Text

D. C. Cook, Units 1 & 2 - Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity
	ML061570157
Person / Time
Site:	Cook
Issue date:	05/26/2006
From:	Jensen J; Indiana Michigan Power Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	AEP:NRC:6449, TSTF-449
	Download: ML061570157 (107)
	v • d • e

Indiana Michigan Power INDIANA Cook Nuclear Plant MICHIGAN One Cook Place POWER' Bridgman, MI 49106 AERcom A unit of American Electric Power May 26, 2006 AEP:NRC:6449 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P1-17 Washington, DC 20555-0001

SUBJECT:

Donald C. Cook Nuclear Plant Units 1 and 2 Docket Nos. 50-315 and 50-316 Application for Technical Specification (TS) Improvement Regarding Steam Generator Tube Integrity

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, proposes to amend Facility Operating Licenses DPR-58 and DPR-74.The proposed amendment would revise the TS requirements related to steam generator tube integrity.

The change is consistent with Nuclear Regulatory Commission-approved Revision 4 to TS Task Force (TSTF) Standard TS Change Traveler, TSTF-449, "Steam Generator Tube Integrity." The availability of the TS improvement was announced in the Federal Register on May 6, 2005 (70 FR 24126) as part of the consolidated line item improvement process.Enclosure 1 provides an affirmation statement pertaining to this letter. Enclosure 2 provides a description of the proposed change and confirmation of applicability.

Attachments IA and lB provide TS pages marked to show changes for Unit 1 and Unit 2, respectively.

Attachments 2A and 2B provide TS pages with the proposed changes incorporated.

Attachmnwnt 3 provides draft TS Bases for Unit 1 only. Unit 2 TS Bases changes are consistent with Unit 1 draft changes.I&M requests approval of the proposed amendment prior to June 1, 2007, to support implementation prior to the Unit 2 Fall 2007 refueling outage. I&M requests a 60-day implementation period following approval.Copies of this letter and its attachments are being transmitted to the Michigan Public Service Commission and Michigan Department of Environmental Quality, in accordance with the requirements of 10 CFR 50.91.

U. S. Nuclear Regulatory Commission AEP:NRC:6449 Page 2 This letter contains no new regulatory commitments.

Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Supervisor, at (269) 466-2649.Sincerely, Veph N. Jensen Site Vice President KS/dmb

Enclosures:

1. Affirmation

2. Indiana Michigan Power Company's Evaluation of the Proposed Changes Attachments:

IA. Donald C. Cook Nuclear Plant Unit 1 Technical Specification Pages Marked To Show Changes lB. Donald C. Cook Nuclear Plant Unit 2 Technical Specification Pages Marked To Show Changes 2A. Donald C. Cook Nuclear Plant Unit 1 Technical Specification Pages With the Proposed Changes Incorporated 2B. Donald C. Cook Nuclear Plant Unit 2 Technical Specification Pages With the Proposed Changes Incorporated

3. Donald C. Cook Nuclear Plant Unit 1 Technical Specification Bases Pages Marked to Show Changes c: J. L. Caldwell, NRC Region III K. D. Curry, Ft. Wayne AEP, w/o enclosures/attachments J. T. King, MPSC MDEQ- WHMD/RPMWS NRC Resident Inspector P. S. Tam, NRC Washington, DC Enclosure 1 to AEP:NRC:6449 AFFIRMATION I, Joseph N. Jensen, being duly sworn, state that I am Site Vice President of Indiana Michigan Power Company (I&M), that I am authorized to sign and file this request with the Nuclear Regulatory Commission on behalf of I&M, and that the statements made and the matters set forth herein pertaining to I&M are true and correct to the best of my knowledge, information, and belief.Indiana Michigan Power Company VN. Jensen Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THIS __ DAY OF _,2006 My ComotnEyxp blic My Commission Expires ce/1-o/&C22-7 17- .

Enclosure 2 to AEP:NRC:6449 INDIANA MICHIGAN POWER COMPANY'S EVALUATION OF THE PROPOSED CHANGES

Subject:

Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity

1.0 INTRODUCTION

2.0 DESCRIPTION

OF PROPOSED AMENDMENT

3.0 BACKGROUND

4.0 REGULATORY REQUIREMENTS AND GUIDANCE 5.0 TECHNICAL ANALYSIS 6.0 REGULATORY ANALYSIS 7.0 NO SIGNIFICANT HAZARDS CONSIDERATION 8.0 ENVIRONMENTAL EVALUATION 9.0 PRECEDENT

10.0 REFERENCES

Enclosure 2 to AEP:NRC:6449 Page 2

1.0 INTRODUCTION

The proposed license amendment is a request by Indiana Michigan Power Company (I&M) to amend Facility Operating Licenses DPR-58 and DPR-74 for the Donald C. Cook Nuclear Plant (CNP) Units 1 and 2 to revise the requirements in Technical Specifications (TS) related to steam generator tube integrity.

The changes are consistent with Nuclear Regulatory Commission (NRC)-approved TS Task Force (TSTF) Standard TS Change Traveler, TSTF-449, "Steam Generator Tube Integrity," Revision 4. The availability of this TS improvement was announced in the Federal Register (FR) on May 6, 2005 (Reference 1), as part of the consolidated line item improvement process (CLIIP).

2.0 DESCRIPTION

OF PROPOSED AMENDMENT Consistent with the NRC-approved Revision 4 of TSTF-449, the proposed TS changes include: " Revised TS definition of LEAKAGE," Revised TS 3.4.13, Reactor Coolant System Operational Leakage," New TS 3.4.17, Steam Generator (SG) Tube Integrity," Revised TS 5.5.7, SG Program," Revised TS 5.6.7, SG Tube Inspection Report," Revised Table of Contents to reflect the proposed changes above.Proposed changes to the Unit 1 TS Bases are also included in this application.

The proposed changes for the Unit 2 TS Bases are consistent with the proposed changes to the Unit 1 TS Bases. As discussed in the NRC's model safety evaluation (SE) published in Reference 2, adoption of the revised TS Bases associated with TSTF-449, Revision 4 is an integral part of implementing this TS improvement.

The changes to the affected TS Bases pages will be incorporated in accordance with the TS Bases Control Program.

3.0 BACKGROUND

The background for this application is adequately addressed by the Reference 1 NRC Notice of Availability, the Reference 2 NRC Notice for Comment, and TSTF-449, Revision 4.4.0 REGULATORY REQUIREMENTS AND GUIDANCE The applicable regulatory requirements and guidance associated with this application are adequately addressed by the Reference 1 NRC Notice of Availability, the Reference 2 NRC Notice for Comment, and TSTF-449, Revision 4.

Enclosure 2 to AEP:NRC:6449 Page 3 5.0 TECHNICAL ANALYSIS I&M has reviewed the SE published in Reference 2 as part of the CLIIP Notice for Comment.This included the NRC staffs SE, the information provided to support TSTF-449, and the changes associated with Revision 4 to TSTF-449.

I&M has concluded that the justifications presented in the TSTF proposal and the SE prepared by the NRC staff are applicable to CNP Unit 1 and Unit 2, and justify this amendment for the incorporation of the changes to the CNP Unit I and Unit 2 TS.6.0 REGULATORY ANALYSIS A description of this proposed change and its relationship to applicable regulatory requirements and guidance was provided in the Reference I NRC Notice of Availability, the Reference 2 NRC Notice for Comment, and TSTF-449, Revision 4.6.1 Verification and Commitments The following information is provided to support the NRC staff's review of this amendment application:

Plant Name, Unit Number Donald C. Cook, Unit 1 Steam Generator Model(s) Babcock & Wilcox 51R Effective Full Power Years (EFPY) 5.0 of Service for Currently Installed (Projected to end of current Cycle 20 -Sept5.be0 2006)SGs Tubing Material (e.g., 600M, 690TT 600TT, 660TT)Number of Tubes per SG 3496 Number and Percentage of Tubes SG 1 SG2 SG 3 SG4 Plugged in Each SG 2 0 1 1 0.057% 0.00% 0.029% 0.029%Number of Tubes Repaired in Each None (excludes tube plugging)SG Degradation Mechanism(s)

Fan Bar Wear Identified Per SG: 150 gallons per day (gpd) for any SG Total: 600 gpd from 4 SGs Current Primary-to-secondary Leakage is evaluated at what temperature Leakage Limits: condition?

