division 2, Consisting of:abc4160-VAC Bus:480-VAC Load Center:480-VAC Motor Control Centers:d) 120-VAC Distribution Panels:3. Unit 2 Division 3, Consisting of:D21 (20A115)D214 (20B201)D214-R-C (20B213)D214-R-G (20B211)D214-R-G1 (20B215)D214-D-G (20B515)20Y10120Y206D22 (20A116)D224 (20B202)D224-R-C (20B214)D224-R-G (20B212)D224-R-G1 (20B216)D224-D-G (20B516)20Y10220Y207D23 (20Al17)D234 (20B203)D234-R-H1 (20B221)D234-R-H (20B217)D234-R-E (20B223)D234-D-G (20B517)20Y10320Y163D24 (20A118)D244 (20B204)D244-R-H1 (20B222)D244-R-H (20B218)D244-R-E (20B224)D244-D-G (20B518)20Y10420Y164a)b)c)4160-VAC Bus:480-VAC Load Center:480-VAC Motor Control Centers:d) 120-VAC Distribution Panels:4. Unit 2 Division 4, Consisting of:a) 4160-VAC Bus:b) 480-VAC Load Center:c) 480-VAC Motor Control Centers:d) 120-VAC Distribution Panels:5. Unit 1 and Common Division 1, Consisting of:a)b)4160-VAC Bus:480-VAC Load Center:D11D114(1OA115)( 1OB201 )LIMERICK

-UNIT 23/4 8-18 LIMICNTRICOA I POWER FYOT OEM RCLIMITING CONDITION FOR OPERATION (Continued)

C) 480-VAC Motor Control Centers:d) 120-VAC Distribution Panels:6. Unit 1 and Common Division 2, Consisting of:a)b)c)4160-VAC Bus:480-VAC Load Center:480-VAC Motor Control Centers:d) 120-VAC Distribution Panels:7. Unit 1 and Common Division 3, Consisting of:D114-R-CD114-R-C1 D114-D-GD114-S-L10Y10110Y20601Y501D12D124D124-R-CD124-R-C1 D124-D-GD124-S-Ll0Y10210Y20702Y501D13D134D134-R-ED134-C-BD134-D-GD234-S-L10Y103lOY16303Y501D14D144D144-R-ED144-C-BD144-D-GD244-S-LlOY10410Y16404Y501(IOB213)(10B219)(IOB515)(00B519)(10A116)(IOB202)(1OB214)(1OB220)(IOB516)(00B520)a)b)c)4160-VAC Bus:480-VAC Load Center:480-VAC Motor Control Centers:(1OA117)(10B203)(10B223)(00B131)(10B517)(00B521)d) 120-VAC Distribution:

8. Unit 1 and Common Division 4, Consisting of:a)b)c)4160-VAC Bus:480-VAC Load Center:480-VAC Motor Control Centers:(10A118)(10B204)(10B224)(00B132)(10B518)(00B522)d) 120-VAC Distribution:

b. D.C. power distribution:

1. Unit 2 Division 1, Consisting of:a)b)250-V DC Fuse Box:250-V DC Motor Control Center:2FA2DA(2AD105)(20D201)LIMERICK

-UNIT 23/4 8-18a THIS PAGE INTENTIONALLY LEFT BLANK ELECTRICAL POWER SYSTEMSLIMITING CONDITION FOR OPERATION (Continued) c) 125-V DC Distribution Panels:2. Unit 2 Division 2, Consisting of:a) 250-V DC Fuse Box:b) 250-V DC Motor Control Centers:c) 125-V DC Distribution Panels:3. Unit 2 Division 3, Consisting of:a)b)125-V DC Fuse Box:125-V DC Distribution Panels:4. Unit 2 Division 4, Consisting of:a)b)125-V DC Fuse Box:125-V DC Distribution Panels:2PPA12PPA22PPA32FB2DB-12DB-22PPB12PPB22PPB32FC2PPC12PPC22PPC32FD2PPD12PPD22PPD31FAIPPA11PPA21FB1PPB11PPB21FC1PPC11PPC21FDIPPD11PPD2(2AD102)(2AD501)(2AD162)(2BD105)(20D202)(20D203)(2BD102)(2BD501)(2BD162)(2CD105)(2CD102)(2CD501)(2CD162)(2DD105)(2DD102)(2DD501)(2DD162)(1AD105)(AD102)(1AD501)(1BD105)(1BD102)(IBD501)(1CD105)(ICD102)1CD501)(1DD105)(DD102)(1DD501)5. Unit 1 and Common Division 1, Consisting of:a) 250-V DC Fuse Box:b) 125-V DC Distribution Panels:6. Unit 1 and Common Division 2, Consisting of-a)b)250-V DC Fuse Box:125-V DC Distribution Panels:7. Unit 1 and Common Division 3, Consisting of:a)b)125-V DC Fuse Box:125-V DC Distribution Panels:8. Unit 1 and Common Division 4, Consisting of:a)b)125-V DC Fuse Box:125-V DC Distribution Panels:APPLICABILITY:

OPERATIONAL CONDITIONS 4, 5, and *ACTION:a. With less than two divisions of the above required Unit 2 A.C.distribution systems energized, suspend CORE ALTERATIONS, handling ofirradiated fuel in the secondary containment and operations with apotential for draining the reactor vessel.*When handling irradiated fuel in the secondary containment.

LIMERICK

-UNIT 23/4 8-19Amendment No. 102 ELECTRICAL POWER SYSTEMSLIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

b. With less than two divisions of the above required Unit 2 D.C.distribution systems energized, suspend CORE ALTERATIONS, handling ofirradiated fuel in the secondary containment and operations with apotential for draining the reactor vessel.c. With any of the above required Unit 1 and common AC and/or DCdistribution system divisions not energized, declare the associated common equipment inoperable, and take the appropriate ACTION for thatsystem.d. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REOUIREMENTS 4.8.3.2 At least the above required power distribution system divisions shallbe determined energized in accordance with the Surveillance Frequency ControlProgram by verifying correct breaker alignment and voltage on thebusses/MCCs/panels.

LIMERICK

-UNIT 23/4 8-20Amendment No. 147 Section 3/4 8.4.1 (Deleted)

THE INFORMATION FROM THIS TECHNICAL SPECIFICATION SECTIONHAS BEEN RELOCATED TO THE TRM. TECHNICAL SPECIFICATIONS PAGES 3/4 8-21 THROUGH 3/4 8-26 OF THIS SECTION HAVE BEENINTENTIONALLY OMITTED.LIMERICK

-UNIT 23/4 8-21Amendment No. -34, 5-7, 153 PAGE INTENTIONALLY LEFT BLANK Section 3/4 8.4.2 (Deleted)

PAGE INTENTIONALLY LEFT BLANKLIMERICK

-UNIT 23/4 8-27Amendment No. -34, 44-;, 170 ELECTRICAL POWER SYSTEMSREACTOR PROTECTION SYSTEM ELECTRICAL POWER MONITORING LIMITING CONDITION FOR OPERATION 3.8.4.3 Two reactor protection system (RPS) electric power monitoring channelsfor each inservice RPS Inverter or alternate power supply shall be OPERABLE.

APPLICABILITY:

At all times.ACTION:a. With one RPS electric power monitoring channel for an inservice RPSInverter or alternate power supply inoperable, restore the inoperable power monitoring channel to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or remove theassociated RPS Inverter or alternate power supply from service.b. With both RPS electric power monitoring channels for an inservice RPSInverter or alternate power supply inoperable, restore at least oneelectric power monitoring channel to OPERABLE status within 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months sor remove the associated RPS Inverter or alternate power supply fromservice.SURVEILLANCE REOUIREMENTS 4.8.4.3 The above specified RPS electric power monitoring channels shall bedetermined OPERABLE:

a. By performance of a CHANNEL FUNCTIONAL TEST each time the plantis in COLD SHUTDOWN for a period of more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , unlessperformed in the previous 6 months.b. In accordance with the Surveillance Frequency Control Program bydemonstrating the OPERABILITY of overvoltage, undervoltage, andunderfrequency protective instrumentation by performance of a CHANNELCALIBRATION including simulated automatic actuation of the protective relays, tripping logic, and output circuit breakers and verifying thefollowing Allowable Values.1. Overvoltage s 127.6 VAC,2. Undervoltage 110.7 VAC,3. Underfrequency 57.05 Hz.LIMERICK

-UNIT 23/4 8-28Amendment No. -34, 3, 46, 147 3.4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCHLIMITING CONDITION FOR OPERATION 3.9.1 The reactor mode switch shall be OPERABLE and locked in the Shutdown orRefuel position.

When the reactor mode switch is locked in the Refuel position:

a. The Refuel position one-rod-out interlock shall be OPERABLE.

b. The following Refuel position interlocks shall be OPERABLE:

1. All rods in.2. Refuel Platform (over-core) position.

3. Refuel Platform hoists fuel-loaded.

4. Service Platform hoist fuel-loaded (with Service Platform installed).

APPLICABILITY:

OPERATIONAL CONDITION 5* **, OPERATIONAL CONDITIONS 3 AND 4 when thereactor mode switch is in the Refuel position.

ACTION:a. With the reactor mode switch not locked in the Shutdown or Refuelposition as specified, suspend CORE ALTERATIONS and lock the reactormode switch in the Shutdown or Refuel position.

b. With the one-rod-out interlock inoperable, verify all control rods arefully inserted and disable withdraw capabilities of all control rods ***,or lock the reactor mode switch in the Shutdown position.

c. With any of the above required Refuel Platform Refuel position interlocks inoperable, take one of the ACTIONS listed below, or suspend COREALTERATIONS.

1. Verify control rods are fully inserted and disable withdrawcapabilities of all control rods***,

or2. Verify Refuel Platform is not over-core (limit switches notreached) and disable Refuel Platform travel over-core, or3. Verify that no Refuel Platform hoist is loaded and disable allRefuel Platform hoists from picking up (grappling) a load.d. With the Service Platform installed over the vessel and any of the aboverequired Service Platform Refuel position interlocks inoperable, take oneof the ACTIONS listed below, or suspend CORE ALTERATIONS.

1. Verify all control rods are fully inserted and disable withdrawcapabilities of all control rods***,

or2. Verify Service Platform hoist is not loaded and disable ServicePlatform hoist from picking up (grappling) a load.* See Special Test Exceptions 3.10.1 and 3.10.3.** The reactor shall be maintained in OPERATIONAL CONDITION 5 whenever fuel isin the reactor vessel with the vessel head closure bolts less than fullytensioned or with the head removed.* Except control rods removed per Specification 3.9.10.1 or 3.9.10.2.

LIMERICK

-UNIT 2 3/4 9-1 Amendment No. 7-6, 112 REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS 4.9.1.1 The reactor mode switch shall be verified to be locked in theShutdown or Refuel position as specified in accordance with the Surveillance Frequency Control Program.4.9.1.2 Each of the above required reactor mode switch Refuel positioninterlocks*

shall be demonstrated OPERABLE by performance of a CHANNELFUNCTIONAL TEST in accordance with the Surveillance Frequency ControlProgram during control rod withdrawal or CORE ALTERATIONS, as applicable.

4.9.1.3 Each of the above required reactor mode switch Refuel positioninterlocks*

that is affected shall be demonstrated OPERABLE by performance of aCHANNEL FUNCTIONAL TEST prior to resuming control rod withdrawal or COREALTERATIONS, as applicable, following repair, maintenance or replacement ofany component that could affect the Refuel position interlock.

The reactor mode switch may be placed in the Run or Startup/Hot Standbyposition to test the switch interlock functions provided that all controlrods are verified to remain fully inserted by a second licensed operator orother technically qualified member of the unit technical staff.LIMERICK

-UNIT 23/4 9-2Amendment No. ;-4, 147 REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 At least two source range monitor (SRM) channels*

shall be OPERABLEand inserted to the normal operating level with:a. Continuous visual indication in the control room,b. At least one with audible alarm in the control room,c. One of the required SRM detectors located in the quadrant where COREALTERATIONS are being performed and the other required SRM detectorlocated in an adjacent

quadrant, andd. Unless adequate SHUTDOWN MARGIN has been demonstrated, the "shorting links" shall be removed from the RPS circuitry prior to and duringthe time any control rod is withdrawn.**

APPLICABILITY:

OPERATIONAL CONDITION 5.***ACTION:With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS and insert all insertable control rods.SURVEILLANCE REQUIREMENTS 4.9.2 Each of the above required SRM channels shall be demonstrated OPERABLE by:a. In accordance with the Surveillance Frequency Control Program:1. Performance of a CHANNEL CHECK,2. Verifying the detectors are inserted to the normal operating level, and3. During CORE ALTERATIONS, verifying that the detector of anOPERABLE SRM channel is located in the core quadrant where COREALTERATIONS are being performed and another is located in anadjacent quadrant.

These channels are not required when sixteen or fewer fuel assemblies, ad-jacent to the SRMs, are in the core. The use of special movable detectors during CORE ALTERATIONS in place of the normal SRM nuclear detectors is per-missible as long as these special detectors are connected to the normal SRMci rcui ts.**Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.

- - See Special Test Exception, Specification 3/4.10.7.

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-UNIT 23/4 9-3Amendment No. 147 REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued)

b. Performance of a CHANNEL FUNCTIONAL TEST in accordance with theSurveillance Frequency Control Program.c. Verifying that the channel count rate is at least 3.0 cps:*1. Prior to control rod withdrawal,

2. Prior to and in accordance with the Surveillance Frequency Control Program during CORE ALTERATIONS, and3. In accordance with the Surveillance Frequency Control Program.d. Verifying, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to and in accordance with theSurveillance Frequency Control Program, that the RPS circuitry

'shorting links" have been removed during:1. The time any control rod is withdrawn**,

unless adequateshutdown margin has been demonstrated, or2. Shutdown margin demonstrations.

May be reduced, provided the source range monitor has an observed count rateand signal-to-noise ratio on or above the curve shown in Figure 3.3.6-1.These channels are not required when sixteen or fewer fuel assemblies, adjacent to the SRMs, are in the core.**Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.

LIMERICK

-UNIT 23/4 9-4Amendment No. 3, 45-3, 4-1-3, 147 REFUELING OPERATIONS 3/4.9.3 CONTROL ROD POSITIONLIMITING CONDITION FOR OPERATION 3.9.3 All control rods shall be inserted.*

APPLICABILITY:

OPERATIONAL CONDITION 5, during CORE ALTERATIONS.**

ACTION:With all control rods not inserted, suspend all other CORE ALTERATIONS, exceptthat one control rod may be withdrawn under control of the reactor mode switchRefuel position one-rod-out interlock.

SURVEILLANCE REQUIREMENTS 4.9.3 All control rods shall be verified to be inserted, except as abovespecified in accordance with the Surveillance Frequency Control Program.*Except control rods removed per Specification 3.9.10.1 or 3.9.10.2.

- See Special Test Exception 3.10.3.LIMERICK

-UNIT 23/4 9-5Amendment No. -, 147 REFUELING OPERATIONS 3/4.9.4 DECAY TIMELIMITING CONDITION FOR OPERATION 3.9.4 The reactor shall be subcritical for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .APPLICABILITY:

OPERATIONAL CONDITION the reactor pressure vessel.5, during movement of irradiated fuel inACTION:With the reactor subcritical for less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , suspend all operations involving movement of irradiated fuel in the reactor pressure vessel.SURVEILLANCE REQUIREMENTS 4.9.4 The reactor shall be determined to have been subcritical for at least24 hours by verification of the date and time of subcriticality prior tomovement of irradiated fuel in the reactor pressure vessel.LIMERICK

-UNIT 23/4 9-6 REFUELING OPERATIONS 3/4.9.5 COMMUNICATIONS LIMITING CONDITION FOR OPERATION 3.9.5 Direct communication shall be maintained between the control room andrefueling floor personnel.

APPLICABILITY:

OPERATIONAL CONDITION 5, during CORE ALTERATIONS.*

ACTION:When direct communication between the control room and refueling floorpersonnel cannot be maintained, immediately suspend CORE ALTERATIONS.*

SURVEILLANCE REQUIREMENTS 4.9.5 Directpersonnel shallControl Programcommunication between the control room and refueling floorbe demonstrated in accordance with the Surveillance Frequency during CORE ALTERATIONS.*

Except movement of control rods with their normal drive system.LIMERICK

-UNIT 23/4 9-7Amendment No. 7-a, 147 REFUELING OPERATIONS 3/4.9.6 REFUELING PLATFORMLIMITING CONDITION FOR OPERATION 3.9.6 The refueling platform shall be OPERABLE and used for handling fuelassemblies or control rods within the reactor pressure vessel.APPLICABILITY:

During handling of fuel assemblies or control rods within thereactor pressure vessel.ACTION:With the requirements for refueling platform OPERABILITY not satisfied, suspenduse of any inoperable refueling platform equipment from operations involving the handling of control rods and fuel assemblies within the reactor pressurevessel after placing the load in a safe condition.

SURVEILLANCE REQUIREMENTS 4.9.6.1 The refueling platform main hoist used for handling of fuel assemblies within the reactor pressure vessel shall be demonstrated OPERABLE within 7 daysprior to the start of such operations by:a. Demonstrating operation of the overload cutoff on the main hoist whenthe load exceeds 1150 +/- 50 pounds.b. Demonstrating operation of the hoist loaded control rod block interlock on the main hoist when the load exceeds 485 +/- 50 pounds.c. Demonstrating operation of the redundant loaded interlock on the mainhoist when the load exceeds 550 + 0, -115 pounds.d. Demonstrating operation of the uptravel interlock when uptravel bringsthe top of the active fuel to not less than 8 feet 0 inches below thenormal water level.LIMERICK

-UNIT 23/4 9-8Amendment No. 8 REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued) 4.9.6.2 The refueling platform frame-mounted auxiliary hoist used forhandling of control rods within the reactor pressure vessel shall be demon-strated OPERABLE within 7 days prior to the use of such equipment by:a. Demonstrating operation of the overload cutoff on the frame mountedhoist when the load exceeds 500 +/- 50 pounds.b. Demonstrating operation of the uptravel mechanical stopmounted hoist when uptravel brings the top of a controlless than 6 feet 6 inches below the normal fuel storagelevel.on the framerod to notpool water4.9.6.3 The refueling platform monorail mounted auxiliary hoist used forhandling of control rods within the reactor pressure vessel shall be demonstra-ted OPERABLE within 7 days prior to the use of such equipment by:a. Demonstrating operation of the overload cutoff on the monorail hoistwhen the load exceeds 500 +/- 50 pounds.b. Demonstrating operation of the uptravel mechanical stop on themonorail hoist when uptravel brings the top of a control rod to notless than 6 feet 6 inches below the normal fuel storage pool waterlevel.LIMERICK

-UNIT 23/4 9-9Amendment No. 8 REFUELING OPERATIONS 3/4.9.7 CRANE TRAVEL-SPENT FUEL STORAGE POOLLIMITING CONDITION FOR OPERATION 3.9.7 Loads in excess of 1200 pounds shall be prohibited from travel overfuel assemblies in the spent fuel storage pool racks.APPLICABILITY:

With fuel assemblies in the spent fuel storage pool racks.ACTION:With the requirements of the above specification not satisfied, place the craneload in a safe condition.

The provisions of Specification 3.0.3 are notapplicable.

SURVEILLANCE REQUIREMENTS 4.9.7 Crane interlocks which prevent crane travel over fuel assemblies inthe spent fuel storage pool racks shall be demonstrated OPERABLE within 7 daysprior to and in accordance with the Surveillance Frequency Control Programduring crane operation.

LIMERICK

-UNIT 23/4 9-10Amendment No. 147 REFUELING OPERATIONS 3/4.9.8 WATER LEVEL -REACTOR VESSELLIMITING CONDITION FOR OPERATION 3.9.8 At least 22 feet of water shall be maintained over the top of thereactor pressure vessel flange.APPLICABILITY:

During handling of fuel assemblies or control rods within thereactor pressure vessel while in OPERATIONAL CONDITION 5 when the fuel assemblies being handled are irradiated or the fuel assemblies seated within the reactorvessel are irradiated.

ACTION:With the requirements of the above specification not satisfied, suspend alloperations involving handling of fuel assemblies or control rods within thereactor pressure vessel after placing all fuel assemblies and control rods ina safe condition.

SURVEILLANCE REOUIREMENTS 4.9.8 The reactor vessel water level shall be determined to be at least itsminimum required depth in accordance with the Surveillance Frequency ControlProgram during handling of fuel assemblies or control rods within the reactorpressure vessel.LIMERICK

-UNIT 23/4 9-11Amendment No. ;-2, 147 REFUELING OPERATIONS 3/4.9.9 WATER LEVEL -SPENT FUEL STORAGE POOLLIMITING CONDITION FOR OPERATION 3.9.9 At least 22 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks.APPLICABILITY:

Whenever irradiated fuel assemblies are in the spent fuel storagepool.ACTION:With the requirements of the above specification not satisfied, suspend allmovement of fuel assemblies and crane operations with loads in the spent fuelstorage pool area after placing the fuel assemblies and crane load in a safecondition.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REOUIREMENTS 4.9.9 The water level in the spent fuel storage pool shall be determined tobe at least at its minimum required depth in accordance with the Surveillance Frequency Control Program.LIMERICK

-UNIT 23/4 9-12Amendment No. 147 REFUELING OPERATIONS 3/4.9.10 CONTROL ROD REMOVALSINGLE CONTROL ROD REMOVALIMITIG CONDITION FOR OPERATION 3.9.10.1 One control rod and/or the associated control rod drive mechanism may be removed from the core and/or reactor pressure vessel provided that atleast the following requirements are satisfied until a control rod and associ-ated control rod drive mechanism are reinstalled and the control rod is fullyinserted in the core.a. The reactor mode switch is OPERABLE and locked in the Shutdown positionor in the Refuel position per Table 1.2 and Specification 3.9.1.b. The source range monitors (SRM) are OPERABLE per Specification 3.9.2.c. The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied, except that the control rod selected to be removed;1. May be assumed to be the highest worth control rod required tobe assumed to be fully withdrawn by the SHUTDOWN MARGIN test,and2. Need not be assumed to be immovable or untrippable.

d. All other control rods in a five-by-five array centered on the controlrod being removed are inserted and electrically or hydraulically disarmed or the four fuel assemblies surrounding the control rod orcontrol rod drive mechanism to be removed from the core and/orreactor vessel are removed from the core cell.e. All other control rods are inserted.

APPLICABILITY:

OPERATIONAL CONDITIONS 4 and 5.ACTION:With the requirements of the above specification not satisfied, suspend removalof the control rod and/or associated control rod drive mechanism from the coreand/or reactor pressure vessel and initiate action to satisfy the aboverequirements.

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-UNIT 23/4 9-13 REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS 4.9.10.1 Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to the start of removal of a control rod and/orthe associated control rod drive mechanism from the core and/or reactor pressurevessel and in accordance with the Surveillance Frequency Control Programthereafter until a control rod and associated control rod drive mechanism arereinstalled and the control rod is inserted in the core, verify that:a. The reactor mode switch is OPERABLE per Surveillance Requirement 4.3.1.1or 4.9.1.2, as applicable, and locked in the Shutdown position or inthe Refuel position with the "one rod out" Refuel position interlock OPERABLE per Specification 3.9.1.b. The SRM channels are OPERABLE per Specification 3.9.2.c. The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied per Specification 3.9.10.1c.

d. All other control rods in a five-by-five array centered on the controlrod being removed are inserted and electrically or hydraulically disarmed or the four fuel assemblies surrounding the control rod orcontrol rod drive mechanism to be removed from the core and/or reactorvessel are removed from the core cell.e. All other control rods are inserted.

LIMERICK

-UNIT 23/4 9- 14Amendment No. 147 REFUELING OPERATIONS MULTIPLE CONTROL ROD REMOVALLIMITING CONDITION FOR OPERATION 3.9.10.2 Any number of control rods and/or control rod drive mechanisms maybe removed from the core and/or reactor pressure vessel provided that at leastthe following requirements are satisfied until all control rods and controlrod drive mechanisms are reinstalled and all control rods are inserted in thecore.a. The reactor mode switch is OPERABLE and locked in the Shutdown positionor in the Refuel position per Specification 3.9.1, except that theRefuel position "one-rod-out" interlock may be bypassed, as required, for those control rods and/or control rod drive mechanisms to beremoved, after the fuel assemblies have been removed as specified below.b. The source range monitors (SRM) are OPERABLE per Specification 3.9.2.c. The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied.

d. All other control rods are either inserted or have the surrounding four fuel assemblies removed from the core cell.e. The four fuel assemblies surrounding each control rod or control roddrive mechanism to be removed from the core and/or reactor vesselare removed from the core cell.APPLICABILITY:

OPERATIONAL CONDITION 5.ACTION:With the requirements of the above specification not satisfied, suspend removalof control rods and/or control rod drive mechanisms from the core and/or reactorpressure vessel and initiate action to satisfy the above requirements.

LIMERICK

-UNIT 23/4 9-15 REFUELING OPERATIONS SURVEILLANCE REOUIREMENTS 4.9.10.2.1 Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to the start of removal of control rods and/orcontrol rod drive mechanisms from the core and/or reactor pressure vessel andin accordance with the Surveillance Frequency Control Program thereafter untilall control rods and control rod drive mechanisms are reinstalled and allcontrol rods are inserted in the core, verify that:a. The reactor mode switch is OPERABLE per Surveillance Requirement 4.3.1.1or 4.9.1.2, as applicable, and locked in the Shutdown position or inthe Refuel position per Specification 3.9.1.b. The SRM channels are OPERABLE per Specification 3.9.2.c. The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied.

d. All other control rods are either inserted or have the surrounding four fuel assemblies removed from the core cell.e. The four fuel assemblies surrounding each control rod and/or controlrod drive mechanism to be removed from the core and/or reactor vesselare removed from the core cell.4.9.10.2.2 Following replacement of all control rods and/or control rod drivemechanisms removed in accordance with this specification, perform a functional test of the "one-rod-out" Refuel position interlock, if this function had beenbypassed.

LIMERICK

-UNIT 23/4 9-16Amendment No. 147 REFUELING OPERATIONS 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION HIGH WATER LEVELLIMITING CONDITION FOR OPERATION 3.9.11.1 One (1) RHR shutdown cooling subsystem shall be OPERABLE and inoperation.

APPLICABILITY:

OPERATIONAL CONDITION 5, when irradiated fuel is in the reactorvessel and the water level is greater than or equal to 22 feet above the topof the reactor pressure vessel flange.ACTION:a. With the required RHR shutdown cooling subsystem inoperable:

1. Within one (1) hour, and once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, verify analternate method of decay heat removal is available.

b. With the required action and associated completion time of Action "a"above not met.1. Immediately suspend loading of irradiated fuel assemblies into thereactor pressure vessel; and2. Immediately initiate action to restore REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY to OPERABLE status; and3. Immediately initiate action to restore one (1) Standby Gas Treatment subsystem to OPERABLE status; and4. Immediately initiate action to restore isolation capability in eachrequired Refueling Floor secondary containment penetration flow pathnot isolated.

c. With no RHR shutdown cooling subsystem in operation:

1. Within one (1) hour from discovery of no reactor coolantcirculation, and once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter, verify reactorcoolant circulation by an alternate method; and2. Once per hour monitor reactor coolant temperature.

SURVEILLANCE REQUIREMENTS 4.9.11.1 At least one (1) RHR shutdown cooling subsystem, or an alternate method,shall be verified to be in operation and circulating reactor coolant inaccordance with the Surveillance Frequency Control Program.The required RHR shutdown cooling subsystem may be removed from operation for upto two (2) hours per eight (8) hour period.LIMERICK

-UNIT 23/4 9-17Amendment No. 4-, 82, 147 REFUELING OPERATIONS LOW WATER LEVELLIMITING CONDITION FOR OPERATION 3.9.11.2 Two (2) RHR shutdown cooling subsystems shall be OPERABLE, and one (1)RHR shutdown cooling subsystem shall be in operation.

APPLICABILITY:

OPERATIONAL CONDITION 5, when irradiated fuel is in the reactorvessel and the water level is less than 22 feet above the top ofthe reactor pressure vessel flange.ACTION:a. With one (1) or two (2) required RHR shutdown cooling subsystems inoperable:

1. Within one (1) hour, and once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, verify analternate method of decay heat removal is available for eachinoperable required RHR shutdown cooling subsystem.

b. With the required action and associated completion time of Action "a"above not met:1. Immediately initiate action to restore REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY to OPERABLE status; and2. Immediately initiate action to restore one (1) Standby GasTreatment subsystem to OPERABLE status; and3. Immediately initiate action to restore isolation capability in eachrequired Refueling Floor secondary containment penetration flowpath not isolated.

c. With no RHR shutdown cooling subsystem in operation:

1. Within one (1) hour from discovery of no reactor coolantcirculation, and once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter, verify reactorcoolant circulation by an alternate method; and2. Once per hour monitor reactor coolant temperature.

SURVEILLANCE REOUIREMENTS 4.9.11.2 At least one (1) RHR shutdown cooling subsystem, or an alternate method,shall be verified to be in operation and circulating reactor coolant inaccordance with the Surveillance Frequency Control Program.The required operating shutdown cooling subsystem may be removed from operation for up to two (2) hours per eight (8) hour period.LIMERICK

-UNIT 23/4 9-18Amendment No. &1, 82, 147 3/4.10 SPECIAL TEST EXCEPTIONS 314.10.I PRTMARY CONTAINMFNT INTFGRITY LIMITING CONDITION FOR OPERATION 3.10.1 The provisions of Specifications 3.6.1.1, 3.6.1.3, and 3.9.1 and Table1.2 may be suspended to permit the reactor pressure vessel closure head andthe drywell head to be removed and the primary containment air lock doors tobe open when the reactor mode switch is in the Startup position during low powerPHYSICS TESTS with THERMAL POWER less than 1% of RATED THERMAL POWER andreactor coolant temperature less than 200'F.APPLICABILITY:

OPERATIONAL CONDITION 2, during low power PHYSICS TESTS.ACTION:With THERMAL POWER greater than or equal to 1% of RATED THERMAL POWER or withthe reactor coolant temperature greater than or equal to 2000F, immediately place the reactor mode switch in the Shutdown position.

SURVEILLANCE REQUIREMENTS 4.10.1 The THERMAL POWER and reactor coolant temperature shall be verified tobe within the limits in accordance with the Surveillance Frequency ControlProgram during low power PHYSICS TESTS.LIMERICK

-UNIT 23/4 10-1Amendment No. 147 SPECIAL TEST EXCEPTIONS 3/4.10.2 ROD WORTH MINIMIZER LIMITING CONDITION FOR OPERATION 3.10.2 The sequence constraints imposed on control rod groups by the rod worthminimizer (RWM) per Specification 3.1.4.1 may be suspended for the following tests provided that control rod movement prescribed for this testing is verifiedby a second licensed operator or other technically qualified member of the unittechnical staff present at the reactor console:a. Shutdown margin demonstration, Specification 4.1.1.b. Control rod scram, Specification 4.1.3.2.c. Control rod friction measurements.

d. Startup Test ProgramAPPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2 when THERMAL POWER is less thanor equal to 10% of RATED THERMAL POWER.ACTION:With the requirements of the above specifications not satisfied, verify thatthe RWM is OPERABLE per Specification 3.1.4.1.SURVEILLANCE REQUIREMENTS 4.10.2 When the sequence constraints imposed by the RWM are bypassed, verify:a. That movement of control rods is blocked or limited to the approvedcontrol rod withdrawal sequence during scram and friction tests.b. That movement of control rods during shutdown margin demonstrations is limited to the prescribed sequence per Specification 3.10.3.c. Conformance with this specification and test procedures by a secondlicensed operator or other technically qualified member of the unittechnical staff.LIMERICK

-UNIT 23/4 10-2 SPECIAL TEST EXCEPTIONS 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS LIMITING CONDITION FOR OPERATION 3.10.3 The provisions of Specification 3.9.1, Specification 3.9.3, and Table1.2 may be suspended to permit the reactor mode switch to be in the Startupposition and to allow more than one control rod to be withdrawn for shutdownmargin demonstration, provided that at least the following requirements aresatisfied.

a. The source range monitors are OPERABLE with the RPS circuitry "shorting links" removed per Specification 3.9.2.b. The rod worth minimizer is OPERABLE per Specification 3.1.4.1 and isprogrammed for the shutdown margin demonstration, or conformance withthe shutdown margin demonstration procedure is verified by a secondlicensed operator or other technically qualified member of the unittechnical staff.c. The "continuous rod withdrawal" control shall not be used duringout-of-sequence movement of the control rods.d. No other CORE ALTERATIONS are in progress.

APPLICABILITY:

OPERATIONAL CONDITION 5, during shutdown margin demonstrations.

ACTION:With the requirements of the above specification not satisfied, immediately place the reactor mode switch in the Shutdown or Refuel position.

SURVEILLANCE REQUIREMENTS 4.10.3 Within 30 minutes prior to and in accordance with the Surveillance Frequency Control Program during the performance of a shutdown margindemonstration, verify that;a. The source range monitors are OPERABLE per Specification 3.9.2,b. The rod worth minimizer is. OPERABLE with the required program perSpecification 3.1.4.1 or a second licensed operator or other techni-cally qualified member of the unit technical staff is present andverifies compliance with the shutdown margin demonstration procedures, andc. No other CORE ALTERATIONS are in progress.

LIMFRICK

-UNIT 23/4 10-3Amendment No. 147 SPECIAL TEST EXCEPTIONS 3/4.10.4 RECIRCULATION LOOPSLIMITING CONDITION FOR OPERATION 3.10.4 The requirements of Specifications 3.4.1.1 and 3.4.1.3 thatrecirculation loops be in operation may be suspended for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months forthe performance of:a. PHYSICS TESTS, provided that THERMAL POWER does not exceed 5% ofRATED THERMAL POWER, orb. The Startup Test Program.APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2, during PHYSICS TESTS and theStartup Test Program.ACTION:a. With the above specified time limit exceeded, insert all control rods.b. With the above specified THERMAL POWER limit exceeded during PHYSICSTESTS, immediately place the reactor mode switch in the Shutdownposition.

SURVEILLANCE REQUIREMENTS 4.10.4.1 The time during which the above specified requirement has been suspended shall be verified to be less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in accordance with the Surveillance Frequency Control Program during PHYSICS TESTS and the Startup Test Program.4.10.4.2 THERMAL POWER shall be determined to be less than 5% of RATED THERMALPOWER in accordance with the Surveillance Frequency Control Program during PHYSICSTESTS.LIMERICK

-UNIT 23/4 10-4Amendment No. 147 SPECIAL TEST EXCEPTIONS 3/4.10.5 OXYGEN CONCENTRATION LIMITING CONDITION FOR OPERATION 3.10.5 The provisions of Specification 3.6.6.3 may be suspended untilcompletion of the Startup Test Program or the reactor has operated for 120Effective Full Power Days.APPLICABILITY:

OPERATIONAL CONDITION 1.ACTIONWith the requirements of the above specification not satisfied, be in at leastSTARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .SURVEILLANCE REQUIREMENTS 4.10.5 The Effective Full Power Days of operation shall be verified to beless than 120, by calculation, in accordance with the Surveillance Frequency Control Program during the Startup Test Program.LIMERICK

-UNIT 23/4 10-5Amendment No. 147 SPECIAL TEST EXCEPTIONS 3/4.10.6 TRAINING STARTUPSLIMITING CONDITION FOR OPERATION 3.10.6 The provisions of Specification 3.5.1 may be suspended to permit oneRHR subsystem to be aligned in the shutdown cooling mode during trainingstartups provided that the reactor vessel is not pressurized, THERMAL POWERis less than or equal to 1% of RATED THERMAL POWER and reactor coolanttemperature is less than 200'F.APPLICABILITY:

OPERATIONAL CONDITION 2, during training startups.