Reactor coolant system (RCS)temperature of 615.2 degrees Fahrenheit

(°F) at the reactor vessel outlet.+

Enclosure 2 to AEP:NRC:6449 Page 4 Plant Name, Unit Number Donald C. Cook, Unit 1 Approved Alternate Tube Repair None Criteria Approved SG Tube Repair None Methods The primary-to-secondary leakage is not to exceed 1 gallon per minute (gpm) for all SGs.Leakage This leakage value is the total leak rate for all SGs assumed in the Control Room Habitability and Offsite dose analyses.Plant Name, Unit Number Donald C. Cook, Unit 2 Steam Generator Model(s) Westinghouse 51F/54F*Effective Full Power Years (EFPY) 10.8 of Service for Currently Installed (As of the end of Cycle 15 -March 2006)SGs Tubing Material (e.g., 600M, 690TT 600TT, 660TT)Number of Tubes per SG 3592 Number and Percentage of Tubes SG 1 SG 2 SG 3 SG 4 Plugged in Each SG 1 5 6 4 0.028% 0.139% 0.167% 0.111%Number of Tubes Repaired in Each None (excludes tube plugging)SG Degradation Mechanism(s)

Support Plate Wear / Foreign Object Wear Identified Per SG: 500 gpd for any SG Total: I gpm total through all SGs Current Primary-to-secondary Leakage Is evaluated at what temperature Leakage Limits: condition?

RCS temperature of 615.2 0 F at the reactor vessel outlet.+Approved Alternate Tube Repair None Criteria (ARC)Approved SG Tube Repair None Methods Enclosure 2 to AEP:NRC:6449 Page 5 Plant Name, Unit Number Donald C. Cook, Unit 2 The primary-to-secondary leakage is not to exceed 1 gpm for all SGs. This leakage value Performance Criteria for Accident is the total leak rate for all SGs assumed in the Leakage Control Room Habitability and Offsite dose analyses.+ Measurements taken at room temperature per TS Bases Surveillance Requirement 3.4.13.2 are adjusted to ensure the TS volumetric leak rate limit is not exceeded at RCS operating temperatures.

The Westinghouse Model 51 design SGs originally installed in CNP Unit 2 were replaced in 1988. The lower assembly (including the tube bundle) was replaced with a lower assembly of a Model 54 design. The upper shell and internals remain the original Model 51 design with upgraded internals.

SG design documents refer to the Unit 2 replacement SG as being Westinghouse Model 51F; however, as a result of the Westinghouse convention where the model number reflects the approximate secondary side surface area of the tubing (i.e. 54,000 square feet), the appropriate reference to the model number for the CNP Unit 2 SGs is now considered to be 54F.7.0 NO SIGNIFICANT HAZARDS CONSIDERATION I&M has reviewed the proposed no significant hazards consideration determination published in Reference 2 as part of the CLIIP. I&M has concluded that the proposed determination presented in the notice is applicable to CNP Unit 1 and Unit 2 and the determination is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.9 1(a).8.0 ENVIRONMENTAL EVALUATION I&M has reviewed the environmental evaluation included in the model SE published in Reference 2 as part of the CLIIP. I&M has concluded that the staff's findings presented in that evaluation are applicable to CNP Unit I and Unit 2 and the evaluation is hereby incorporated by reference for this application.

9.0 PRECEDENT This application is being made in accordance with the CLIIP. I&M is not proposing variations or deviations from the TS changes described in TSTF-449, Revision 4, or the NRC staff's model SE published in Reference

2.

Enclosure 2 to AEP:NRC:6449 Page 6

10.0 REFERENCES

1. Federal Register Notice: Notice of Availability of Model Application Concerning Technical Specification Improvement to Modify Requirements Regarding Steam Generator Tube Integrity Using the Consolidated

'Line Item Improvement Process, published May 6, 2005 (70 FR 24126).2. Federal Register Notice: Notice of Opportunity for Comment on Model Safety Evaluation on Technical Specification Improvement to Modify Requirements Regarding the Addition of LCO 3.4.[17] on Steam Generator Tube Integrity Using the Consolidated Line Item Improvement Process, published March 2, 2005 (70 FR 10298).

Attachment 1A to AEP:NRC:6449 DONALD C. COOK NUCLEAR PLANT UNIT 1 TECHNICAL SPECIFICATION PAGES MARKED TO SHOW CHANGES Table of Contents, Page 2 of 5 Table of Contents, Page 5 of 5 1.1-3 3.4.13-1 3.4.13-2 3.4.13-3 3.4.17-1 (new)3.4.17-2 (new)5.5-5 5.5-6 5.5-7 5.5-8 5.5-9 5.5-10 5.6-4 UNIT I APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Chapter/Specification Paae 3.3 INSTRUMENTATION 3.3.1 Reactor Trip System (RTS) Instrumentation

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

3.3.1-1 Table 3.3.1-1, Reactor Trip System Instrumentation

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

3.3.1-11 3.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation

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

3.3.2-1 Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation

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

3.3.2-7 3.3.3 Post Accident Monitoring (PAM) Instrumentation

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

3.3.3-1 Table 3.3.3-1, Post Accident Monitoring Instrumentation

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

3.3.3-4 3.3.4 Remote Shutdown Monitoring Instrumentation

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

3.3.4-1 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation

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

3.3.5-1 3.3.6 Containment Purge Supply and Exhaust System Isolation Instrumentation

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

3.3.6-1 Table 3.3.6-1, Containment Purge Supply and Exhaust System Isolation Instrumentation

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

3.3.6-4 3.3.7 Control Room Emergency Ventilation (CREV) System Actuation Instrumentation

..........

3.3.7-1 Table 3.3.7-1, CREV Actuation Instrumentation

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

3.3.7-3 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)Lim its .............................................................................................................................

3 .4 .1-1 3.4.2 RCS Minimum Temperature for Criticality

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

3.4.2-1 3.4.3 RCS Pressure and Temperature (P/T) Limits ..................................................................

3.4.3-1 Figure 3.4.3-1, Reactor Coolant System Pressure versus Temperature Limits -Heatup Limit, Criticality Limit, and Leak Test Limit (Applicable for service period up to 32 EFPY) ...........................................................................................................

3.4.3-3 Figure 3.4.3-2, Reactor Coolant System Pressure versus Temperature Limits -, Various Cooldown Rates Limits (Applicable for service period up to 32 EFPY) ........ 3.4.3-4 3.4.4 RCS Loops -MODES 1 and 2 .........................................................................................

3.4.4-1 3.4.5 RCS Loops -MODE 3 ......................................................................................................

3.4.5-1 3.4.6 RCS Loops -MODE 4 ......................................................................................................

3.4.6-1 3.4.7 RCS Loops -MODE 5, Loops Filled ................................................................................

3.4.7-1 3.4.8 RCS Loops -MODE 5, Loops Not Filled ..........................................................................

3.4.8-1 3.4.9 Pressurizer

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

3.4.9-1 3.4.10 Pressurizer Safety Valves ................................................................................................

3.4.10-1 3.4.11 Pressurizer Power Operated Relief Valves (PORVs) ......................................................

3.4.11-1 3.4.12 Low Temperature Overpressure Protection (LTOP) System ...........................................

3.4.12-1 3.4.13 RCS Operational LEAKAGE ............................................................................................

3.4.13-1 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage ..................................................................

3.4.14-1 3.4.15 RCS Leakage Detection Instrumentation

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

3.4.15-1 3.4.16 RCS Specific Activity .........................................................................................................

3.4.16-1 Figure 3.4.16-1, Reactor Coolant DOSE EQUIVALENT 1-131 Specific Activity Limit Versus Percent of RATED THERMAL POWER .........................................................

3.4.16-3 Cook Nuclear Plant Unit I Page 2 of 5 Amendment No. 287 Cook Nuclear Plant Unit 1 Page 2 of 5 Amendment No. 287 UNIT 1 APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Chapter/Specification Pae 5.0 ADMINISTRATIVE CONTROLS 5.1 R esponsibility

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

5.1-1 5.2 O rganization

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

5.2-1 5.2.1 Onsite and Offsite Organizations

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

5.2-1 5.2 .2 U nit S taff ..........................................................................................................................

5.2-1 5.3 Unit Staff Qualifications

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

5.3-1 5.4 P rocedures

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

5.4-1 5.5 Programs and Manuals .......................................................................................................

5.5-1 5.5.1 Offsite Dose Calculation Manual (ODCM) .......................................................................

5.5-1 5.5.2 Leakage Monitoring Program ...........................................................................................

5.5-2 5.5.3 Radioactive Effluent Controls Program ............................................................................

5.5-2 5.5.4 Component Cyclic or Transient Limits .............................................................................

5.5-3 5.5.5 Reactor Coolant Pump Flywheel Inspection Program .....................................................

5.5-4 5.5.6 Inservice Testing Program ...............................................................................................

5.5-4 5.5.7 Steam Generator (SG) Program ......................................................................................

5.5-5 Table 6.5.7 1, Minimum Number of Steam Generatoem to be Inspectcd During....r.i.e In.pection

..........

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

.........5. 9 Table 5.5.7 2, Steam Gcnc-rator (VSO) Tube lnspection

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

5.5 1 5.5.8 Secondary Water Chemistry Program .............................................................................

5.5-74 5.5.9 Ventilation Filter Testing Program (VFTP) .......................................................................

5.5-74 5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Program ...............................

5.5-5 b.44 5.5.11 Diesel Fuel Oil Testing Program ......................................................................................

5.5ý-I u 5.5.12 Technical Specifications (TS) Bases Control Program ....................................................

5.5-414-5 5.5.13 Safety Function Determination Program (SFDP) .............................................................

5.5-1.A-6 5.5.14 Containment Leakage Rate Testing Program ............................

5 5.5-17 5.5.15 Battery Monitoring and Maintenance Program .................................................................

5.5-MF 5.6 Reporting Requirements

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

5.6-1 5.6.1 Occupational Radiation Exposure Report ........................................................................

5.6-1 5.6.2 Annual Radiological Environmental Operating Report .....................................................

5.6-1 5.6.3 Radioactive Effluent Release Report ...............................................................................

5.6-2 5.6.4 Monthly Operating Reports ..............................................................................................

5.6-2 5.6.5 CORE OPERATING LIMITS REPORT (COLR) ...............................................................

5.6-2 5.6.6 Post Accident Monitoring Report ......................................................................................

5.6-4 5.6.7 Steam Generator Tube Inspection Report .......................................................................

5.6-4 5.7 High Radiation Area ....................

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

5.7-1 Cook Nuclear Plant Unit I Page 5 of 5 Amendment No. 287 Definitions 1.1 1.1 Definitions ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable.

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

In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.LEAKAGE LEAKAGE shall be: a. Identified LEAKAGE 1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank, 2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE, or 3. Reactor Coolant System (RCS) LEAKAGE through a steam generator ( to the Secondary System b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE; and c. Pressure Boundary LEAKAGE LEAKAGE (except SG nina toseo' LEAKAGE)through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.Cook Nuclear Plant Unit I 1.1-3 Amendment No. 287 Cook Nuclear Plant Unit 1 1.1-3 Amendment No. 287 RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to: a. No pressure boundary LEAKAGE;b. 0.8 gpm unidentified LEAKAGE;c. 10 gpm identified LEAKAGE; Ld--ll-.3--------------

-. r J ~-'-.m n p fpr.~+,%re

0 C~p'. ,n A Ii ~ 5 fle. 150 gallons per day primary to secondary LEAKAGE through any one EI'i5g r SGl.APPLICABILITY:

MODES 1. 2, 3. and 4.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Unidentified LEAKAGE A.1 Verify source of unidentified 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s> 0.8 gpm. LEAKAGE is not the pressurizer surge line.OR A.2 Reduce unidentified 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months LEAKAGE to within limit.B. Unidentified LEAKAGE B.1 Reduce unidentified 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s> 1.0 gpm. LEAKAGE to < 1.0 gpm.Cook Nuclear Plant Unit I 3.4.13-1 Amendment No. 287 RCS Operational LEAKAGE 3.4.13 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Identified LEAKAGE not C.1 Reduce LEAKAGE to within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months within limits, limits.OR Primar,' to cocon~dar', LEAKA.GE not within limits.D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A, B, AND or C not met.D.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR Pressure boundary LEAKAGE exists.OR SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1 NOTES-E Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Verify RCS operational r is within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months limits by performance of RCS water inventory balance.Cook Nuclear Plant Unit 1 3.4.13-2 Amendment No. 287 RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.13.2 --ac-eTdana e ql ~ ~ ~ ~ ~ ~ ~ wt th Steamdytho~q n~oeS~t generator-f tube integrity is in accordancre with the Steam Gonreator Program.Cook Nuclear Plant Unit 1 3.4.13-3 Amendment No. 287 SG Tube Integrity 3.4.17ý.j i7_EAC.Tbý._cooLANT

$Y.ýTEM (RCS)ý'AA'7- ' ____8te6-rn'

ý6ýýeýner'at'oýlý,(S(:ý)"Tu,6-e-in"te-,grity CO 3.4ý117 SG tubp integrity shall býp alaýNQ--- --- AllSG tub s'-aifig h uberp i riei shaIbe lg ed n aq&qýc ihteSemGnrtrPorm

ýPP'CtA4l CiT-ý:_ MbbES 1 "21 3, and 4, V46 N UIT I CTN_ FE001RED ACTION ý-OgiplE-TIO-N TIM-E--F,---b-n 6 o ý m__ -o-re SG tu__b e-s Verify tLibe)-ni-tegrity of the di-ýatisfying the tube repair tube(s) is'riteria and not plugged -naintained until the next n accordance with the, efueling outage or SGýtearn Generator ube inskpction.

roqr a m, Plug the a fected ube(s) in rior to entering ccordance with the Steam ODE 4 follow-in'9 tor Proqram, ext refueling outage Ir SG tube inspq.ýfjon U,_.__'ý,e'q'_u`Jred'Actio'n an-d Ljý:- aeýin-qdijýj".

0 ur ssociated Completion

.e of Con ion A n Llvmt jo FN D L2 b-e __n_ T60 E E,_!n tT'ý_6L-1 rs 5g SýGý. t zj, i n i ejilyý_91ýiaintained.

Cook Nuclear Plant Unit I 3.4.17-1 SG Tube Integrity 3.4.17ý-OýF-ILýý-NýE-fEC)UR~t~NT*

+'34i'_ '- V r'i'Gtubeýtegly i "aU-od neviht Ptem G~eator Porm Cook Nuclear Plant Unit I 3.4.17-2 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Proaram Steti G neatr P rormsalb esalse an im e etdto nue htS tub inegriyimanandIna l ,thSem Geeator rg r halinldete olo ig rvi50nl Cook Nuclear Plant Unit I 5.5-5 Amendment No. 287 Cook Nuclear Plant Unit 1 5.5-5 Amendment No. 287 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Program (continued) t.41-, 'CI Oereatioa LAKG.Proisonsfo S tue epir rieraTbsfudy nevcispton p C 1 Z -fthe be"'i6G I ingthefjs*r-u l'go tq, Ths rgram providcs rcguirmcnts for ctcamR gcncrator tube sample selectio n an inspctio.

Each steam generator shall be determnined OPERABLE during Cook Nuclear Plant Unit 1 5.5-6 Amendment No. 287 Programs and Manuals 5.5 shutdown by selecating and inspocting at least the minimumn number of steam generators specifled in Table S5.57 1. The steam; genreatOr tube Minimuim sample size, finspection result cGsifctin and the eor:odiato required shall be as specified Ri Table 5.5.7- 2. The -Gnuu.wu inwpection ou.t..m g..c=ator tubes shall be pe- -d at the FrFequencies specified in'SpecificationS 557c AnAd the inspectcd tubes shall be verified acceptable per the awccptauv ee ri GF 01 of SpeKmUaf I 6i.: 7 .a.The tubes selected for each inser.'ice inspection shall include at least 3%4 ot the total numbehPr o-f t-ub-es in -all steamm generators.