ACTION:With the requirements of the above specification not satisfied, immediately place the reactor mode switch in the Shutdown position.

SURVEILLANCE REQUIREMENTS 4.10.6 The reactor vessel shall be verified to be unpressurized and theTHERMAL POWER and reactor coolant temperature shall be verified to be withinthe limits in accordance with the Surveillance Frequency Control Program duringtraining startups.

10LIMERICK

-UNIT 23/4 10-6Amendment No. 147 SPECIAL TEST EXCEPTIONS 3/4.10.7 SPECIAL INSTRUMENTATION

-INITIAL CORE LOADINGLIMITING CONDITION FOR OPERATION 3.10.7 During initial core loading within the Startup Test Program theprovisions of Specification 3.9.2 may be suspended provided that at least twosource range monitor (SRM) channels with detectors inserted to the normaloperating level are OPERABLE with:a. One of the required SRM channels continuously indicating*

in thecontrol room,b. One of the required SRM detectors located in the quadrant where COREALTERATIONS are being performed and the other required SRM detectorlocated in an adjacent quadrant,**

c. The RPS "shorting links" shall be removed prior to and during fuelloading,d. The reactor mode switch is OPERABLE and locked in the REFUELposition.

APPLICABILITY:

OPERATIONAL CONDITION 5ACTION:With the requirements of the above specifications notsuspend all operations involving CORE ALTERATIONS andcontrol rods.satisfied, immediately insert all insertable SURVEILLANCE REQUIREMENTS 4.10.7 Each of the above required SRM channels shallby:a. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months prior to and at least onceALTERATIONS:

be demonstrated OPERABLEper 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE1. Performance of a CHANNEL CHECK***2. Confirming that the above requirednormal operating level and locatedSpecification 3.10.7.SRM detectors are at thein the quadrants required by*Up to 16 fuel bundles may be loaded without a visual indication of countrate.**The use of special movable detectors during CORE ALTERATIONS in place of thenormal SRM nuclear detectors is permissible as long as these specialdetectors are connected to the normal SRM circuits.

- - Check may be performed by use of movable neutron source. Movement of themovable neutron source is not a CORE ALTERATION.

LIMERICK

-UNIT 23/4 10-7 SPFCIAI TFST FYCFPTTINS SURVEILLANCE REQUIREMENTS (Continued) 4.10.7 (Continued)

3. The RPS "shorting links" are removed.4. The reactor mode switch is locked in the REFUEL position.

b. Performance of a CHANNEL FUNCTIONAL TEST within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior tothe start and at least once per 7 days during CORE ALTERATIONS.

c. Verifying for at least one SRM channel that the count rate is atleast 0.7 cps*:1. Immediately following the loading of the first 16 fuel bundles.2. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter during CORE ALTERATIONS.

Provided signal-to-noise is 2 (for initial startup only). Otherwise, 3 cps.LIMERICK

-UNIT 23/4 10-8 SPECIAL TEST EXCEPTIONS 3/4.10.8 INSERVICE LEAK AND HYDROSTATIC TESTINGLIMITING CONDITION FOR OPERATION 3.10.8 When conducting inservice leak or hydrostatic

testing, the average reactorcoolant temperature specified in Table 1.2 for OPERATIONAL CONDITION 4 may beincreased to 212'F, and operation considered not to be in OPERATIONAL CONDITION 3,to allow performance of an inservice leak or hydrostatic test provided the following OPERATIONAL CONDITION 3 Specifications are met:a. 3.3.2b.C.d.3.6.5.1.1 3.6.5.1.2 3.6.5.2.1 ISOLATION ACTUATION INSTRUMENTATION, Functions 7.a, 7.c.1, 7.c.2and 7.d of Table 3.3.2-1;REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY; REFUELING AREA SECONDARY CONTAINMENT INTEGRITY; REACTOR ENCLOSURE SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVES;REFUELING AREA SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVES;andSTANDBY GAS TREATMENT SYSTEM.e. 3.6.5.2.2

f. 3.6.5.3APPLICABILITY:

OPERATIONAL CONDITION 4, with average reactor coolant temperature greater than 200°F and less than or equal to 212'F.ACTION:With the requirements of the above Specifications not satisfied:

1. Immediately enter the applicable (OPERATIONAL CONDITION

3) action theaffected Specification; or2. Immediately suspend activities that could increase the average reactor coolanttemperature or pressure and reduce the average reactor coolant temperature to200°F or less within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .SURVEILLANCE REOUIREMENTS 4.10.8 Verify applicable OPERATIONAL CONDITION Specifications listed in 3.10.8 are met.3 surveillances for theLIMERICK

-UNIT 23/4 10-9Amendment No. 95 I INTENTIONALLY LEFT BLANK Section 3/4 11.1.1 through 3/4 11.1.4 (Deleted)

THE INFORMATION FROM THESE TECHNICAL SPECIFICATIONS SECTIONS HAS BEENRELOCATED TO THE ODCM. TECHNICAL SPECIFICATIONS PAGES 3/4 11-2 THROUGH3/4 11-6 OF THIS SECTION HAVEBEEN INTENTIONALLY OMITTED.LIMERICK

-UNIT 23/4 11-1Amendment No. 11 PAGE INTENTIONALLY LEFT BLANK RADIOACTIVE EFFLUENTS LIQUID HOLDUP TANKSLIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in any outsidetemporary tanks shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases.APPLICABILITY:

At all times.ACTION:a. With the quantity of radioactive material in any of the above tanksexceeding the above limit, immediately suspend all additions ofradioactive material to the tank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce thetank contents to within the limit and describe the events leading tothis condition in the next Annual Radioactive Effluent ReleaseReport.b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.4 The quantity of radioactive material contained in each of the abovetanks shall be determined to be within the above limit by analyzing a repre-sentative sample of the tank's contents in accordance with the Surveillance Frequency Control Program when radioactive materials are being added to the tank.LIMERICK

-UNIT 23/4 11-7Amendment No. -, 147 Section 3/4 11.2.1 through Section 3/4 11.2.4 (Deleted)

THE INFORMATION FROM THESE TECHNICAL SPECIFICATIONS SECTIONS HAS BEENRELOCATED TO THE ODCM. TECHNICAL SPECIFICATIONS PAGES 3/4 11-9 THROUGH3/4 11-14 OF THESE SECTIONS HAVEBEEN INTENTIONALLY OMITTED.LIMERICK

-UNIT 23/4 11-8Amendment No. 11 RADIOACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURELIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of hydrogen in the main condenser offgas treatment system shall be limited to less than or equal to 4% by volume.APPLICABILITY:

Whenever the main condenser air ejector system is in operation.

ACTION:a. With the concentration of hydrogen in the main condenser offgastreatment system exceeding the limit, restore the concentration towithin the limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.5 The concentration of hydrogen in the main condenser offgas treatment system shall be determined to be within the above limits by continuously monitoring the waste gases in the main condenser offgas treatment system withthe hydrogen monitors required OPERABLE by Table 3.3.7.12-1 of Specifica-tion 3.3.7.12.

LIMERICK

-UNIT 23/4 11-15 RADIOACTIVE EFFLUENTS MAIN CONDENSER LIMITING CONDITION FOR OPERATION 3.11.2.6 The rate of the sum of the activities of the noble gases Kr-85m, Kr-87,Kr-88, Xe-133, Xe-135, and Xe-138 measured at the recombiner after-condenser discharge shall be limited to less than or equal to 330 millicuries/second.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2*, and 3*.ACTION:With the rate of the sum of the activities of the specified noble gasesat the recombiner after-condenser discharge exceeding 330 millicuries/second, restore the gross radioactivity rate to within its limit within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or bein at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .SURVEILLANCE REQUIREMENTS 4.11.2.6.1 recombiner dance withThe rate of the sum of the activities of noble gases at theafter-condenser discharge shall be continuously monitored in accor-Specification 3.3.7.12.

4.11.2.6.2 The rate of the sum of the activities of the specified noble gasesfrom the recombiner after-condenser discharge shall be determined to be withinthe limits of Specification 3.11.2.6 at the following frequencies by performing an isotopic analysis of a representative sample of gases taken at the recombiner after condenser discharge:

a. In accordance with the Surveillance Frequency Control Program.b. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following an increase, as indicated by the MainCondenser Off-Gas Pretreatment Radioactivity

Monitor, of greaterthan 50%, after factoring out increases due to changes in THERMALPOWER level or air in-leakage, in the nominal steady-state fissiongas release from the primary coolant.c. The provisons of Specification 4.0.4 are not applicable.

When the main condenser air ejector is in operation.

LIMERICK

-UNIT 23/4 11-16Amendment No. 147 Section 3/4 11-2.7 (Deleted)

THE INFORMATION FROM THISTECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE ODCM.LIMERICK

-UNIT 23/4 11-17Amendment No. 11 1 Section 3/4 11.3 through 3/4 11.4 (Deleted)

THE INFORMATION FROM THESE TECHNICAL SPECIFICATIONS SECTIONS HAS BEEN RELOCATED TO THE PCP OR ODCM. TECHNICAL SPECIFICATIONS PAGES 3/4 11-19 THROUGH 3/4 11-20 OF THESESECTIONS HAVE BEEN INTENTIONALLY OMITTED.LIMERICK

-UNIT 23/4 11-18Amendment No. 11 I Section 3/4.12 (Deleted)

THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEENRELOCATED TO THE ODCM. TECHNICAL SPECIFICATIONS PAGES 3/4 12-2 THROUGH3/4 12-14 OF THIS SECTION HAVEBEEN INTENTIONALLY OMITTED.LIMERICK

-UNIT 23/4 12-1Amendment No. 11 1 PAGE INTENTIONALLY LEFT BLANK BASES FORSECTIONS 3.0 AND 4.0LIMITING CONDITIONS FOR OPERATION ANDSURVEILLANCE REQUIREMENTS NOTEThe BASES contained in succeeding pages summarize the reasons for the Specifications in Sections 3.0and 4.0, but in accordance with 10 CFR 50.36 arenot part of these Technical Specifications.

3/4.0 APPLICABILITY BASESSpecifications 3.0.1 through 3.0.4 establish the general requirements applicable to Limiting Conditions for Operation.

These requirements are basedon the requirements for Limiting Conditions for Operation stated in the Codeof Federal Regulations, 10 CFR 50.36(c)(2):

"Limiting Conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of thefacility.

When a limiting condition for operation of a nuclear reactoris not met, the licensee shall shut down the reactor or follow anyremedial action permitted by the technical specification until thecondition can be met."Specification 3.0.1 establishes the Applicability statement within eachindividual specification as the requirement for when (i.e., in whichOPERATIONAL CONDITIONS or other specified conditions) conformance to theLimiting Conditions for Operation is required for safe operation of thefacility.

The ACTION requirements establish those remedial measures that mustbe taken within specified time limits when the requirements of a LimitingCondition for Operation are not met. It is not intended that the shutdownACTION requirement be used as an operation convenience which permits (routine) voluntary removal of a system(s) or component(s) from service in lieu of otheralternatives that would not result in redundant systems or components beinginoperable.

There are two basic types of ACTION requirements.

The first specifies theremedial measures that permit continued operation of the facility which is notfurther restricted by the time limits of the ACTION requirements.

In thiscase, conformance to the ACTION requirements provides an acceptable level ofsafety for unlimited continued operation as long as the ACTION requirements continue to be met. The second type of ACTION requirement specifies a timelimit in which conformance to the conditions of the Limiting Condition forOperation must be met. This time limit is the allowable outage time torestore an inoperable system or component to OPERABLE status or for restoring parameters within specified limits. If these actions are not completed withinthe allowable outage time limits, a shutdown is required to place the facilityin an OPERATIONAL CONDITION or other specified condition in which thespecification no longer applies.The specified time limits of the ACTION requirements are applicable from thepoint of time it is identified that a Limiting Condition for Operation is notmet. The time limits of the ACTION requirements are also applicable when asystem or component is removed from service for surveillance testing orinvestigation of operational problems.

Individual specifications may includea specified time limit for the completion of a Surveillance Requirement whenequipment is removed from service.

In this case, the allowable outage timelimits of the ACTION requirements are applicable when this limit expires ifthe surveillance has not been completed.

When a shutdown is required tocomply with ACTION requirements, the plant may have entered an OPERATIONAL CONDITION in which a new specification becomes applicable.

In this case, thetime limits of the ACTION requirements would apply from the point in time thatthe new specification becomes applicable if the requirements of the LimitingCondition for Operation are not met.LIMERICK

-UNIT 2B 3/4 0-1 APPLICABILITY BASESSpecification 3.0.2 establishes that noncompliance with a specification existswhen the requirements of the Limiting Condition for Operation are not met andthe associated ACTION requirements have not been implemented within thespecified time interval.

The purpose of this specification is to clarify that(1) implementation of the ACTION requirements within the specified timeinterval constitutes compliance with a specification and (2) completion of theremedial measures of the ACTION requirements is not required when compliance with a Limiting Condition of Operation is restored within the time intervalspecified in the associated ACTION requirements.

Specification 3.0.3 establishes the shutdown ACTION requirements that must beimplemented when a Limiting Condition for Operation is not met and thecondition is not specifically addressed by the associated ACTION requirements.

The purpose of this specification is to delineate the time limits for placingthe unit in a safe shutdown CONDITION when plant operation cannot be maintained within the limits for safe operation defined by the Limiting Conditions forOperation and its ACTION requirements.

It is not intended to be used as anoperational convenience which permits (routine) voluntary removal of redundant systems or components from service in lieu of other alternatives that wouldnot result in redundant systems or components being inoperable.

One hour isallowed to prepare for an orderly shutdown before initiating a change in plantoperation.

This time permits the operator to coordinate the reduction inelectrical generation with the load dispatcher to ensure the stability andavailability of the electrical grid. The time limits specified to reach lowerCONDITIONS of operation permit the shutdown to proceed in a controlled andorderly manner that is well within the specified maximum cooldown rate andwithin the cooldown capabilities of the facility assuming only the minimumrequired equipment is OPERABLE.

This reduces thermal stresses on components of the primary coolant system and the potential for a plant upset that couldchallenge safety systems under conditions for which this specification applies.If remedial measures permitting limited continued operation of the facilityunder the provisions of the ACTION requirements are completed, the shutdown maybe terminated.

The time limits of the ACTION requirements are applicable from the point in time there was a failure to meet a Limiting Condition forOperation.

Therefore, the shutdown may be terminated if the ACTION requirements have been met or time limits of the ACTION requirements have not expired, thusproviding an allowance for the completion of the required actions.The time limits of Specification 3.0.3 allow 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months for the plant to be inCOLD SHUTDOWN when a shutdown is required during POWER operation.

If theplant is in a lower CONDITION of operation when a shutdown is required, thetime limit for reaching the next lower CONDITION of operation applies.However, if a lower CONDITION of operation is reached in less time thanallowed, the total allowable time to reach COLD SHUTDOWN, or other OPERATIONAL CONDITION, is not reduced.

For example, if STARTUP is reached in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , thetime allowed to reach HOT SHUTDOWN is the next 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months because the total timeto reach HOT SHUTDOWN is not reduced from the allowable limit of 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months .Therefore, if remedial measures are completed that would permit a return toPOWER operation, a penalty is not incurred by having to reach a lowerCONDITION of operation in less than the total time allowed.0LIMERICK

-UNIT 2B 3/4 0-2 APPLICABILITY BASESThe same principle applies with regard to the allowable outage time limits ofthe ACTION requirements, if compliance with the ACTION requirements for onespecification results in entry into an OPERATIONAL CONDITION or condition ofoperation for another specification in which the requirements of the LimitingCondition for Operation are not met. If the new specification becomesapplicable in less time than specified, the difference may be added to theallowable outage time limits of the second specification.

However, theallowable outage time of ACTION requirements for a higher CONDITION ofoperation may not be used to extend the allowable outage time that isapplicable when a Limiting Condition for Operation is not met in a lowerCONDITION of operation.

The shutdown requirements of Specification 3.0.3 do not apply in CONDITIONS 4and 5, because the ACTION requirements of individual specifications define theremedial measures to be taken.Specification 3.0.4 establishes limitations on changes in OPERATIONAL CONDITIONS or other specified conditions in the Applicability when a Limiting Condition forOperation is not met. It allows placing the unit in an OPERATIONAL CONDITION orother specified condition stated in that Applicability (e.g., the Applicability desired to be entered) when unit conditions are such that the requirements of theLimiting Condition for Operation would not be met, in accordance withSpecification 3.0.4.a, Specification 3.0.4.b, or Specification 3.0.4.c.Specification 3.0.4.a allows entry into an OPERATIONAL CONDITION or otherspecified condition in the Applicability with the Limiting Condition for Operation not met when the associated ACTION requirements to be entered permit continued operation in the OPERATIONAL CONDITION or other specified condition in theApplicability for an unlimited period of time. Compliance with ACTIONrequirements that permit continued operation of the unit for an unlimited periodof time in an OPERATIONAL CONDITION or other specified condition provides anacceptable level of safety for continued operation.

This is without regard to thestatus of the unit before or after the OPERATIONAL CONDITION change. Therefore, in such cases, entry into an OPERATIONAL CONDITION or other specified condition inthe Applicability may be made in accordance with the provisions of the ACTIONrequirements.

Specification 3.0.4.b allows entry into an OPERATIONAL CONDITION or otherspecified condition in the Applicability with the Limiting Condition for Operation not met after performance of a risk assessment addressing inoperable systems andcomponents, consideration of the results, determination of the acceptability ofentering the OPERATIONAL CONDITION or other specified condition in theApplicability, and establishment of risk management

actions, if appropriate.

The risk assessment may use quantitative, qualitative, or blended approaches, andthe risk assessment will be conducted using the plant program, procedures, andcriteria in place to implement 10 CFR 50.65(a)(4),

which requires that riskimpacts of maintenance activities be assessed and managed.

The risk assessment, for the purposes of Specification 3.0.4.b, must take into account all inoperable Technical Specification equipment regardless of whether the equipment is includedin the normal 10 CFR 50.65(a)(4) risk assessment scope. The risk assessments willbe conducted using the procedures and guidance endorsed by Regulatory Guide 1.182,"Assessing and Managing Risk Before Maintenance Activities at Nuclear PowerPlants."

Regulatory Guide 1.182 endorses the guidance in Section 11 of NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at NuclearPower Plants."

These documents address general guidance for conduct of the riskassessment, quantitative and qualitative guidelines for establishing riskLIMERICK

-UNIT 2B 3/4 0-3Amendment No. 12-4, 132 APPLICABILITY BASESmanagement

actions, and example risk management actions.

These include actions toplan and conduct other activities in a manner that controls overall risk,increased risk awareness by shift and management personnel, actions to reduce theduration of the condition, actions to minimize the magnitude of risk increases (establishment of backup success paths or compensatory measures),

anddetermination that the proposed OPERATIONAL CONDITION change is acceptable.

Consideration should also be given to the probability of completing restoration such that the requirements of the Limiting Condition for Operation would be metprior to the expiration of the ACTION requirement's specified time interval thatwould require exiting the Applicability.

Specification 3.0.4.b may be used with single, or multiple systems and components unavailable.

NUMARC 93-01 provides guidance relative to consideration ofsimultaneous unavailability of multiple systems and components.

The results of the risk assessment shall be considered in determining theacceptability of entering the OPERATIONAL CONDITION or other specified condition inthe Applicability, and any corresponding risk management actions.

TheSpecification 3.0.4.b risk assessments do not have to be documented.

The Technical Specifications allow continued operation with equipment unavailable in OPERATIONAL CONDITION 1 for the duration of the specified time interval.

Sincethis is allowable, and since in general the risk impact in that particular OPERATIONAL CONDITION bounds the risk of transitioning into and through theapplicable OPERATIONAL CONDITIONS or other specified conditions in theApplicability of the Limiting Condition for Operation, the use of theSpecification 3.0.4.b allowance should be generally acceptable, as long as the riskis assessed and managed as stated above. However, there is a small subset ofsystems and components that have been determined to be more important to risk anduse of the Specification 3.0.4.b allowance is prohibited.

The Limiting Condition for Operations governing these system and components contain Notes prohibiting theuse of Specification 3.0.4.b by stating that Specification 3.0.4.b is notapplicable.

Specification 3.0.4.c allows entry into a OPERATIONAL CONDITION or other specified condition in the Applicability with the Limiting Condition for Operation not metbased on a Note in the Specification which states Specification 3.0.4.c isapplicable.

These specific allowances permit entry into OPERATIONAL CONDITIONS orother specified conditions in the Applicability when the associated ACTIONrequirements to be entered do not provide for continued operation for an unlimited period of time and a risk assessment has not been performed.

This allowance mayapply to all the ACTION requirements or to a specific ACTION requirement of aSpecification.

The risk assessments performed to justify the use ofSpecification 3.0.4.b usually only consider systems and components.

For thisreason, Specification 3.0.4.c is typically applied to Specifications whichdescribe values and parameters (e.g., Reactor Coolant Specific Activity),

and maybe applied to other Specifications based on NRC plant-specific approval.

The provisions of this Specification should not be interpreted as endorsing thefailure to exercise the good practice of restoring systems or components toOPERABLE status before entering an associated OPERATIONAL CONDITION or otherspecified condition in the Applicability.

The provisions of Specification 3.0.4 shall not prevent changes in OPERATIONAL CONDITIONS or other specified conditions in the Applicability that are required tocomply with ACTION requirements.

In addition, the provisions of Specification 3.0.4 shall not prevent changes in OPERATIONAL CONDITIONS or other specified conditions in the Applicability that result from any unit shutdown.

In thisLIMERICK

-UNIT 2B 3/40O-3aAmendment No. 1-24, 132 I APPLICABILITY BASEScontext, a unit shutdown is defined as a change in OPERATIONAL CONDITION or otherspecified condition in the Applicability associated with transitioning fromOPERATIONAL CONDITION 1 to OPERATIONAL CONDITION 2, OPERATIONAL CONDITION 2 toOPERATIONAL CONDITION 3, and OPERATIONAL CONDITION 3 to OPERATIONAL CONDITION 4.Upon entry into an OPERATIONAL CONDITION or other specified condition in theApplicability with the Limiting Condition for Operation not met, Specification 3.0.1 and Specification 3.0.2 require entry into the applicable Conditions andACTION requirements until the Condition is resolved, until the Limiting Condition for Operation is met, or until the unit is not within the Applicability of theTechnical Specification.

Surveillances do not have to be performed on the associated inoperable equipment (or on variables outside the specified limits),

as permitted by Specification 4.0.1. Therefore, utilizing Specification 3.0.4 is not a violation ofSpecification 4.0.1 or Specification 4.0.4 for any Surveillances that have notbeen performed on inoperable equipment.

However, SRs must be met to ensureOPERABILITY prior to declaring the associated equipment OPERABLE (or variablewithin limits) and restoring compliance with the affected Limiting Condition forOperation.

Specification 4.0.1 through 4.0.5 establish the general requirements applicable to Surveillance Requirements.

These requirements are based on the Surveillance Requirements stated in the Code of Federal Regulations 10 CFR 50.36(c)(3):

"Surveillance requirements are requirements relating to test, calibration, or inspection to ensure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and thatthe limiting conditions of operation will be met."Specification 4.0.1 establishes the requirement that SRs must be met during theOPERATIONAL CONDITIONS or other specified conditions in the Applicability forwhich the requirements of the Limiting Condition for Operation apply, unlessotherwise specified in the individual SRs. This Specification is to ensure thatSurveillances are performed to verify the OPERABILITY of systems and components, and that variables are within specified limits. Failure to meet a Surveillance within the specified Surveillance time interval and allowed extension, inaccordance with Specification 4.0.2, constitutes a failure to meet the LimitingCondition for Operation.

Systems and components are assumed to be OPERABLE when the associated SRs havebeen met. Nothing in this Specification,

however, is to be construed as implyingthat systems or components are OPERABLE when:a. The systems or components are known to be inoperable, although stillmeeting the SRs; orb. The requirements of the Surveillance(s) are known to be not metbetween required Surveillance performances.

Surveillances do not have to be performed when the unit is in an OPERATIONAL CONDITION or other specified condition for which the requirements of theassociated Limiting Condition for Operation are not applicable, unless otherwise specified.

The SRs associated with a Special Test Exception Limiting Condition for Operation are only applicable when the Special Test Exception LimitingCondition for Operation is used as an allowable exception to the requirements ofa Specification.

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-UNIT 2B 3/40O-3bAmendment No. 4-24, 132 APPLICABILITY BASESUnplanned events may satisfy the requirements (including applicable acceptance criteria) for a given SR. In this case, the unplanned event may be credited asfulfilling the performance of the SR. This allowance includes those SRs whoseperformance is normally precluded in a given OPERATIONAL CONDITION or otherspecified condition.

Surveillances, including Surveillances invoked by ACTION requirements, do nothave to be performed on inoperable equipment because the ACTIONS define theremedial measures that apply. Surveillances have to be met and performed inaccordance with Specification 4.0.2, prior to returning equipment to OPERABLEstatus.Upon completion of maintenance, appropriate post maintenance testing is requiredto declare equipment OPERABLE.

This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance withSpecification 4.0.2. Post maintenance testing may not be possible in thecurrent OPERATIONAL CONDITION or other specified conditions in the Applicability due to the necessary unit parameters not having been established.

In thesesituations, the equipment may be considered OPERABLE provided testing has beensatisfactorily completed to the extent possible and the equipment is nototherwise believed to be incapable of performing its function.

This will allowoperation to proceed to an OPERATIONAL CONDITION or other specified condition where other necessary post maintenance tests can be completed.

Some examples of this process are:a. Control Rod Drive maintenance during refueling that requires scramtesting at > 950 psi. However, if other appropriate testing issatisfactorily completed and the scram time testing ofSpecification 4.1.3.2 is satisfied, the control rod can beconsidered OPERABLE.

This allows startup to proceed to reach950 psi to perform other necessary testing.b. High pressure coolant injection (HPCI) maintenance during shutdownthat requires system functional tests at a specified pressure.

Provided other appropriate testing is satisfactorily completed, startup can proceed with HPCI considered OPERABLE.

This allowsoperation to reach the specified pressure to complete the necessary post maintenance testing.LIMERICK

-UNIT 2B 3/4 0-3cAmendment No. 4-2-4, 132 THIS PAGE INTENTIONALLY LEFT BLANK APPLICABILITY BASESSpecification 4.0.2 establishes the limit for which the specified time intervalfor Surveillance Requirements may be extended.

It permits an allowable extension of the normal surveillance interval to facilitate surveillance scheduling andconsideration of plant operating conditions that may not be suitable forconducting the surveillance; e.g., transient conditions or other ongoingsurveillance or maintenance activities.

It also provides flexibility toaccommodate the length of a fuel cycle for surveillances that are performed ateach refueling outage and are specified with an 24-month surveillance interval.

It is not intended that this provision be used repeatedly as a convenience toextend the surveillance intervals beyond that specified for surveillances thatare not performed during refueling outages.

Likewise, it is not the intent thatREFUELING INTERVAL surveillances be performed during power operation unless it isconsistent with safe plant operation.

The limitation of Specification 4.0.2 is based onengineering judgment and the recognition that the most probable result of any particular surveillance being performed is the verification of conformance with the Surveillance Requirements.

This provision is sufficient to ensure that the reliability ensured throughsurveillance activities is not significantly degraded beyond that obtained from the specified surveillance interval.

Specification 4.0.3 establishes the flexibility to defer declaring affectedequipment inoperable or an affected variable outside the specified limits when aSurveillance has not been completed within the specified Surveillance timeinterval and allowed extension.

A delay period of up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to thelimit of the specified Surveillance time interval, whichever is greater, appliesfrom the point in time that it is discovered that the Surveillance has not beenperformed in accordance with Specification 4.0.2, and not at the time that thespecified Surveillance time interval and allowed extension was not met.This delay period provides adequate time to complete Surveillances that havebeen missed. This delay period permits the completion of a Surveillance beforecomplying with ACTION requirements or other remedial measures that mightpreclude completion of the Surveillance.

The basis for this delay period includes consideration of unit conditions, adequate

planning, availability of personnel, the time required to perform theSurveillance, the safety significance of the delay in completing the requiredSurveillance, and the recognition that the most probable result of anyparticular Surveillance being performed is the verification of conformance withthe requirements.

When a Surveillance with a Surveillance time interval basednot on time intervals, but upon specified unit conditions, operating situations, or requirements of regulations (e.g., prior to entering OPERATIONAL CONDITION 1after each fuel loading, or in accordance with 10 CFR 50, Appendix J, asmodified by approved exemptions, etc.) is discovered to have not been performed when specified, Specification 4.0.3 allows for the full delay period of up tothe specified Surveillance time interval to perform the Surveillance.

However,since there is not a time interval specified, the missed Surveillance should beperformed at the first reasonable opportunity.

Specification 4.0.3 provides a time limit for, and allowances for theperformance of, Surveillances that become applicable as a consequence ofOPERATIONAL CONDITION changes imposed by ACTION requirements.

Failure to comply with specified Surveillance time intervals and allowedextensions for SRs is expected to be an infrequent occurrence.

Use of the delayperiod established by Specification 4.0.3 is a flexibility which is not intendedto be used as an operational convenience to extend Surveillance intervals.

LIMERICK

-UNIT 2B 3/4 0-4Amendment No. 4, -4, 124 APPLICABILITY BASESWhile up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or the limit of the specified Surveillance time interval isprovided to perform the missed Surveillance, it is expected that the missedSurveillance will be performed at the first reasonable opportunity.

Thedetermination of the first reasonable opportunity should include consideration of the impact on plant risk (from delaying the Surveillance as well as any plantconfiguration changes required or shutting the plant down to perform theSurveillance) and impact on any analysis assumptions, in addition to unitconditions,

planning, availability of personnel, and the time required toperform the Surveillance.

This risk impact should be managed through theprogram in place to implement 10 CFR 50.65(a)(4) and its implementation

guidance, NRC Regulatory Guide 1.182, 'Assessing and Managing Risk BeforeMaintenance Activities at Nuclear Power Plants.'

This Regulatory Guideaddresses consideration of temporary and aggregate risk impacts, determination of risk management action thresholds, and risk management action up to andincluding plant shutdown.

The missed Surveillance should be treated as anemergent condition as discussed in the Regulatory Guide. The risk evaluation may use quantitative, qualitative, or blended methods.

The degree of depth andrigor of the evaluation should be commensurate with the importance of thecomponent.

Missed Surveillances for important components should be analyzedquantitatively.

If the results of the risk evaluation determine the riskincrease is significant, this evaluation should be used to determine the safestcourse of action. All missed Surveillances will be placed in the Corrective Action Program.If a Surveillance is not completed within the allowed delay period, then theequipment is considered inoperable or the variable is considered outside thespecified limits and the ACTION requirements for the applicable LimitingCondition for Operation begin immediately upon expiration of the delay period.If a Surveillance is failed within the delay period or the variable is outsidethe specified limits, then the equipment is inoperable and the Completion Timesof the Required Actions for the applicable LCO Conditions begin immediately upon the failure of the Surveillance.

Completion of the Surveillance within the delay period allowed by thisSpecification, or within the allowed times specified in the ACTION requirements, restores compliance with Specification 4.0.1.Specification 4.0.4 establishes the requirement that all applicable SRs must bemet before entry into an OPERATIONAL CONDITION or other specified condition in theApplicability.

This Specification ensures that system and component OPERABILITY requirements andvariable limits are met before entry into OPERATIONAL CONDITIONS or otherspecified conditions in the Applicability for which these systems and components ensure safe operation of the unit. The provisions of this Specification shouldnot be interpreted as endorsing the failure to exercise the good practice ofrestoring systems or components to OPERABLE status before entering an associated OPERATIONAL CONDITION or other specified condition in the Applicability.

A provision is included to allow entry into an OPERATIONAL CONDITION or otherspecified condition in the Applicability when a Limiting Condition for Operation is not met due to a Surveillance not being met in accordance with Specification 3.0.4.However, in certain circumstances, failing to meet an SR will not result inSpecification 4.0.4 restricting an OPERATIONAL CONDITION change or other specified LIMERICK

-UNIT 2B 3/4 0-5 Amendment No. h-21, 94, 4-24, 132 APPLICABILITY BASEScondition change. When a system, subsystem,

division, component, device, orvariable is inoperable or outside its specified limits, the associated SR(s) arenot required to be performed, per Specification 4.0.1, which states thatsurveillances do not have to be performed on inoperable equipment.

When equipment is inoperable, Specification 4.0.4 does not apply to the associated SR(s) sincethe requirement for the SR(s) to be performed is removed.

Therefore, failing toperform the Surveillance(s) within the specified Surveillance time interval doesnot result in a Specification 4.0.4 restriction to changing OPERATIONAL CONDITIONS or other specified conditions of the Applicability.

However, since the LimitingCondition for Operation is not met in this instance, Specification 3.0.4 willgovern any restrictions that may (or may not) apply to OPERATIONAL CONDITION orother specified condition changes.

Specification 4.0.4 does not restrict changingOPERATIONAL CONDITIONS or other specified conditions of the Applicability when aSurveillance has not been performed within the specified Surveillance timeinterval, provided the requirement to declare the Limiting Condition for Operation not met has been delayed in accordance with Specification 4.0.3.The provisions of Specification 4.0.4 shall not prevent entry into OPERATIONAL CONDITIONS or other specified conditions in the Applicability that are required tocomply with ACTION requirements.

In addition, the provisions of Specification 4.0.4 shall not prevent changes in OPERATIONAL CONDITIONS or other specified conditions in the Applicability that result from any unit shutdown.

In thiscontext, a unit shutdown is defined as a change in OPERATIONAL CONDITION or otherspecified condition in the Applicability associated with transitioning fromOPERATIONAL CONDITION 1 to OPERATIONAL CONDITION 2, OPERATIONAL CONDITION 2 toOPERATIONAL CONDITION 3, and OPERATIONAL CONDITION 3 to OPERATIONAL CONDITION 4.Specification 4.0.5 establishes the requirement that inservice inspection of ASMECode Class 1, 2 and 3 components and inservice testing of ASME Code Class 1, 2 and3 pumps and valves shall be performed in accordance with a periodically updatedversion of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda, andthe ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) andapplicable Addenda as required by 10 CFR 50.55a. Additionally, the Inservice Inspection Program conforms to the NRC staff positions identified in NRC GenericLetter 88-01, "NRC Position on IGSCC in BWR Austinetic Stainless Steel Piping,"

asapproved in NRC Safety Evaluations dated March 6, 1990 and October 22, 1990, or inaccordance with alternate measures approved by the NRC staff.This specification includes a clarification of the frequencies for performing the inservice inspection and testing activities required by Section XI of theASME Boiler and Pressure Vessel Code and applicable

Addenda, and the ASME Codefor Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda.

This clarification is provided to ensure consistency in surveillance intervals throughout the Technical Specifications and to remove anyambiguities relative to the frequencies for performing the required inservice inspection and testing activities.

Under the terms of this specification, the more restrictive requirements ofthe Technical Specifications take precedence over the ASME Code and applicable Addenda.

The requirements of Specification 4.0.4 to perform surveillance activities before entry into an OPERATIONAL CONDITION or other specified condition takes precedence over the ASME Code provision that allows pumps and valves to betested up to one week after return to normal operation.

The Technical Specification definition of OPERABLE does not allow a grace period before acomponent, which is not capable of performing its specified

function, is declaredinoperable and takes precedence over the ASME Code provision that allows a valveto be incapable of performing its specified function for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months beforebeing declared inoperable.

LIMERICK

-UNIT 2B 3/4 0-6Amendment No. 4-321, 4-3-, 155 3/4.1 REACTIVITY CONTROL SYSTEMSBASES3/4.1.1 SHUTDOWN MARGINA sufficient SHUTDOWN MARGIN ensures that (1) the reactor can be madesubcritical from all operating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable withinacceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

Since core reactivity values will vary through core life as a function offuel depletion and poison burnup, the demonstration of SHUTDOWN MARGIN will beperformed in the cold, xenon-free condition and shall show the core to besubcritical by at least R + 0.38% A k/k or R + 0.28% A k/k, as appropriate.

The 0.38% A k/k includes uncertainties and calculation biases. The value of Rin units of % A k/k is the difference between the calculated value of minimumshutdown margin during the operating cycle and the calculated shutdown marginat the time of the shutdown margin test at the beginning of cycle. The valueof R must be positive or zero and must be determined for each fuel loading cycle.Two different values are supplied in the Limiting Condition for Operation to provide for the different methods of demonstration of the SHUTDOWN MARGIN.The highest worth rod may be determined analytically or by test. The SHUTDOWNMARGIN is demonstrated by (an insequence) control rod withdrawal at thebeginning of life fuel cycle conditions, and, if necessary, at any future timein the cycle if the first demonstration indicates that the required margin couldbe reduced as a function of exposure.

Observation of subcriticality in thiscondition assures subcriticality with the most reactive control rod fullywithdrawn.

This reactivity characteristic has been a basic assumption in the analysisof plant performance and can be best demonstrated at the time of fuel loading,but the margin must also be determined anytime a control rod is incapable ofinsertion.

3/4.1.2 REACTIVITY ANOMALIES Since the SHUTDOWN MARGIN requirement for the reactor is small, a carefulcheck on actual conditions to the predicted conditions is necessary, and thechanges in reactivity can be inferred from these comparisons of core keffective (keff). Since the comparisons are easily done, frequent checks are not animposition on normal operations.

A 1% change is larger than is expected fornormal operation so a change of this magnitude should be thoroughly evaluated.

A change as large as 1% would not exceed the design conditions of the reactorand is on the safe side of the postulated transients.

LIMERICK

-UNIT 2B 3/4 1-1Associated with Amendment No. 168 REACTIVITY CONTROL SYSTEMSBASES3/4.1.3 CONTROL RODSThe specification of this section ensure that (1) the minimum SHUTDOWNMARGIN is maintained, (2) the control rod insertion times are consistent withthose used in the accident

analysis, and (3) the potential effects of the roddrop accident are limited.

The ACTION statements permit variations from the basicrequirements but at the same time impose more restrictive criteria for continued operation.

A limitation on inoperable rods is set such that the resultant effecton total rod worth and scram shape will be kept to a minimum.

The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.Damage within the control rod drive mechanism could be a generic problem,therefore with a control rod immovable because of excessive friction ormechanical interference, operation of the reactor is limited to a time periodwhich is reasonable to determine the cause of the inoperability and at the sametime prevent operation with a large number of inoperable control rods.Control rods that are inoperable for other reasons are permitted to betaken out of service provided that those in the nonfully-inserted position areconsistent with the SHUTDOWN MARGIN requirements.

The number of control rods permitted to be inoperable could be more thanthe eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor must be shutdownfor investigation and resolution of the problem.The control rod system is designed to bring the reactor subcritical at arate fast enough to prevent the MCPR from becoming less than the fuel claddingsafety limit during the limiting power transient analyzed in Section 15.2 ofthe FSAR. This analysis shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in thespecifications, provided the required protection and MCPR remains greater thanthe fuel cladding safety limit. The occurrence of scram times longer thenthose specified should be viewed as an indication of a systemic problem withthe rod drives and therefore the surveillance interval is reduced in order toprevent operation of the reactor for long periods of time with a potentially serious problem.Scram time testing at zero psig reactor coolant pressure is adequate toensure that the control rod will perform its intended scram function during startupof the plant until scram time testing at 950 psig reactor coolant pressure isperformed prior to exceeding 40% rated core thermal power.The scram discharge volume is required to be OPERABLE so that it will beavailable when needed to accept discharge water from the control rods during areactor scram and will isolate the reactor coolant system from the containment when required.

The OPERABILITY of all SDV vent and drain valves ensures that the SDV ventand drain valves will close during a scram to contain reactor water discharged tothe SDV piping. The SDV has one common drain line and one common vent line.Since the vent and drain lines are provided with two valves in series, the singleLIMERICK

-UNIT 2B 3/4 1-2Amendment No. 49-, 131 REACTIVITY CONTROL SYSTFMSBASESCONTROL RODS (Continued) failure of one valve in the open position will not impair the isolation functionof the system. Additionally, the valves are required to open on scram reset toensure that a path is available for the SDV piping to drain freely at othertimes.When one SDV vent or drain valve is inoperable in one or more lines, thevalves must be restored to OPERABLE status within 7 days. The allowable outagetime is reasonable, given the level of redundancy in the lines and the lowprobability of a scram occurring while the valve(s) are inoperable.

The SDV isstill isolable since the redundant valve in the affected line is OPERABLE.

Duringthese periods, the single failure criterion may not be preserved, and a higherrisk exists to allow reactor water out of the primary system during a scram.If both valves in a line are inoperable, the line must be isolated tocontain the reactor coolant during a scram. When a line is isolated, thepotential for an inadvertent scram due to high SDV level is increased.

ACTION "e" is modified by a note ("****")

that allows periodic draining andventing of the SDV when a line is isolated.

During these periods, the line may beunisolated under administrative control.

This allows any accumulated water in theline to be drained, to preclude a reactor scram on SDV high level. This isacceptable since the administrative controls ensure the valve can be closedquickly, by a dedicated

operator, if a scram occurs with the valve open. The 8hour allowable outage time to isolate the line is based on the low probability ofa scram occurring while the line is not isolated and the unlikelihood ofsignificant CRD seal leakage.Control rods with inoperable accumulators are declared inoperable andSpecification 3.1.3.1 then applies.

This prevents a pattern of inoperable accumulators that would result in less reactivity insertion on a scram thanhas been analyzed even though control rods with inoperable accumulators maystill be inserted with normal drive water pressure.

The drive water pressurenormal operating range is specified in system operating procedures which provideranges for system alignment and control rod motion (exercising).

Operability ofthe accumulator ensures that there is a means available to insert the controlrods even under the most unfavorable depressurization of the reactor.

A controlrod is considered trippable if it is capable of fully inserting as a result of ascram signal.LIMERICK

-UNIT 2B 3/4 1-2aAmendment No. 3, 4-314, 140 THIS PAGE INTENTIONALLY LEFT BLANK REACTIVITY CONTROL SYSTEMSBASESCONTROL RODS (Continued)

Control rod coupling integrity is required to ensure compliance with theanalysis of the rod drop accident in the FSAR. The overtravel position featureprovides the only positive means of determining that a rod is properly coupledand therefore this check must be performed prior to achieving criticality aftercompleting CORE ALTERATIONS that could have affected the control rod couplingintegrity.

The subsequent check is performed as a backup to the initial demon-stration.

In order to ensure that the control rod patterns can be followed and there-fore that other parameters are within their limits, the control rod positionindication system must be OPERABLE.

The control rod housing support restricts the outward movement of a controlrod to less than 3 inches in the event of a housing failure.

The amount ofrod reactivity which could be added by this small amount of rod withdrawal isless than a normal withdrawal increment and will not contribute to any damageto the primary coolant system. The support is not required when there is nopressure to act as a driving force to rapidly eject a drive housing.The required surveillances are adequate to determine that the rods are OPERABLEand not so frequent as to cause excessive wear on the system components.

3/4.1.4 CONTROL ROD PROGRAM CONTROLSControl rod withdrawal and insertion sequences are established to assurethat the maximum insequence individual control rod or control rod segments whichare withdrawn at any time during the fuel cycle could not be worth enough toresult in a peak fuel enthalpy greater than 280 cal/gm in the event of a controlrod drop accident.

The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal.

When THERMAL POWER is greaterthan 10% of RATED THERMAL POWER, there is no possible rod worth which, ifdropped at the design rate of the velocity

limiter, could result in a peakenthalpy of 280 cal/gm. Thus requiring the RWM to be OPERABLE whenTHERMAL POWER is less than or equal to 10% of RATED THERMAL POWER providesadequate control.The RWM provides automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.

The analysis of the rod drop accident is presented in Section 15.4.9 ofthe FSAR and the techniques of the analysis are presented in a topical report,Reference 1, and two supplements, References 2 and 3. Additional pertinent analysis is also contained in Amendment 17 to the Reference 4 Topical Report.The RBM is designed to automatically prevent fuel damage in the event oferroneous rod withdrawal from locations of high power density over the range ofpower operation.

Two channels are provided.

Tripping one of the channels willblock erroneous rod withdrawal to prevent fuel damage. This system backs up thewritten sequence used by the operator for withdrawal of control rods. RBM OPERA-BILITY is required when the limiting condition described in Specification 3.1.4.3 exists.LIMERICK

-UNIT 2B 3/4 1-3Amendment No. 49, 147 REACTIVITY CONTROL SYSTEMS3/4.1.5 STANDBY LIQUID CONTROL SYSTEMThe standby liquid control system provides a backup capability for bringingthe reactor from full power to a cold, Xenon-free

shutdown, assuming that thewithdrawn control rods remain fixed in the rated power pattern.

To meet thisobjective it is necessary to inject a quantity of boron which produces a concen-tration of 660 ppm in the reactor core and other piping systems connected to thereactor vessel. To allow for potential leakage and improper mixing, this con-centration is increased by 25%. The required concentration is achieved by havingavailable a minimum quantity of 3,160 gallons of sodium pentaborate solutioncontainng a minimum of 3,754 lbs of sodium pentaborate having therequisite Boron-10 atom % enrichment of 29% as determined from Reference 5.This quantity of solution is a net amount which is above the pump suctionshutoff level setpoint thus allowing for the portion which cannot be injected.

The above quantities calculated at 29% Boron-lO enrichment have beendemonstrated by analysis to provide a Boron-lO weight equivalent of 185 lbs inthe sodium pentaborate solution.

Maintaining this Boron-lO weight in the nettank contents ensures a sufficient quantity of boron to bring the reactor to acold, Xenon-free shutdown.

The pumping rate of 41.2 gpm provides a negative reactivity insertion rate overthe permissible solution volume range, which adequately compensates for thepositive reactivity effects due to elimination of steam voids, increased waterdensity from hot to cold, reduced doppler effect in uranium, reduced neutronleakage from boiling to cold, decreased control rod worth as the moderator cools,and xenon decay. The temperature requirement ensures that the sodium pentaborate always remains in solution.

With redundant pumps and explosive injection valves and with a highlyreliable control rod scram system, operation of the reactor is permitted tocontinue for short periods of time with the system inoperable or for longerperiods of time with one of the redundant components inoperable.

The SLCS system consists of three separate and independent pumps andexplosive valves. Two of the separate and independent pumps and explosive valvesare required to meet the minimum requirements of this technical specification and, where applicable, satisfy the single failure criterion.

To ensure that SLCSpump discharge pressure does not exceed the SLCS relief valve setpoint duringoperation following an anticipated transient without scram (ATWS) event, no morethan two pumps shall be aligned for automatic operation in OPERATIONAL CONDITIONS 1, 2, and 3. This maintains the equivalent control capacity to satisfy 10 CFR50.62 (Requirements for reduction of risk from anticipated transients withoutscram (ATWS). With three pumps aligned for automatic operation, the system isinoperable and ACTION statement (b) applies.The SLCS must have an equivalent control capacity of 86 gpm of 13% weightsodium pentaborate in order to satisfy 10 CFR 50.62. As part of the ARTS/MELLL program the ATWS analysis was updated to reflect the new rod line. As a resultof this it was determined that the Boron 10 enrichment was required to beincreased to 29% to prevent exceeding a suppression pool temperature of 1900F. This equivalency requirement is fulfilled by having a systemwhich satisfies the equation given in 4.1.5.b.2.

The upper limit concentration of 13.8% has been established as a reasonable limit to prevent precipitation of sodium pentaborate in the event of a loss oftank heating, which allow the solution to cool.LIMERICK

-UNIT 2B 3/4 1-4Amendment No. 48, 446,Associated with Amendment 163 REACTTVTTY CONTROL SYSTEMSBASESSTANDBY LIQUID CONTROL SYSTEM (Continued)

Surveillance requirements are established on a frequency that assures a highreliability of the system. Once the solution is established, boron concentration will not vary unless more boron or water is added, thus a check on the temperature and volume assures that the solution is available for use.Replacement of the explosive charges in the valves will assure that these valveswill not fail because of deterioration of the charges.The Standby Liquid Control System also has a post-DBA LOCA safety function tobuffer Suppression Pool pH in order to maintain bulk pH above 7.0. The buffering ofSuppression Pool pH is necessary to prevent iodine re-evolution to satisfy themethodology for Alternative Source Term. Manual initiation is used, and the minimumamount of total boron required for Suppression Pool pH buffering is 240 lbs. Giventhat at least 185 lbs of Boron-lO is maintained in the tank, the total boron in thetank will be greater than 240 lbs for the range of enrichments from 29% to 62%.ACTION Statement (a) applies only to OPERATIONAL CONDITIONS 1 and 2 because asingle pump can satisfy both the reactor control function and the post-DBA LOCAfunction to control Suppression Pool pH since boron injection is not required until13 hours post-LOCA.

ACTION Statement (b) applies to OPERATIONAL CONDITIONS 1, 2 and3 to address the post-LOCA safety function of the SLC system.1. C. J. Paone, R. C. Stirn and J. A. Woolley, "Rod Drop Accident Analysisfor Large BWR's," G. E. Topical Report NEDO-10527, March 1972.2. C. J. Paone, R. C. Stirn, and R. M. Young, Supplement 1 to NEDO-10527, July1972.3. J. M. Haun, C. J. Paone, and R. C. Stirn, Addendum 2, "Exposed Cores,"Supplement 2 to NEDO-10527, January 1973.4. Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel".5. "Maximum Extended Load Line Limit and ARTS Improvement Program Analyses forLimerick Generating Station Units I and 2," NEDC-32193P, Revision 2, October1993.LIMERICK

-UNIT 2B 3/4 1-5Amendment No. 489, 4-4-6, 147 THIS PAGE INTENTIONALLY LEFT BLANK 3/4 2 PnWFR OISTRTRIITTmN ITMTTSBASES3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATEThis specification assures that the peak cladding temperature (PCT)following the postulated design basis Loss-of-Coolant Accident (LOCA) will notexceed the limits specified in 10 CFR 50.46 and that the fuel design analysislimits specified in NEDE-24011-P-A (Reference

2) will not be exceeded.

Mechanical Design Analysis:

NRC approved methods (specified inReference

2) are used to demonstrate that all fuel rods in a lattice operating at the bounding power history, meet the fuel design limits specified inReference

2. No single fuel rod follows, or is capable of following, thisbounding power history.

This bounding power history is used as the basis forthe fuel design analysis MAPLHGR limit.LOCA Analysis:

A LOCA analysis is performed in accordance with 1OCFR50Appendix K to demonstrate that the permissible planar power (MAPLHGR) limitscomply with the ECCS limits specified in 10 CFR 50.46. The analysis is performed for the most limiting break size, break location, and single failure combination for the plant, using the evaluation model described in Reference 9.The MAPLHGR limit as showm in the CORE OPERATING LIMITS REPORT is themost limiting composite of the fuel mechanical design anaylsis MAPLHGR and theECCS MAPLHGR limit.Only the most limiting MAPLHGR values are shown in the CORE OPERATING LIMITS REPORT for multiple lattice fuel. Compliance with the specific latticeMAPLHGR operating limits, which are available in Reference 3, is ensured by useof the process computer.

As a result of no longer utilizing an APRM trip setdown requirement, additional constraints are placed on the MAPLHGR limits to assure adherence to the fuel-mechanical design bases. These constraints are introduced throughthe MAPFAC(P) and MAPFAC(F) factors as defined in the COLR.LIMERICK

-UNIT 2B 3/4 2-1Amendment No. 4, --4, 48 POWER DISTRIBUTION LIMITSBASES3/4.2.2 (DELETED)

INFORMATION CONTAINED ONTHIS PAGE HAS BEENDELETEDLIMERICK

-UNIT 2B 3/4 2-2Amendment No. 4, 4, 48 LEFT INTENTIONALLY BLANKAmendment No.14 ýLIMERICK

-UNIT 2B 3/4 2-3 POWER DISTRIBUTION LIMITSBASES3/4.2.3 MINIMUM CRITICAL POWER RATIOThe required operating limit MCPRs at steady-state operating conditions as specified in Specification 3.2.3 are derived from the established fuelcladding integrity Safety Limit MCPR, and an analysis of abnormal operational transients.

For any abnormal operating transient analysis evaluation with theinitial condition of the reactor being at the steady-state operating limit, itis required that less than 0.1% of fuel rods in the core are susceptible totransition boiling or that the resulting MCPR does not decrease below the SafetyLimit MCPR at any time during the transient assuming instrument trip settinggiven in Specification 2.2.To assure that the fuel cladding integrity Safety Limit is not exceededduring any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss offlow, increase in pressure and power, positive reactivity insertion, andcoolant temperature decrease.

The evaluation of a given transient begins with the system initial para-meters shown in FSAR Table 15.0-2 that are input to a BWR system dynamicbehavior transient computer program.

The codes used to evaluate transients arediscussed in Reference 2.The MCPR operating limits derived from the transient analysis are dependent on the operating core flow and power state (MCPR(F),

and MCPR(P),

respectively) toensure adherence to fuel design limits during the worst transient that occurswith moderate frequency (Ref. 6). Flow dependent MCPR limits (MCPR(F))

aredetermined by steady state thermal hydraulic methods with key physics responseinputs benchmarked using the three dimensional BWR simulator code (Ref. 7) toanalyze slow flow runout transients.

Power dependent MCPR limits (MCPR(P))

are determined by the codes used toevaluate transients as described in Reference

2. Due to the sensitivity of thetransient response to initial core flow levels at power levels below those atwhich the turbine stop valve closure and turbine control valve fast closurescrams are bypassed, high and low flow MCPR(P),

operating limits are providedfor operating between 25% RTP and 30% RTP.The MCPR operating limits specified in the COLR are the result of theDesign Basis Accident (DBA) and transient analysis.

The operating limit MCPRis determined by the larger of the MCPR(F),

and MCPR(P) limits.LIMERICK

-UNIT 2 B 3/4 2-4 Amendment No. 4, 489ECR LG 99 01138, ECR LG 12-00035 PnWFR lISTRIBIITIMN IIMTTSBASESMINIMUM CRITICAL POWER RATIO (Continued)

At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER,the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which maybe employed at this point, operating plant experience indicates that the resulting MCPR value is in excess of requirements by a considerable margin. During initialstartup testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMALPOWER level with minimum recirculation pump speed. The MCPR margin will thus bedemonstrated such that future MCPR evaluation below this power level will be shown tobe unnecessary.

The daily requirement for calculating MCPR when THERMAL POWER isgreater than or equal to 25% of RATED THERMAL POWER is sufficient since powerdistribution shifts are very slow when there have not been significant power orcontrol rod changes.

The requirement for calculating MCPR when a limiting controlrod pattern is approached ensures that MCPR will be known following a change inTHERMAL POWER or power shape, regardless of magnitude, that could place operation ata thermal limit.3/4.2.4 LINEAR HEAT GENERATION RATEThis specification assures that the Linear Heat Generation Rate (LHGR)in any rod is less than the design linear heat generation even if fuel pelletdensification is postulated.

Reference:

1. Deleted.2. "General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A (latest approved revision).

3. "Basis of MAPLHGR Technical Specifications for Limerick Unit 2," NEOC-31930P (as amended).

4. Deleted5. Increased Core Flow and Partial Feedwater Heating Analysis for LimerickGenerating Station Unit 2 Cycle 1, NEDC-31578P, MARCH 1989 including Errata and Addenda Sheet No. 1 dated MAY 31, 1989.6. NEDC-32193P, "Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Limerick Generating Station Units 1 and 2,"Revision 2, October 1993.7. NEDO-30130-A, "Steady State Nuclear Methods,"

May 1985.8. NEDO-24154, "Qualification of the One-Dimensional Core Transient Model forBoiling Water Reactors,"

October 1978.9. NEDC-32170P, "Limerick Generating Station Units 1 and 2 SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis,"

June 1993.LIMERICK

-UNIT 2B 3/4 2-5Amendment No. 4, --4, 48 THIS PAGE INTENTIONALLY LEFT BLANK 3/4.3 INSTRUMENTATION BASES3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:a. Preserve the integrity of the fuel cladding.

b. Preserve the integrity of the reactor coolant system.c. Minimize the energy which must be adsorbed following aloss-of-coolant

accident, andd. Prevent inadvertent criticality.

This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intendedfunction even during periods when instrument channels may be out of servicebecause of maintenance.

When necessary, one channel may be made inoperable for brief intervals to conduct required surveillance.

The reactor protection system is made up of two independent trip systems.There are usually four channels to monitor each parameter with two channels in eachtrip system. The outputs of the channels in a trip system are combined in a logic sothat either channel will tripthat trip system. The tripping of both trip systemswill produce a reactor scram. The APRM system is divided into four APRM channels andfour 2-Out-Of-4 Voter channels.

Each APRM channel provides inputs to each of thefour voter channels.

The four voter channels are divided into two groups of twoeach, with each group of two providing inputs to one RPS trip system. The system isdesigned to allow one APRM channel, but no voter channels, to be bypassed.

The system meets the intent of IEEE-279 for nuclear power plant protection systems.

Surveillance intervals are determined in accordance with the Surveillance Frequency Control Program and maintenance outage times have been determined inaccordance with NEDC-30851P-A, "Technical Specification Improvement Analyses for BWRReactor Protection System" and NEDC-32410P-A, "Nuclear Measurement Analysis andControl Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function."

The bases for the trip settings of the RPS are discussed in thebases for Specification 2.2.1.The APRM Functions include five Functions accomplished by the four APRMchannels (Functions 2.a, 2.b, 2.c, 2.d, and 2.f) and one accomplished by the four 2-Out-Of-4 Voter channels (Function 2.e). Two of the five Functions accomplished bythe APRM channels are based on neutron flux only (Functions 2.a and 2.c), oneFunction is based on neutron flux and recirculation drive flow (Function 2.b) and oneis based on equipment status (Function 2.d). The fifth Function accomplished by theAPRM channels is the Oscillation Power Range Monitor (OPRM) Upscale trip Function2.f, which is based on detecting oscillatory characteristics in the neutron flux.The OPRM Upscale Function is also dependent on average neutron flux (Simulated Thermal Power) and recirculation drive flow, which are used to automatically enablethe output trip.The Two-Out-Of-Four Logic Module includes 2-Out-Of-4 Voter hardware and theAPRM Interface hardware.

The 2-Out-Of-4 Voter Function 2.e votes APRM Functions 2.a,2.b, 2.c, and 2.d independently of Function 2.f. This voting is accomplished by the2-Out-Of-4 Voter hardware in the Two-Out-Of-Four Logic Module. The voter includesseparate outputs to RPS for the two independently voted sets of Functions, each ofwhich is redundant (four total outputs).

The analysis in Reference 2 took credit forthis redundancy in the justification of the 12-hour allowed out-of-service time forLIMERICK

-UNIT 2B 3/4 3-1 Amendment No. 4-7,-15,2-9-3,4-4,4-39,147 3/4.3 INSTRUMENTATION BASES3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued)

Action b, so the voter Function 2.e must be declared inoperable if any of itsfunctionality is inoperable.

The voter Function 2.e does not need to be declaredinoperable due to any failure affecting only the APRM Interface hardware portion ofthe Two-Out-Of-Four Logic Module.Three of the four APRM channels and all four of the voter channels arerequired to be OPERABLE to ensure that no single failure will preclude a scram on avalid signal. To provide adequate coverage of the entire core, consistent with thedesign bases for the APRM Functions 2.a, 2.b, and 2.c, at least 20 LPRM inputs, withat least three LPRM inputs from each of the four axial levels at which the LPRMs arelocated, must be operable for each APRM channel.

In addition, no more than 9 LPRMsmay be bypassed between APRM calibrations (weekly gain adjustments).

For the OPRMUpscale Function 2.f, LPRMs are assigned to "cells" of 3 or 4 detectors.

A minimumof 23 cells (Reference 9), each with a minimum of 2 OPERABLE LPRMs, must be OPERABLEfor each APRM channel for the OPRM Upscale Function 2.f to be OPERABLE in thatchannel.

LPRM gain settings are determined from the local flux profiles measured bythe TIP system. This establishes the relative local flux profile for appropriate representative input to the APRM System. The 2000 EFPH frequency is based onoperating experience with LPRM sensitivity changes.References 4, 5 and 6 describe three algorithms for detecting thermal-hydraulic instability related neutron flux oscillations:

the period based detection algorithm, the amplitude based algorithm, and the growth rate algorithm.

All threeare implemented in the OPRM Upscale Function, but the safety analysis takes creditonly for the period based detection algorithm.

The remaining algorithms providedefense in depth and additional protection against unanticipated oscillations.

OPRMUpscale Function OPERABILITY for Technical Specification purposes is based only onthe period based detection algorithm.

An OPRM Upscale trip is issued from an APRM channel when the period baseddetection algorithm in that channel detects oscillatory changes in the neutron flux,indicated by the combined signals of the LPRM detectors in any cell, with periodconfirmations and relative cell amplitude exceeding specified setpoints.

One or morecells in a channel exceeding the trip conditions will result in a channel trip. AnOPRM Upscale trip is also issued from the channel if either the growth rate oramplitude based algorithms detect growing oscillatory changes in the neutron flux forone or more cells in that channel.The OPRM Upscale Function is required to be OPERABLE when the plant is at25% RATED THERMAL POWER. The 25% RATED THERMAL POWER level is selected to providemargin in the unlikely event that a reactor power increase transient occurring whilethe plant is operating below 29.5% RATED THERMAL POWER causes a power increase to orbeyond the 29.5% RATED THERMAL POWER OPRM Upscale trip auto-enable point withoutoperator action. This OPERABILITY requirement assures that the OPRM Upscale tripautomatic-enable function will be OPERABLE when required.

Actions a, b and c define the Action(s) required when RPS channels arediscovered to be inoperable.

For those Actions, separate entry condition is allowedfor each inoperable RPS channel.

Separate entry means that the allowable timeclock(s) for Actions a, b or c start upon discovery of inoperability for thatspecific channel.

Restoration of an inoperable RPS channel satisfies only the actionstatements for that particular channel.

Action statement(s) for remaining inoperable channel(s) must be met according to their original entry time.Because of the diversity of sensors available to provide trip signals and theredundancy of the RPS design, an allowable out of service time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months has beenshown to be acceptable (NEDC-30851P-A and NEDC-32410P-A) to permit restoration of anyinoperable channel to OPERABLE status. However, this out of service time is onlyacceptable provided that the associated Function's (identified as a "Functional Unit"in Table 3.3.1-1) inoperable channel is in one trip system and the Function stillmaintains RPS trip capability.

LIMERICK

-UNIT 2B 3/4 3-1aAmendment No. 4-,-5-29,4,1-O,44-39,4-56, Associated with Amendment 163 3/4.3 INSTRUMENTATION BASES3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued)

The requirements of Action a are intended to ensure that appropriate actionsare taken if multiple, inoperable, untripped channels within the same trip system forthe same Function result in the Function not maintaining RPS trip capability.

AFunction is considered to be maintaining RPS trip capability when sufficient channelsare OPERABLE or in trip (or the associated trip system is in trip), such that bothtrip systems will generate a trip signal from the given Function on a valid signal.For the typical Function with one-out-of-two taken twice logic, including the IRMFunctions and APRM Function 2.e (trip capability associated with APRM Functions 2.a,2.b, 2.c, 2.d, and 2.f are discussed below), this would require both trip systems tohave one channel OPERABLE or in trip (or the associated trip system in trip).For Function 5 (Main Steam Isolation Valve--Closure),

this would require bothtrip systems to have each channel associated with the MSIVs in three main steam lines(not necessarily the same main steam lines for both trip systems)

OPERABLE or in trip(or the associated trip system in trip).For Function 9 (Turbine Stop Valve-Closure),

this would require both tripsystems to have three channels, each OPERABLE or in trip (or the associated tripsystem in trip).The completion time to satisfy the requirements of Action a is intended toallow the operator time to evaluate and repair any discovered inoperabilities.

The 1hour Completion Time is acceptable because it minimizes risk while allowing time forrestoration or tripping of channels.

With trip capability maintained, i.e., Action a satisfied, Actions b and c asapplicable must still be satisfied.

If the inoperable channel cannot be restored toOPERABLE status within the allowable out of service time, Action b requires that thechannel or the associated trip system must be placed in the tripped condition.

Placing the inoperable channel in trip (or the associated trip system in trip) wouldconservatively compensate for the inoperability, restore capability to accommodate asingle failure, and allow operation to continue.

As noted, placing the trip system in trip is not applicable to satisfy Actionb for APRM Functions 2.a, 2.b, 2.c, 2.d, or 2.f. Inoperability of one required APRMchannel affects both trip systems.

For that condition, the Action b requirements canonly be satisfied by placing the inoperable APRM channel in trip. Restoring OPERABILITY or placing the inoperable APRM channel in trip are the only actions thatwill restore capability to accommodate a single APRM channel failure.

Inoperability of more than one required APRM channel of the same trip function results in loss oftrip capability and the requirement to satisfy Action a.The requirements of Action c must be satisfied when, for any one or moreFunctions, at least one required channel is inoperable in each trip system. In thiscondition, provided at least one channel per trip system is OPERABLE, normally theRPS still maintains trip capability for that Function, but cannot accommodate asingle failure in either trip system (see additional bases discussion above relatedto loss of trip capability and the requirements of Action a, and special cases forFunctions 2.a, 2.b, 2.c, 2.d, 2.f, 5 and 9).LIMERICK

-UNIT 2B 3/4 3-1bAmendment No. 4-44, 139 3/4.3 INSTRUMENTATION BASES3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued)

The requirements of Action c limit the time the RPS scram logic, for anyFunction, would not accommodate single failure in both trip systems (e.g., one-out-of-one and one-out-of-one arrangement for a typical four channel Function).

Thereduced reliability of this logic arrangement was not evaluated in NEDC-30851P-A forthe 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Completion Time. Within the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the associated Functionmust have all required channels OPERABLE or in trip (or any combination) in one tripsystem.Completing the actions required by Action c restores RPS to a reliability level equivalent to that evaluated in NEDC-30851P-A, which justified a 12 hourallowable out of service time as allowed by Action b. To satisfy the requirements ofAction c, the trip system in the more degraded state should be placed in trip or,alternatively, all the inoperable channels in that trip system should be placed intrip (e.g., a trip system with two inoperable channels could be in a more degradedstate than a trip system with four inoperable channels if the two inoperable channelsare in the same Function while the four inoperable channels are all in different Functions).

The decision of which trip system is in the more degraded state shouldbe based on prudent judgment and take into account current plant conditions (i.e.,what OPERATIONAL CONDITION the plant is in). If this action would result in a scramor RPT, it is permissible to place the other trip system or its inoperable channelsin trip.The 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowable out of service time is judged acceptable based on theremaining capability to trip, the diversity of the sensors available to provide thetrip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of a scram.As noted, Action c is not applicable for APRM Functions 2.a, 2.b, 2.c, 2.d, or2.f. Inoperability of an APRM channel affects both trip systems and is not associated with a specific trip system as are the APRM 2-Out-Of-4 voter and other non-APRMchannels for which Action c applies.

For an inoperable APRM channel, the requirements of Action b can only be satisfied by tripping the inoperable APRM channel.

Restoring OPERABILITY or placing the inoperable APRM channel in trip are the only actions thatwill restore capability to accommodate a single APRM channel failure.If it is not desired to place the channel (or trip system) in trip to satisfythe requirements of Action a, Action b or Action c (e.g., as in the case whereplacing the inoperable channel in trip would result in a full scram), Action drequires that the Action defined by Table 3.3.1-1 for the applicable Function beinitiated immediately upon expiration of the allowable out of service time.Table 3.3.1-1, Function 2.f, references Action 10, which defines the actionrequired if OPRM Upscale trip capability is not maintained.

Action lOb isrequired to address identified equipment failures.

Action lOa is to addresscommon mode vendor/industry identified issues that render all four OPRM channelsinoperable at once. For this condition, References 2 and 3 justified use ofalternate methods to detect and suppress oscillations for a limited period oftime, up to 120 days. The alternate methods are procedurally established consistent with the guidelines identified in Reference 7 requiring manualoperator action to scram the plant if certain predefined events occur. The 12-hour allowed completion time to implement the alternate methods is based onengineering judgment to allow orderly transition to the alternate methods whilelimiting the period of time during which no automatic or alternate detect andLIMERICK

-UNIT 2B 3/4 3-1cAmendment No. 4-94, 139 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued) suppress trip capability is formally in place. The 120-day period during whichuse of alternate methods is allowed is intended to be an outside limit to allowfor the case where design changes or extensive analysis might be required tounderstand or correct some unanticipated characteristic of the instability detection algorithms or equipment.

The evaluation of the use of alternate methods concluded, based on engineering

judgment, that the likelihood of aninstability event that could not be adequately handled by the alternate methodsduring the 120-day period was negligibly small. Plant startup may continue whileoperating within the allowed completion time of Action lOa. The primary purposeof this is to allow an orderly completion, without undue impact on plantoperation, of design and verification activities in the event of a requireddesign change to the OPRM Upscale function.

This exception is not intended as analternative to restoring inoperable equipment to OPERABLE status in a timelymanner.Action lOa is not intended and was not evaluated as a routine alternative to returning failed or inoperable equipment to OPERABLE status. Correction ofroutine equipment failure or inoperability is expected to be accomplished withinthe completion times allowed for LCO 3.3.1 Action a or Action b, as applicable.

Action 10b applies when routine equipment OPERABILITY cannot be restored withinthe allowed completion times of LCO 3.3.1 Actions a or b, or if a common modeOPRM deficiency cannot be corrected and OPERABILITY of the OPRM Upscale Functionrestored within the 120-day allowed completion time of Action lOa.The OPRM Upscale trip output shall be automatically enabled (not-bypassed) when APRM Simulated Thermal Power is 29.5% and recirculation drive flow is < 60%as indicated by APRM measured recirculation drive flow. NOTE: 60% recirculation drive flow is the recirculation drive flow that corresponds to 60% of rated coreflow. This is the operating region where actual thermal-hydraulic instability andrelated neutron flux oscillations may occur. As noted in Table 4.3.1.1-1, Notec, CHANNEL CALIBRATION for the OPRM Upscale trip Function 2.f includes confirming that the auto-enable (not-bypassed) setpoints are correct.