The tuber, selected for these inspections shall be sceceted on a random basis except:.1. 'Ace eprine in similar plants with similar water chemictr,'

indcats citial rea t be inspected, then at least 50% of the tubes i~ntaocted shall be from thee= critical areas: 2. Thne first? pcemofo tubes selet~ tor ;eFean inservlce inne ieton (subsequent to the preerie inpetion) of each steam generao shall OGIude-a) All nonplugged tubes that previously had detectable wall penetrations greater than Or equal to 29%;b) Tubes in those areas where experfiencGe has indicated potential p~eblemse a G) A tube inspection pursuant to Specification 5.5.7-.d.4.h) shall be.&pnc~rmon on ea~n selentn tubl. II uny tube does not prit the passage of the eddy current probe for a tubhethis shall be recorded and an adjaent tube shalll be selected and subjected to a tube

3. The tubes selected as the se~cnd andd thirdd samples (if required by Table 5.5.7 2) duinIg each inserice inspection may be subjected to a partial tube inspection provided-a) The tubes selected for there samples include the tubs fro those areas of the tube sheet aIay where tubes with impoections weepeiusly found; and b) The inspections include those portions of the tubes where-imecrfections were ereviously found.Cook Nuclear Plant Unit I 5.5-7 Amendment No. 287 Cook Nuclear Plant Unit 1 5.5-7 Amendment No. 287 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Stoam; Generator (SG) Program; (continued)

b. The resu.ts of each sampleinspect;on Shall be classified into one of the folloWing thrcc eategorics:

99!22M wInsoecieon Results C I Less than 5.% of the total tubes inspected are dcgradcd tubes an. d none of the in .-pected tubes a~e defetve-.C 2 Greater than or equal to 5% and loss than Or equal to 10% of the total tubes inspected are degraded tubes or one or mor~e tubes, but not MOre thAn 1%of the total tubes inspected, are defective.

G 3 1o4re than 10% of the total tubes inspected are degraded tubes or mor~e than 1 % of theinpce tubes are defective.

Note: in all inpcinpeiuly degraded tubes must exhibit significant (greater than Or equal to1%uther wall penetrations to be included in h G.The above required *.se...e iRspection perfo~med at the following Frequencies:.

s of steam qeneratGF tubes shall be 1. The first inser'.'ice inspeetior I-shall be performed after 6 EffcctV;a!endar months Of initial criticalit e Full,1-Q-IFP-r lfn -RnT .. ..............

...replacemnent of steam generators.

Subsequent inser.ie inpetions shall be performned at interv;als of not less than 12-noir monre thfan 21 calendar months aftef the pr-evious inspecton.

if tw.o conseautive inspections following service under All V.olatile Treatment renditions, ntincluding the preserVce inspection, result in all inspection results falling Into the C I catogor', or if two consecutive inspecations and no additional degradation has oscurred, the inspecti mnay be e~dended to a maximum of once per 40 moneths.onFFrequeinev

-2. If. the results inservice inspectian of a steam gen;erator conducted i ac~ordancGe with Table 5.5.7 2 at 40 month intew.als fall in Catogor; C 3, the inspeefion rFrequency shall be inrGeased to once per 20 mneths. The inrGease in inspection Frequency shall apply until a subsequent inspection satisfies the of Speclfcatio c at which time the rFrequencGy m~ay be e~dended to a maximum Of once per 40 months;, and Cook Nuclear Plant Unit 1 5.5-8 Amendment No. 287 Cook Nuclear Plant Unit 1 5.5-8 Amendment No. 287 Programs and Manuals 5.5~ rrograms ana manuals.AA.b.b4 bteam uvneawo (SG~) i'roaram (continued)

3. AdditioRal.

uRsrheduled rnser, ,;e inspertions shall be pr-formcd on da.ne with the first sample in.pecti.

n eacah steam generator in :-QACVrg sptea ine R Table 6.5. 2 auwing the snutdown subsequent to any-of the following conditions:

a) to tube leaks (not includ;ng leaks -riginating fromn tube to tube sheet welds) in eXceSS GOf the limisat b) A seism~ic occurrencGe greater than the Operating Basis Ea~thquae G) A less Of coolant accidcnt rcquiFrng actuation of the engineered safet'y features; o 14 m ain Rc;laml o'r fe'awa'er line P'eaK.d. Acceptance Criteria I. As used in this Specification:

a) imporfection mneans an exception to the dimensions, finisho contourIF Of -Atube from that Fequired by fablllatiol drawings or specifications.

Eddy current testing indications below 20% of the noiAl tu be wall1 thickness,, if detectable, ma" be considered as W aC..u1.1jL:I

i A=lL = 90i11iu 10HUMUcj cricr=MrIM.

wastageou i==F nrcorrosn ourrig on eithe ;r ;-;;d( or outside of a tube;v.G) Degraded Tube men animerfetion greater than or equal to 20% ofthe nomlinaml wall t-hickness caused by degradation:

d) Percent Degradation moans the percentage of the tube wl thickness afce o eoe by do gradatio;e) P@pr mneans an mefeto of such severit; that it exceeds the plugging limit. A tube containing a defect is defective;, SPlugging Limit m~ean;s the imperfection depth at or beyond which the tube shall be removed from sricde. Any tube which, upon inseto, exhibits tube wall degradation of 40% or mr~ee of the nominal tube wall thickness shall be plugged prior to return;ing the steam aenergatr-to sewlie:=Cook Nuclear Plant Unit I 5.5-9 Amendment No. 287 Cook Nuclear Plant Unit I 5.5-9 Amendment No. 287 Programs and Manuals 5.5 Ia.o I-'oaa Iana

/1InuIs"I' 5.5.7 Steam Geneto (S) rarm (continued) g) Uservicablc deScribes the condition of a tube if it leaks or c.ntains a defeGt large enough to affect its stru.tural integrity in the event of an Opcratinig Basis Earthquake, a less of coolant accident, Or a m~ain steam Wle or fecdwater line break, as pcified OFn Spccification; 5.5.7-.r..3 above;h) Tube inspection; menF n inpetion; of the steam gen;erator tube from; the point of enr, (oleside)

GOmplctcly aroun~d the U bond to the top support to the cold leg- and v-i) Preserwice Inspection mneans an inspection of the full lengho each tube fin the steam gcncr~atOr pcrformod by eddy current tec~hniques prior to sewrvie to establish a baseline condition of the tubing. This inspectio shall be performned after the field hydrostaticG test and prior to initial en;t~' into- M.ODE 1 using the eqiment and techniques expected to be used during 2. The steam generator shall be determined OPE-R-ABLE after coempleting the corresponding actions (plugging all tubes exceeding the plugging limit and all tubes containing through wall c~raks) required by# AM A A A i AM A A A II AA noe provisions of SR 3.. and SI 3.0.3 are appiica~e to thle SG WF9gram test FFequ Cook Nuclear Plant Unit I 5.5-10 Amendment No. 287 Cook Nuclear Plant Unit 1 5.5-10 Amendment No. 287 Programs and Manuals 5.5 Table 5.5.7 1 (page 1 of 1)Minimum ,-umber of tteam Genreatore to be inspected .uring ,,ser-W'ie inspection Prcscr~ioa inspcctIon yes Numnber of Steam GcneratGrS per Un~it 4 First lnser'..ie Inspectien 2 Second and Subsequent

!nser:ice 1nspections W _(a) The thir-d and fourth steam generatorS not inspected duFrin the first isr.ieiseto shall be inspected durin~g the second and third inspections, rospecti':el.Tefuh and subsequent inspeGWtions a" be limited to one steam generator on; a rotating Schedule encomass;g 3 N% of the tubes (Where- N' is tho number of steam generatbr-s in the plant)!--J !--=L-- LL --L --II --t*7 Ynfl rflr~. sing fly ynfl vire.. fir r.rfl~,.flI sin .ne.r.flflr.flme rr.flrfl.~Yn

,fl~ ~ nanI~rflYflr~

nr r ........... .........-vs ..conmDins in one Or more steam generators may be Iound to be mere severe than those in w other steam generators.

U nder sucih circumstances the sample scguence shallb vJ .....modmedC to incspecth inc -Y m swcovoro nconoimens.

Cook Nuclear Plant Unit 1 5.5-11 Amendment No. 287 Cook Nuclear Plant Unit 1 5.5-11 Amendment No. 287 Programs and Manuals 5.5 Tnpfi;7- 1 Ia', -I of '" sw~am (cieenaWo (SG) T ube hnspection First Sample In;specto Second Sample InspcctionR Thii4Sam~ple Reuie Reuie Result A miniFAmF peFS NAoe NA NA NA NA 4 .4- 4 4~- e.fenstve tubes-R insperA tubes4R in None NA NA PG4fet, a 4 Noele tuibes aRd 4 u ad-fion-ak 49 tubes tu~bes n4his Q4 e9F Perform of firet sample.NA NA-4 1 --t *1-tubes iR4i tubes, iepeet 28 tubee e--nhath-ef SG, aRd noif NR p~wsuant to Spe~ifieG Allethtz SGs aee Nehe NA NA SemeSGs Perform action NA NA a~ , f9F G 2 additional sampe SGs-are Q43 Additeenal SG is G tubes in ear*FepaiF defeetive tubes, and RetifyNRG SperGifiration NA NA Where. S -3 (NWn)%;, N isthe numbeF of SGs in the unit; and n isthe number Of SGS inspected durn aninpetion.