Other surveillances ensure that the APRM Simulated Thermal Power properly correlates with THERMALPOWER (Table 4.3.1.1-1, Note d) and that recirculation drive flow properlycorrelates with core flow (Table 4.3.1.1-1, Note g).If any OPRM Upscale trip auto-enable setpoint is exceeded and the OPRMUpscale trip is not enabled, i.e., the OPRM Upscale trip is bypassed when APRMSimulated Thermal Power is 29.5% and recirculation drive flow is < 60%, then theaffected channel is considered inoperable for the OPRM Upscale Function.

Alternatively, the OPRM Upscale trip auto-enable setpoint(s) may be adjusted toplace the channel in the enabled condition (not-bypassed).

If the OPRM Upscaletrip is placed in the enabled condition, the surveillance requirement is met andthe channel is considered OPERABLE.

As noted in Table 4.3.1.1-1, Note g, CHANNEL CALIBRATION for the APRMSimulated Thermal Power -Upscale Function 2.b and the OPRM Upscale Function 2.f,includes the recirculation drive flow input function.

The APRM Simulated ThermalPower -Upscale Function and the OPRM Upscale Function both require a valid driveflow signal. The APRM Simulated Thermal Power -Upscale Function uses drive flowto vary the trip setpoint.

The OPRM Upscale Function uses drive flow toautomatically enable or bypass the OPRM Upscale trip output to RPS. A CHANNELCALIBRATION of the APRM recirculation drive flow input function requires bothcalibrating the drive flow transmitters and establishing a valid drive flow /LIMERICK

-UNIT 2B 3/4 3-idAmendment No. 4,-3-,Associated with Amendment 163 3/4.3 INSTRUMENTATION BASES3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued) core flow relationship.

The drive flow / core flow relationship is established once per refuel cycle, while operating within 10% of rated core flow and within10% of RATED THERMAL POWER. Plant operational experience has shown that thisflow correlation methodology is consistent with the guidance and intent inReference

8. Changes throughout the cycle in the drive flow / core flowrelationship due to the changing thermal hydraulic operating conditions of thecore are accounted for in the margins included in the bases or analyses used toestablish the setpoints for the APRM Simulated Thermal Power -Upscale Functionand the OPRM Upscale Function.

For the Simulated Thermal Power -Upscale Function (Function 2.b), theCHANNEL CALIBRATION surveillance requirement is modified by two Notes. Thefirst Note requires evaluation of channel performance for the condition wherethe as-found setting for the channel setpoint is outside its as-found tolerance but conservative with respect to the Allowable Value. Evaluation of channelperformance will verify that the channel will continue to behave in accordance with safety analysis assumptions and the channel performance assumptions in thesetpoint methodology.

The purpose of the assessment is to ensure confidence inthe channel performance prior to returning the channel to service.

For channelsdetermined to be OPERABLE but degraded, after returning the channel to servicethe performance of these channels will be evaluated under the plant Corrective Action Program.

Entry into the Corrective Action Program will ensure requiredreview and documentation of the condition.

The second Note requires that theas-left setting for the channel be within the as-left tolerance of the TripSetpoint.

The as-left and as-found tolerances, as applicable, will be appliedto the surveillance procedure setpoint.

This will ensure that sufficient marginto the Safety Limit and/or Analytical Limit is maintained.

If the as-leftchannel setting cannot be returned to a setting within the as-left tolerance ofthe Trip Setpoint, then the channel shall be declared inoperable.

The as-lefttolerance for this function is calculated using the square-root-sum-of-squares of the reference accuracy and the measurement and test equipment error(including readability).

The as-found tolerance for this function is calculated using the square-root-sum-of-squares of the reference

accuracy, instrument drift, and the measurement and test equipment error (including readability).

As noted in Table 3.3.1-2, Note "*", the redundant outputs from the2-Out-Of-4 Voter channel are considered part of the same channel, but theOPRM and APRM outputs are considered to be separate

channels, so N = 8 todetermine the interval between tests for application of Specification 4.3.1.3 (REACTOR PROTECTION SYSTEM RESPONSE TIME). The note furtherrequires that testing of OPRM and APRM outputs shall be alternated.

Each test of an OPRM or APRM output tests each of the redundant outputsfrom the 2-Out-Of-4 Voter channel for that function, and each of thecorresponding relays in the RPS. Consequently, each of the RPS relays is testedevery fourth cycle. This testing frequency is twice the frequency justified byReferences 2 and 3.Automatic reactor trip upon receipt of a high-high radiation signalfrom the Main Steam Line Radiation Monitoring System was removed as the resultof an analysis performed by General Electric in NEDO-31400A.

The NRC approvedthe results of this analysis as documented in the SER (letter to George J. Beck,BWR Owner's Group from A.C. Thadani, NRC, dated May 15, 1991).LIMERICK

-UNIT 2 B 3/4 3-le Amendment No. 444-4--3-9.-4-4.

Associated with Amendment 163 314.3 TNSTRIJMFNTATTON BASES3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (continued)

The measurement of response time at the frequencies specified in theSurveillance Frequency Control Program provides assurance that the protective functions associated with each channel are completed within the time limitassumed in the safety analyses.

No credit was taken for those channels withresponse times indicated as not applicable except for the APRM Simulated ThermalPower -Upscale and Neutron Flux -Upscale trip functions and the OPRM Upscaletrip function (Table 3.3.1-2, Items 2.b, 2.c, and 2.f). Response time may bedemonstrated by any series of sequential, overlapping or total channel testmeasurement, provided such tests demonstrate the total channel response time asdefined.

Sensor response time verification may be demonstrated by either (1)inplace, onsite or offsite test measurements, or (2) utilizing replacement sensors with certified response times. Response time testing for the sensors asnoted in Table 3.3.1-2 is not required based on the analysis in NEDO-32291-A.

Response time testing for the remaining channel components is required as noted.For the digital electronic portions of the APRM functions, performance characteristics that determine response time are checked by a combination ofautomatic self-test, calibration activities, and response time tests of the 2-Out-Of-4 Voter (Table 3.3.1-2, Item 2.e).LIMERICK

-UNIT 2B 3/4 3-1fAmendment No. 4-99,4-3-9,-4-7, Associated with Amendment 163 INSTRUMENTATION BASES3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used tomitigate the consequences of accidents by prescribing the OPERABILITY tripsetpoints and response times for isolation of the reactor systems.

Whennecessary, one channel may be inoperable for brief intervals to conduct requiredsurveillance.

Surveillance intervals are determined in accordance with the Surveillance Frequency Control Progam and maintenance outage times have been determined inaccordance with NEDC-30851P, Supplement 2, "Technical Specification Improvement Analysis for BWR Instrumentation Common to RPS and ECCS Instrumentation" asapproved by the NRC and documented in the NRC Safety Evaluation Report (SER)(letter to D.N. Grace from C.E. Rossi dated January 6, 1989) and NEDC-31677P-A, "Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation,"

as approved by the NRC and documented in the NRC SER (letter toS.D. Floyd from C.E. Rossi dated June 18, 1990).Automatic closure of the MSIVs upon receipt of a high-high radiation signal from the Main Steam Line Radiation Monitoring System was removed as theresult of an analysis performed by General Electric in NEDO-31400A.

The NRCapproved the results of this analysis as documented in the SER (letter toGeorge J. Beck, BWR Owner's Group from A.C. Thadani, NRC, dated May 15, 1991).Some of the trip settings may have tolerances explicitly stated where boththe high and low values are critical and may have a substantial effect onsafety. The setpoints of other instrumentation, where only the high or low endof the setting have a direct bearing on safety, are established at a level awayfrom the normal operating range to prevent inadvertent actuation of the systemsinvolved.

Except for the MSIVs, the safety analysis does not address individual sensorresponse times or the response times of the logic systems to which the sensorsare connected.

For D.C. operated valves, a 3 second delay is assumed before thevalve starts to move. For A.C. operated valves, it is assumed that the A.C.power supply is lost and is restored by startup of the emergency dieselgenerators.

In this event, a time of 13 seconds is assumed before the valvestarts to move. In addition to the pipe break, the failure of the D.C. operatedvalve is assumed; thus the signal delay (sensor response) is concurrent with the10-second diesel startup and the 3 second load center loading delay. The safetyanalysis considers an allowable inventory loss in each case which in turndetermines the valve speed in conjunction with the 13-second delay. It followsthat checking the valve speeds and the 13-second time for emergency powerestablishment will establish the response time for the isolation functions.

Response time testing for sensors are not required based on the analysis inNEDO-32291-A.

Response time testing of the remaining channel components isrequired as noted in Table 3.3.2-3.Operation with a trip set less conservative than its Trip Setpoint butwithin its specified Allowable Value is acceptable on the basis that thedifference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safetyanalyses.

Primary containment isolation valves that are actuated by the isolation signals specified in Technical Specification Table 3.3.2-1 are identified inTechnical Requirements Manual Table 3.6.3-1.3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation instrumentation is provided toinitiate actions to mitigate the consequences of accidents that are beyond theability of the operator to control.

This specification provides the OPERABILITY requirements, trip setpoints and response times that will ensure effectiveness of the systems to provide the design protection.

Although the instruments are 0listed by system, in some cases the same instrument may be used to send theactuation signal to more than one system at the same time.LIMERICK

-UNIT 2B 3/4 3-2 Amendment No. 4483-5-2,9-3,-

, 147 INSTRUMENTATION BASES3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION (Continued)

Surveillance intervals are determined in accordance with the Surveillance Frequency Control Program and maintenance outage times have been determined inaccordance with NEDC-30936P, Parts 1 and 2, "Technical Specification Improvement Methodology (with Demonstration for BWR ECCS Actuation Instrumentation),"

asapproved by the NRC and documented in the SER (letter to 0. N. Grace from A. C.Thadani dated December 9, 1988 (Part 1) and letter to D. N. Grace from C. E.Rossi dated December 9, 1988 (Part 2)).Successful operation of the required safety functions of the Emergency CoreCooling Systems (ECCS) is dependent upon the availability of adequate power forenergizing various components such as pump motors, motor operated valves, and theassociated control components.

If the loss of power instrumentation detects thatvoltage levels are too low, the buses are disconnected from the offsite powersources and connected to the onsite diesel generator (DG) power sources.

The lossof power relays in each channel have sufficient overlapping detection characteristics and functionality to permit operation subject to the conditions inAction Statement

37. Bases 3/4.8.1, 3/4.8.2, and 3/4.8.3 provide discussion regarding parametric bounds for determining operability of the offsite sources.Those Bases assume that the loss of power relays are operable.

With an inoperable 127Z-11XOX relay, the grid voltage is monitored to 230kV (for the 101 Safeguard Bus Source) or 525kV (for the 201 Safeguard Bus Source) to increase the margin forthe operation of the 127Z-11XOX relay.Operation with a trip set less conservative than its Trip Setpoint butwithin its specified Allowable Value is acceptable on the basis that thedifference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safetyanalyses.

3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION The anticipated transient without scram (ATWS) recirculation pump tripsystem provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient.

The response of theplant to this postulated event falls within the envelope of study events inGeneral Electric Company Topical Report NEDO-10349, dated March 1971, NEDO-24222, dated December 1979, and Section 15.8 of the FSAR.The end-of-cycle recirculation pump trip (EOC-RPT) system is a supplement tothe reactor trip. During turbine trip and generator load rejection events, theEOC-RPT will reduce the likelihood of reactor vessel level decreasing to level2. Each EOC-RPT system trips both recirculation pumps, reducing coolant flow inorder to reduce the void collapse in the core during two of the most limitingpressurization events. The two events for which the EOC-RPT protective featurewill function are closure of the turbine stop valves and fast closure of theturbine control valves.A fast closure sensor from each of two turbine control valves provides inputto the EOC-RPT system; a fast closure sensor from each of the other two turbinecontrol valves provides input to the second EOC-RPT system. Similarly, aposition switch for each of two turbine stop valves provides input to one EOC-RPT system; a position switch from each of the other two stop valves providesinput to the other EOC-RPT system. For each EOC-RPT system, the sensor relaycontacts are arranged to form a 2-out-of-2 logic for the fast closure of turbinecontrol valves and a 2-out-of-2 logic for the turbine stop valves. Theoperation of either logic will actuate the EOC-RPT system and trip bothrecirculation pumps.LIMERICK

-UNIT 2B 3/4 3-3 Amendment No. -I-7,33,4-2-O, 147 INSTRUMENTATION BASES3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION (Continued)

Each EOC-RPT system may be manually bypassed by use of a keyswitch which isadministratively controlled.

The manual bypasses and the automatic Operating Bypass at less than 29.5% of RATED THERMAL POWER are annunciated in the controlroom.The EOC-RPT system response time is the time assumed in the analysis betweeninitiation of valve motion and complete suppression of the electric arc, i.e.,175 ms. Included in this time are: the response time of the sensor, the timeallotted for breaker arc suppression, and the response time of the system logic.LIMERICK

-UNIT 2B 3/4 3-3aAmendment No. 4-2-0,Associated with Amendment 163 INSTRUMENTATION BASESSurveillance intervals are determined in accordance with the Surveillance Frequency Control Program and maintenance outage times have been determined inaccordance with GENE-770-06-1, "Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications,"

as approved by the NRC and documented in the SER (letter to R.D.Binz, IV, from C.E. Rossi dated July 21, 1992).Operation with a trip set less conservative than its Trip Setpoint butwithin its specified Allowable Value is acceptable on the basis that thedifference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safetyanalyses.

3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation isprovided to initiate actions to assure adequate core cooling in the event ofreactor isolation from its primary heat sink and the loss of feedwater flow tothe reactor vessel. This instrumentation does not provide actuation of any ofthe emergency core cooling equipment.

Surveillance intervals are determined in accordance with the Surveillance Frequency Control Program and maintenance outage times have been specified inaccordance with recommendations made by GE in their letter to the BWR Owner'sGroup dated August 7, 1989,

SUBJECT:

"Clarification of Technical Specification changes given in ECCS Actuation Instrumentation Analysis."

Operation with a trip set less conservative than its Trip Setpoint butwithin its specified Allowable value is acceptable on the basis that thedifference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safetyanalyses.

3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION The control rod block functions are provided consistent with therequirements of the specifications in Section 3/4.1.4, Control Rod ProgramControls and Section 3/4.2 Power Distribution Limits and Section 3/4.3Instrumentation.

The trip logic is arranged so that a trip in any one of theinputs will result in a control rod block.Surveillance intervals are determined in accordance with the Surveillance Frequency Control Program and maintenance outage time have been determined inaccordance with NEDC-30851P, Supplement 1, "Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation,"

as approved by the NRC anddocumented in the SER (letter to D. N. Grace from C. E. Rossi dated September 22,1988).Operation with a trip set less conservative than its Trip Setpoint butwithin its specified Allowable Value is acceptable on the basis that thedifference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safetyanalyses.

LIMERICK

-UNIT 2B 3/4 3-4Amendment No. 4-4, 4-7, 3, 147 INTENTIONALLY LEFT BLANK INSTRUMENTATION BASES3/4.3.7 MONITORING INSTRUMENTATION 3/4.3.7.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring instrumentation ensures that:(1) the radiation levels are continually measured in the areas served by theindividual

channels, and (2) the alarm or automatic action is initiated when theradiation level trip setpoint is exceeded, and (3) sufficient information isavailable on selected plant parameters to monitor and assess these variables following an accident.

This capability is consistent with 10 CFR Part 50,Appendix A, General Design Criteria 19, 41, 60, 61, 63, and 64.The surveillance interval for the Main Control Room Normal Fresh Air SupplyRadiation Monitor is determined in accordance.

with the Surveillance Frequency ControlProgram.3/4.3.7.2 (Deleted)

INFORMATION FROM THIS SECTION RELOCATED TO THE UFSAR.3/4.3.7.3 (Deleted)

-INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM.3/4.3.7.4 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION AND CONTROLSThe OPERABILITY of the remote shutdown system instrumentation and controlsensures that sufficient capability is available to permit shutdown and maintenance ofHOT SHUTDOWN of the unit from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent withGeneral Design Criterion 19 of 10 CFR Part 50, Appendix A. The Unit 1 RHR transferswitches are included only due to their potential impact on the RHRSW system, which iscommon to both units.3/4.3.7.5 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures thatsufficient information is available on selected plant parameters to monitor andassess important variables following an accident.

This capability is consistent withthe recommendations of Regulatory Guide 1.97, "Instrumentation for Light Water CooledNuclear Power Plants to Assess Plant Conditions During and Following an Accident,"

December 1975 and NUREG-0737, "Clarification of TMI Action Plan Requirements,"

November 1980.Table 3.3.7.5-1, Accident Monitoring Instrumentation, Item 2, requires twoOPERABLE channels of Reactor Vessel Water Level monitoring from each of twooverlapping instrumentation loops to ensure monitoring of Reactor Vessel Water Levelover the range of -350 inches to +60 inches. Each channel is comprised of oneOPERABLE Wide Range Level instrument loop (-150 inches to +60 inches) and one OPERABLEFuel Zone Range instrument loop (-350 inches to -100 inches).

Both instrument loops,Wide Range and Fuel Zone Range, are required by Tech. Spec. 3.3.7.5 to providesufficient overlap to bound the required range as described in UFSAR Section 7.5.Action 80 is applicable if the required number of instrument loops per channel (WideRange and Fuel Zone Range) is not maintained.

LIMERICK

-UNIT 2 B 3/4 3-5 Amendment No. 4-1-,4-,3-3,46,4-15, ECR 02 00470,-31-5,-4-7, ECR LG 09-00585 INSTRUMENTATION BASES3/4.3.7.5 ACCIDENT MONITORING INSTRUMENTATION (continued)

Table 3.3.7.5-1, Accident Monitoring Instrumentation, Item 13, requires twoOPERABLE channels of Neutron Flux monitoring from each of three overlapping instrumentation loops to ensure monitoring of Neutron Flux over the range of 10-6% to100% full power. Each channel is comprised of one OPERABLE SRM (10-'% to 10-3% power),one OPERABLE IRM (10-% to 40% power) and one OPERABLE APRM (0% to 125% power). Allthree instrument loops, SRM, IRM and APRM, are required by Tech. Spec..3.3.7.5 toprovide sufficient overlap to bound the required range as described in UFSAR Section7.5. Action 80 is applicable if the required number of instrument loops per channel(SRM, IRM, and APRM) is not maintained.

3/4.3.7.6 SOURCE RANGE MONITORSThe source range monitors provide the operator with information of the statusof the neutron level in the core at very low power levels during startup and shutdown.

At these power levels, reactivity additions shall not be made without this flux levelinformation available to the operator.

When the intermediate range monitors are onscale, adequate information is available without the SRMs and they can be retracted.

LIMERICK

-UNIT 2B 3/4 3-5aAmendment No. 44,4-7,-3,4,44-5 ECR LG 09-00585 INSTRUMENTATION BASES3/4.3.7.7 (Deleted)

-INFORMATION FROM THIS SECTION RELOCATED TO THE TRM.3/4.3.7.8 CHLORINE AND TOXIC GAS DETECTION SYSTEMSThe OPERABILITY of the chlorine and toxic gas detection systems ensuresthat an accidental chlorine and/or toxic gas release will be detected promptlyand the necessary protective actions will be automatically initiated for chlo-rine and manually initiated for toxic gas to provide protection for controlroom personnel.

Upon detection of a high concentration of chlorine, the controlroom emergency ventilation system will automatically be placed in the chlorineisolation mode of operation to provide the required protection.

Upon detection of a high concentration of toxic gas, the control room emergency ventilation system will manually be placed in the chlorine isolation mode of operation toprovide the required protection.

The detection systems required by this speci-fication are consistent with the recommendations of Regulatory Guide 1.95, "Pro-tection of Nuclear Power Plant Control Room Operators against an Accidental Chlorine Release,"

February 1975.There are three toxic gas detection subsystems.

The high toxic chemicalconcentration alarm in the Main Control Room annunciates when two of the threesubsystems detect a high toxic gas concentration.

An Operate/Inop keylock switch isprovided for each subsystem which allows an individual subsystem to be placed in thetripped condition.

Placing the keylock switch in the INOP position initiates one ofthe two inputs required to initiate the alarm in the Main Control Room.Surveillance intervals are determined in accordance with the Surveillance Frequency Control Program and maintenance outage times have been determined inaccordance with GENE-770-06-1, "Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications,"

as approved by the NRC and documented in the SER (letter to R.D.Binz, IV, from C.E. Rossi dated July 21, 1992).3/4.3.7.9 (Deleted)

-INFORMATION FROM THIS SECTION RELOCATED TO THE TRM.LIMERICK

-UNIT 2B 3/4 3-6 Amendment No. 44,2-5,3,4-5,&8,7-9, 147 (INTENTIONALLY LEFT BLANK)

INSTRUMENTATION BASES3/4.3.7.10 (Deleted) 3/4.3.7.11 (Deleted)

-INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM.3/4.3.7.12 OFFGAS MONITORING INSTRUMENTATION This instrumentation includes provisions for monitoring the concentrations of potentially explosive gas mixtures and noble gases in the off-gas system.3/4.3.8 (Deleted)

-INFORMATION FROM THIS SECTION RELOCATED TO THE UFSAR.3/4.3.9 FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION The feedwater/main turbine trip system actuation instrumentation isprovided to initiate action of the feedwater system/main turbine trip systemin the event of failure of feedwater controller under maximum demand.

REFERENCES:

1. NEDC-30851P-A, "Technical Specification Improvement Analyses for BWRReactor Protection System,"

March 1988.2. NEDC-32410P-A, "Nuclear Measurement Analysis and Control Power RangeNeutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability TripFunction,"

October 1995.3. NEDC-32410P-A, Supplement 1, "Nuclear Measurement Analysis andControl Power Range Neutron Monitor (NUMAC PRNM) Retrofit PlusOption III Stability Trip Function,"

November 1997.4. NEDO-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology,"

November 1995.5. NEDO-31960-A, Supplement 1, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology,"

November 1995.6. NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications,"

August 1996.7. Letter, L. A. England (BWROG) to M. J. Virgilio, "BWR Owners' GroupGuidelines for Stability Interim Corrective Action,"

June 6, 1994.8. GE Service Information Letter No. 516, "Core Flow Measurement

-GEBWR/3, 4, 5 and 6 Plants,"

July 26, 1990.9. GE Letter NSA 00-433, Alan Chung (GE) to Sujit Chakraborty (GE),"Minimum Number of Operable OPRM Cells for Option III Stability atLimerick 1 & 2," May 02, 2001.LIMERICK

-UNIT 2 B 3/4 3-7 Amendment No. 4-4, 2-5, , 464, 48,44-1, 139 WAIIER LEVEL NOME?4CLATLOM NO.C2)(1)VESSEL ZERO REDN(OL)So". 239.0557.5 3 0.04".5 -3"3" -12LOWIDE RANGE NARRlOW RANGE"" )54.0 IRC Nd 44) -.01T IS IM ) REACTOINITIATE RCIC.HPCL TRIP RECRC.PULPS. CONT. OSOATON.129 (0) -10D?drrLlAE C.S.. 4T,.,START DIESEL ANDCONTRIBUTE TO A.D.S. -tTAF .161CLOSE AMSWFUIELZONEMi(NDICATION I ONLY)-161 jA- .311 X-101 L -350mTr", 5-~g S isw~ fee=-, isTp~mat wonv.b. Thecaftrbon is Vmu reaf at eandconamoe.

The Wvorn ernwat Jaw prmsuwes ctengee.&ves)

Is boonlde by vie safty WWJJMh~eh~ nafl*= Vi. wogt0 VWoS @vnm W Up. WO nfitevc svel.BASES FIGURE B 3/4.3-1REACTOR VESSEL WATER LEVELLIMERICK

-UNIT 2B 3/4 3-8 BASES FIGURE B 3/4 3-2APRM CONFIGURATION LIMERICK

-UNIT 2B 3/4 3-9Amendment No. 109 THIS PAGE INTENTIONALLY LEFT BLANK 3/4_4 RFACTOR COOL ANT SYSTFMBASES3/4.4.1 RECIRCULATION SYSTEMThe impact of single recirculation loop operation upon plant safety isassessed and shows that single-loop operation is permitted if the MCPR fuelcladding safety limit is increased as noted by Specification 2.1.2, APRM scramand control rod block setpoints are adjusted as noted in Tables 2.2.1-1 and3.3.6-2, respectively.

An inoperable jet pump is not, in itself, a sufficient reason to declarea recirculation loop inoperable, but it does, in case of a design-basis-accident, increase the blowdown area and reduce the capability of reflooding the core; thus,the requirement for shutdown of the facility with a jet pump inoperable.

Jet pumpfailure can be detected by monitoring jet pump performance on a pre-scribed schedule for significant degradation.

Additionally, surveillance on the pump speed of the operating recirculation loop is imposed to exclude the possibility of excessive internals vibration.

The surveillance on differential temperatures below 30%RATED THERMAL POWER or 50% rated recirculation loop flow is to mitigate theundue thermal stress on vessel nozzles, recirculation pump and vessel bottomhead during the extended operation of the single recirculation loop mode.Surveillance of recirculation loop flow, total core flow, and diffuser-to-lower plenum differential pressure is designed to detect significant degradation injet pump performance that precedes jet pump failure.

This surveillance is requiredto be performed only when the loop has forced recirculation flow since surveillance checks and measurements can only be performed during jet pump operation.

The jetpump failure of concern is a complete mixer displacement due to jet pump beamfailure.

Jet pump plugging is also of concern since it adds flow resistance to therecirculation loop. Significant degradation is indicated if the specified criteriaconfirm unacceptable deviations from established patterns or relationships.

Sincerefueling activities (fuel assembly replacement or shuffle, as well as anymodifications to fuel support orifice size or core plate bypass flow) can affectthe relationship between core flow, jet pump flow, and recirculation loop flow,these relationships may need to be re-established each cycle. Similarly, initialentry into extended single loop operation may also require establishment of theserelationships.

During the initial weeks of operation under such conditions, whilebase-lining new "established patterns,"

engineering judgment of the dailysurveillance results is used to detect significant abnormalities which couldindicate a jet pump failure.The recirculation pump speed operating characteristics (pump flow and loopflow versus pump speed) are determined by the flow resistance from the loopsuction through the jet pump nozzles.

A change in the relationship indicates aplug, flow restriction, loss in pump hydraulic performance,

leakage, or new flowpath between the recirculation pump discharge and jet pump nozzle. For thiscriterion, the pump flow and loop flow versus pump speed relationship must beverified.

LIMERICK

-UNIT 2 B 3/4 4-1 Amendment No. 4,43-94,Associated with Amendment 157 REACTOR COOLANT SYSTEMBASES3/4.4.1 RECIRCULATION SYSTEM (continued)

Individual jet pumps in a recirculation loop normally do not have the sameflow. The unequal flow is due to the drive flow manifold, which does notdistribute flow equally to all risers. The flow (or jet pump diffuser to lowerplenum differential pressure) pattern or relationship of one jet pump to the loopaverage is repeatable.

An appreciable change in this relationship is anindication that increased (or reduced) resistance has occurred in one of the jetpumps. This may be indicated by an increase in the relative flow for a jet pumpthat has experienced beam cracks.The deviations from normal are considered indicative of a potential problemin the recirculation drive flow or jet pump system. Normal flow ranges andestablished jet pump flow and differential pressure patterns are established byplotting historical data.Recirculation pump speed mismatch limits are in compliance with the ECCSLOCA analysis design criteria for two recirculation loop operation.

The limitswill ensure an adequate core flow coastdown from either recirculation loopfollowing a LOCA. In the case where the mismatch limits cannot be maintained during two loop operation, continued operation is permitted in a singlerecirculation loop mode.In order to prevent undue stress on the vessel nozzles and bottom headregion, the recirculation loop temperatures shall be within 50°F of each otherprior to startup of an idle loop. The loop temperature must also be within50'F of the reactor pressure vessel coolant temperature to prevent thermal shockto the recirculation pump and recirculation nozzles.

Sudden equalization of atemperature difference

> 145°F between the reactor vessel bottom head coolantand the coolant in the upper region of the reactor vessel by increasing coreflow rate would cause undue stress in the reactor vessel bottom head.3/4.4.2 SAFETY/RELIEF VALVESThe safety valve function of the safety/relief valves operates to preventthe reactor coolant system from being pressurized above the Safety Limit of1325 psig in accordance with the ASME Code. A total of 12 OPERABLE safety/relief valves is required to limit reactor pressure to within ASME III allow-able values for the worst case upset transient.

Demonstration of the safety/relief valve lift settings will occur onlyduring shutdown.

The safety/relief valves will be removed and either setpressure tested or replaced with spares which have been previously set pres-sure tested and stored in accordance with manufacturers recommendations at thefrequency specified in the Surveillance Frequency Control Program.LIMERICK

-UNIT 2 B 3/4 4-2 Amendment No. 9,4--94,44-,

Associated with Amendment 157 REACTOR COOLANT SYSTEMBASES.3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE3/4.4.3.1 LEAKAGE DETECTION SYSTEMSBACKGROUND UFSAR Safety Design Basis (Ref. 1), requires means for detecting and, to the extentpractical, identifying the location of the source of Reactor Coolant System (RCS)RESSURE BOUNDARY LEAKAGE.

Regulatory Guide 1.45, Revision 0, (Ref. 2) describes acceptable methods for selecting leakage detection systems.Limits on leakage from the reactor coolant pressure boundary (RCPB) are required sothat appropriate action can be taken before the integrity of the RCPB is impaired (Ref.2). Leakage detection systems for the RCS are provided to alert the operators whenleakage rates above normal background levels are detected and also to supplyquantitative measurement of leakage rates. In addition to meeting the OPERABILITY requirements, the monitors are typically set to provide the most sensitive responsewithout causing an excessive number of spurious alarms.Systems for quantifying the leakage are necessary to provide prompt and quantitative information to the operators to permit them to take immediate corrective action.Leakage from the RCPB inside the drywell is detected by at least one of four (4)independently monitored variables which include drywell sump flow monitoring equipment with the required RCS leakage detection instrumentation (i.e., the drywell floor drainsump flow monitoring system, or, the drywell equipment drain sump monitoring systemwith the drywell floor drain sump overflowing to the drywell equipment drain sump),drywell gaseous radioactivity, drywell unit cooler condensate flow rate and drywellpressure/temperature levels. The primary means of quantifying leakage in the drywellis the drywell sump monitoring system for UNIDENTIFIED LEAKAGE and the drywellequipment drain tank flow monitoring system for IDENTIFIED LEAKAGE.

IDENTIFIED leakageis not germane to this Tech Spec and the associated drywell equipment drain tank flowmonitoring system is not included.

.The drywell floor drain sump flow monitoring system monitors UNIDENTIFIED LEAKAGEcollected in the floor drain sump. UNIDENTIFIED LEAKAGE consists of leakage from RCPBcomponents inside the drywell which are not normally subject to leakage and otherwise routed to the drywell equipment drain sump. The primary containment floor drain sumphas transmitters that supply level indication to the main control room via the plantmonitoring system. The floor drain sump level transmitters are associated withHigh/Low level switches that open/close the sump tank drain valves automatically.

Thelevel instrument processing unit calculates an average leak rate (gpm) for a givenmeasurement period which resets whenever the sump drain valve closes. The levelprocessing unit provides an alarm to the main control room each time the average leakrate changes by a predetermined value since the last time the alarm was reset. For thedrywell floor drain sump flow monitoring system, the setpoint basis is a 1 gpm changein UNIDENTIFIED LEAKAGE.An alternate to the drywell floor drain sump flow monitoring system for quantifying UNIDENTIFIED LEAKAGE is the drywell equipment drain sump monitoring system, if thedrywell floor drain sump is overflowing to the drywell equipment drain sump. In thisconfiguration, the drywell equipment drain sump collects all leakage into the drywellequipment drain sump and the overflow from the drywell floor drain sump. Therefore, ifthe drywell floor drain sump is overflowing to the drywell equipment drain sump, thedrywel equipment drain sump monitoring system can be used to quantify UNIDENT FIEDLEAKAGE.

In this condition, all leakage measured by the drywell equipment drain sumpmonitoring system is assumed to be UNIDENTIFIED LEAKAGE.

The leakage determination

process, inc uding the transition to and use of the alternate method is described instation procedures.

The alternate method would only be used when the drywell floordrain sump flow monitoring system is unavailable.

In addition to the drywell sump monitoring system described above, the discharge ofeach sump is monitored by an independent flow element.

The measured flow rate from theflow element is integrated and recorded.

A main control room alarm is also provided toindicate an excessive sump discharge rate measured via the flow element.

This system,referred to as the "drywell floor drain flow totalizer",

is not credited for drywellfloor drain sump flow monitoring system operability.

LIMERICK

-UNIT 2 B 3/4 4-3 Amendment No. 4-2, 4-03,Associated with Amendment No. 4-6;-, 169 REACTOR COOLANT SYSTEMBASESBACKGROUND (Continued)

The primary containment atmospheric gaseous radioactivity monitoring systemcontinuously monitors the primary containment atmosphere for gaseous radioactivity levels. A sudden increase of radioactivity, which may be attributed to RCPB steam orreactor water leakage, is annunciated in the main control room.Condensate from the eight drywell air coolers is routed to the drywell floor drain sumpand is monitored by a series of flow transmitters that provide indication and alarms inthe main control room. The outputs from the flow transmitters are added together bysumming units to provide a total continuous condensate drain flow rate. The high flowalarm setpoint is based on condensate drain flow rate in excess of I gpm over thecurrently identified preset leak rate. The drywell air cooler condensate flow ratemonitoring system serves as an added indicator, but not quantifier, of RCS UNIDENTIFIED LEAKAGE (Ref. 4).The drywell temperature and pressure monitoring systems provide an indirect method fordetecting RCPB leakage.

A temperature and/or pressure rise in the drywell above normallevels may be indicative of a reactor coolant or steam leakage (Ref. 5).APPLICABLE SAFETY ANALYSESA threat of significant compromise to the RCPB exists if the barrier contains a crackthat is large enough to propagate rapidly.

Leakage rate limits are set low enough todetect the leakage emitted from a single crack in the RCPB (Refs. 6 and 7).A control room alarm allow the operators to evaluate the significance of the indicated leakage and, if necessary, shut down the reactor for further investigation andcorrective action. The allowed leakage rates are well below the rates predicted forcritical crack sizes (Ref. 7). Therefore, these actions provide adequate responsebefore a significant break in the RCPB can occur.RCS leakage detection instrumentation satisfies Criterion 1 of the NRC PolicyStatement.

LIMITING CONDITION FOR OPERATION (LCO)This LCO requires instruments of diverse monitoring principles to be OPERABLE toprovide confidence that small amounts of UNIDENTIFIED LEAKAGE are detected in time toallow actions to place the plant in a safe condition, when RCS leakage indicates possible RCPB degradation.

The LCO requires four instruments to be OPERABLE.

The required instrumentation to quantify UNIDENTIFIED LEAKAGE from the RCS consists ofeither the drywell floor drain sump flow monitoring system, or, the drywell equipment drain sump monitoring system with the drywell floor drain sump overflowing to thedrywell equipment drain sump. For either system to be considered

operable, the flowmonitoring portion of the system must be operable.

The identification of an increasein UNIDENTIFIED LEAKAGE will be delayed by the time required for the UNIDENTIFIED LEAKAGE to travel to the drywell floor drain sump and it may take longer than one hourto detect a 1 gpm increase in UNIDENTIFIED

LEAKAGE, depending on the origin andmagnitude of the leakage.