Cook Nuclear Plant Unit 1 5.5-12 Amendment No. 287 Reporting Requirements 5.6 I 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.5.6.6 Post Accident Monitoring Report When a report is required by Condition B or H of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.Steam Generator Tube Inspection Report 5.6.7 T esc60-e-oi- -Sr)e--ti

-- for e doneo hSG Fa.-- o 'naithin 1 dac~fe follo ing -inthe Icmpluet Io of .jied ch icr iekohd. ipecr -tion o gcneatioorienhatio be f repo ar),d to d thsue d NRCaaiale fsevc b.~ Th copetei rcult of th sa eeao tb ne~c iccto hl bersubmte tof the NRCprior to Mrinch 1 for the inspection that wa~lsýý n completed in the presiuts cfte taenda geerar. thie rzo hal nlue Cook Nuclear Plant Unit 1 5.6-4 Amendment No. 2-Sq-, 288 Reporting Requirements 5.6 1. Number and extent of tubecs inspc*ted:;

2. Location and percGent of of an; imnperftion;j and Wall thickncss penetration for each Rndication 2- Id-entification of tubes, pluoed.A. Results of steam genreator tub9 inepections Which fall into CategoryC shall be reported to the NRC in accordance with 10 GCFR 50.72. A L~icens6o Event Report shall be submitted in accordance with 1 0 C FR 5-0.7-3 and shall providc a deccription of investigatines conducted to determine the cause oG the tube degradation and corrective m~easurec taken to prevent recuF~rren.

Cook Nuclear Plant Unit I 5.6-5 Amendment No. 2-87,288 Attachment 1B to AEP:NRC:6449 DONALD C. COOK NUCLEAR PLANT UNIT 2 TECHNICAL SPECIFICATION PAGES MARKED TO SHOW CHANGES Table of Contents, Page 2 of 5 Table of Contents, Page 5 of 5 1.1-3 3.4.13-1 3.4.13-2 3.4.17-1 (new)3.4.17-2 (new)5.5-5 5.5-6 5.5r7 5.5-8 5.5-9 5.5-10 5.6-4 UNIT 2 APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Chapter/Specification Page 3.3 INSTRUMENTATION 3.3.1 Reactor Trip System (RTS) Instrumentation

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3.3.1-1 Table 3.3.1-1, Reactor Trip System Instrumentation

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3.3.1-11 3.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation

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

3.3.2-1 Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation

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3.3.2-7 3.3.3 Post Accident Monitoring (PAM) Instrumentation

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

3.3.3-1 Table 3.3.3-1, Post Accident Monitoring Instrumentation

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

3.3.3-4 3.3.4 Remote Shutdown Monitoring Instrumentation

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

3.3.4-1 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation

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

3.3.5-1 3.3.6 Containment Purge Supply and Exhaust System Isolation Instrumentation

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

3.3.6-1 Table 3.3.6-1, Containment Purge Supply and Exhaust System Isolation Instrum entation ..........................................................................................................

3.3.6-4 3.3.7 Control Room Emergency Ventilation (CREV) System Actuation Instrumentation

..........

3.3.7-1 Table 3.3.7-1, CREV Actuation Instrumentation

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

3.3.7-3 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)Lim its .............................................................................................................................

3 .4 .1-1 3.4.2 RCS Minimum Temperature for Criticality

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

3.4.2-1 3.4.3 RCS Pressure and Temperature (P/T) Limits ..................................................................

3.4.3-1 Figure 3.4.3-1, Reactor Coolant System Pressure versus Temperature Limits -Heatup Limit, Criticality Limit, and Leak Test Limit (Applicable for service period up to 32 E FPY ) ...........................................................................................................

3.4 .3-3 Figure 3.4.3-2, Reactor Coolant System Pressure versus Temperature Limits -Various Cooldown Rates Limits (Applicable for service period up to 32 EFPY) ........ 3.4.3-4 3.4.4 RCS Loops -MODES 1 and 2 .........................................................................................

3.4.4-1 3.4.5 RCS Loops -MODE 3 ......................................................................................................

3.4.5-1 3.4.6 RCS Loops -MODE 4 ......................................................................................................

3.4.6-1 3.4.7 RCS Loops -MODE 5, Loops Filled ................................................................................

3.4.7-1 3.4.8 RCS Loops -MODE 5, Loops Not Filled ..........................................................................

3.4.8-1 3.4.9 P ressurizer

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

3.4.9-1 3.4.10 Pressurizer Safety Valves ................................................................................................

3.4.10-1 3.4.11 Pressurizer Power Operated Relief Valves (PORVs) ......................................................

3.4.11-1 3.4.12 Low Temperature Overpressure Protection (LTOP) System ...........................................

3.4.12-1 3.4.13 RCS Operational LEAKAGE ............................................................................................

3.4.13-1 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage ..................................................................

3.4.14-1 3.4.15 RCS Leakage Detection Instrumentation...;

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3.4.15-1 3.4.16 RCS Specific Activity .......................................................................................................

3.4.16-1 Figure 3.4.16-1, Reactor Coolant DOSE EQUIVALENT 1-131 Specific Activity Limit Versus Percent of RATED THERMAL POWER .................

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

3.4.16-3 b- ue- L ....... ...:.... ..17~Cook Nuclear Plant Unit 2 Page 2 of 5 Amendment No. 269 Cook Nuclear Plant Unit 2 Page 2 of 5 Amendment No. 269 UNIT 2 APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Chapter/Specification Page 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility

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5.1-1 5.2 Organization

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

5.2-1 5.2.1 Onsite and Offsite Organizations

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5.2-1 5.2.2 Unit Staff ..........................................................................................................................

5.2-1 5.3 Unit Staff Qualifications

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5.3-1 5.4 Procedures

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5.4-1 5.5 Programs and Manuals .......................................................................................................

5.5-1 5.5.1 Offsite Dose Calculation Manual (ODCM) .......................................................................

5.5-1 5.5.2 Leakage Monitoring Program ...........................................................................................

5.5-2 5.5.3 Radioactive Effluent Controls Program ............................................................................

5.5-2 5.5.4 Component Cyclic or Transient Limits ............................................................................

5.5-3 5.5.5 Reactor Coolant Pump Flywheel Inspection Program .....................................................

5.5-4 5.5.6 Inservice Testing Program ..............................................................................................

5.5-4 5.5.7 Steam Generator (SG) Program .....................................................................................

5.5-5 Table 5.5.7 1, Minimum Number of Steam Gcnerators to be inspected DuFrig , Rser.'ic lncpectieR....

......n***...............

- - - *'* ................

- ... -*........

5.5 9 Tablo 5.54 2 7 2 Steamr (SG) Tube Inhep cti ..,...-.........................

5.5 10 5.5.8 Secondary W ater Chemistry Program .............................................................................

5.5-74 -5.5.9 Ventilation Filter Testing Program (VFTP) .......................................................................

5.5-17 5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Program ..............................

5.5-'04 5.5.11 Diesel Fuel Oil Testing Program ......................................................................................

5.-,1 44 5.5.12 Technical Specifications (TS) Bases Control Program ..............................................

5 5.5-5 45 5.5.13 Safety Function Determination Program (SFDP) ......................................................

5 5. -I 4-6 5.5.14 Containment Leakage Rate Testing Program ..................................................................

5.5-,--7 5.5.15 Battery Monitoring and Maintenance Program .............................................................

5... 5.5 u 5.6 Reporting Requirements

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5.6-1 5.6.1 ' Occupational Radiation Exposure Report ........................................................................

5.6-1 5.6.2 Annual Radiological Environmental Operating Report ....................................................

5.6-1 5.6.3 Radioactive Effluent Release Report ...............................................................................

5.6-2 5.6.4 Monthly Operating Reports .............................................................................................

5.6-2 5.6.5 CORE OPERATING LIMITS REPORT (COLR) ..............................................................

5.6-2 5.6.6 Post Accident Monitoring Report .....................................................................................

5.6-4 5.6.7 Steam Generator Tube Inspection Report ......................................................................

5.6-4 5.7 High Radiation Area ............................................................................................................

5.7-1 Cook Nuclear Plant Unit 2 Page SofS Amendment No. 269 Cook Nuclear Plant Unit 2 Page 5 of 5 Amendment No. 269 Definitions 1.1 1.1 Definitions ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable.

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

In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.LEAKAGE LEAKAGE shall be: a. Identified LEAKAGE 1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank, 2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE, or 3. Reactor Coolant System (RCS) LEAKAGE through a steam generator (SG} to the Secondary System_i to- secon-r a .G )b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE; and c. Pressure Boundary LEAKAGE LEAKAGE (except SG E f6 senaL-ry LEAKAGE)through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.Cook Nuclear Plant Unit 2 1.1-3 Amendment No. 269 RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to: a. No pressure boundary LEAKAGE;b. 1 gpm unidentified LEAKAGE;c. 10 gpm identified LEAKAGE; EL d. 1 gpm total prfimar', to cccondary LEr=A.K-A.GE through all stem generatorc (SGs); and Re. MW gallonsper day primary to secondary LEAKAGE through any one Lt6 m tor(SG1.APPLICABILITY:

MODES 1, 2, 3, and 4.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RCS [p-e-ioni A.1 Reduce LEAKAGE to within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months LEAKAGE not within limits.limits for reasons other than pressure bounda LEAKAGE p-rimaiyt Le-on'aryiIEAKAGE.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not AND met.B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR Pressure boundary LEAKAGE exists.OR[EAKAGE notwithin

__________

______Cook Nuclear Plant Unit 2 3.4.13-1 Amendment No. 269 RCS Operational LEAKAGE 3.4.13 CONDITION REQUIRED ACTION COMPLETION TIME Cook Nuclear Plant Unit 2 3.4.13 -2 Amendment No. 269 RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1 Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Verify RCS operational

=leakage is within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months limits by performance of RCS water inventory balance.SR 3.4.13.2aodane with thc Steam generator tube integrity is in; acrdance with the Steam Generator Program.Cook Nuclear Plant Unit 2 3.4.13-3 Amendment No. 269 SG Tube Integrity 3.4.17 0-4--,'St-e--a'

... m G e"n- --- e- r -a t o r (S G) --f 66 e n- fý'g-hfy'G tube intýýjrityshall bp ma Liýiýb-ýA-A'7-S int-AND Fpp0ýAODf'Y:'

1,'2, 3,-an'd-4, V a Y-I --;.- -A, , , -ý -, -i .. ....[A. Z rIUY LI1U alleULCU ILJDUýýS)

III ccordance with the Steam enerator Prog ra.m,, FOh-b ,- neg faitaind Cook Nuclear Plant Unit 2 3.4.17-1 SG Tube Integrity 3.4.17 LURVEILL -ANCE RE .0 1 U1 I REMEN , TSýUkVýJICLýNCE rftQANCY?pce with the ance rR ---V-e-r-iý'ý6'-f-u'b"e'i'n'teg'rit-y' i'n"-a'c6o-rd-Fac6-6rd.team Ge-nerator-Program., 6th the Steamýenerator'

ýrogram F-R -7 'Ve'r'if"y-fh--a-'t e'a'-c'h' i 'ns*pe'-c't-e"d-S'G--t'Li-b-e-t-h'-a--t-'s'a-ti'sfie"-s--t-h-e-io f6 -ente ring 1ýbe repair criteria is PlUgged in accordance with. the 0 DE 4 following pý'teqm Generator Prq FSG tube_gram. inspection Cook Nuclear Plant Unit 2 3.4.17-2 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Steam Gener ator (SG) Programý~~ ýearn6en'afý-5gramsa11be salshd-ar ipe-n n'e"'toe'sure ha Gtb n tgiys manand S I n diin h tamGeea rga Cook Nuclear Plant Unit 2 5.5-5 Amendment No. 269 Cook Nuclear Plant Unit 2 5.5-5 Amendment No. 269 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Program (continued) aoiin forSG tuerpiartri.T bsfudb ise~ieiset t D0oti lw ihadpheultoo xedn Qý,o h o ia ubewal ti-cness~ha i1epýge, I 9~'p-e' -f d fthe u'b ' i'eac 'G'd'i This program; proVidcs reguircmcnts for steam gcrcrator tube sample selcstion and inspcction.

Each steam genrA-a-tor shall bce dctcrm~ncd OPER ABLE drmUing Cook Nuclear Plant Unit 2 5.5-6 Amendment No. 269 Programs and Manuals 5.5 shutdown by selecting and inspeiting at least the meinimum; number of steam generators specified in Table 5.5.7- 1. The steam generator tube minimu sample size, inspection reSult classification, and the crrsFsponding action requredhall be as spec.ified in Table 5.5.7 2. The inser':icc inspection; Of steamA geerto tubes shall be performned at the Frqnce specified in Spcfication; 5.5.7-.G and the inspec-tedd tb--es slhffalll be veriifiled acceptable per the accepance criteria of Specification 5.5.74d-A. The tub,-es selected for each iericinptonshall inclu--de at least 3%o the total number of tubes in all steam generators.

The tubes selected for these inspections shall be selected on -a random-P basis except: 1. experience in similar srirn;lr water heMistr;, indicates ctical afreas to be inspccted, then at least 5t% of the tubes ins~eeted shall be from; these criFtical areas:*~~~~~~~.~~~~~

nofrtcmicO ue cicc o ac neciseto (subsequent to the prese.. r....e ispet.) o each. steam a) All nonplugged tubes that prcviously had detectable wall penetrations greater than or equal to 20%;-b) Tubes in those areas where experieneo has indicated potentia!p Iebiems; apI4 G) A tube inspection pursuant to Specification; 5.5.7.d.1 .h) shall he permnit the passage of the eddycuenprb for a tube inspection, this, shall be recorded an ndaent tube shall be selected and subiected to a tubeineci:

3. The tubes sel.pa~taltbeins I I i reAed as me second anna thlrd- Samples (IT rsguired by during each iRsevrice inspecti-n may be subjected to a npection nmrovided a) The tubes selested for these samples include the tubes from those areas of the tube sheet arly where tubes w imefections were previously found ;and b) The ins:,pections in those pVltir impeI ineR were previously found.s, of the tubes where Cook Nuclear Plant Unit 2 5.5-7 Amendment No. 269 Programs and Manuals 5.5 5.5 Programns and ManualS&.&.7 Steam Generator (SG) Proaramf (GOntinued) bh The results of each sample folloWing three categorie&l:l Ca G.Ineroton Results" 1 Less than 5% of the total tubes inspertcd are degraded tubes and none of the inspected tubes a~e defee~ive-.

C 2 Greater than or equal to 5% an~d less than or equal to I% -of the total tubes inspected are degraded tubes or one or mor~e tubes, but noet more than 1%of the total tubes i ..vpected, are defective.

C 3 ore than; 10% of the total tubes in~spected are degraded tubes or mor~e than 1 % of theinpce tubes are defective.

Note- In all inpcinpeiuly degraded tubes must exhibit significan (gr-eater than or eql to 1% fu rt ther wall penetrations to be inlue i the aoove ocrccntaae calculations.

Q.The above required ;.e m inspection PFrFmed at the following Frequen-ies:

s of steam generator tubes shall be I. The first inr.,vai ,inpectiv r sehallbe perfoFrmed after 6 Effectivep Full Pe-I.r ... M9tH U" t W H... ...ýI Ga"ea"-- -e. -niva:O ; .... ,-a,. tyn-,-i, replacement of steam generatorts.

Subsequent inrie isectfions shall be performedIat intvals of not less than 2IInor morze tha-2 calendar months aftir the previous inspection.

If "Ao consecutive Inspecr.tions follwing senvi-e

.Volatile Tr-eatment c-o-nditions, not inlludIng the preservie inspection, result in all inspetioRn results falling in.to the C 1 +atego,', or if two consecutive inspectionas demon~strate that previously observed degradation has Rot continued and no additional degradation has occurred, the inspertien FrequencY m~ay be e~dended to a maximum of once per 10 months.2. if the results inRser'c inspection of a steam generator conductedi accor~dance with Table 5.5.7 2 at 10 month in;tcr.als fall in Catogor,'

C 3, the inspection FrFequency shall be inrGeased to once per 20 moneths. The inrGease in inspection Frequency shall apply until a subsequent inspection satisfies the criteria of Specification 5.5.7-.G.4, at which time the Frequency may be extended to a maximum of once per 40 months; and Cook Nuclear Plant Unit 2 5.5-8 Amendment No. 269 Cook Nuclear Plant Unit 2 5.5-8 Amendment No. 269 Programs and Manuals 5.5 5.5 Programs and Man.uals 5.6.7- Steam Generator (SG) DProaram (Gontinucd)

3. Additional, unScheduled iico inpectIonsshall be peFormlled on each steam generator in accordancGe with the firSt samplc inspection spccified in Table 5.5.7- 2 dur1Ing the shu'tdoIwn subsequent to an- of tho followi n conditions:I a) Primua; to secondar, tube leaks (not includ ilng leaks originating from tube to tube sheet welds) in excess of thel b) A sciramic occurrencAe greatcr than the Operating Basis Ea~thqakej; c.) A loss of coolant accident requiring actuation of the engineered saft* features; or I! .... ! J I L I.. .... L---- ; .. ....--#, main steam iinc or rccawater nnc DreaK.d. Accrseptance Criteria I A used in this SVqecifIatioV:

a)Imeeroctinmeans an eXception to the dimension~s, finish, or contourofR atubefronmthat required byfabrication drawin~gsor specifiations.