This sensitivity is acceptable for containment sump monitorOPERABILITY.

The reactor coolant contains radioactivity that, when released to the primarycontainment, can be detected by the gaseous primary containment atmospheric radioactivity monitor.

A radioactivity detection system is included for monitoring gaseous activities because of its sensitivity and rapid response to RCS leakage, but ithas recognized limitations.

Reactor coolant radioactivity levels will be low duringinitial reactor startup and for a few weeks thereafter, until activated corrosion products have been formed and fission products appear from fuel element claddingcontamination or cladding defects.

If there are few fuel element cladding defects andlow levels of activation

products, it may not be possible for the gaseous primaryLIMERICK

-UNIT 2 B 3/4 4-3a Amendment No. 1-03,Associated with Amendment No. 4--7, 169 REACTOR COOLANT SYSTEMBASESLIMITING CONDITION FOR OPERATION (LCO) (Continued) containment atmospheric radioactivity monitor to detect a 1 gpm increase within 1 hourduring normal operation.

However, the gaseous primary containment atmospheric radioactivity monitor is OPERABLE when it is capable of detecting a 1 gpm increase inUNIDENTIFIED LEAKAGE within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months given an RCS activity equivalent to that assumed inthe design calculations for the monitors (Reference 9).The LCO is satisfied when monitors of diverse measurement means are available.

Thus,the drywell floor drain sump monitoring system in combination with a gaseous primarycontainment atmospheric radioactivity

monitor, a primary containment air coolercondensate flow rate monitoring system, and a primary containment pressure andtemperature monitoring system provides an acceptable minimum.APPLICABILITY In OPERATIONAL CONDITIONS 1, 2, and 3, leakage detection systems are required to beOPERABLE to support LCO 3.4.3.2.

This applicability is consistent with that for LCO3.4.3.2.ACTIONSA. With the primary containment atmosphere gaseous monitoring system inoperable, grabsamples of the primary containment atmosphere must be taken and analyzed to provideperiodic leakage information.

[Provided a sample is obtained and analyzed once every12 hours, the plant may be operated for up to 30 days to allow restoration of theradioactivity monitoring system. The plant may continue operation since other formsof drywell leakage detection are available.]

The 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval provides periodic information that is adequate to detectleakage.

The 30 day Completion Time for Restoration recognizes other forms ofleakage detection are available.

B. With required drywell sump monitoring system inoperable, no other form of samplingcan provide the equivalent information to quantify leakage at the required 1gpm/hour sensitivity.

However, the primary containment atmospheric gaseousmonitor [and the primary containment air cooler condensate flow rate monitor]

willprovide indication of changes in leakage.With required drywell sump monitoring system inoperable, drywell condensate flowrate monitoring frequency increased from 12 to every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and UNIDENTIFIED LEAKAGE and total leakage being determined every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (Ref: SR 4.4.3.2.1.b) operation may continue for 30 days. To the extent practical, the surveillance frequency change associated with the drywell condensate flow rate monitoring system,makes up for the loss of the drywell floor drain sump monitoring system which had anormal surveillance requirement to monitor leakage every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Also note that inthis instance, the drywell floor drain tank flow totalizer will be used to complywith SR 4.4.3.2.1.b.

The 30 day Completion Time of the required ACTION isacceptable, based on operating experience, considering the multiple forms of leakagedetection that are still available.

C. With the required primary containment air cooler condensate flow rate monitoring system inoperable, SR 4.4.3.1.a must be performed every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to provide periodicinformation of activity in the primary containment at a more frequent interval thanthe routine frequency of every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months interval provides periodicinformation that is adequate to detect leakage and recognizes that other forms ofleakage detection are available.

The required ACTION has been clarified to stateLIMERICK

-UNIT 2 B 3/4 4-3b Amendment No. 4-G3, 4-32,Associated with Amendment No. 44-6, 169 REACTOR COOLANT SYSTFMBASESACTIONS (Continued) that the additional surveillance requirement is not applicable if the requiredprimary containment atmosphere gaseous radioactivity monitoring system is alsoinoperable.

Consistent with SR 4.0.3, surveillances are not required to beperformed on inoperable equipment.

In this case, ACTION Statement A. and E.requirements apply.D. With the primary containment pressure and temperature monitoring systeminoperable, operation may continue for up to 30 days given the system's indirectcapability to detect RCS leakage.

However, other more limiting Tech Specrequirements associated with the primary containment pressure/temperature monitoring system will still apply.E. With both the primary containment atmosphere gaseous radioactivity monitor and theprimary containment air cooler condensate flow rate monitor inoperable, the onlymeans of detecting leakage is the drywell floor drain sump monitor and the drywellpressure/temperature instrumentation.

This condition does not provide the requireddiverse means of leakage detection.

The required ACTION is to restore either of theinoperable monitors to OPERABLE status within 30 days to regain the intended leakagedetection diversity.

The 30 day Completion Time ensures that the plant will not beoperated in a degraded configuration for a lengthy time period. While the primarycontainment atmosphere gaseous radioactivity monitor is INOPERABLE, primarycontainment atmospheric grab samples will be taken and analyzed every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months sinceACTION Statement A. requirements also apply.F. With the drywell floor drain sump monitoring system inoperable and the drywell unitcoolers condensate flow rate monitoring system inoperable, one of the two remaining means of detecting leakage is the primary containment atmospheric gaseous radiation monitor.

The primary containment atmospheric gaseous radiation monitor typically cannot detect a 1 gpm leak within one hour when RCS activity is low. Indirectmethods of monitoring RCS leakage must be implemented.

Grab samples of the primarycontainment atmosphere must be taken and analyzed and monitoring of RCS leakage byadministrative means must be performed every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to provide alternate periodicinformation.

Administrative means of monitoring RCS leakage include monitoring and trendingparameters that may indicate an increase in RCS leakage.

There are diversealternative mechanisms from which appropriate indicators may be selected based onplant conditions.

It is not necessary to utilize all of these methods, but amethod or methods should be selected considering the current plant conditions andhistorical or expected sources of UNIDENTIFIED LEAKAGE.

The administrative methodsare the drywell cooling fan inlet/outlet temperatures, drywell equipment drain sumptemperature indicator, drywell equipment drain tank hi temperature indicator, anddrywell equipment drain tank flow indicator.

These indications, coupled with theatmospheric grab samples, are sufficient to alert the operating staff to anunexpected increase in UNIDENTIFIED LEAKAGE.In addition to the primary containment atmospheric gaseous radiation monitor andindirect methods of monitoring RCS leakage, the primary containment pressure andtemperature monitoring system is also available to alert the operating staff to anunexpected increase in UNIDENTIFIED LEAKAGE.LIMERICK

-UNIT 2 B 3/4 4-3c Amendment 4-03, 4-3-2, 444Associated with Amendment No. 167 REACTOR COOLANT SYSTEMBASES _ACTIONS (Continued)

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval is sufficient to detect increasing RCS leakage.

The RequiredAction provides 7 days to restore another RCS leakage monitor to OPERABLE statusto regain the intended leakage detection diversity.

The 7-day Completion Timeensures that the plant will not be operated in a degraded configuration for alengthy time period.G. If any required ACTION of Conditions A, B, C, D, E or F cannot be met within theassociated Completion Time, the plant must be brought to an OPERATIONAL CONDITION in which the LCO does not apply. To achieve this status, the plant must be broughtto at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and COLD SHUTDOWN within the next 24hours. The allowed Completion Times are reasonable, based on operating experience, to perform the ACTIONS in an orderly manner and without challenging plant systems.SURVEILLANCE REQUIREMENTS SR 4.4.3.1.a This SR is for the performance of a CHANNEL CHECK of the required primary containment atmospheric monitoring system. The check gives reasonable confidence that the channelis operating properly.

SR 4.4.3.1.b This SR is for the performance of a CHANNEL FUNCTIONAL TEST of the required RCS leakagedetection instrumentation.

The test ensures that the monitors can perform theirfunction in the desired manner. The test also verifies the alarm setpoint and relativeaccuracy of the instrument string.SR 4.4.3.1.c This SR is for the performance of a CHANNEL CALIBRATION of required leakage detection instrumentation channels.

The calibration verifies the accuracy of the instrument string, including the instruments located inside containment.

SR 4.4.3.1.d This SR provides a routine check of primary containment pressure and temperature forindirect evidence of RCS leakage.REFERENCES

1. LGS UFSAR, Section 5.2.5.1.2. Regulatory Guide 1.45, Revision 0, "Reactor Coolant Pressure Boundary LeakageDetection Systems,"

May 1973.3. LGS UFSAR, Section 5.2.5.2.1.3

4. LGS UFSAR, Section 5.2.5.2.1.4

5. LGS UFSAR, Section 5.2.5.2.1.1(2)

6. GEAP-5620, April 1968.7. NUREG-75/067, October 1975.8. LGS UFSAR, Section 5.2.5.6.9. LGS UFSAR, Section 5.2.5.2.1.5 LIMERICK

-UNIT 2 B 3/4 4-3d Amendment 4-44, 147Associated with Amendment No. 167 REACTOR COOIANT SYSTEMBASES3/4.4.3.2 OPERATIONAL LEAKAGEThe allowable leakage rates from the reactor coolant system have been based on thepredicted and experimentally observed behavior of cracks in pipes. The normallyexpected background leakage due to equipment design and the detection capability of theinstrumentation for determining system leakage was also considered.

The evidenceobtained from experiments suggests that for leakage somewhat greater than thatspecified for UNIDENTIFIED LEAKAGE the probability is small that the imperfection orcrack associated with such leakage would grow rapidly.

However, in all cases, if theleakage rates exceed the values specified or the leakage is located and known to bePRESSURE BOUNDARY

LEAKAGE, the reactor will be shutdown to allow further investigation and corrective action. The limit of 2 gpm increase in UNIDENTIFIED LEAKAGE over a 24-hour period and the monitoring of drywell floor drain sump and drywell equipment draintank flow rate at least once every eight (8) hours conforms with NRC staff positions specified in NRC Generic Letter 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping,"

as revised by NRC Safety Evaluation dated March 6, 1990. TheACTION requirement for the 2 gpm increase in UNIDENTIFIED LEAKAGE limit ensures thatsuch leakage is identified or a plant shutdown is initiated to allow furtherinvestigation and corrective action. Once identified, reactor operation may continuedependent upon the impact on total leakage.The function of Reactor Coolant System Pressure Isolation Valves (PIVs) is toseparate the high pressure Reactor Coolant System from an attached low pressure system.The ACTION requirements for pressure isolation valves are used in conjunction with thesystem specifications for which PIVs are listed in The Technical Requirements Manualand with primary containment isolation valve requirements to ensure that plantoperation is appropriately limited.The Surveillance Requirements for the RCS pressure isolation valves provide addedassurance of valve integrity thereby reducing the probability of gross valve failureand consequent intersystem LOCA. Leakage from the RCS pressure isolation valves isnot included in any other allowable operational leakage specified in Section 3.4.3.2.3/4.4.4 (Deleted)

INFORMATION FROM THIS SECTION RELOCATED TO THE TRMLIMERICK

-UNIT 2B 3/4 4-3eAmendment No. 4-0-3, 4-34, 4-3-3, 144Associated with Amendment No. 167 REACTOR COOLANT SYSTEMBASES3/4.4.5 SPECIFIC ACTIVITYThe limitations on the specific activity of the primary coolant ensurethat the 2-hour thyroid and whole body doses resulting from a main steam linefailure outside the containment during steady state operation will not exceedsmall fractions of the dose guidelines of 10 CFR Part 100. The values for thelimits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations.

These values are conservative in that specific site parameters, such as SITE BOUNDARY location and meteoro-logical conditions, were not considered in this evaluation.

The ACTION statement permitting POWER OPERATION to continue for limitedtime periods with the primary coolant's specific activity greater than 0.2microcurie per gram DOSE EQUIVALENT 1-131, but less than or equal to 4 micro-curies per gram DOSE EQUIVALENT 1-131, accommodates possible iodine spikingphenomenon which may occur following changes in THERMAL POWER. This action ismodified by a Note that permits the use of the provisions of Specification 3.0.4.c.

This allowance permits entry into the applicable OPERATIONAL CONDITION (S) while relying on the ACTION requirements.

Operation with specific activitylevels exceeding 0.2 microcurie per gram DOSE EQUIVALENT 1-131 but less than orequal to 4 microcuries per gram DOSE EQUIVALENT 1-131 must be restricted sincethese activity levels increase the 2-hour thyroid dose at the SITE BOUNDARYfollowing a postulated steam line rupture.Closing the main steam line isolation valves prevents the release ofactivity to the environs should a steam line rupture occur outside containment.

The surveillance requirements provide adequate assurance that excessive specificactivity levels in the reactor coolant will be detected in sufficient time totake corrective action.3/4.4.6 PRESSURE/TEMPERATURE LIMITSAll components in the reactor coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.These cyclic loads are introduced by normal load transients, reactor trips,and startup and shutdown operations.

The various categories of load cyclesused for design purposes are provided in Section 3.9 of the FSAR. Duringstartup and shutdown, the rates of temperature and pressure changes are limitedso that the maximum specified heatup and cooldown rates are consistent withthe design assumptions and satisfy the stress limits for cyclic operation.

LIMERICK

-UNIT 2B 3/4 4-4Amendment No. 132 REACTOR COOLANT SYSTFMBASESPRESSURE/TEMPERATURE LIMITS (Continued)

The operating limit curves of Figure 3.4.6.1-1 are derived from the fracturetoughness requirements of 10 CFR 50 Appendix G and ASME Code Section XI,Appendix G. The curves are based on the RTNDT and stress intensity factorinformation for the reactor vessel components.

Fracture toughness limits andthe basis for compliance are more fully discussed in FSAR Chapter 5, Para-graph 5.3.1.5, "Fracture Toughness."

The reactor vessel materials have been tested to determine their initialRTNDT. The results of these tests are shown in Table B 3/4.4.6-1.

Reactoroperation and resultant fast neutron, E greater than 1 MeV, irradiation willcause an increase in the RTJDT. Therefore, an adjusted reference temperature, based upon the fluence, nickel content and copper content of the materialin question, can be predicted using Bases Figure B 3/4.4.6-1 and the recommenda-tions of Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of ReactorVessel Materials."

The pressure/temperature limit curve, Figure 3.4.6.1-1, curves A, B and C, includes an assumed shift in RTNDT for the conditions at32 EFPY. In addition, an intermediate A curve has been provided for 22 EFPY.The A, B and C limit curves are predicted to be bounding for all areasof the RPV until 32 EFPY.The pressure-temperature limit lines shown in Figures 3.4.6.1-1, curves C,and A, for reactor criticality and for inservice leak and hydrostatic testing havebeen provided to assure compliance with the minimum temperature requirements ofAppendix G to 10 CFR Part 50 for reactor criticality and for inservice leak andhydrostatic testing.LIMERICK

-UNIT 2B 3/4 4-5Amendment No. , 80, 4-4-1, 130 REACTOR COOLANT SYSTEMBASES3/4.4.7 MAIN STEAM LINE ISOLATION VALVESDouble isolation valves are provided on each of the main steam lines tominimize the potential leakage paths from the containment in case of a line break.Only one valve in each line is required to maintain the integrity of thecontainment,

however, single failure considerations require that two valves beOPERABLE.

The surveillance requirements are based on the operating history ofthis type valve. The maximum closure time has been selected to contain fissionproducts and to ensure the core is not uncovered following line breaks. Theminimum closure time is consistent with the assumptions in the safety analyses toprevent pressure surges.3/4.4.8 (DELETED) 3/4.4.9 RESIDUAL HEAT REMOVALThe RHR system is required to remove decay heat and sensible heat in order tomaintain the temperature of the reactor coolant.

RHR shutdown cooling is comprised offour (4) subsystems which make two (2) loops. Each loop consists of two (2) motordriven pumps, a heat exchanger, and associated piping and valves. Both loops have acommon suction from the same recirculation loop. Two (2) redundant, manuallycontrolled shutdown cooling subsystems of the RHR System can provide the required decayheat removal capability.

Each pump discharges the reactor coolant, after it has beencooled by circulation through the respective heat exchangers, to the reactor via theassociated recirculation loop or to the reactor via the low pressure coolant injection pathway.

The RHR heat exchangers transfer heat to the RHR Service Water System. TheRHR shutdown cooling mode is manually controlled.

An OPERABLE RHR shutdown cooling subsystem consists of an RHR pump, a heatexchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path.In HOT SHUTDOWN condition, the requirement to maintain OPERABLE two (2) independent RHRshutdown cooling subsystems means that each subsystem considered OPERABLE must beassociated with a different heat exhanger loop, i.e., the "A" RHR heat exchanger withthe "A" RHR pump or the "C" RHR pump, and the "B" RHR heat exchanger with the "B" RHRpump or the "D" RHR pump are two (2) independent RHR shutdown cooling subsystems.

Onlyone (1) of the two (2) RHR pumps associated with each RHR heat exchanger loop isLIMERICK

-UNIT 2 B 3/4 4-6 Amendment No. 4,64,8,8,4

,Associated with Amendment 160 3/4.4.9 RESIDUAL HEAT REMOVAL (Cont'd)required to be OPERABLE for that independent subsystem to be OPERABLE.

During COLDSHUTDOWN and REFUELING conditions,

however, the subsystems not only have a commonsuction source, but are allowed to have a common heat exchanger and common discharge piping. To meet the LCO of two (2) OPERABLE subsystems, both pumps in one (1) loop orone (1) pump in each of the two (2) loops must be OPERABLE.

Since the piping and heatexchangers are passive components, that are assumed not to fail, they are allowed tobe common to both subsystems.

Additionally, each RHR shutdown cooling subsystem isconsidered OPERABLE if it can be manually aligned (remote or local) in the shutdowncooling mode for removal of decay heat. Operation (either continuous or intermittent) of one (1) subsystem can maintain and reduce the reactor coolant temperature asrequired.

However, to ensure adequate core flow to allow for accurate average reactorcoolant temperature monitoring, nearly continuous operation is required.

Alternate decay heat removal methods are available to operators.

These alternate methods of decay heat removal can be verified available either by calculation (whichincludes a review of component and system availability to verify that an alternate decay heat removal method is available) or by demonstration, and that a method ofcoolant mixing be operational.

Decay heat removal capability by ambient losses can beconsidered in evaluating alternate decay heat removal capability.

LIMERICK

-UNIT 2B 3/4 4-6aAmendment No. 82 BASES TABLE B 3/4.4.6-1 REACTOR VESSEL TOUGHNESS*

LIMITINGBELTLINECOMPONENT PlateWeldWELD SEAM I.D.OR MAT'L TYPESA-533 Gr. B,CL.AB (Field Weld)HEAT/SLAB ORHEAT/LOTB 3416-1640892/J424B27AE STARTINGCU (%) Ni (%) RTNDT (OF) ARTNDT **(OF).14 .65 +40 +48.09 1.0 -60 +58MIN.UPPER SHELF(LFT-LBS)

ART (OF)NA +122NA +541NOTES:

Based on 110% of** These values areNON-BELTLINE COMPONENT Top Shell RingBottom Head DomeBottom Head TorusTop Head TorusTop Head FlangeVessel FlangeFeedwater NozzleWeldLPCI Nozzle***

Closure Studsoriginal power.given only for the benefitMT'L TYPE ORWELD SEAM I.D.SA 533, Gr. B, CL.1of calculating the end-of-life (EOL/32 EFPY) RTNDTHEAT/SLAB OR HIGHEST STARTINGHEAT/LOT RTNDT ('F)C9800-2 -16C9245-2 +22C9362-2 +28C9646-2 -20123B300 +10SA-508, CL. 2Non-Beltline SA-508, CL. 2SA-540, Gr. B-242L2058Q2Q29WAl 1Q2Q33WAl 1+100-12-4Meet requirements of 45 ft-lbsand 25 mils Lat. Exp. at +10°FThe design of the LPCI nozzles results in their experiencing 10'1 N/Cm2which predicts an EOL (32 EFPY) RTINT of +35°F.an EOL fluence in excess ofLIMERICK

-UNIT 2B 3/4 4-7Amendment No. -54, 111 INTENTIONALLY LEFT BLANK 1.2Co 1.00.8CD0.4.-0.20.010 20 30 40Service Life (Years*)BASES FIGURE B 3/4.4.6-1 FAST NEUTRON FLUENCE (E> 1 MeV) AT 1/4 T AS A FUNCTIONOF SERVICE LIFE** At 90% of Rated Thermal Power and 90% availability

.LIMERICK

-UNIT 2 B 3/4 4-8Amendment No. 51 0PAGE INTENTIONALLY LEFT BLANK 3/4.5 EMERGENCY CORE COOLING SYSTEMBASES3/4.5.1 and 3/4.5.2 ECCS -OPERATING and SHUTDOWNThe core spray system (CSS), together with the LPCI mode of the RHR system,is provided to assure that the core is adequately cooled following a loss-of-coolant accident and provides adequate core cooling capacity for all breaksizes up to and including the double-ended reactor recirculation line break,and for smaller breaks following depressurization by the ADS.The CSS is a primary source of emergency core cooling after the reactorvessel is depressurized and a source for flooding of the core in case ofaccidental draining.

The surveillance requirements provide adequate assurance that the CSS willbe OPERABLE when required.

Although all active components are testable andfull flow can be demonstrated by recirculation through a test loop duringreactor operation, a complete functional test requires reactor shutdown.

Thepump discharge piping is maintained full to prevent water hammer damage topiping and to start cooling at the earliest moment.The low pressure coolant injection (LPCI) mode of the RHR system isprovided to assure that the core is adequately cooled following a loss-of-coolant accident.

Four subsystems, each with one pump, provide adequate coreflooding for all break sizes up to and including the double-ended reactorrecirculation line break, and for small breaks following depressurization bythe ADS.The surveillance requirements provide adequate assurance that the LPCIsystem will be OPERABLE when required.

Although all active components aretestable and full flow can be demonstrated by recirculation through a testloop during reactor operation, a complete functional test requires reactorshutdown.

The pump discharge piping is maintained full to prevent waterhammer damage to piping and to start cooling at the earliest moment.The high pressure coolant injection (HPCI) system is provided to assurethat the reactor core is adequately cooled to limit fuel clad temperature inthe event of a small break in the reactor coolant system and loss of coolantwhich does not result in rapid depressurization of the reactor vessel. TiheHPCI system permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized.

TheHCPI system continues to operate until reactor vessel pressure is below thepressure at which CSS operation or LPCI mode of the RHR system operation maintains core cooling.The capacity of the system is selected to provide the required core cooling.The HPCI pump is designed to deliver greater than or equal to 5600 gpm at reactorpressures between 1182 and 200 psig and is capable of delivering at least 5000 gpmbetween 1182 and 1205 psig. In the system's normal alignment, water from thecondensate storage tank is used instead of injecting water from the suppression pool into the reactor, but no credit is taken in the safety analyses forthe condensate storage tank water.LIMERICK

-UNIT 2 B 3/4 5-1 Amendment No. -.ECR 00-00177 EMERGENCY CORE COOLING SYSTEMBASESECCS -OPERATING and SHUTDOWN (Continued)

With the HPCI system inoperable, adequate core cooling is assured by theOPERABILITY of the redundant and diversified automatic depressurization systemand both the CS and LPCI systems.

In addition, the reactor core isolation cooling (RCIC) system, a system for which no credit is taken in the safetyanalysis, will automatically provide makeup at reactor operating pressures ona reactor low water level condition.

The HPCI out-of-service period of 14 daysis based on the demonstrated OPERABILITY of redundant and diversified lowpressure core cooling systems and the RCIC system. The HPCI system, and one LPCIsubsystem, and/or one CSS subsystem out-of-service period of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ensures thatsufficient ECCS, comprised of a minimum of one CSS subsystem, three LPCI subsystems, and all of the ADS will be available to 1) provide for safe shutdown of the facility, and 2) mitigate and control accident conditions within the facility.

A Note prohibits the application of Specification 3.0.4.b to an inoperable HPCI subsystem.

There is anincreased risk associated with entering an OPERATIONAL CONDITION or other specified condition in the Applicability with an inoperable HPCI subsystem and the provisions ofSpecification 3.0.4.b, which allow entry into an OPERATIONAL CONDITION or otherspecified condition in the Applicability with the Limiting Condition for Operation notmet after performance of a risk assessment addressing inoperable systems andcomponents, should not be applied in this circumstance.

The surveillance requirements provide adequate assurance that the HPCIsystem will be OPERABLE when required.

Although all active components aretestable and full flow can be demonstrated by recirculation through a test loopduring reactor operation, a complete functional test with reactor vesselinjection requires reactor shutdown.

The pump discharge piping is maintained full to prevent water hammer damage and to provide cooling at the earliestmoment.Upon failure of the HPCI system to function properly after a small breakloss-of-coolant

accident, the automatic depressurization system (ADS) automa-tically causes selected safety/relief valves to open, depressurizing the reactorso that flow from the low pressure core cooling systems can enter the core intime to limit fuel cladding temperature to less than 2200'F. ADS is conserva-tively required to be OPERABLE whenever reactor vessel pressure exceeds 100 psig.This pressure is substantially below that for which the low pressure core cool-ing systems can provide adequate core cooling for events requiring ADS.ADS automatically controls five selected safety-relief valves. The safetyanalysis assumes all five are operable.

The allowed out-of-service time for onevalve for up to fourteen days is determined in a similar manner to other ECCSsub-system out-of-service time allowances.

Verification that ADS accumulator gas supply header pressure is 90psig ensures adequate gas pressure for reliable ADS operation.

The accumulator on each ADS valve provides pneumatic pressure for valve actuation.

The designpneumatic supply pressure requirements for the accumulator are such that,following a failure of the pneumatic supply to the accumulator at least twovalve actuations can occur with the drywell at 70% of design pressure.

TheECCS safety analysis assumes only one actuation to achieve thedepressurization required for operation of the low pressure ECCS. This minimumrequired pressure of 90 psig is provided by the PCIG supply.LIMERICK

-UNIT 2 B 3/4 5-2 Amendment No. 4-1-6,-1-32, 147 EMERGENCY CORE COOLING SYSTEMBASESECCS -OPERATING and SHUTDOWN (Continued) 3/4.5.3 SUPPRESSION CHAMBERThe suppression chamber is required to be OPERABLE as part of the ECCS toensure that a sufficient supply of water is available to the HPCI, CS andLPCI systems in the event of a LOCA. This limit on suppression chamber minimumwater volume ensures that sufficient water is available to permit recirculation cooling flow to the core. The OPERABILITY of the suppression chamber inOPERATIONAL CONDITION 1, 2, or 3 is also required by Specification 3.6.2.1.Repair work might require making the suppression chamber inoperable.

Thisspecification will permit those repairs to be made and at the same time giveassurance that the irradiated fuel has an adequate cooling water supply whenthe suppression chamber must be made inoperable, including

draining, inOPERATIONAL CONDITION 4 or 5.In OPERATIONAL CONDITION 4 and 5 the suppression chamber minimum requiredwater volume is reduced because the reactor coolant is maintained at or below200'F. Since pressure suppression is not required below 212'F, the minimumwater volume is based on NPSH, recirculation volume and vortex prevention plusa safety margin for conservatism.

LIMERICK

-UNIT 2 B 3/4 5-2a Amendment No. 116 INTENTIONALLY LEFT BLANK 3/4.6 CONTAINMENT SYSTEMSBASES3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY ensures that the release of radioactive mate-rials from the containment atmosphere will be restricted to those leakage pathsand associated leak rates assumed in the safety analyses.

This restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARYradiation doses to within the limits of 10 CFR Part 100 during accident conditions.

3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGEThe limitations on primary containment leakage rates ensure that the totalcontainment leakage volume will not exceed the value calculated in the safetyanalyses at the design basis LOCA maximum peak containment pressure of 44 psig, Pa. Asan added conservatism, the measured overall integrated leakage rate (Type A Test) isfurther limited to less than or equal to 0.75 La during performance of the periodictests to account for possible degradation of the containment leakage barriers betweenleakage tests.Operating experience with the main steam line isolation valves hasindicated that degradation has occasionally occurred in the leak tightness ofthe valves; therefore the special requirement for testing these valves.The surveillance testing for measuring leakage rates is consistent withthe Primary Containment Leakage Rate Testing Program.3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCKThe limitations on closure and leak rate for the primary containment airlock are required to meet the restrictions on PRIMARY CONTAINMENT INTEGRITY and the Primary Containment Leakage Rate Testing Program.

Only one closed door inthe air lock is required to maintain the integrity of the containment.

3/4.6.1.4 MSIV LEAKAGE ALTERNATE DRAIN PATHWAYCalculated doses resulting from the maximum leakage allowances for themain steamline isolation valves in the postulated LOCA situations will notexceed the criteria of 10 CFR Part 100 guidelines, provided the main steam linesystem from the isolation valves up to and including the turbine condenser remainsintact. Operating experience has indicated that degradation has occasionally occurred in the leak tightness of the MSIVs such that the specified leakagerequirements have not always been continuously maintained.

The requirement forthe MSIV Leakage Alternate Drain Pathway serves to reduce the offsite dose.LIMERICK

-UNIT 2 B 3/4 6-1 Amendment No. , 3,ECR 11-00395 CONTAINMENT SYSTEMSBASES3/4.6.1.5 PRIMARY CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life ofthe unit. Structural integrity is required to ensure that the containment willwithstand the maximum calculated pressure in the event of a LOCA. A visualinspection in accordance with the Primary Containment Leakage Rate TestingProgram is sufficient to demonstrate this capability.

3/4.6.1.6 DRYWELL AND SUPPRESSION CHAMBER INTERNAL PRESSUREThe limitations on drywell and suppression chamber internal pressure ensurethat the calculated containment peak pressure does not exceed the designpressure of 55 psig during LOCA conditions or that the external pressure differ-ential does not exceed the design maximum external pressure differential of5.0 psid. The limit of -1.0 to + 2.0 psig for initial containment pressurewill limit the total pressure to < 44 psig which is less than the designpressure and is consistent with the safety analysis.

3/4.6.1.7 DRYWELL AVERAGE AIR TEMPERATURE The limitation on drywell average air temperature ensures that the con-tainment peak air temperature does not exceed the design temperature of 340°Fduring steam line break conditions and is consistent with the safety analysis.

3/4.6.1.8 DRYWELL AND SUPPRESSION CHAMBER PURGE SYSTEMThe drywell and suppression chamber purge supply and exhaust isolation valves are required to be closed during plant operation except as required forinerting, deinerting, pressure

control, ALARA or air quality considerations forpersonnel entry, or Surveillances that require the valves to be open. Limitingthe use of the drywell and suppression chamber purge system to specific criteriais imposed to protect the integrity of the SGTS filters.

Analysis indicates that should a LOCA occur while this pathway is being utilized, the associated pressure surge through the (18 or 24") purge lines will adversely affect theintegrity of SGTS. This condition is not imposed on the 1 and 2 inch valves usedfor pressure control since a surge through these lines does not threaten theoperability of SGTS.Surveillance requirement 4.6.1.8 ensures that the primary containment purgevalves are closed as required or, if open, open for an allowable reason. Ifa purge valve is open in violation of this SR, the valve is considered inoperable.

The SR is modified by a Note stating that primary containment purge valvesare only required to be closed in OPERATIONAL CONDITIONS 1, 2 and 3. The SRis also modified by a Note stating that the SR is not required to be met whenthe purge valves are open for the stated reasons.

The Note states that thesevalves may be opened for inerting, deinerting, pressure

control, ALARA or airquality considerations for personnel entry, or Surveillances that require thevalves to be open. The 18 or 24 inch purge valves are capable of closing inthe environment following a LOCA. Therefore, these valves are allowed to beopen for limited periods of time.LIMERICK

-UNIT 2B 3/4 6-2Amendment No. ,#-4,8,9-,

147 CONTAINMENT SYSTEMSBASES3/4.6.2 DEPRESSURIZATION SYSTEMSThe specifications of this section ensure that the primary containment pressure will not exceed the design pressure of 55 psig during primary systemblowdown from full operating pressure.

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

Since all of the gases in the drywell are purged into the suppression chamber airspace during a loss-of-coolant

accident, the pressure of the suppression chamberair space must not exceed 55 psig. The design volume of the suppression chamber,water and air, was obtained by considering that the total volume of reactorcoolant is discharged to the suppression chamber and that the drywell volume ispurged to the suppression chamber.Using the minimum or maximum water volumes given in this specification, suppression pool pressure during the design basis accident is below the designpressure.

Maximum water volume of 134,600 ft3 results in a downcomer submergence of 12'3" and the minimum volume of 122,120 ft3 results in a submergence approximately 2'3" less. The majority of the Bodega tests were run with a submerged length of 4feet and with complete condensation.

Thus, with respect to the downcomer submergence, this specification is adequate.

The maximum temperature at the end of theblowdown tested during the Humboldt Bay and Bodega Bay tests was 170°F and thisis conservatively taken to be the limit for complete condensation of the reactorcoolant, although condensation would occur for temperature above 1700F.Should it be necessary to make the suppression chamber inoperable, this shallonly be done as specified in Specification 3.5.3.Under full power operating conditions, blowdown through safety/relief valvesassuming an initial suppression chamber water temperature of 95°F results in abulk water temperature of approximately 140°F immediately following blowdownwhich is below the 190°F bulk temperature limit used for complete condensation via T-quencher devices.

At this temperature and atmospheric

pressure, theavailable NPSH exceeds that required by both the RHR and core spray pumps, thusthere is no dependency on containment overpressure during the accident injection phase. If both RHR loops are used for containment

cooling, there is no dependency on containment overpressure for post-LOCA operations.

LIMERICK

-UNIT 2B 3/4 6-3Amendment No. --3, 48, 51 3/4.6.2 DEPRESSURIZATION SYSTEMS (Cont.)One of the surveillance requirements for the suppression pool cooling (SPC)mode of the RHR system is to demonstrate that each RHR pump develops a flow rate110,000 gpm while operating in the SPC mode with flow through the heatexchanger and its associated closed bypass valve, ensuring that pump performance has not degraded during the cycle and that the flow path is operable.

This testconfirms one point on the pump design curve and is indicative of overallperformance.

Such inservice inspections confirm component operability, trendperformance and detect incipient failures by indicating abnormal performance.

TheRHR heat exchanger bypass valve is used for adjusting flow through the heatexchanger, and is not designed to be a tight shut-off valve. With the bypassvalve closed, a portion of the total flow still travels through the bypass, whichcan affect overall heat transfer.

However, no heat transfer performance requirement of the heat exchanger is intended by the current Technical Specification surveillance requirement.

This is confirmed by the lack of any flowrequirement for the RHRSW system in Technical Specifications Section 3/4.7.1.Verifying an RHR flowrate through the heat exchanger does not demonstrate heatremoval capability in the absence of a requirement for RHRSW flow. LGS doesperform heat transfer testing of the RHR heat exchangers as part of its responseto Generic Letter 89-13, which verified the commitment to meet the requirements ofGDC 46.Experimental data indicate that excessive steam condensing loads can beavoided if the peak local temperature of the suppression pool is maintained below200'F during any period of relief valve operation for T-quencher devices.Specifications have been placed on the envelope of reactor operating conditions sothat the reactor can be depressurized in a timely manner to avoid the regime ofpotentially high suppression chamber loadings.