Eddy current testing indications below:20%

of the nom~in~al tub-e wall thickness, if detectable, may be considered as a pwe~et&ie~q hI% pewagavein eanns a secr'Geio 1dUoedi GraGKing.

wastage. war or general corrosion occurring on either inside or otside of a tube-;v.c)- Derae Tbmo iperfection greater than o equal to 20 4f t-he nmalW-all t-hic-kness caused by degradation:

d) Percent Degradation moeans the percentage .of the tube wl thickness affected Or removed by degradation; e) De@fect maeans an imspe~fectien of such severity that it exceeds the plugging limit. A tube containing a defect us defective;, f) Plugging Limit moans the imefcindepth at Or beyond dwhich the tube shall be removed from 50R'iCC. Any tube which, upon inspectioni, exhibits tube wall degradation of 40% or more of thetube Wall thickness shall be plugged prIor to returning the steam geRnFator to serfice;V Cook Nuclear Plant Unit 2 5.5-9 Amendment No. 269 Cook Nuclear Plant Unit 2 5.5-9 Amendment No. 269 Programs and Manuals 5.5 b., IWGgramS anAG Manua.S 5.5.7 Steam Gen;eratOr (8G) Programf (con~tinued) g) Unsewr'.ieable dcScribcs the condition of a tube if it leakso con~tains a defcct large enough to affoct its structural integrity On the event of an Operating Basis Earthquake, a laes of coolant acciden;t, or a main ste-am line Or feedw.fator line br~eak, as"pcfied in Specification; 5.5.7-.G.3 above;h) Tube inspectioni moans an; in~specation; of the steam gen~erator tube from the poin~t of en;try (hot leg side) rompletely, around the U bend to the top support to the cold leg; and v.i) Preserv.'ic Insvection mnean;s an inspection of the full lengho e acGh tub h e inA th e Asteam geeao performed by eddy curr-ent techniques prior to servicne to establish a baseline condition of the tubing. This inspecation; shall be perforFmed after the field-hydFrosatic test and prier to initial entry int MODE 1 using the equipment and techniques expected to be used during subsequenti inse.ie ipections.

2. The steam gen~erator shall be determined OPErRA4.BL-E a;fter com~pleting the corresponding actions (plugging all tubes exceeding the plugging limit and all tubes containing through wall cracks) required by The provisions of SR 3.0.2 and SIR 3.0.3 are applicable to the SG ProGram test Cook Nuclear Plant Unit 2 5.5-10 Amendment No. 269 Cook Nuclear Plant Unit 2 5.5-10 Amendment No. 269 Programs and Manuals 5.5'r-L.1- C C '7 4 / --4 -IV 4 %I/o I II Minimum NUmDcr or ~rcam uenorarorc to ne incoectad vurina IncoR'fcc incoection

.... r...-.... ...--Preservice Inopection

~e NubrofScmGpeaoe pr Unt4 FiSot InseFient

!nsperctieo 2 Second and Subsequent tnser'.ice Inspectmios, ____ __________(a) The third and fourth steam acnreators not insnected duFrin the first ine)ieicectiOn

%v J shall be inspected durOing the second an;d third inispectAions, respectively.

The fourFth and subseeuent inspectionS may be limited to one steeam generator 9n a rotatin;g schedule encomassig 3 N% of the tubes (whore hN is the number of steam generators in the plant)ifte rutsof the firSt Or previu .npetions iniacthat all steam generators are performing in a like mnanner. Note thati under _om circmtances, the operating Genditions in one or mr;Ge steam generators may be found to be mor~e severe than these i otheFr seamn genrwator.

UndeFr such circumstancaes the sample soguen~e shalLtbe v* o moditie to Onspect the mneet severe conditions.

Cook Nuclear Plant Unit 2 5.5-11 Amendment No. 269 Cook Nuclear Plant Unit 2 5.5-11 Amendment No. 269 Programs and Manuals 5.5 Table 5.5.72 (page 1 of 1)Steam Gen;erator (SG) Tube Insepotion Frest Sample Inepeco Socond Sample Inspectionf ThFd Sample Same Sie Result Aetion Result Aten Result A A-minrum G4 NOPe NA NA NA NA G-e Pe-de-feet G4 None NA NA peF-SG tubes-and G-ude e G4 None iepeet tube a-4 add-tifeal 2S inspert defeetive tubes additional 4S tubes ,, tubes-O4Rhs Q4 PeGe SG catiGR for G-3--result-A Q- Perfrm ctine NA NA G4 1spert Allethe NoRe NA NA tubes4R-te SGs-ae defeetive SemeSGs PefGorm actio NA NA tubessperA aeG 2, fer-G-24esult.

284ubes- but- e fer seGRd eaes-ethef additfinal sampe SGr, and...ti.

SGs- e NRG pussuant Q4 _ _tG Additoal Ispee-all NA NA Speifia_ C-2 tubes,, OneaGh&66 SG, pug er fepa&defeetieve tubes, and notify NRG puwsua44 e 8peeifieatioR WAhAe:. S -3 (N!n)%;O N is the lnumber of SG25- the unitN and n *ec the number of SGs inspectAed durn aniepcion.

Cook Nuclear Plant Unit 2 5.5-12 Amendment No. 269 Reporting Requirements 5.6 I 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.5.6.6 Post Accident Monitoring Report When a report is required by Condition B or H of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.Steam Generator Tube Inspection Report 5.6.7... The n eacý SGj pe of inspections perfori-ned Aci-ive',d'eg-rad-ation mech'anisms found, incansm 2k iainjcnqe elii-g fos ac h erdt oatx n ore tati n (if lier, n es dszs(fa aibl)osevc A. V-ihia 1 days following the cimpletion of each insew-iceiinspection ot suea g tubes, the number of tubes plugged i eah steam generator shall be r-eperted to the NRC-.h. The complete results of the steam generator tubc inc'ceiseto shall be subm~ited-to- the NRC- prior to March 1 for thenpecinta was-- I II I -I compicica in me preyleius caicn~aa ycaF. +IG mceport snail inciluac: Cook Nuclear Plant Unit 2 5.6-4 Amendment No. 2460, 270 Reporting Requirements 5.6 I 1. Nwumber and e)ent of tubes incpeted;2. Loc-ation and percent of of an impcrfcction:

and wall .ration for each 2- Ilont-i-ticaln

,? mtu o plugged.G. Results of steam gcncrator tube finspcctianc which fall into Categow' C shall be reportcd to the hIRC in accodance with 1 0 CMR 50.72. A Licencec Event Report shall be submitted in accr.Gdance with 10- CF-R 5-0.7-3 and shall provide a decc~iPtiGR of wnetgtescnductcd to dcteFrminc the cause ot the tube dcgradation and corrective mneasures taken to prcvent recurrcncc.

Cook Nuclear Plant Unit 2 5.6-5 Amendment No. 269, 270 Attachment 2A to AEP:NRC:6449 DONALD C. COOK NUCLEAR PLANT UNIT 1 TECHNICAL SPECIFICATION PAGES WITH THE PROPOSED CHANGES INCORPORATED Table of Contents, Page 2 of 5 Table of Contents, Page 5 of 5 1.1-3 3.4.13-1 3.4.13-2 3.4.13-3 3.4.17-1 (new)3.4.17-2 (new)5.5-5 5.5-6 5.5-7 5.5-8 5.5-9 5.5-10 5.5-11 5.5-12 5.5-13 5.5-14 5.6-4 UNIT I APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Chapter/Specification Paae 3.3 INSTRUMENTATION 3.3.1 Reactor Trip System (RTS) Instrumentation

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3.3.1-1 Table 3.3.1-1, Reactor Trip System Instrumentation

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3.3.1-11 3.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation

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3.3.2-1 Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation

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3.3.2-7 3.3.3 Post Accident Monitoring (PAM) Instrumentation

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3.3.3-1 Table 3.3.3-1, Post Accident Monitoring Instrumentation

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3.3.3-4 3.3.4 Remote Shutdown Monitoring Instrumentation

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3.3.4-1 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation

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3.3.5-1 3.3.6 Containment Purge Supply and Exhaust System Isolation Instrumentation

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3.3.6-1 Table 3.3.6-1, Containment Purge Supply and Exhaust System Isolation Instrum entation ..........................................................................................................