Because of the large volume and thermal capacity of the suppression pool,the volume and temperature normally changes very slowly and monitoring theseparameters daily is sufficient to establish any temperature trends. By requiring the suppression pool temperature to be frequently recorded during periods ofsignificant heat addition, the temperature trends will be closely followed sothat appropriate action can be taken.In addition to the limits on temperature of the suppression chamber poolwater, operating procedures define the action to be taken in the event a safety-relief valve inadvertently opens or sticks open. As a minimum this action shallinclude:

(1) use of all available means to close the valve, (2) initiate suppres-sion pool water cooling, (3) initiate reactor shutdown, and (4) if other safety-relief valves are used to depressurize the reactor, their discharge shall beseparated from that of the stuck-open safety/relief valve to assure mixing anduniformity of energy insertion to the pool.During a LOCA, potential leak paths between the drywell and suppression chamberairspace could result in excessive containment pressures, since the steam flow intothe airspace would bypass the heat sink capabilities of the chamber.

Potential sourcesof bypass leakage are the suppression chamber-to-drywell vacuum breakers (VBs),penetrations in the diaphragm floor, and cracks in the diaphragm floor and/or liner plate anddowncomers located in the suppression chamber airspace.

The containment pressure responseto the postulated bypass leakage can be mitigated by manually actuating the suppression chamber sprays. An analysis was performed for a design bypass leakage area of A/, equalto 0.0500 ft2 to verify that the operator has sufficient time to initiate the sprays priorto exceeding the containment design pressure of 55 psig. The limit of 10% of the designvalue of 0.0500 ft2ensures that the design basis for the steam bypass analysis is metLIMERICK

-UNIT 2B 3/4 6-3aAmendment No. 2-3, 31 CONTAINMENT SYSTEMSBASESDEPRESSURIZATION SYSTEMS (Continued)

The drywell-to-suppression chamber bypass test at a differential pressure ofat least 4.0 psi verifies the overall bypass leakage area for simulated LOCAconditions is less than the specified limit. For those outages where thedrywell-to-suppression chamber bypass leakage test in not conducted, the VB leakagetest verifies that the VB leakage area is less than the bypass limit, with a76% margin to the bypass limit to accommodate the remaining potential leakage areathrough the passive structural components.

Previous drywell-to-suppression chamberbypass test data indicates that the bypass leakage through the passive structural components will be much less than the 76% margin. The VB leakage limit, combinedwith the negligible passive structural leakage area, ensures that the drywell-to-suppression chamber bypass leakage limit is met for those outages for which thedrywell-to-suppression chamber bypass test is not scheduled.

3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVESThe OPERABILITY of the primary containment isolation valves ensures thatthe containment atmosphere will be isolated from the outside environment inthe event of a release of radioactive material to the containment atmosphere or pressurization of the containment and is consistent with the requirements of GDC 54 through 57 of Appendix A of 10 CFR Part 50. Containment isolation within the time limits specified for those isolation valves designed to closeautomatically ensures that the release of radioactive material to the environ-ment will be consistent with the assumptions used in the analyses for a LOCA.The scram discharge volume vent and drain valves serve a dual function, oneof which is primary containment isolation.

Since the other safety functions of thescram discharge volume vent and drain valves would not be available if the normalPCIV actions were taken, actions are provided to direct the user to the scramdischarge volume vent and drain operability requirements contained in Specification 3.1.3.1.

However, since the scram discharge volume vent and drain valves are PCIVs,the Surveillance Requirements of Specification 4.6.3 still apply to these valves.The opening of a containment isolation valve that was locked or sealed closedto satisfy Technical Specification 3.6.3 Action statements, may be reopened on anintermittent basis under administrative controls.

These controls consist ofstationing a dedicated individual at the controls of the valve, who is in continuous communication with the control room. In this way, the penetration can be rapidlyisolated when a need for primary containment isolation is indicated.

Primary containment isolation valves governed by this Technical Specification are identified in Table 3.6.3-1 of the TRM.This Surveillance Requirement requires a demonstration that a representative sample of reactor instrument line excess flow check valves (EFCVs) is OPERABLE byverifying that the valve actuates to the isolation position on a simulated instrument line break signal. The representative sample consists of an approximately equalnumber of EFCVs, such that each EFCV is tested in accordance with the Surveillance Frequency Control Program.

In addition, the EFCVs in the sample are representative olthe various plant configurations, models, sizes, and operating environments.

Thisensures that any potentially common problem with a specific type or application ofEFCV is detected at the earliest possible time. This Surveillance Requirement providesassurance that the instrumentation line EFCVs will perform so that predicted radiological consequences will not be exceeded during a postulated instrument linebreak event. Furthermore, any EFCV failures will be evaluated to determine ifadditional testing in the test interval is warranted to ensure overall reliability ismaintained.

Operating experience has demonstrated that these components are highlyreliable and that failures to isolate are very infrequent.

Therefore, testing of arepresentative sample was concluded to be acceptable from a reliability standpoint.

For some EFCVs, this Surveillance can be performed with the reactor at power.LIMERICK

-UNIT 2B 3/4 6-4 Amendment No.

147 CONTAINMENT SYSTEMSBASES3/4.6.4 VACUUM RELIEFVacuum relief valves are provided to equalize the pressure between the 0suppression chamber and drywell.

This system will maintain the structural integrity of the primary containment under conditions of large differential pressures.

The vacuum breakers between the suppression chamber and the drywell mustnot be inoperable in the open position since this would allow bypassing of thesuppression pool in case of an accident.

Two pairs of valves are required toprotect containment structural integrity.

There are four pairs of valves(three to provide minimum redundancy) so that operation may continue for up to72 hours with no more than two pairs of vacuum breakers inoperable in the closedposition.

Each vacuum breaker valve's position indication system is of great enoughsensitivity to ensure that the maximum steam bypass leakage coefficient ofA,rk = 0.05 ft2for the vacuum relief system (assuming one valve fully open) will not be exceeded.

LIMERICK

-UNIT 2B 3/4 6-4aAmendment No. 1101 CONTAINMENT SYSTEMSBASES3/4.6.5 SECONDARY CONTAINMENT Secondary containment is designed to minimize any ground level release ofradioactive material which may result from an accident.

The Reactor Enclosure and associated structures provide secondary containment during normal operation when the drywell is sealed and in service.

At other times the drywell may beopen and, when required, secondary containment integrity is specified.

Establishing and maintaining a vacuum in the reactor enclosure secondary containment with the standby gas treatment system in accordance with the Surveillance Frequency Control Program, along with the surveillance of the doors, hatches, dampersand valves, is adequate to ensure that there are no violations of the integrity ofthe secondary containment.

The OPERABILITY of the reactor enclosure recirculation system and the standbygas treatment systems ensures that sufficient iodine removal capability willbe available in the event of a LOCA. The reduction in containment iodine inventory reduces the resulting SITE BOUNDARY and Control Room radiation doses associated withcontainment leakage.

The operation of these systems and resultant iodine removalcapacity are consistent with the assumptions used in the LOCA analysis.

Provisions have been made to continuously purge the filter plenums with instrument air when thefilters are not in use to prevent buildup of moisture on the adsorbers and the HEPAfilters.As a result of the Alternative Source Term (AST) project, secondary containment integrity of the refueling area is not required during certainconditions when handling irradiated fuel or during CORE ALTERATIONS and alignment of the Standby Gas Treatment System to the refueling area is not required.

Thecontrol room dose analysis for the Fuel Handling Accident (FHA) is based onunfiltered releases from the South Stack and therefore, does not require theStandby Gas Treatment System to be aligned to the refueling area.However, when handling RECENTLY IRRADIATED FUEL or during operations with apotential for draining the reactor vessel with the vessel head removed and fuel inthe vessel, secondary containment integrity of the refueling area is required andalignment of the Standby Gas Treatment System to the refueling area is required.

The AST fuel handling analysis does not include an accident involving RECENTLYIRRADIATED FUEL or an accident involving draining the reactor vessel.The Standby Gas Treatment System is required to be OPERABLE when handlingirradiated fuel, handling RECENTLY IRRADIATED FUEL, during CORE ALTERATIONS andduring operations with a potential to drain the vessel with the vessel headremoved and fuel in the vessel. Fuel Handling Accident releases from the NorthStack must be filtered through the Standby Gas Treatment System to maintaincontrol room doses within regulatory limits. The OPERABILITY of the Standby GasTreatment System assures that releases, if made through the North Stack, arefiltered prior to release.LIMERICK

-UNIT 2 B 3/4 6-5 Amendment No. -4,5-4,,86,4-46,4-4-7-,

ECR LG 09-00052 CONTAINMENT SYSTEMSBASESSECONDARY CONTAINMENT (Continued)

Although the safety analyses assumes that the reactor enclosure secondary containment draw down time will take 930 seconds, these surveillance require-ments specify a draw down time of 916 seconds.

This 14 second difference isdue to the diesel generator starting and sequence loading delays which is notpart of this surveillance requirement.

The reactor enclosure secondary containment draw down time analyses assumesa starting point of 0.25 inch of vacuum water gauge and worst case SGTS dirtyfilter flow rate of 2800 cfm. The surveillance requirements satisfy this as-sumption by starting the drawdown from ambient conditions and connecting theadjacent reactor enclosure and refueling area to the SGTS to split the exhaustflow between the three zones and verifying a minimum flow rate of 2800 cfm fromthe test zone. This simulates the worst case flow alignment and verifies ade-quate flow is available to drawdown the test zone within the required time.The Technical Specification Surveillance Requirement 4.6.5.3.b.3 is intendedto be a multi-zone air balance verification without isolating any test zone.The SGTS is common to Unit 1 and 2 and consists of two independent subsystems.

The power supplies for the common portions of the subsystems arefrom Unit 1 safeguard busses, therefore the inoperability of these Unit 1supplies are addressed in the SGTS ACTION statements in order to ensure adequateonsite power sources to SGTS for its Unit 2 function during a loss of offsitepower event. The allowable out of service times are consistent with those inthe Unit 1 Technical Specifications for SGTS and AC electrical power supply outof service condition combinations.

LIMERICK

-UNIT 2B 3/4 6-5aAmendment No. -4,-4,86,-46,,-14-7, ECR LG 09-00052 CONTAINMENT SYSTEMSBASESSECONDARY CONTAINMENT (Continued)

The SGTS fans are sized for three zones and therefore, when aligned to asingle zone or two zones, will have excess capacity to more quickly drawdownthe affected zones. There is no maximum flow limit to individual zones orpairs of zones and the air balance and drawdown time are verified when allthree zones are connected to the SGTS.The three zone air balance verification and drawdown test will be doneafter any major system alteration, which is any modification which will havean effect on the SGTS flowrate such that the ability of the SGTS to drawdownthe reactor enclosure to greater than or equal to 0.25 inch of vacuum watergage in less than or equal to 916 seconds could be affected.

The field tests for bypass leakage across the SGTS charcoal adsorber andHEPA filter banks are performed at a flow rate of 5764 +/- 10% cfm. The laboratory analysis performed on the SGTS carbon samples will be tested at a velocity of66 fpm based on the system residence time.The SGTS filter train pressure drop is a function of air flow rate andfilter conditions.

Surveillance testing is performed using either the SGTS ordrywell purge fans to provide operating convenience.

Each reactor enclosure secondary containment zone and refueling areasecondary containment zone is tested independently to verify the design leaktightness.

A design leak tightness of 2500 cfm or less for each reactorenclosure and 764 cfm or less for the refueling area at a 0.25 inch of vacuumwater gage will ensure that containment integrity is maintained at an acceptable level if all zones are connected to the SGTS at the same time.The Reactor Enclosure Secondary Containment Automatic Isolation Valvesand Refueling Area Secondary Containment Automatic Isolation Valves can befound in the UFSAR.The post-LOCA offsite dose analysis assumes a reactor enclosure secondary containment post-draw down leakage rate of 2500 cfm and certain post-accident X/Q values. While the post-accident X/Q values represent a statistical inter-pretation of historical meteorological data, the highest ground level windspeed which can be associated with these values is 7 mph (Pasquill-Gifford stability Class G for a ground level release).

Therefore, the surveillance requirement assures that the reactor enclosure secondary containment is verifiedunder meteorological conditions consistent with the assumptions utilized in thedesign basis analysis.

Reactor Enclosure Secondary Containment leakage teststhat are successfully performed at wind speeds in excess of 7 mph would alsosatisfy the leak rate surveillance requirements, since it shows compliance with more conservative test conditions.

LIMERICK

-UNIT 2B 3/4 6-6Amendment No. -4, 6-9, 86 CONTAINMENT SYSTEMSBASES3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROLThe primary containment atmospheric mixing system is provided to ensureadequate mixing of the containment atmosphere to prevent localized accumulations of hydrogen and oxygen from exceeding the lower flammability limit during post-LOCA conditions.

All nuclear reactors must be designed to withstand events that generatehydrogen either due to the zirconium metal water reaction in the core or due toradiolysis.

The primary method to control hydrogen is to inert the primarycontainment.

With the primary containment inert, that is, oxygen concentration

<4.0volume percent (v/o), a combustible mixture cannot be present in the primarycontainment for any hydrogen concentration.

The capability to inert the primarycontainment and maintain oxygen <4.0 v/o works together with Drywell Hydrogen MixingSystem to provide redundant and diverse methods to mitigate events that producehydrogen.

LIMERICK

-UNIT 2B 3/4 6-7Amendment No.ECR n0 n03, 135 3/4.7 PLANT SYSTEMSBASES3/4.7.1 SERVICE WATER SYSTEMS -COMMON SYSTEMSThe OPERABILITY of the service water systems ensures that sufficient coolingcapacity is available for continued operation of safety-related equipment duringnormal and accident conditions.

The redundant cooling capacity of these systems,assuming a single failure, is consistent with the assumptions used in the accidentconditions within acceptable limits.The RHRSW and ESW systems are common to Units 1 and 2 and consist of twoindependent subsystems each with two pumps. One pump per subsystem (loop) ispowered from a Unit 1 safeguard bus and the other pump is powered from a Unit 2safeguard bus. In order to ensure adequate onsite power sources to the systemsduring a loss of offsite power event, the inoperability of these supplies arerestricted in system ACTION statements.

RHRSW is a manually operated system used for core and containment heatremoval.

Each of two RHRSW subsystems has one heat exchanger per unit. EachRHRSW pump provides adequate cooling for one RHR heat exchanger.

By limitingoperation with less than three OPERABLE RHRSW pumps with OPERABLE DieselGenerators, each unit is ensured adequate heat removal capability for the designscenario of LOCA/LOOP on one unit and simultaneous safe shutdown of the otherunit.Each ESW pump provides adequate flow to the cooling loads in its associated loop. With only two divisions of power required for LOCA mitigation of oneunit and one division of power required for safe shutdown of the other unit,one ESW pump provides sufficient capacity to fulfill design requirements.

ESWpumps are automatically started upon start of the associated Diesel Generators.

Therefore, the allowable out of service times for OPERABLE ESW pumps and theirassociated Diesel Generators is limited to ensure adequate cooling during aloss of offsite power event.3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM -COMMON SYSTEMThe OPERABILITY of the control room emergency fresh air supply systemensures that the control room will remain habitable for occupants during andfollowing an uncontrolled release of radioactivity, hazardous chemicals, or smoke.Constant purge of the system at 1 cfm is sufficient to reduce the buildup ofmoisture on the adsorbers and HEPA filters.

The OPERABILITY of this system inconjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less Total Effective Dose Equivalent.

This limitation is consistent with the requirements of 10 CFRPart 50.67, Accident Source Term.Since the Control Room Emergency Fresh Air Supply System is not credited forfiltration in OPERATIONAL CONDITIONS 4 and 5, applicability to 4 and 5 is onlyrequired to support the Chlorine and Toxic Gas design basis isolation requirements.

The CREFAS is common to Units I and 2 and consists of two independent subsystems.

The power supplies for the system are from Unit 1 Safeguard busses, therefore, the inoperability of these Unit 1 supplies are addressed inthe CREFAS ACTION statements in order to ensure adequate onsite power sourcesto CREFAS during a loss of offsite power event. The allowable out of serviceLIMERICK

-UNIT 2B 3/4 7-1Amendment No. 4-46, 149 PLANT SYSTEMSBASESCONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM -COMMON SYSTEM (Continued) times are consistent with those in the Unit 1 Technical Specifications forCREFAS and AC electrical power supply out of service condition combinations.

The Control Room Envelope (CRE) is the area within the confines of the CREboundary that contains the spaces that control room occupants inhabit to controlthe unit during normal and accident conditions.

This area encompasses the controlroom, and other noncritical areas including adjacent support offices, toilet andutility rooms. The CRE is protected during normal operation, natural events, andaccident conditions.

The CRE boundary is the combination of walls, floor,ceiling,

ducting, valves, doors, penetrations and equipment that physically formthe CRE. The OPERABILITY of the CRE boundary must be maintained to ensure thatthe inleakage of unfiltered air into the CRE will not exceed the inleakage assumedin the licensing basis analysis of design basis accident (DBA) consequences to CREoccupants.

The CRE and its boundary are defined in the Control Room EnvelopeHabitability Program.In addition, the CREFAS System provides protection from radiation, smoke andhazardous chemicals to the CRE occupants.

The analysis of hazardous chemicalreleases demonstrates that the toxicity limits are not exceeded in the CREfollowing a hazardous chemical release (Ref. 1). The evaluation of a smokechallenge demonstrates that it will not result in the inability of the CREoccupants to control the reactor either from the control room or from the remoteshutdown panels (Ref. 2).In order for the CREFAS subsystems to be considered

OPERABLE, the CREboundary must be maintained such that the CRE occupant dose from a largeradioactive release does not exceed the calculated dose in the licensing basisconsequence analyses for DBAs, and that CRE occupants are protected from hazardous chemicals and smoke.The LCO is modified by a Note allowing the CRE boundary to be openedintermittently under administrative controls.

This Note only applies to openingsin the CRE boundary that can be rapidly restored to the design condition, such asdoors, hatches, floor plugs, and access panels. For entry and exit through doors,the administrative control of the opening is performed by the person(s) enteringor exiting the area. For other openings, these controls should be proceduralized and consist of stationing a dedicated individual at the opening who is incontinuous communication with the operators in the CRE. This individual will havea method to rapidly close the opening and to restore the CRE boundary to acondition equivalent to the design condition when a need for CRE isolation isindicated.

If the unfiltered inleakage of potentially contaminated air past the CREboundary and into the CRE can result in CRE occupant radiological dose greaterthan the calculated dose of the licensing basis analyses of DBA consequences (allowed to be up to 5 rem TEDE), or inadequate protection of CRE occupants fromhazardous chemicals or smoke, the CRE boundary is inoperable.

Actions must betaken to restore an OPERABLE CRE boundary within 90 days.LIMERICK

-UNIT 2B 3/4 7-1aAmendment No. 4-32-, 149 PLANT SYSTEMSBASESCONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM -COMMON SYSTEM (Continued)

During the period that the CRE boundary is considered inoperable, actionmust be initiated immediately to implement mitigating actions to lessen the effecton CRE occupants from the potential hazards of a radiological or chemical event ora challenge from smoke. Actions must be taken within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to verify that inthe event of a DBA, the mitigating actions will ensure that CRE occupantradiological exposures will not exceed the calculated dose of the licensing basisanalyses of DBA consequences, and that CRE occupants are protected from hazardous chemicals and smoke. These mitigating actions (i.e., actions that are taken tooffset the consequences of the inoperable CRE boundary) should be preplanned forimplementation upon entry into the condition, regardless of whether entry isintentional or unintentional.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is reasonable based onthe low probability of a DBA occurring during this time period, and the use ofmitigating actions.

The 90 day Completion Time is reasonable based on thedetermination that the mitigating actions will ensure protection of CRE occupants within analyzed limits while limiting the probability that CRE occupants will haveto implement protective measures that may adversely affect their ability tocontrol the reactor and maintain it in a safe shutdown condition in the event of aDBA. In addition, the 90 day Completion Time is a reasonable time to diagnose, plan and possibly repair, and test most problems with the CRE boundary.

SR 4.7.2.2 verifies the OPERABILITY of the CRE boundary by testing forunfiltered air inleakage past the CRE boundary and into the CRE. The details ofthe testing are specified in the Control Room Envelope Habitability Program.The CRE is considered habitable when the radiological dose to CRE occupants calculated in the licensing basis analyses of DBA consequences is no more than 5rem Total Effective Dose Equivalent and the CRE occupants are protected fromhazardous chemicals and smoke. SR 4.7.2.2 verifies that the unfiltered airinleakage into the CRE is no greater than the flow rate assumed in the licensing basis analyses of DBA consequences.

When unfiltered air inleakage is greater thanthe assumed flow rate, Required Action 3.7.2.a.2 must be entered.

Required Action3.7.2.a.2.c allows time to restore the CRE boundary to OPERABLE status providedmitigating actions can ensure that the CRE remains within the licensing basishabitability limits for the occupants following an accident.

Compensatory measures are discussed in Regulatory Guide 1.196, Section C.2.7.3, (Ref. 3) whichendorses, with exceptions, NEI 99-03, Section 8.4 and Appendix F (Ref. 4). Thesecompensatory measures may also be used as mitigating actions as required byRequired Action 3.7.2.a.2.b.

Temporary analytical methods may also be used ascompensatory measures to restore OPERABILITY (Ref. 5). Options for restoring theCRE boundary to OPERABLE status include changing the licensing basis DBAconsequence

analysis, repairing the CRE boundary, or a combination of theseactions.

Depending upon the nature of the problem and the corrective action, afull scope inleakage test may not be necessary to establish that the CRE boundaryhas been restored to OPERABLE status.LIMERICK

-UNIT 2B 3/4 7-1bAmendment No. 149 PLANT SYSTEMSBASESCONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM -COMMON SYSTEM (Continued)

REFERENCES

1. UFSAR Section 6.42. UFSAR Section 9.53. Regulatory Guide 1.1964. NEI 99-03, "Control Room Habitability Assessment Guidance,"

June2001.5. Letter from Eric J. Leeds (NRC) to James W. Davis (NEI) dated January30, 2004, "NEI Draft White Paper, Use of Generic Letter 91-18 Processand Alternative Source Terms in the Context of Control RoomHabitability."

(ADAMS Accession No. ML040300694).

3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEMThe reactor core isolation cooling (RCIC) system is provided to assureadequate core cooling in the event of reactor isolation from its primary heatsink and the loss of feedwater flow to the reactor vessel without requiring actuation of any of the emergency core cooling system equipment.

The RCICsystem is conservatively required to be OPERABLE whenever reactor pressure ex-ceeds 150 psig. This pressure is substantially below that for which lowpressure core cooling systems can provide adequate core cooling.The RCIC system specifications are applicable during OPERATIONAL CONDITIONS 1, 2, and 3 when reactor vessel pressure exceeds 150 psig because RCIC is theprimary non-ECCS source of emergency core cooling when the reactor ispressurized.

With the RCIC system inoperable, adequate core cooling is assured by theOPERABILITY of the HPCI system and justifies the specified 14 day out-of-service period. A Note prohibits the application of Specification 3.0.4.b to aninoperable RCIC system. There is an increased risk associated with entering anOPERATIONAL CONDITION or other specified condition in the Applicability with aninoperable RCIC subsystem and the provisions of Specification 3.0.4.b, whichallow entry into an OPERATIONAL CONDITION or other specified condition in theApplicability with the Limiting Condition for Operation not met after performance of a risk assessment addressing inoperable systems and components, should not beapplied in this circumstance.

The surveillance requirements provide adequate assurance that RCIC willbe OPERABLE when required.

Although all active components are testable andfull flow can be demonstrated by recirculation during reactor operation, acomplete functional test requires reactor shutdown.

The pump discharge pipingis maintained full to prevent water hammer damage and to start cooling at theearliest possible moment.LIMERICK

-UNIT 2B 3/4 7-1cAmendment No. 4-321, 149 PLANT SYSTEMSBA-SES3/4.7.4 SNUBBERSAll snubbers are required OPERABLE to ensure that the structural integrity ofthe reactor coolant system and all other safety related systems is maintained during and following a seismic or other event initiating dynamic loads. Snubbersexcluded from this inspection program are those installed on nonsafety-related systems and then only if their failure or failure of the system on which they areinstalled would have no adverse effect on any safety related system.Snubbers are classified and grouped by design and manufacturer but not bysize. For example, mechanical snubbers utilizing the same design features of the2-kip, lO-kip, and 100-kip capacity manufactured by Company "A" are of the sametype. The same design mechanical snubbers manufactured by Company "B" for thepurposes of this Technical Specification would be of a different type, as wouldhydraulic snubbers from either manufacturer.

A list of individual snubbers with detailed information of snubber locationand size and of system affected shall be available at the plant in accordance withSection 50.71(c) of 10 CFR Part 50. The accessibility of each snubber shall bedetermined and approved by the Plant Operations Review Committee.

Thedetermination shall be based upon the existing radiation levels and the expectedtime to perform a visual inspection in each snubber location as well as otherfactors associated with accessibility during plant operations (e.g., temperature, atmosphere,

location, etc.), and the recommendations of Regulatory Guides 8.8 and8.10. The addition or deletion of any snubber shall be made in accordance withSection 50.59 of 10 CFR Part 50.The visual inspection schedule is based on the number of unacceptable snubbersfound during the previous inspection in proportion to the sizes of the varioussnubber populations or categories.

The visual inspection interval is based on afuel cycle of up to 24 months and may be as long as two fuel cycles or 48 monthsdepending on the number of unacceptable snubbers found during the previous visualinspection.

The visual inspection schedule provides confidence that a constantlevel of snubber protection is being maintained and generally allows performance of visual inspections and corrective actions during plant outages.

Inspections performed before that interval has elapsed may be used as a new reference point todetermine the next inspection.

However, the results of such early inspections performed before the original required time interval has elapsed (nominal timeless 25%) may not be used to lengthen the required inspection interval.

Anyinspection whose results required a shorter inspection interval will override theprevious schedule.

A snubber is considered inoperable if it fails the acceptance criteria of thevisual inspection.

LIMERICK

-UNIT 2 B 3/4 7-2 Amendment No. 15 THIS PAGE INTENTIONALLY LEFT BLANK PLANT SYSTEMSBASESSNUBBERS (Continued)

To provide assurance of snubber functional reliability one of two functional testing methods is used with the stated acceptance criteria:

1. Functionally test 13.3% sample of a type of snubber with an additional 1/2sample tested for each functional testing failure, or2. Functionally test 37 snubbers and determine sample acceptance usingFigure 4.7.4-1.Functional Testing sample plans are based on ASME/ANSI OMc-1990 Addenda toASME/ANSI OM-1987, Part 4.Figure 4.7.4-1 was developed using "Wald's Sequential Probability Ratio Plan"as described in Quality Control and Industrial Statistics" by Acheson J. Duncan.Permanent or other exemptions from the surveillance program for individual snubbers may be granted by the Commission if a justifiable basis for exemption ispresented and, if applicable, snubber life destructive testing was performed toqualify the snubbers for the applicable design conditions at either the completion of their fabrication or at a subsequent date. Snubbers so exempted shall belisted in the list of individual snubbers indicating the extent of the exemptions.

The service life of a snubber is evaluated via manufacturer input andinformation through consideration of the snubber service conditions and associated installation and maintenance records (i.e., newly installed

snubber, sealreplaced, spring replaced, in high radiation area, in high temperature area,etc.). The requirement to monitor the snubber service life is included to ensurethat the snubbers periodically undergo a performance evaluation in view of theirage and operating conditions.

These records will provide statistical bases -forfuture consideration of snubber service life.3/4.7.5 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing,including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium.

Thislimitation will ensure that leakage from byproduct, source, and special nuclearmaterial sources will not exceed allowable intake values. Sealed sources areclassified into three groups according to their use, with surveillance requirements commensurate with the probability of damage to a source in thatgroup. Those sources which are frequently handled are required to be tested moreoften than those which are not. Sealed sources which are continuously enclosedwithin a shielded mechanism, i.e,, sealed sources within radiation monitoring

devices, are considered to be stored and need not be tested unless they areremoved from the shielded mechanism.

LIMERICK

-UNIT 2B 3/4 7-3Amendment No. -!-a, 42 PLANT SYSTEMSBASESSEALED SOURCE CONTAMINATION (Continued)

The testing frequency for start-up sources and fission detectors is basedupon physical limitations in leak testing.

For example, the Californium 252start-up neutron source must be leak tested by the manufacturer remotely in ahot cell facility.

Due to the physical design of this source, a six monthfrequency for contamination testing provides reasonable assurance that theradioactive material is properly contained.

LIMERICK

-UNIT 2B 3/4 7-3a PLANT SYSTEMSBASES3/4.7.6 (Deleted)

-INFORMATION FROM THIS SECTION RELOCATED TO THE TRM.3/4.7.1 (Deleted)

-INFORMATION FROM THIS SECTION RELOCATED TO THE TRM.LIMERICK

-UNIT 2B 3/4 7-4Amendment No. 68 PLANT SYSTEMSBASES3/4 7.8 MAIN TURBINE BYPASS SYSTEM 0The required OPERABILITY of the main turbine bypass system is consistent with theassumptions of the feedwater controller failure analysis in the cycle specifictransient analysis.

The main turbine bypass system is required to be OPERABLE to limit peak pressurein the main steam lines and to maintain reactor pressure within acceptable limitsduring events that cause rapid pressurization such that the Safety Limit MCPR isnot exceeded.

With the main turbine bypass system inoperable, continued operation is based on the cycle specific transient analysis which has been performed for thefeedwater controller

failure, maximum demand with bypass failure.LIMERICK

-UNIT 2B 3/4 7-5Amendment No. 16 I 3/4.8 ELECTRICAL POWER SYSTEMSBASES3/4.8.1.

3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and ONSITE POWERDISTRIBUTION SYSTEMSThe OPERABILITY of the A.C. and D.C. power sources and associated distribution systems during operation ensures that sufficient power will beavailable to supply the safety-related equipment required for (1) the safeshutdown of the facility and (2) the mitigation and control of accidentconditions within the facility.

The minimum specified independent andredundant A.C. and D.C. power sources and distribution systems satisfy therequirements of General Design Criterion 17 of Appendix A to 10 CFR Part 50.An offsite power source consists of all breakers, transformers,

switches, interrupting

devices, cabling, and controls required to transmit power from theoffsite transmission network to the onsite Class 1E emergency bus or buses. Thedetermination of the OPERABILITY of an offsite source of power is dependent upongrid and plant factors that, when taken together, describe the design basiscalculation requirements for voltage regulation.

The combination of these factorsensures that the offsite source(s),

which provide power to the plant emergency buses, will be fully capable of supporting the equipment required to achieve andmaintain safe shutdown during postulated accidents and transients.

The plant factors consist of the status of the Startup Transformer

(#10 and#20) load tap changers (LTCs), the status of the Safeguard Transformer

(#101 and#201) load tap changers (LTCs), and the alignment of emergency buses on theSafeguard Buses (101-Bus and 201-Bus).

For an offsite source to be considered

operable, both of its respective LTCs (#10 AND #101 for the source to the 101-Bus,#20 AND #201 for the source to the 201-Bus) must be in service, and in automatic.

For the third offsite source (from 66 kV System) to be considered

operable, theconnected Safeguard Transformer

(#101 or #201) LTC must be in service and inautomatic.

There is a dependency between the alignment of the emergency buses andthe allowable post contingency voltage drop percentage.

The grid factors consist of actual grid voltage levels (real time) and thepost trip contingency voltage drop percentage value.The minimum offsite source voltage levels are established by the voltageregulation calculation.

The transmission system operator (TSO) will notify LGSwhen an agreed upon limit is approached.

The post trip contingency percentage voltage drop is a calculated valuedetermined by the TSO that would occur as a result of the tripping of one of theLimerick generators.

The TSO will notify LGS when an agreed upon limit isexceeded.

The voltage regulation calculation establishes the acceptable percentage voltage drop based upon plant configuration; the acceptable value isdependent upon plant configuration.

Due to the 20 Source being derived from the tertiary of the 4A and 4Btransformer, its operability is influenced by both the 230 kV system and the 500kV system. The 10 Source operability is only influenced by the 230 kV system.LIMERICK

-UNIT 2 B 3/4 8-1 ECR-0 0- 937, ECR 99 00682,ECR 05-00297Amendment No. 1-2-6 3/4-R Fl FC.TRTC.AJ PnWFR SYSTFMSBASES3/4.8.1.

3/4.8.2.

and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and ONSITE POWERDISTRIBUTION SYSTEMSThe anticipated post trip contingency voltage drop for the 66 kV Source(Transformers 8A/8B) is calculated to be less than the 230 kV and 500 kV systems.This is attributed to the electrical distance between the output of the Limerickgenerators and the input to the 8A/8B transformers.

Additionally, the UnitAuxiliary Buses do not transfer to the 8A/8B transformers; this provides margin tothe calculated post trip contingency voltage drop limit.There are various means of hardening the 10 and 20 Sources to obtainadditional margin to the post trip contingency voltage drop limits. These meansinclude, but are not limited to, source alignment of the 4 kV buses, preventing transfer of 13 kV buses, limiting transfer of selected 13 kV loads, and operation with 13 kV buses on the offsite sources.

The specific post trip contingency voltage drop percentage limits for these alignments are identified in the voltageregulation calculation, and controlled via plant procedures.

There are alsoadditional restrictions that can be applied to these limits in the event that anLTC is taken to manual, or if the bus alignment is outside the Two Source ruleset.LGS unit post trip contingency voltage drop percentage calculations areperformed by the PJM Energy Management System (EMS). The PJM EMS consists of aprimary and backup system. LGS will be notified if the real time contingency analysis capability of PJM is lost. Upon receipt of this notification, LGS is torequest PJM to provide an assessment of the current condition of the grid based onthe tools that PJM has available.

The determination of the operability of theoffsite sources would consider the assessment provided by PJM and whether thecurrent condition of the grid is bounded by the grid studies previously performed for LGS.Based on specific design analysis, variations to any of these parameters canbe determined, usually at the sacrifice of another parameter, based on plantconditions.

Specifics regarding these variations must be controlled by plantprocedures or by operability determinations, backed by specific designcalculations.

The ACTION requirements specified for the levels of degradation of thepower sources provide restriction upon continued facility operation commensurate with the level of degradation.

The OPERABILITY of the power sources are con-sistent with the initial condition assumptions of the safety analyses and arebased upon maintaining at least two of the onsite A.C. and the corresponding D.C. power sources and associated distribution systems OPERABLE during accidentconditions coincident with an assumed loss-of-offsite power and single failureof the other onsite A.C. or D.C. source. At least two onsite A.C. and theircorresponding D.C. power sources and distribution systems providing power forat least two ECCS divisions (1 Core Spray loop, 1 LPCl pump and 1 RHR pump insuppression pool cooling) are required for design basis accident mitigation asdiscussed in UFSAR Table 6.3-3.LIMERICK

-UNIT 2 B 3/4 8-1a E GR-0-0937 ECR 99 062Amendment No. 4-2-6, ECR 09-00284 3/4.8 ELECTRICAL POWER SYSTEMSBASESA.C. SOURCES.

D.C. SOURCES, and ONSITE POWER DISTRIBUTION SYSTEMS (Continued)

Onsite A.C. operability requirements for common systems such as CREFAS, SGTS,RHRSW and ESW are addressed in the appropriate system specification actionstatements.

A.C. SourcesAs required by Specification 3.8.1.1, Action e, when one or more diesel generators are inoperable, there is an additional ACTION requirement to verify that all remaining required

systems, subsystems, trains, components, and devices, that depend on theOPERABLE diesel generators as a source of emergency power, are also OPERABLE.

The LPCImode of the RHR system is considered a four train system, of which only two trains arerequired.

The verification for LPCI is not required until two diesel generators areinoperable.

This requirement is intended to provide assurance that a loss-of-offsite power event will not result in a complete loss of safety function of critical systemsduring the period when one or more of the diesel generators are inoperable.

The termverify as used in this context means to administratively check by examining logs orother information to determine if certain components are out-of-service for maintenance or other reasons.

It does not mean to perform the surveillance requirements needed todemonstrate the OPERABILITY of the component.

Specification 3.8.1.1, Action i, prohibits the application of Specification 3.0.4.b to an inoperable diesel generator.

There is an increased risk associated withentering an OPERATIONAL CONDITION or other specified condition in the Applicability with.an inoperable diesel generator subsystem and the provisions of Specification 3.0.4.b,which allow entry into an OPERATIONAL CONDITION or other specified condition in theApplicability with the Limiting Condition for Operation not met after performance of arisk assessment addressing inoperable systems and components, should not be applied inthis circumstance.

If it can be determined that the cause of the inoperable EDG does not exist on theremaining operable EDG(s), based on a common-mode evaluation, then the EDG start test(SR 4.8.1.1.2.a.4) does not have to be performed.

If it cannot otherwise be determined that the cause of the initial inoperable EDG does not exist on the remaining EDG(s),then satisfactory performance of the start test suffices to provide assurance ofcontinued operability of the remaining EDG(s). If the cause of the initialinoperability exists on the remaining operable EDG(s), the EDG(s) shall be declaredinoperable upon discovery and the appropriate action statement for multiple inoperable EDGs shall be entered.

In the event the inoperable EDG is restored to operable statusprior to completing the EDG start test (SR 4.8.1.1.2.a.4) or common-mode failureevaluation as required in Specification 3.8.1.1, the plant corrective action programshall continue to evaluate the common-mode failure possibility.

However, this continued evaluation is not subject to the time constraint imposed by the action statement.

Theprovisions contained in the inoperable EDG action requirements that avoid unnecessary EDG testing are based on Generic Letter 93-05, "Line-Item Technical Specifications Improvement to Reduce Surveillance Requirements for Testing During Power Operation,"

dated September 27, 1993.LIMERICK

-UNIT 2 B 3/4 8-1b ECR 00 00937, ECR 99 00682,Amendment No. 4-2-6,4-32,440, ECR 09-00284 3/4.8 ELECTRICAL POWER SYSTEMSBASESA.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION SYSTEMS (Continued)

The time, voltage, and frequency acceptance criteria specified for the EDG singlelargest post-accident load rejection test (SR 4.8.1.1.2.e.2) are derived from Regulatory Guide 1.9, Rev. 2, December 1979, recommendations.

The test is acceptable if the EDGspeed does not exceed the nominal (synchronous) speed plus 75% of the difference betweennominal speed and the overspeed trip setpoint, or 115% of nominal, whichever is lower.This computes to be 66.5 Hz for the LGS EDGs. The RHR pump motor represents the singlelargest post-accident load. The 1.8 seconds specified is equal to 60% of the 3-secondload sequence interval associated with sequencing the next load following the RHR pumpsin response to an undervoltage on the electrical bus concurrent with a LOCA. Thisprovides assurance that EDG frequency does not exceed predetermined limits and thatfrequency stability is sufficient to support proper load sequencing following a rejection of the largest single load.D.C. SourcesWith one division with one or two battery chargers inoperable (e.g., the voltagelimit of 4.8.2.1.a.2 is not maintained),

the ACTIONS provide a tiered response thatfocuses on returning the battery to the fully charged state and restoring a fullyqualified charger to OPERABLE status in a reasonable time period. Action a.1 requiresthat the battery terminal voltage be restored to greater than or equal to the minimumestablished float voltage within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . This time provides for returning theinoperable charger to OPERABLE status or providing an alternate means of restoring battery terminal voltage to greater than or equal to the minimum established floatvoltage.

Restoring the battery terminal voltage to greater than or equal to theminimum established float voltage provides good assurance that, within 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months , thebattery will be restored to its fully charged condition (Action a.2) from any discharge that might have occurred due to the charger inoperability.

A discharged battery having terminal voltage of at least the minimum established float voltage indicates that the battery is on the exponential charging current portion(the second part) of its recharge cycle. The time to return a battery to its fullycharged state under this condition is simply a function of the amount of the previousdischarge and the recharge characteristic of the battery.

Thus there is good assurance of fully recharging the battery within 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months , avoiding a premature shutdown with itsown attendant risk.If established battery terminal float voltage cannot be restored to greater thanor equal to the minimum established float voltage within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and the charger isnot operating in the current-limiting mode, a faulty charger is indicated.

A faultycharger that is incapable of maintaining established battery terminal float voltagedoes not provide assurance that it can revert to and operate properly in the currentlimit mode that is necessary during the recovery period following a battery discharge event that the DC system is designed for.LIMERICK

-UNIT 2 B 3/4 8-Ic ECR 00 00937, ECR 99 00682,Amendment No. 4-2-6, , 150 3/4.R Fl FUTRC.AI P(1WFR SYTEMSBASESA.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION SYSTEMS (Continued)

If the charger is operating in the current limit mode after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months that is anindication that the battery is partially discharged and its capacity margins will bereduced.

The time to return the battery to its fully charged condition in this case is afunction of the battery charger capacity, the amount of loads on the associated DC system,the amount of the previous discharge, and the recharge characteristic of the battery.

Thecharge time can be extensive, and there is not adequate assurance that it can be recharged within 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months (Action a.2).Action a.2 requires that the battery float current be verified for Divisions 1 and2 as < 2 amps, and for Divisions 3 and 4 as < 1 amp. This indicates that, if thebattery had been discharged as the result of the inoperable battery charger, it has nowbeen fully recharged.

If at the expiration of the initial 18 hour2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months period the batteryfloat current is not within limits this indicates there may be additional batteryproblems.

Action a.3 limits the restoration time for the inoperable battery charger to 7days. This action is applicable if an alternate means of restoring battery terminalvoltage to greater than or equal to the minimum established float voltage has been used(e.g., balance of plant non-Class 1E battery charger).

The 7 days reflects a reasonable time to effect restoration of the qualified battery charger to OPERABLE status.With one or more cells in one or more batteries in one division

< 2.07 V, thebattery cell is degraded.

Per Action b.1, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , verification of the requiredbattery charger OPERABILITY is made by monitoring the battery terminal voltage(4.8.2.1.a.2) and of the overall battery state of charge by monitoring the battery floatcharge current (4.8.2.1.a.1).

This assures that there is still sufficient batterycapacity to perform the intended function.

Therefore, with one or more cells in one ormore batteries

< 2.07 V, continued operation is permitted for a limited period up to 24hours.Division 1 or 2 with float current > 2 amps, or Division 3 or 4 with float current> 1 amp, indicates that a partial discharge of the battery capacity has occurred.

Thismay be due to a temporary loss of a battery charger or possibly due to one or morebattery cells in a low voltage condition reflecting some loss of capacity.

Per Actionb.2, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months verification of the required battery charger OPERABILITY is made bymonitoring the battery terminal voltage.Since Actions b.1 and b.2 only specify "perform,"

a failure of 4.8.2.1.a.1 or4.8.2.1.a.2 acceptance criteria does not result in this Action not being met. However,if one of the Surveillance Requirements is failed the appropriate Action(s),

depending onthe cause of the failures, is also entered.If the Action b.2 condition is due to one or more cells in a low voltage condition but still greater than 2.07 V and float voltage is found to be satisfactory, this is notindication of a substantially discharged battery and 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months is a reasonable time priorto declaring the battery inoperable.

With one or more batteries in one division with one or more cells electrolyte level above the top of the plates, but below the minimum established design limits, (i.e.,greater than the minimum level indication mark), the battery still retains sufficient capacity to perform the intended function.

Per Action b.3, within 31 days the minimumestablished design limits for electrolyte level must be re-established.

LIMERICK

-UNIT 2B 3/4 8-1dAmendment No. 126 1 3/4.8 ELECTRICAL POWER SYSTEMSBASESA.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION SYSTEMS (Continued)

With electrolyte level below the top of the plates there is a potential for dryoutand plate degradation.

Action b.3 addresses this potential (as well as provisions inSpecification 6.8.4.h, "Battery Monitoring and Maintenance Program").

Within 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months slevel is required to be restored to above the top of the plates. The Action requirement to verify that there is no leakage by visual inspection and the Specification 6.8.4.hitem to initiate action to equalize and test in accordance with manufacturer's recommendation are taken from Annex D of IEEE Standard 450-1995.

They are performed following the restoration of the electrolyte level to above the top of the plates. Basedon the results of the manufacturer's recommended testing the battery may have to bedeclared inoperable and the affected cell(s) replaced.

Per Action b.4, with one or more batteries in one division with pilot celltemperature less than the minimum established design limits, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is allowed torestore the temperature to within limits. A low electrolyte temperature limits thecurrent and power available.

Since the battery is sized with margin, while batterycapacity is degraded, sufficient capacity exists to perform the intended function and theaffected battery is not required to be considered inoperable solely as a result of thepilot cell temperature not met.Per Action b.5, with one or more batteries in more than one division with batteryparameters not within limits there is not sufficient assurance that battery capacity hasnot been affected to the degree that the batteries can still perform their requiredfunction, given that multiple divisions are involved.

With multiple divisions

involved, this potential could result in a total loss of function on multiple systems that relyupon the batteries.

The longer restoration times specified for battery parameters on onedivision not within limits are therefore not appropriate, and the parameters must berestored to within limits on all but one division within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .When any battery parameter is outside the allowances of Actions b.1, b.2, b.3,b.4, or b.5, sufficient capacity to supply the maximum expected load requirement is notensured and a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months restoration time is appropriate.

Additionally, discovering one ormore batteries in one division with one or more battery cells float voltage less than2.07 V and float current greater than limits indicates that the battery capacity may notbe sufficient to perform the intended functions.

The battery must therefore be restoredwithin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .The OPERABILITY of the minimum specified A.C. and D.C. power sources andassociated distribution systems during shutdown and refueling ensures that (1) thefacility can be maintained in the shutdown or refueling condition for extended timeperiods and (2) sufficient instrumentation and control capability is available formonitoring and maintaining the unit status.The surveillance requirements for demonstrating the OPERABILITY of thediesel generators are in accordance with the recommendations of Regulatory Guide 1.9, "Selection of Diesel Generator Set Capacity for Standby PowerLIMERICK

-UNIT 2 B 3/4 8-le Amendment No. 4-2-6,ECR 0 093, ER -99 0-06-82,-

ECR 09-00284 3/4.8 ELECTRICAL POWER SYSTEMSBASESA.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION SYSTEMS (Continued)

Supplies, March 10, 1971, Regulatory Guide 1.137 "Fuel-Oil Systems for StandbyDiesel Generators,"

Revision 1, October 1979 and Regulatory Guide 1.108, "Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at NuclearPower Plants,"

Revision 1, August 1977 except for paragraphs C.2.a(3),

C.2.c(1),

C.2.c(2),

C.2.d(3) and C.2.d(4),

and the periodic testing will be performed inaccordance with the Surveillance Frequency Control Program.

The exceptions toRegulatory Guide 1.108 allow for gradual loading of diesel generators during testingand decreased surveillance test frequencies (in response to Generic Letter 84-15).The single largest post-accident load on each diesel generator is the RHR pump.The Surveillance Requirement for removal of accumulated water from the fuel oilstorage tanks is for preventive maintenance.

The presence of water does notnecessarily represent failure of the Surveillance Requirement, provided theaccumulated water is removed during performance of the Surveillance.

Accumulated water in the fuel oil storage tanks constitutes a collection of waterat a level that can be consistently and reliably measured.

The minimum level atwhich accumulated water can be consistently and reliably measured in the fueloil storage tank sump is 0.25 inches. Microbiological fouling is a major cause offuel oil degradation.

There are numerous bacteria that can grow in fuel oil andcause fouling, but all must have a water environment in order to survive.Removal of accumulated water from the fuel storage tanks once every (31) dayseliminates the necessary environment for bacterial survival.

This is the mosteffective means of controlling microbiological fouling.

In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation.

Water may.come from any of several sources, including condensation, ground water, rainwater, contaminated fuel oil, and from breakdown of the fuel oil by bacteria.

Frequent checking for and removal of accumulated water minimizes fouling andprovides data regarding the watertight integrity of the fuel oil system. TheSurveillance Frequencies are established by Regulatory Guide 1.137.The surveillance requirements for demonstrating the OPERABILITY of theunits batteries are in accordance with the recommendations of IEEE Standard 450-1995, "IEEE Recommended Practice for Maintenance,

Testing, and Replacement of VentedLead-Acid Batteries for Stationary Applications."

Verifying battery float current while on float charge (4.8.2.1.a.1) is used todetermine the state of charge of the battery.

Float charge is the condition in whichthe charger is supplying the continuous charge required to overcome the internal lossesof a battery and maintain the battery in a charged state. The float currentrequirements are based on the float current indicative of a charged battery.

Use offloat current to determine the state of charge of the battery is consistent with IEEE-450-1995.

This Surveillance Requirement states the float current requirement is notrequired to be met when battery terminal voltage is less than the minimum established float voltage of 4.8.2.1.a.2.

When this float voltage is not maintained the Actionsof LCO 3.8.2.1, Action b., are being taken, which provide the necessary andappropriate verifications of the battery condition.

Furthermore, the float currentlimits are established based on the float voltage range and is not directly applicable when this voltage is not maintained.

B LIMERICK

-UNIT 2 B 3/4 8-2 Amendment No. 44,8-5,49,4-2-4, 147ECR 97 01067 THIS PAGE INTENTIONALLY LEFT BLANKECR 00-00937 3/4.8 ELECTRICAL POWER SYSTEMSBASESA.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION SYSTEMS (Continued)

Verifying, per 4.8.2.1.a.2, battery terminal voltage while on float charge forthe batteries helps to ensure the effectiveness of the battery chargers, which supportthe ability of the batteries to perform their intended function.

Float charge is thecondition in which the charger is supplying the continuous charge required to overcomethe internal losses of a battery and maintain the battery in a fully charged statewhile supplying the continuous steady state loads of the associated DC subsystem.

Onfloat charge, battery cells will receive adequate current to optimally charge thebattery.

The voltage requirements are based on the minimum float voltage established by the battery manufacturer (2.20 Vpc, average, or 132 V at the battery terminals).

This voltage maintains the battery plates in a condition that supports maintaining thegrid life (expected to be approximately 20 years).Surveillance Requirements 4.8.2.1.b.1 and 4.8.2.1.c require verification that thecell float voltages are equal to or greater than 2.07 V.The limit specified in 4.8.2.1.b.2 for electrolyte level ensures that the platessuffer no physical damage and maintains adequate electron transfer capability.

Surveillance Requirement 4.8.2.1.b.3 verifies that the pilot cell temperature isgreater than or equal to the minimum established design limit (i.e., 60 degreesFahrenheit).

Pilot cell electrolyte temperature is maintained above this temperature to assure the battery can provide the required current and voltage to meet the designrequirements.

Temperatures lower than assumed in battery sizing calculations act toinhibit or reduce battery capacity.

Surveillance Requirement 4.8.2.1.d.1 verifies the design capacity of the batterychargers.

According to Regulatory Guide 1.32, the battery charger supply isrecommended to be based on the largest combined demands of the various steady stateloads and the charging capacity to restore the battery from the design minimum chargestate to the fully charged state, irrespective of the status of the unit during thesedemand occurrences.

The minimum required amperes and duration ensures that theserequirements can be satisfied.

Surveillance Requirement 4.8.2.1.d.1 requires that each battery charger be capableof supplying the amps listed for the specified charger at the minimum established floatvoltage for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The ampere requirements are based on the output rating of thechargers.

The voltage requirements are based on the charger voltage level after aresponse to a loss of AC power. This time period is sufficient for the chargertemperature to have stabilized and to have been maintained for at least 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .A battery service test, per 4.8.2.1.d.2, is a special test of the battery's capability, as found, to satisfy the design requirements (battery duty cycle) of the DCelectrical power system. The discharge rate and test length corresponds to the designduty cycle requirements as specified in the UFSAR.LIMERICK

-UNIT 2 B 3/4 8-2a Amendment No. 99,4-2-6, 147ECR 97 01067 3/4.8 ELECTRICAL POWER SYSTEMSBASESA.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION SYSTEMS (Continued)

A battery performance discharge test (4.8.2.1.e and f) is a test of constantcurrent capacity of a battery, normally done in the as found condition, after havingbeen in service, to detect any change in the capacity determined by the acceptance test. The test is intended to determine overall battery degradation due to age andusage. Degradation (as used in 4.8.2.1.f) is indicated when the battery capacity dropsmore than 10% from its capacity on the previous performance test, or is below 90% ofthe manufacturer's rating.Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying 4.8.2.1.e and 4.8.2.1.f;

however, only themodified performance discharge test may be used to satisfy the battery service testrequirements of 4.8.2.1.d.2.

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-UNIT 2B 3/4 8-2bAmendment No. 126 I ELECTRICAL POWER SYSTEMSBASES3/4.8.4 ELECTRICAL EOUIPMENT PROTECTIVE DEVICESThe RPS Electric Power Monitoring System is provided to isolate the RPS busfrom the RPS/UPS inverter or an alternate power supply in the event ofovervoltage, undervoltage, or underfrequency.

This system protects the loadsconnected to the RPS bus from unacceptable voltage and frequency conditions.

Theessential equipment powered from the RPS buses includes the RPS logic, scramsolenoids, and valve isolation logic.The Allowable Values are derived from equipment design limits, corrected forcalibration and instrument errors. The trip setpoints are then determined, accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived in this manner provide adequate protection and include allowances forinstrumentation uncertainties, calibration tolerances, and instrument drift.The Allowable Values for the instrument settings are based on the RPSproviding power within the design ratings of the associated RPS components (e.g.,RPS logic, scram solenoids).

The most limiting voltage requirement and associated line losses determine the settings of the electric power monitoring instrument channels.

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-UNIT 2B 3/4 8-3Amendment No. -5;,Bases Ltr 11/18/98,

-6,Associated with Amendment No. 170 THIS PAGE INTENTIONALLY LEFT BLANK 3/4.9 REFUELING OPERATIONS

.BASES3/4.9.1 REACTOR MODE SWITCHLocking the OPERABLE reactor mode switch in the Shutdown or Refuel position, as specified, ensures that the restrictions on control rod withdrawal and refueling platform movement during the refueling operations are properly activated.

Theseconditions reinforce the refueling procedures and reduce the probability ofinadvertent criticality, damage to reactor internals or fuel assemblies, andexposure of personnel to excessive radioactivity.

3/4.9.2 INSTRUMENTATION The OPERABILITY of at least two source range monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core. The minimum count rate is not required when sixteen or fewer fuelassemblies are in the core. During a typical core reloading, two, three or fourirradiated fuel assemblies will be loaded adjacent to each SRM to produce greaterthan the minimum required count rate. Loading sequences are selected to providefor a continuous multiplying medium to be established between the required oper-able SRMs and the location of the core alteration.

This enhances the abilityof the SRMs to respond to the loading of each fuel assembly.

During a core un-loading, the last fuel to be removed is that fuel adjacent to the SRMs..3/4.9.3 CONTROL ROD POSITIONThe requirement that all control rods be inserted during other COREALTERATIONS ensures that fuel will not be loaded into a cell without a controlrod.3/4.9.4 DECAY TIMEThe minimum requirement for reactor subcriticality prior to fuel movementensures that sufficient time has elapsed to allow the radioactive decay of theshort lived fission products.

This decay time is consistent with the assump-tions used in the accident analyses.

3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling stationpersonnel can be promptly informed of significant changes in the facility statusor core reactivity condition during movement of fuel within the reactor pressurevessel.LIMERICK

-UNIT 2 B 3/4 9-1 REFUELING OPERATIONS BASES3.4.9.6 REFUELING PLATFORMThe OPERABILITY requirements ensure that (1) the refueling platform willbe used for handling control rods and fuel assemblies within the reactor pressurevessel, (2) each hoist has sufficient load capacity for handling fuel assemblies and control rods, (3) the core internals and pressure vessel are protected fromexcessive lifting force in the event they are inadvertently engaged duringlifting operations, and (4) inadvertent criticality will not occur due to fuelbeing loaded into a unrodded cell.3/4.9.7 CRANE TRAVEL -SPENT FUEL STORAGE POOLThe restriction on movement of loads in excess of the nominal weight of afuel assembly and associated lifting device over other fuel assemblies in thestorage pool ensures that in the event this load is dropped 1) the activityrelease will be limited to that contained in a single fuel assembly, and 2) anypossible distortion of fuel in the storage racks will not result in a criticalarray. This assumption is consistent with the activity release assumed in thesafety analyses.

3/4.9.8 and 3/4.9.9 WATER LEVEL -REACTOR VESSEL and WATER LEVEL -SPENT FUELSTORAGE POOLThe restrictions on minimum water level ensure that sufficient water depthis available to remove 99% of the assumed 10% iodine gap activity releasedfrom the rupture of an irradiated fuel assembly.

This minimum water depth isconsistent with the assumptions of the accident analysis.

3/4.9.10 CONTROL ROD REMOVALThese specifications ensure that maintenance or repair of control rods orcontrol rod drives will be performed under conditions that limit the probability of inadvertent criticality.

The requirements for simultaneous removal of morethan one control rod are more stringent since the SHUTDOWN MARGIN specification provides for the core to remain subcritical with only one control rod fullywithdrawn.

3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION Irradiated fuel in the shutdown reactor core generates heat during the decay offission products and increases the temperature of the reactor coolant.

This decay heatmust be removed by the RHR system to maintain adequate reactor coolant temperature.

RHR shutdown cooling is comprised of four (4) subsystems which make two (2)loops. Each loop consists of two (2) motor driven pumps, a heat exchanger, andassociated piping and valves. Both loops have a common suction from the samerecirculation loop. Two (2) redundant, manually controlled shutdown cooling subsystems of the RHR system provide decay heat removal.

Each pump discharges the reactorcoolant, after circulation through the respective heat exchanger, to the reactor viathe associated recirculation loop. The RHR heat exchangers transfer heat to the RHRService Water System.An OPERABLE RHR shutdown cooling subsystem consists of one (1) OPERABLE RHR pump,one (1) heat exchanger, and the associated piping and valves. The requirement forLIMERICK

-UNIT 2B 3/4 9-2Amendment No. 5-0, 64, 82 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION (Cont'd)having one (1) RHR shutdown cooling subsystem OPERABLE ensures that 1) sufficient cooling capacity is available to remove decay heat and maintain the water in thereactor pressure vessel below 140'F, and 2) sufficient coolant circulation would beavailable through the reactor core to assure accurate temperature indication.

The requirement to have two (2) RHR shutdown cooling subsystems OPERABLE whenthere is less than 22 feet of water above the reactor vessel flange ensures that asingle failure of the operating loop will not result in a complete loss of residualheat removal capability.

With the reactor vessel head removed and 22 feetof water above the reactor vessel flange, a large heat sink is available forcore cooling.

Thus, in the event of a failure of the operating RHR subsystem, adequatetime is provided to initiate alternate methods capable of decay heat removal oremergency procedures to cool the core.To meet the LCO of the two (2) subsystems OPERABLE when there is less than 22feet of water above the reactor vessel flange, both pumps in one (1) loop or one (1)pump in each of the two (2) loops must be OPERABLE.

The two (2) subsystems have acommon suction source and are allowed to have a common heat exchanger and commondischarge piping. Additionally, each shutdown cooling subsystem can provide therequired decay heat removal capability;

however, ensuring operability of the othershutdown cooling subsystem provides redundancy.

The required cooling capacity of an alternate method of decay heat removal shouldbe ensured by verifying its capability to maintain or reduce reactor coolanttemperature either by calculation (which includes a review of component and systemavailability to verify that an alternate decay heat removal method is available) or bydemonstration.

Decay heat removal capability by ambient losses can be considered inevaluating alternate decay heat removal capability.

With the required decay heat removal subsystem(s) inoperable and the requiredalternate method(s) of decay heat removal not available in accordance with Action "a",additional actions are required to minimize any potential fission product release tothe environment.

This includes ensuring Refueling Floor Secondary Containment isOPERABLE; one (1) Standby Gas Treatment subsystem is OPERABLE; and Secondary Containment isolation capability (i.e., one (1) Secondary Containment isolation valveand associated instrumentation are OPERABLE or other acceptable administrative controlsto assure isolation capability) in each associated penetration not isolated that isassumed to be isolated to mitigate radioactive releases.

This may be performed as anadministrative check, by examining logs or other information to determine whether thecomponents are out of service for maintenance or other reasons.

It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the components.

If, however, any required component is inoperable, then it must be restored to OPERABLEstatus. In this case, the surveillance may need to be performed to restore thecomponent to OPERABLE status. Actions must continue until all required components areOPERABLE.

If no RHR subsystem is in operation, an alternate method of coolant circulation isrequired to be established within one (1) hour. The Completion Time is modified suchthat one (1) hour is applicable separately for each occurrence involving a loss ofcoolant circulation.

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-UNIT 2 B 3/4 9-2a Amendment No. 82,ECR 01-00386 INTENTIONALLY LEFT BLANK 3/4.10 SPECIAL TFST FXCEPTTONS BASES3/4.10.1 PRIMARY CONTAINMENT INTEGRITY The requirement for PRIMARY CONTAINMENT INTEGRITY is not applicable duringthe period when open vessel tests are being performed during the low powerPHYSICS TESTS.3/4.10.2 ROD WORTH MINIMIZER In order to perform the tests required in the technical specifications it is necessary to bypass the sequence restraints on control rod movement.

Theadditional surveillance requirements ensure that the specifications on heatgeneration rates and shutdown margin requirements are not exceeded during theperiod when these tests are being performed and that individual rod worths donot exceed the values assumed in the safety analysis.

3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS Performance of shutdown margin demonstrations with the vessel head removedrequires additional restrictions in order to ensure that criticality does notoccur. These additional restrictions are specified in this LCO.3/4.10.4 RECIRCULATION LOOPSThis special test exception permits reactor criticality under no flowconditions and is required to perform certain startup and PHYSICS TESTS whileat low THERMAL POWER levels.3/4.10.5 OXYGEN CONCENTRATION Relief from the oxygen concentration specifications is necessary in orderto provide access to the primary containment during the initial startup andtesting phase of operation.

Without this access the startup and test programcould be restricted and delayed.3/4.10.6 TRAINING STARTUPSThis special test exception permits training startups to be performed withthe reactor vessel depressurized at low THERMAL POWER and temperature whilecontrolling RCS temperature with one RHR subsystem aligned in the shutdowncooling mode in order to minimize contaminated water discharge to theradioactive waste disposal system.3/4.10.7 SPECIAL INSTRUMENTATION

-INITIAL CORE LOADINGThis special test exception permits relief from the requirements for aminimum count rate while loading the first 16 fuel bundles to allow sufficient source-to-detector coupling such that minimum count rate can be achieved on anSRM. This is acceptable because of the significant margin to criticality while loading the initial 16 fuel bundles.LIMERICK

-UNIT 2B 3/4 10-1 3/4.10 SPECIAL TEST EXCEPTIONS BASES3/4.10.8 INSERVICE LEAK AND HYDROSTATIC TESTINGThis special test exception permits certain reactor coolant pressure tests to beperformed in OPERATIONAL CONDITION 4 when the metallurgical characteristics of thereactor pressure vessel (RPV) or plant temperature control capabilities during thesetests require the pressure testing at temperatures greater than 200'F and less thanor equal to 212'F (normally corresponding to OPERATIONAL CONDITION 3). Theadditionally imposed OPERATIONAL CONDITION 3 requirements for SECONDARY CONTAINMENT INTEGRITY provide conservatism in response to an operational event.Invoking the requirement for Refueling Area Secondary Containment Integrity alongwith the requirement for Reactor Enclosure Secondary Containment Integrity appliesthe requirements for Reactor Enclosure Secondary Containment Integrity to anextended area encompassing Zones 2 and 3. Operations with the Potential forDraining the Vessel, Core alterations, and fuel handling are prohibited in thissecondary containment configuration.

Drawdown and inleakage testing performed forthe combined zone system alignment shall be considered adequate to demonstrate integrity of the combined zones.Inservice hydrostatic testing and inservice leak pressure tests required by SectionXI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure VesselCode are performed prior to the reactor going critical after a refueling outage. Theminimum temperatures (at the required pressures) allowed for these tests aredetermined from the RPV pressure and temperature (P/T) limits required by LCO 3.4.6,Reactor Coolant System Pressure/Temperature Limits. These limits are conservatively based on the fracture toughness of the reactor vessel, taking into accountanticipated vessel neutron fluence.

With increased reactor fluence over time, theminimum allowable vessel temperature increases at a given pressure.

liLIMERICK

-UNIT 2B 3/4 10-2Amendment No. -5ECR 99 00864, 130 3/4.11 RADIOACTIVE EFFLUENTS BASES3/4.11.1.1 and 3/4.11.1.2 (Deleted)

THE INFORMATION FROM THESE SECTIONSHAS BEEN RELOCATED TO THE ODCM.LIMERICK

-UNIT 2B 3/4 11-1Amendment No. 11 THIS PAGE INTENTIONALLY LEFT BLANK RADIOACTIVE EFFLUENTS BASES3/4.11.1.3 (Deleted)

-INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM.3/4.11.1.4 LIQUID HOLDUP TANKSThe tanks listed in this specification include all those outdoor radwastetanks that are not surrounded by liners, dikes, or walls capable of holdingthe tank contents and that do not have tank overflows and surrounding areadrains connected to the liquid radwaste treatment system.Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of thetanks' contents, the resulting concentrations would be less than 10 times thelimits of 10 CFR Part 20, Appendix B, Table 2, Column 2, at the nearest potablewater supply and the nearest surface water supply in an UNRESTRICTED AREA.3/4.11.2.1 (Deleted)

-INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM.LIMERICK

-UNIT 2B 3/4 11-2Amendment No. 44Associated with Amendment No. 148 RADIOACTIVE EFFLUENTS BASES3/4.11.2.2 through 3/4 11.2.4 (Deleted)

THE INFORMATION FROM THESE SECTIONSHAS BEEN RELOCATED TO THE ODCM.-7LIMERICK

-UNIT 2B 3/4 11-3Amendment No. 11 RADIOACTIVE EFFLUENTS BASES.ý14_11_2_ý FXPImTSVF UAS MIXTIIRFThis specification is provided to ensure that the concentration of poten-tially explosive gas mixtures contained in the main condenser offgas treat-ment system is maintained below the flammability limits of hydrogen and oxygen.Maintaining the concentration of hydrogen below its flammability limit providesassurance that the releases of radioactive materials will be controlled inconformance with the requirements of General Design Criterion 60 of Appendix Ato 10 CFR Part 50.LIMERICK

-UNIT 2B 3/4 11-4Amendment No.11 I BASES3/4.11.2.6 MAIN CONDENSER Restricting the gross radioactivity rate of noble gases from the main condenser provides reasonable assurance that the total body exposure to an individual at theexclusion area boundary will not exceed a small fraction of the limits of 10 CFR Part100 in the event this effluent is inadvertently discharged directly to theenvironment without treatment.

This specification implements the requirements ofGeneral Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50.3/4.11.2.7, 3/4.11.3, and 3/4.11.4 (Deleted)

-INFORMATION FROM THESE SECTIONS RELOCATED TO THE ODCM OR PCP.LIMERICK

-UNIT 2B 3/4 11-5Amendment No. 11 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASESSection 3/4.12 (Deleted)

THE INFORMATION FROM THIS SECTIONHAS BEEN RELOCATED TO THE ODCM.BASES PAGE B 3/4 12-2 HAS BEENINTENTIONALLY OMITTED.LIMERICK

-UNIT 2B 3/4 12-1Amendment No. 11 1 THIS PAGE INTENTIONALLY LEFT BLANK SECTION 5.0DESIGN FEATURES THIS PAGE INTENTIONALLY LEFT BLANK 5.0 DESIGN FEATURES5.1 SITEEXCLUSION AREA5.1.1 The exclusion area shall be as shown in Figure 5.1.1-1.LOW POPULATION ZONE5.1.2 The low population zone shall be as shown in Figure 5.1.2-1.MAPS DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS ANDLIQUID EFFLUENTS 5.1.3 Information regarding radioactive gaseous and liquid effluents, which willallow identification of structures and release points as well as definition ofUNRESTRICTED AREAS within the SITE BOUNDARY that are accessible to MEMBER OF THEPUBLIC, shall be as shown in Figures 5.1.3-la and 5.1.3-lb.

5.1.4 (Deleted)

5.2 CONTAINMENT

CONFIGURATION 5.2.1 The primary containment is a steel lined reinforced concrete structure consisting of a drywell and suppression chamber.

The drywell is a steel-lined reinforced concrete vessel in a shape of a truncated cone on top of a water filledsuppression chamber and is separated by a diaphragm slab and connected to thesuppression chamber through a series of downcomer vents. The drywell has amaximum free air volume of 243,580 cubic feet at a minimum suppression pool levelof 22 feet. The suppression chamber has a maximum air region of 159,540 cubicfeet and a minimum water region of 122,120 cubic feet.DESIGN TEMPERATURE AND PRESSURE5.2.2 The primary containment is designed and shall be maintained for:a. Maximum internal pressure 55 psig.b. Maximum internal temperature:

drywell 340'F.suppression pool 2200F.c. Maximum external to internal differential pressure 5 psid.d. Maximum floor differential pressure:

30 psid, downward.

20 psid, upward.LIMERICK

-UNIT 25-1Amendment No. 11 1 irk°oo=,wi,FIGURE 5.1.1-1EXCLUSION AREALIMERICK

-UNIT 25-2 FIGURE 5.1.2-1LOW POPULATION ZONELIMERICK

-UNIT 25-3 FIGURE 5.1.3-laMAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARYFOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS LIMERICK

-UNIT 25-4 LIGUIDm~Art. FA*'4FIGURE 5.1.3-lbMAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARYFOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS LIMERICK

-UNIT 25-5 THE FIGURE ON THIS PAGE HAS BEEN RELOCATED TO THE ODCM.LIMERICK

-UNIT 25-6Amendment No. 11 DESIGN FEATURESSECONDARY CONTAINMENT 5.2.3 The secondary containment consists of three distinct isolatable zones.Zones I and II are the Unit 1 and Unit 2 reactor enclosures respectively.

Zone III is the common refueling area. Each zone has an independent normalventilation system which is capable of providing secondary containment zoneisolation as required.

Each reactor enclosure (Zone I or II) completely encloses and providessecondary containment for its corresponding primary containment and reactorauxiliary or service equipment, and has a minimum free volume of 1,800,000 cubic feet.The common refueling area (Zone III) completely encloses and providessecondary containment for the refueling servicing equipment and spent fuelstorage facilities for Units 1 and 2, and has a minimum free volume of 2,200,000 cubic feet.5.3 REACTOR COREFUEL ASSEMBLIES 5.3.1 The reactor core shall consist of not more than 764 fuel assemblies and shall be limited to those fuel assemblies which have been analyzed with NRCapproved codes and methods and have been shown to comply with all Safety DesignBases in the Final Safety Analysis Report (FSAR).CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 185 cruciform-shaped control rodassemblies.

5.4 REACTOR COOLANT SYSTEMDESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

a. In accordance with the code requirements specified in Section 5.2of the FSAR, with allowance for normal degradation pursuant to theapplicable Surveillance Requirements, LIMERICK

-UNIT 25-7 DESIGN FEATURESDESIGN PRESSURE AND TEMPERATURE (Continued)

b. For a pressure of:1. 1250 psig on the suction side of the recirculation pump.2. 1500 psig from the recirculation pump discharge to the outletside of the discharge shutoff valve.3. 1500 psig from the discharge shutoff valve to the jet pumps.c. For a temperature of 575°F.VOLUME5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 22,400 cubic feet at a nominal steam dome saturation temperature of 5520F.5.5 FUEL STORAGECRITICALITY 5.5.1.1 The spent fuel storage racks are designed and shall be maintained with:a. A keff equivalent to less than or equal to 0.95 when flooded withunborated water, including all calculational uncertainties andbiases as described in Section 9.1.2 of the FSAR.b. A nominal center-to-center distance between fuel assemblies placed in the storage racks of greater than or equal to 6.244inches.5.5.1.2 The keff for new fuel for the first core loading stored dry in thespent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is assumed.DRAINAGE5.5.2 The spent fuel storage pool is designed and shall be maintained to preventinadvertent draining of the pool below elevation 346'0".CAPACITY5.5.3 The spent fuel storage pool is designed and shall be maintained with astorage capacity limited to no more than 4117 fuel assemblies.

5.6 COMPONENT CYCLIC OR TRANSIENT LIMIT5.6.1 The components identified in Table 5.6.1-1 are designed and shall bemaintained within the cyclic or transient limits of Table 5.6.1-1.LIMERICK

-UNIT 25-8Amendment No. 4-3, 51 TABLE 5.6.1-1COMPONENT CYCLIC OR TRANSIENT LIMITSCOMPONENT CYCLIC ORTRANSIENT LIMITDESIGN CYCLEOR TRANSIENT Reactor120 heatup and cooldown cycles80 step change cycles180 reactor trip cycles130 hydrostatic pressure andleak tests70°F to 560°F to 70°FLoss of feedwater heaters100% to 0% of RATED THERMAL POWERPressurized to 930 and 1250 psigLIMERICK

-UNIT 25-9 THIS PAGE INTENTIONALLY LEFT BLANK SECTION 6.0ADMINISTRATIVE CONTROLS THIS PAGE INTENTIONALLY LEFT BLANK 6.0 ADMINISTRATIVE CONTROLS6.1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall unit operation andshall delegate in writing the succession to this responsibility during hisabsence.6.1.2 The Shift Manager, or during his absence from the control room, adesignated individual shall be responsible for the control room command function.

A management directive to this effect, signed by the Vice President, LimerickGenerating Station shall be reissued to all station personnel on an annualbasis.6.2 ORGANIZATION 6.2.1 OFFSITE AND ONSITE ORGANIZATIONS Onsite and offsite organizations shall be established for unit operation andcorporate management, respectively.

The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclearpower plant.a. Lines of authority, responsibility, and communication shall beestablished and defined for the highest management levels throughintermediate levels to and including all operating organizational positions.

These relationships shall be documented and updated, asappropriate, in the form of organizational charts, functional descriptions of departmental responsibilities and relationships, andjob descriptions for key personnel positions, or in equivalent formsof documentation.

These requirements shall be documented in theLimerick Quality Assurance Program.b. The Plant Manager shall be responsible for overall unit safeoperation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.c. The Vice President, Limerick Generating Station shall have corporate responsibility for overall plant nuclear safety and shall take anymeasures needed to ensure acceptable performance of the staff inoperating, maintaining, and providing technical support to the plantto ensure nuclear safety.d. The individuals who train the operating staff and those who carryout health physics and quality assurance functions may report to theappropriate onsite manager;

however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

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-UNIT 26-1Amendment No. 2, 60 ADMINISTRATIVE CONTROLS6.2.2 UNIT STAFFa. Each on duty shift shall be composed of at least the minimum shiftcrew composition shown in Table 6.2.2-1;b. At least one licensed Operator shall be in the control room when fuelis in the reactor.

In addition, while the unit is in OPERATIONAL CONDITION 1, 2, or 3, at least one licensed Senior Operator shall bein the control room;c. A Health Physics Technician*

shall be on site when fuel is in thereactor;d. ALL CORE ALTERATIONS shall be observed and directly supervised byeither a licensed Senior Operator or licensed Senior Operator Limitedto Fuel Handling who has no other concurrent responsibilities duringthis operation;

e. (Deleted)

-INFORMATION FROM THIS SECTION RELOCATED TO THE TRM.f. (Deleted)

The Health Physics Technician position may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to accommodate unexpected

absence, provided immediate action is taken to fill the required position.

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-UNIT 26-2Amendment No. -9, 60, 68, 4--02, 159 ADMINISTRATIVE CONTROLS6.2.2 UNIT STAFF (Continued)

g. The individual filling the position of Operations Manager as defined byANSI/ANS-3.1-1978 or another Manager in Operations shall hold a SeniorReactor Operator License.LIMERICK

-UNIT 26-3Amendment No. 2,98,41,60,4-24, 159 INTENTIONALLY LEFT BLANKLIMERICK

-UNIT 26-4Amendment No. 2 TABLE 6.2.2-1MINIMUM SHIFT CREW COMPOSITION TWO UNITS WITH A COMMON CONTROL ROOMWITH UNIT 1 IN CONDITION 4 OR 5 OR DEFUELEDPOSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITIONCONDITION 1, 2, or 3 CONDITION 4 OR 5SM 1* 1*SRO 2* 2*RO 2 1NLO 2 2**STA 1*** NoneWITH UNIT 1 IN CONDITION 1, 2, OR 3POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITIONCONDITION 1, 2, or 3 CONDITION 4 or 5SM 1* 1*SRO 2* 2*RO 2** 1NLO 2** 1STA 1*,*** NoneTABLE NOTATIONS

Individual(s) may fill the same position on Unit 1.** One of the two required individuals may fill the same position on Unit 1.***The STA position may be filled by an on-shift SM or SRO provided the individual meets the 1985 NRC Policy Statement on Engineering Expertise on Shift.SM -Shift Manager with a Senior Operator license on Unit 2.SRO -Individual with a Senior Operator license on Unit 2.RO -Individual with an Operator license on Unit 2.NLO -Non-licensed operator properly qualified to support the unit to whichassigned.

STA -Shift Technical AdvisorExcept for the Shift Manager (SM), the shift crew composition may be one lessthan the minimum requirements of Table 6.2.2-1 for a period of time not toexceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on-duty shift crewmembers provided immediate action is taken to restore the shift crew compo-sition to within the minimum requirements of Table 6.2.2-1.

This provision does not permit any shift crew position to be unmanned upon shift change dueto an upcoming shift crewman being late or absent.During any absence of the Shift Manager (SM) from the control room while the unitis in OPERATIONAL CONDITION 1, 2, or 3, an individual with a valid SeniorOperator license shall be designated to assume the control room command function.

During any absence of the Shift Manager (SM) from the control room while the unit isin OPERATIONAL CONDITION 4 or 5, an individual with a valid Senior Operatorlicense or Operator license shall be designated to assume the control roomcommand function.

LIMERICK

-UNIT 26-5Amendment No. 2-9, 60 ADMINISTRATIVE CONTROLS6.2.3 DELETED.

The information from this section is located in the UFSAR.6.2.4 SHIFT TECHNICAL ADVISOR6.2.4.1 The Shift Technical Advisor shall provide advisory technical support to ShiftSupervision in the areas of thermal hydraulics, reactor engineering, and plant analysiswith regard to safe operation of the unit. The Shift Technical Advisor shall meet thequalifications specified by the 1985 NRC Policy Statement on Engineering Expertise on Shift.6.3 [NIT STAFF niJALIFICATIONS 6.3.1 Each member of unit staff shall meet or exceed the minimum qualifications ofANSI/ANS 3.1-1978 for comparable positions, except for the Manager -Radiation Protection whoshall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, and thelicensed operators who shall comply only with the requirements of 1OCFR55.LIMERICK

-UNIT 26-6Amendment No. 2, 2-0, 2-9, 4-0, 171 @Crecptedk by Letter dtedOctober 17, 1995 ADMINISTRATIVE CONTROLS6.4 DELETED6.5 DELETEDTHE INFORMATION FROM SECTION 6.5 HAS BEEN RELOCATED TO THE QATRLIMERICK

-UNIT 26-7 Amendment No. 2-O, 68, -!4O, 4-2-9, 4-24, 138 ADMINISTRATIVE CONTROLSTHE INFORMATION FROM SECTION 6.5 HAS BEEN RELOCATED TO THE QATRLIMERICK

-UNIT 26-8Amendment No. --0, -8, 4-2, 4-1-8, 138 ADMINISTRATIVE CONTROLSTHE INFORMATION FROM SECTION 6.5 HAS BEEN RELOCATED TO THE QATRLIMERICK

-UNIT 26-9Amendment No. -, 49, 42, 138 ADMINISTRATIVE CONTROLSTHE INFORMATION FROM SECTION 6.5 HAS BEEN RELOCATED TO THE QATRLIMERICK

-UNIT 26-10Amendment No. &Q, 138 THIS PAGE IS INTENTIONALLY LEFT BLANK.LIMERICK

-UNIT 26-11Amendment No. 60 ADMINISTRATIVE CONTROLSTHE INFORMATION FROM SECTION 6.5 HAS BEEN RELOCATED TO THE QATRLIMERICK

-UNIT 26-12Amendment No. 2, 0, 4-, 4i-8, 138 ADMINISTRATIVE CONTROLS6.6 REPORTABLE EVENT ACTION6.6.1 The following actions shall be taken for REPORTABLE EVENTS:a. The Commission shall be notified and a report submitted pursuant to therequirements of Section 50.73 to 10 CFR Part 50, andb. Each REPORTABLE EVENT shall be reviewed by the PORC and submitted to theNRB, Plant Manager and the Vice President, Limerick Generating Station.6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. The NRC Operations Center shall be notified by telephone as soon as possibleand in all cases within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The Vice President, Limerick Generating

Station, Plant Manager, and the NRB shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .b. A Safety Limit Violation Report shall be prepared.

The report shall bereviewed by the NRB. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation uponunit components,

systems, or structures, and (3) corrective action taken toprevent recurrence.

c. The Safety Limit Violation Report shall be submitted to the Commission, theNRB, Plant Manager, and the Vice President, Limerick Generating Station,within the 14 days of the violation.

LIMERICK

-UNIT 26-12aAmendment No. 0, 40, 4-1--, 138 INTENTIONALLY LEFT BLANK ADMINISTRATIVE CONTROLSSAFETY LIMIT VIOLATION (Continued)

d. Critical operation of the unit shall not be resumed until authorized bythe Commission.

6.8 PROCEDURES AND PROGRAMS6.8.1coveriWritten procedures shall be established, implemented, and maintained ng the activities referenced below:a. The applicable procedures recommended in Appendix A of Regulatory Guide1.33, Revision 2, February 1978.b. The applicable procedures required to implement the requirements ofNUREG-0737 and Supplement 1 to NUREG-0737.

c. Refueling operations.

d. Surveillance and test activities of safety-related equipment.

e. Security Plan implementation.

f. Emergency Plan implementation.

g. Fire Protection Program implementation.

h. PROCESS CONTROL PROGRAM implementation.

i. OFFSITE DOSE CALCULATION MANUAL implementation.

j. Quality Assurance Program for effluent and environmental monitoring, using the guidance of Regulatory Guide 4.15, February 1979.The information from Section 6.8.2 has been relocated to the QATR.The information from Section 6.8.3 has been relocated to the QATR.6.8.26.8.3LIMERICK

-UNIT 26-13Amendment No. 4-9, 40, 4921, 138 ADMINISTRATIVE CONTROLSPROCEDURES AND PROGRAMS (Continued) 6.8.4 The following programs shall be established, implemented, and maintained:

a. Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outsidecontainment that could contain highly radioactive fluids during aserious transient or accident to as low as practical levels. Thesystems include the core spray, high pressure coolant injection, reactor core isolation

cooling, residual heat removal, post-accident sampling system (until such time as a modification eliminates the PASSsystem as a potential leakage path), safeguard piping fill system, controlrod drive scram discharge system, and containment air monitor systems.

Theprogram shall include the following:

1. Preventive maintenance and periodic visual inspection requirements, and2. Integrated leak test requirements for each system at refueling cycle intervals or less.b. In-Plant Radiation Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accidentconditions.

This program shall include the following:

1. Training of personnel,

2. Procedures for monitoring, and3. Provisions for maintenance of sampling and analysis equipment.

c. DELETEDLIMERICK

-UNIT 26-14Amendment No. 129 ADMINISTRATIVE CONTROLSPROCEDURES AND PROGRAMS (Continued)

d. Radioactive Effluent Controls ProgramA program shall be provided conforming with 10 CFR 50.36a for thecontrol of radioactive effluents and for maintaining the doses toMEMBERS OF THE PUBLIC from radioactive effluents as low asreasonably achievable.

The program (1) shall be contained in theODCM, (2) shall be implemented by operating procedures, and (3)shall include remedial actions to be taken whenever the programlimits are exceeded.

The program shall include the following elements:

1) Limitations on the operability of radioactive liquid andgaseous monitoring instrumentation including surveillance testsand setpoint determination in accordance with the methodology in the ODCM,2) Limitations on the concentrations of radioactive materialreleased in liquid effluents to UNRESTRICTED AREAS conforming to10 times the concentration values in 10 CFR Part 20, Appendix B,Table 2, Column 2,3) Monitoring,

sampling, and analysis of radioactive liquid andgaseous effluents in accordance with 10 CFR 20.1302 and with themethodology and parameters in the ODCM,4) Limitations on the annual and quarterly doses or dosecommitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conforming to Appendix I to 10 CFR Part 50,5) Determination of cumulative dose contributions fromradioactive effluents for the current calendar quarter andcurrent calendar year in accordance with the methodology andparameters in the ODCM at least every 31 days. Determination of projected dose contributions from radioactive effluents inaccordance with the methodology in the ODCM at least every 31days,6) Limitations on the operability and use of the liquid andgaseous effluent treatment systems to ensure that theappropriate portions of these systems are used to reducereleases of radioactivity when the projected doses in a 31-dayperiod would exceed 2 percent of the guidelines for the annualdose or dose commitment conforming to Appendix I to10 CFR Part 50,7) Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at orbeyond the SITE BOUNDARY shall be limited to the following:

a. For noble gases: less than or equal to 500 mrem/yr to the totalbody and less than or equal to 3000 mrem/yr to the skin, andb. For iodine-131, iodine-133,

tritium, and all radionuclides inparticulate form with half-lives greater than 8 days: less thanor equal to 1500 mrem/yr to any organ.LIMERICK

-UNIT 26-14aAmendment No. -4, 4-4-8, 158 1 ADMINISTRATIVE CONTROLSPROCEDURES AND PROGRAMS (Continued)

8) Limitations on the annual quarterly air doses resulting fromnoble gases released in gaseous effluents from each unit toareas beyond the SITE BOUNDARY conforming to Appendix I to10 CFR Part 50,9) Limitations on the annual and quarterly doses to a MEMBER OFTHE PUBLIC from Iodine-131, Iodine-133,

tritium, and allradionuclides in particulate form with half-lives greater than8 days in gaseous effluents released from each unit to areasbeyond the SITE BOUNDARY conforming to Appendix I to10 CFR Part 50,10) Limitations on venting and purging of the Mark II containment through the Standby Gas Treatment System to maintain releasesas low as reasonably achievable, and11) Limitations on the annual dose or dose commitment to any MEMBEROF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.e. Meteorological Monitoring ProgramA program shall be provided to provide meteorological information inthe environs of the plant. The program shall provide sufficient meteorological data for estimating potential radiation doses to thepublic.The program shall (1) be contained in the ODCM, (2) conform to theguidance of Regulatory Guide 1.23, "Safety Guide 23 -Onsite Meteoro-logical Program",

and (3) include limitations on the operability ofmeteorological monitoring instrumentation including surveillance tests in accordance with the methodology in the ODCM.f. Radiological Environmental Monitoring ProgramA program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) represen-tative measurements of radioactivity in the highest potential exposurepathways, and (2) verification of the accuracy of the effluent moni-toring program and modeling of environmental exposure pathways.

Theprogram shall (1) be contained in the ODCM, (2) conform to the guidanceof Appendix I to 10 CFR Part 50, and (3) include the following:

1) Monitoring,

sampling, analysis, and reporting of radiation andradionuclides in the environment in accordance with themethodology and parameters in the ODCM,2) A Land Use Census to ensure that changes in the use of areas atand beyond the SITE BOUNDARY are identified and thatmodifications to the monitoring program are made if required bythe results of this census, and3) Participation in a Interlaboratory Comparison Program to ensurethat independent checks on the precision and accuracy of themeasurements of radioactive materials in environmental samplematrices are performed as part of the quality assurance programfor environmental monitoring.

LIMERICK

-UNIT 26-14bAmendment No. 11 I ADMINISTRATIVE CONTROLSPROCEDURES AND PROGRAMS (Continued)

g. Primary Containment Leakage Rate Testing ProgramA program shall be established to implement the leakage rate testing of thecontainment as required by 10 CFR 50.54 (o) and 10 CFR 50, Appendix J,Option B, as modified by approved exemptions.

This program shall be inaccordance with the guidelines contained in Regulatory Guide 1.163'Performance-Based Containment Leakage Test program,"

dated September 1995, as modified by the following exception to NEI 94-01, Rev. 0,"Industry Guideline for Implementing Performance-Based Option of 10 CFR50, Appendix J":a. Section 9.2.3: The first Type A test performed after May 21, 1999shall be performed no later than May 21, 2014.The peak calculated containment internal pressure for the design basis loss ofcoolant accident, Pa, is 44.0 psig.The maximum allowable primary containment leakage rate, La, at Pa, shall be0.5% of primary containment air weight per day.Leakage rate acceptance criteria are:a. Primary Containment leakage rate acceptance criterion is less than orequal to 1.0 L,. During the first unit startup following testing inaccordance with this program, the leakage rate acceptance criteria areless than or equal to 0.60 La for the Type B and Type C tests and lessthan or equal to 0.75 La for Type A tests;b. Air lock testing acceptance criteria are:1) Overall airlock leakage rate is less than or equal to 0.05 L,when tested at greater than or equal to Pa.2) Seal leakage rate is less than or equal to 5 scf per hour whenthe gap between the door seals is pressurized to 10 psig.The provisions of Specification 4.0.2 do not apply to the test frequencies specified in the Primary Containment Leakage Rate Testing Program.The provisions of Specification 4.0.3 are applicable to the tests described in the Primary Containment Leakage Rate Testing Program.h. Technical Specifications (TS) Bases Control ProgramThis program provides a means for processing changes to the Bases of theseTechnical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.b. Licensees may make changes to Bases without prior NRC approval providedthe changes do not require either of the following:

A change in the TS incorporated in the license; orA change to the UFSAR or Bases that requires NRC approval pursuant to 10 CFR50.59.c. The Bases Control Program shall contain provisions to ensure that theBases are maintained consistent with the UFSAR.d. Proposed changes that meet the criteria of b. above shall be reviewedand approved by the NRC prior to implementation.

Changes to the Basesimplemented without prior NRC approval shall be provided to the NRC ona frequency consistent with 10 CFR 50.71(e).

LIMERICK

-UNIT 26-14cAmendment No. 81, 4-2-4, 151 ADMINISTRATIVE CONTROLSPROCEDURES AND PROGRAMS (Continued)

i. Battery Monitoring and Maintenance ProgramThis Program provides for restoration and maintenance, based on therecommendations of IEEE Standard 450, "IEEE Recommended Practice forMaintenance,

Testing, and Replacement of Vented Lead-Acid Batteries ForStationary Applications,"

of the following:

a. Actions to restore battery cells with float voltage < 2.13 volts,andb. Actions to equalize and test battery cells that have beendiscovered with electrolyte level below the minimum established design limit.j. Surveillance Frequency Control ProgramThis program provides controls for Surveillance Frequencies.

The Programshall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure theassociated Limiting Conditions for Operation are met.a. The Surveillance Frequency Control Program shall contain a list ofFrequencies of those Surveillance Requirements for which theFrequency is controlled by the program.b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies,"

Revision 0.c. The provisions of Surveillance Requirements 4.0.2 and 4.0.3 areapplicable to the Frequencies established in the Surveillance Frequency Control Program.LIMERICK

-UNIT 26- 14dAmendment No. 4-2-6, 147 ADMINISTRATIVE CONTROLS6.9 REPORTING REQUIREMENTS ROUTINE REPORTS6.9.1 In addition to the applicable reporting requirements of Title 10, Codeof Federal Regulations, the following reports shall be submitted to the RegionalAdministrator of the Regional Office of the NRC unless otherwise noted.STARTUP REPORT6.9.1.1 A summary report of plant startup and power escalation testing shallbe submitted following (1) receipt of an Operating

License, (2) amendment tothe license involving a planned increase in power level, (3) installation offuel that has a different design or has been manufactured by a different fuelsupplier, and (4) modifications that may have significantly altered the nuclear,thermal, or hydraulic performance of the unit.6.9.1.2 The startup report shall address each of the tests identified in Sub-section 14.2.12 of the Final Safety Analysis Report and shall include a descrip-tion of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with designpredictions and specifications.

Any corrective actions that were required toobtain satisfactory operation shall also be described.

Any additional specificdetails required in license conditions based on other commitments shall beincluded in this report.6.9.1.3 Startup reports shall be submitted within (1) 90 days following comple-tion of the startup test program, (2) 90 days following resumption or commence-ment of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest.

If the startup report does not cover all three events(i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial operation) supplementary reports shall be submitted at least every 3 months until all three events have been completed.

ANNUAL REPORTS*6.9.1.4 Annual reports covering the activities of the unit as described belowfor the previous calendar year shall be submitted prior to March 1 of each yearunless otherwise noted.6.9.1.5 Reports required on an annual basis shall include:a. Deleted*A single submittal may be made for a multiple unit station.LIMERICK

-UNIT 26-15Amendment No. 4-4-0, 137 ADMINISTRATIVE CONTROLSANNUAL REPORTS (Continued)

b. (Deleted)

c. Any other unit unique reports required on an annual basis.d. The results of specific activity analysis in which the primarycoolant exceeded the limits of Specification 3.4.5. The following information shall be included:

(1) Reactor power history starting48 hours prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit wasexceeded and results of one analysis after the radioiodine activitywas reduced to less than limit. Each result should include date andtime of sampling and the radioiodine concentrations; (3) Cleanupsystem flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample inwhich the limit was exceeded; (4) Graph of the 1-131 concentration andone other radioiodine isotope concentration in microcuries per gramas a function of time for the duration of the specific activity abovethe steady-state level; and (5) The time duration when the specificactivity of the primary coolant exceeded the radioiodine limit.MONTHLY OPERATING REPORTS*6.9.1.6 DeletedANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT*6.9.1.7 The Annual Radiological Environmental Operating Report coveringthe operation of the unit during the previous calendar year shall be submitted before May 1 of each year. The initial report shall be submitted prior toMay 1 of the year following initial criticality.

The report shall includesummaries, interpretations, analysis of trends of the results of theRadiological Environmental Monitoring Program for the reporting period. Thematerial provided shall be consistent with the objectives outlined in (1) theODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.*A single submittal may be made for a multiple unit station.LIMERICK

-UNIT 26-16Amendment No. 44, 4-!0-, 4-9-, 137 ADMINISTRATIVE CONTROLSANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT*6.9.1.8 The Annual Radioactive Effluent Release Report covering theoperation of the unit during the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. Thereport shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.The material provided shall be consistent with the objectives outlinedin the ODCM and Process Control Program and in conformance with 10 CFR50.36a and 10 CFR Part 50, Appendix I, Section IV.B.I.*A single submittal may be made for a multiple unit station.

The submittal should combine those sections that are common to all units at the station.LIMERICK

-UNIT 26-17Amendment No. 4-4, 4-5, 100 INTENTIONALLY LEFT BLANKLIMERICK

-UNIT 26-18Amendment No. 4, 11 I ADMINISTRATIVE CONTROLSCORE OPERATING LIMITS REPORT6.9.1.9 Core Operating Limits shall be established prior to each reloadcycle, or prior to any remaining portion of a reload cycle, and shall bedocumented in the CORE OPERATING LIMITS REPORT for the following:

a. The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) forSpecification 3.2.1,b. MAPFAC(P) and MAPFAC(F) factors for Specification 3.2.1,c. The MINIMUM CRITICAL POWER RATIO (MCPR) for Specification 3.2.3,d. The MCPR(P) and MCPR(F) adjustment factor for specification 3.2.3,e. The LINEAR HEAT GENERATION RATE (LHGR) for Specification 3.2.4,f. The power biased Rod Block Monitor setpoints and the Rod BlockMonitor MCPR OPERABILITY limits of Specification 3.3.6.g. The Reactor Coolant System Recirculation Flow upscale trip setpointand allowable value for Specification 3.3.6,h. The Oscillation Power Range Monitor (OPRM) period based detection algorithm (PBDA) setpoints for Specification 2.2.1,i. The minimum required number of operable main turbine bypass valvesfor Specification 3.7.8 and the TURBINE BYPASS SYSTEM RESPONSE TIMEfor Specification 4.7.8.c.6.9.1.10 The analytical methods used to determine the core operating limitsshall be those previously reviewed and approved by the NRC, specifically thosedescribed in the following documents:

a. NEDE-24011-P-A "General Electric Standard Application for ReactorFuel" (Latest approved revision),

b. NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications,"

August 1996.6.9.1.11 The core operating limits shall be determined such that allapplicable limits (e.g., fuel thermal-mechanical limits, corethermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN,transient analysis limits, and accident analysis limits) of the safetyanalysis are met.6.9.1.12 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to theNRC Document Control Desk with copies to the Regional Administrator andResident Inspector.

SPECIAL REPORTS6.9.2 Special reports shall be submitted to the Regional Administrator of theRegional Office of the NRC within the time period specified for each report.LIMERICK

-UNIT 26-18a Amendment No. 4,49,49,444,4-49, 161 INTENTIONALLY LEFT BLANK ADMINISTRATIVE CONTROLS6.10 DELETEDTHE INFORMATION FROM SECTION 6.10 HAS BEEN RELOCATED TO THE OATRLIMERICK

-UNIT 26-19Amendment No. 9, 138 ADMINISTRATIVE CONTROLS6.11 RADIATION PROTECTION PROGRAM6.11.1 Procedures for personnel radiation protection shall be prepared con-sistent with the requirements of 10 CFR Part 20 and shall be approved, main-tained, and adhered to for all operations involving personnel radiation exposure.

6.12 HIGH RADIATION AREAAs provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controlsshall be applied to high radiation areas in place of the controls required byparagraph 20.1601(a) and (b) of 10 CFR Part 20:6.12.1 High Radiation Areas with dose rates (deep dose equivalent) greaterthan 0.1 rem/hr and not exceeding 1.0 rem/hour (at 30 centimeters from theradiation sources or from any surface penetrated by the radiation):

a. Each accessible entryway to such an area shall be barricaded and conspicuously posted as a High Radiation Area. Suchbarricades may be opened as necessary to permit entry orexit of personnel or equipment.

LIMERICK

-UNIT 26-20Amendment No. 4-4, 00, 138 ADMINISTRATIVE CONTROLSHIGH RADIATION AREA (Continued)

b. Access to, and activities in, each such area shall be controlled by means of a Radiation Work Permit (RWP) or equivalent thatincludes radiation protection instructions, job coverage andmonitoring requirements.

Radiological information (i.e., doserates) is included on the radiation surveys associated with theRWP or equivalent.

c. Individuals qualified in radiation protection procedures andpersonnel continuously escorted by such individuals may be exemptedfrom the requirement for an RWP or equivalent while performing their assigned duties provided that they are following plantradiation protection procedures for entry to, exit from, andwork in such areas.d. Each individual or group entering such an area shall be providedwith or accompanied by one or more of the following:

1. A radiation monitoring device that continuously displaysradiation dose rates in the area ("radiation monitoring and indicating device"),

OR2. A radiation monitoring device with the capability todisplay accumulated dose and which continuously integrates the radiation dose rates in the area andalarms when the device's dose alarm setpoint isreached ("alarming dosimeter"),

OR3. A radiation monitoring device with the capability todisplay accumulated dose and which continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, OR4. A direct reading dosimeter AND:a) A health physics qualified individual (i.e.,qualified in radiation protection procedures) with a radiation dose rate monitoring devicewho is responsible for controlling personnel radiation exposure within the area, ORb) Be under the surveillance, as specified in theRWP or equivalent, by means of closed circuittelevision, of a health physics qualified individual (i.e., qualified in radiation protection procedures),

responsible for controlling personnel radiation exposure in the area.e. Except for individuals qualified in radiation protection procedures, entry into such areas shall be made only afterdose rates in the area have been established and entrypersonnel are knowledgeable of them.LIMERICK

-UNIT 26-20aAmendment No. 100 ADMINISTRATIVE CONTROLSHIGH RADIATION AREA (Continued) 6.12.2 In addition to the requirements of Section 6.12.1, HighRadiation Areas with dose rates (deep dose equivalent) greater than 1.0rem/hour (at 30 centimeters from the radiation source or from any surfacepenetrated by the radiation),

but less than 500 rad/hr (at 1 meter fromthe radiation source or from any surface penetrated by the radiation source) accessible to personnel shall be controlled as follows:a. Each accessible entryway to such an area shall be con-spicously posted as a High Radiation Area and shall beprovided with a locked door, gate, or guard that preventsunauthorized entry, and in addition:

1. All such door and gate keys shall be maintained underthe administrative control of radiation protection supervision.

2. Doors and gates shall remain locked or guarded exceptduring periods of personnel or equipment entry or exit.b. Such individual areas that are within a larger area, suchas containment, that is controlled as a High Radiation Area,where no enclosure exists for purpose of locking and whereno enclosure can reasonable be constructed around theindividual area need not be controlled by a locked door or gate,but shall be barricaded and conspicuously posted as a HighRadiation Area, and a conspicuous, clearly visible flashinglight shall be activated at the area as a warning device.c. Each individual entering such an area shall be provided withor accompanied by one or more of the following:

1. A dose rate survey meter and a radiation monitoring device with the capability to display accumulated doseand an integrating alarm setpoint, OR2. A radiation monitoring device with the capability todisplay accumulated dose and which continuously integrates the radiation dose rates in the area andalarms when the device's dose alarm setpoint isreached ("alarming dosimeter"),

OR3. A radiation monitoring device with the capability todisplay accumulated dose and which continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area AND with the meansto communicate with the individuals in the area, OR4. A direct reading dosimeter AND:LIMERICK

-UNIT 26-21Amendment No. -14, 100 ADMINISTRATIVE CONTROLSHIGH RADIATION AREA (Continued) a) A health physics qualified individual (i.e.,qualified in radiation protection procedures) with a radiation dose rate monitoring device whois responsible for controlling personnel radiation exposure within the area, ORb) Be under the surveillance, as specified in the RWPor equivalent, by means of closed circuittelevision, of a health physics qualified individual (i.e., qualified in radiation protection procedures),

responsible for controlling personnel radiation exposure in the area, and with the means tocommunicate with the individuals in the area.6.13 PROCESS CONTROL PROGRAM (PCP)6.13.1 Changes to the PCP:a. Shall be documented with the following information:

1. Sufficient information to support the change together with theappropriate analyses or evaluations justifying the change(s) andLIMERICK

-UNIT 26-21aAmendment No. 440-, 138 ADMINISTRATIVE CONTROLSPROCESS CONTROL PROGRAM (Continued)

2. A determination that the change did not reduce the overallconformance of the solidified waste product to existingrequirements of Federal, State, or other applicable regulations.

b. Shall become effective upon review and acceptance by the PORC andapproval of the Plant Manager.6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM)6.14.1 Changes to the ODCM:a. Shall be documented with the following information:

1. Sufficient information to support the change together with theappropriate analyses or evaluations justifying the change(s) and2. A determination that the change will maintain the level ofradioactive effluent control required by 10 CFR 20.1302,40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50and not adversely impact the accuracy or reliability ofeffluent, dose, or setpoint calculations.

b. Shall become effective upon review and acceptance by the PORC andthe approval of the Plant Manager.c. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with theAnnual Radioactive Effluent Release Report for the period of thereport in which any change to the ODCM was made. Each change shallbe identified by markings in the margin of the affected pages,clearly indicating the area of the page that was changed, and shallindicate the date (e.g., month/year) the change was implemented.

6.15 (Deleted)

-INFORMATION FROM THIS SECTION RELOCATED TO THE ODCM.6.16 CONTROL ROOM ENVELOPE HABITABILITY PROGRAMA Control Room Envelope (CRE) Habitability Program shall be established andimplemented to ensure that CRE habitability is maintained such that, with anOPERABLE Control Room Emergency Fresh Air Supply (CREFAS)

System, CRE occupants can control the reactor safely under normal conditions and maintain it in a safecondition following a radiological event, hazardous chemical

release, or a smokechallenge.

The program shall ensure that adequate radiation protection isprovided to permit access and occupancy of the CRE under design basis accident(DBA) conditions without personnel receiving radiation exposures in excess of 5rem total effective dose equivalent (TEDE) for the duration of the accident.

Theprogram shall include the following elements:

a. The definition of the CRE and the CRE boundary.

b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance

.LIMERICK

-UNIT 26-22Amendment No. 2,4443-5,4-38,449, 149 ADMINISTRATIVE CONTROLSCONTROL ROOM ENVELOPE HABITABILITY PROGRAM (Continued)

c. Requirements for (i) determining the unfiltered air inleakage past theCRE boundary into the CRE in accordance with the testing methods andat the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at NuclearPower Reactors,"

Revision 0, May 2003, and (ii) assessing CREhabitability at the Frequencies specified in Sections C.1 and C.2 ofRegulatory Guide 1.197, Revision 0.d. Measurement, at designated locations, of the CRE pressure relative toall external areas adjacent to the CRE boundary during thepressurization mode of operation by one train of the CREFAS, operating at the flow rate required by SR 4.7.2.1 c.1, at a Frequency of 24months on a STAGGERED TEST BASIS. The results shall be trended andused as part of the 24 month assessment of the CRE boundary.

e. The quantitative limits on unfiltered air inleakage into the CRE.These limits shall be stated in a manner to allow direct comparison tothe unfiltered air inleakage measured by the testing described inparagraph

c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basisanalyses of DBA consequences.

Unfiltered air inleakage limits forhazardous chemicals must ensure that exposure of CRE occupants tothese hazards will be within the assumptions in the licensing basis.f. The provisions of Specification 4.0.2 are applicable to theFrequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundaryas required by paragraphs c and d, respectively.

LIMERICK

-UNIT 26-23Amendment No. 149 INTENTIONALLY LEFT BLANK

ML14112A474

Contents

Text

Attachment:

Enclosures:

1.0 DEFINITIONS

2.0 INTRODUCTION

Reference:

REFERENCES:

Reference:

SUBJECT:

REFERENCES:

5.2 CONTAINMENT

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