3.3.6-4 3.3.7 Control Room Emergency Ventilation (CREV) System Actuation Instrumentation

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3.3.7-1 Table 3.3.7-1, CREV Actuation Instrumentation

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

3.3.7-3 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)L im its .............................................................................................................................

3 .4 .1-1 3.4.2 RCS Minimum Temperature for Criticality

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3.4.2-1 3.4.3 RCS Pressure and Temperature (P/T) Limits ..................................................................

3.4.3-1 Figure 3.4.3-1, Reactor Coolant System Pressure versus Temperature Limits -Heatup Limit, Criticality Limit, and Leak Test Limit (Applicable for service period up to 32 E FPY ) ...........................................................................................................

3.4 .3-3 Figure 3.4.3-2, Reactor Coolant System Pressure versus Temperature Limits -Various Cooldown Rates Limits (Applicable for service period up to 32 EFPY) ........ 3.4.3-4 3.4.4 RCS Loops -MODES I and 2 .........................................................................................

3.4.4-1 3.4.5 RCS Loops -MODE 3 ......................................................................................................

3.4.5-1 3.4.6 RCS Loops -MODE 4 .....................................................................................................

3.4.6-1 3.4.7 RCS Loops -MODE 5, Loops Filled ................................................................................

3.4.7-1 3.4.8 RCS Loops -MODE 5, Loops Not Filled ..........................................................................

3.4.8-1 3.4.9 Pressurizer

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

3.4.9-1 3.4.10 Pressurizer Safety Valves ................................................................................................

3.4.10-1 3.4.11 Pressurizer Power Operated Relief Valves (PORVs) .....................................................

3.4.11-1 3.4.12 Low Temperature Overpressure Protection (LTOP) System ...........................................

3.4.12-1 3.4.13 RCS Operational LEAKAGE ...........................................................................................

3.4.13-1 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage ..................................................................

3.4.14-1 3.4.15 RCS Leakage Detection Instrumentation

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

3.4.15-1 3.4.16 RCS Specific Activity ......................................

...... 3.4.16-1 Figure 3.4.16-1, Reactor Coolant DOSE EQUIVALENT 1-131 Specific Activity Limit Versus Percent of RATED THERMAL POWER .........................................................

3.4.16-3 3.4.17 Steam Generator (SG) Tube Integrity

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

3.4.17-1 Cook Nuclear Plant Unit 1 Page 2 of 5 Amendment No. 28, Cook Nuclear Plant Unit 1 Page 2 of 5 Amendment No. 2-97-,

UNIT 1 APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Chapter/Specification Page 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility

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5.1-1 5.2 O rganization

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5.2-1 5.2.1 O nsite and Offsite O rganizations

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5.2-1 5.2.2 Unit Staff ..........................................................................................................................

5.2-1 5.3 Unit Staff Q ualifications

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5.3-1 5.4 Procedures

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5.4-1 5.5 Program s and M anuals .......................................................................................................

5.5-1 5.5.1 Offsite Dose Calculation M anual (O DCM ) .......................................................................

5.5-1 5.5.2 Leakage M onitoring Program ...........................................................................................

5.5-2 5.5.3 Radioactive Effluent Controls Program ............................................................................

5.5-2 5.5.4 Com ponent Cyclic or Transient Lim its .............................................................................

5.5-3 5.5.5 Reactor Coolant Pum p Flyw heel Inspection Program .....................................................

5.5-4 5.5.6 Inservice Testing Program ...............................................................................................

5.5-4 5.5.7 Steam G enerator (SG ) Program ......................................................................................

5.5-5 5.5.8 Secondary W ater Chem istry Program .............................................................................

5.5-7 5.5.9 Ventilation Filter Testing Program (VFTP) .......................................................................

5.5-7 5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Program ...............................

5.5-10 5.5.11 Diesel Fuel O il Testing Program ......................................................................................

5.5-10 5.5.12 Technical Specifications (TS) Bases Control Program ....................................................

5.5-11 5.5.13 Safety Function Determ ination Program (SFDP) .............................................................

5.5-12 5.5.14 Containm ent Leakage Rate Testing Program ..................................................................

5.5-13 5.5.15 Battery M onitoring and M aintenance Program .................................................................

5.5-14 5.6 Reporting Requirem ents .....................................................................................................

5.6-1 5.6.1 O ccupational Radiation Exposure Report .......................................................................

5.6-1 5.6.2 Annual Radiological Environm ental O perating Report .....................................................

5.6-1 5.6.3 Radioactive Effl uent Release Report ...............................................................................

5.6-2 5.6.4 M onthly O perating Reports ..............................................................................................

5.6-2 5.6.5 CO RE O PERATING LIM ITS REPO RT (CO LR) ..............................................................

5.6-2 5.6.6 Post Accident M onitoring Report ......................................................................................

5.6-4 5.6.7 Steam G enerator Tube Inspection Report .......................................................................

5.6-4 5.7 High Radiation Area ............................................................................................................

5.7-1 Cook Nuclear Plant Unit 1 Page 5 of 5 Amendment No. 247-,

Definitions 1.1 1.1 Definitions ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable.

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

In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.LEAKAGE LEAKAGE shall be: a. Identified LEAKAGE 1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank, 2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE, or 3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE; and c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE)through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.Cook Nuclear Plant Unit I 1.1-3 Amendment No. 2-87, RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.13 RCS Operational LEAKAGE LCO 3.4.13 APPLICABILITY:

RCS operational LEAKAGE shall be limited to: a. No pressure boundary LEAKAGE;b. 0.8 gpm unidentified LEAKAGE;c., 10 gpm identified LEAKAGE; and d. 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).MODES 1, 2, 3, and 4.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Unidentified LEAKAGE A.1 Verify source of unidentified 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s> 0.8 gpm. LEAKAGE is not the pressurizer surge line.OR A.2 Reduce unidentified 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months LEAKAGE to within limit.B. Unidentified LEAKAGE B.1 Reduce unidentified 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s> 1.0 gpm. LEAKAGE to s 1.0 gpm.Cook Nuclear Plant Unit 1 3.4.13-1 Amendment No. 297-,

RCS Operational LEAKAGE 3.4.13 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Identified LEAKAGE not C.1 Reduce LEAKAGE to within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months within limits, limits.D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A, B, AND or C not met.D.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR Pressure boundary LEAKAGE exists.OR Primary to secondary LEAKAGE not within limit.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1 NOTES 1. Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

2. Not applicable to primary to secondary LEAKAGE.Verify RCS operational LEAKAGE is within limits by 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months performance of RCS water inventory balance.Cook Nuclear Plant Unit 1 3.4.13-2 Amendment No. 2-8-7, RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.13.2 NOTES Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Verify primary to secondary LEAKAGE is < 150 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months gallons per day through any one SG.Cook Nuclear Plant Unit I 3.4.13-3 Amendment No. 2-9-7, SG Tube Integrity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.17 Steam Generator (SG) Tube Integrity LCO 3.4.17 SG tube integrity shall be maintained.

AND All SG tubes satisfying the tube repair criteria shall be plugged in accordance with the Steam Generator Program.APPLICABILITY:

MODES 1, 2, 3, and 4.ACTIONS L ------------

I Separate Condition entry is allowed for each SG tube.CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SG tubes A.1 Verify tube integrity of the 7 days satisfying the tube repair affected tube(s) is criteria and not plugged maintained until the next in accordance with the refueling outage or SG Steam Generator tube inspection.

Program.AND A.2 Plug the affected tube(s) in Prior to entering accordance with the Steam MODE 4 following the Generator Program. next refueling outage or SG tube inspection B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not AND met.B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR SG tube integrity not maintained.

Cook Nuclear Plant Unit 1 3.4.17-1 Amendment No.

SG Tube Integrity 3.4.17 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify SG tube integrity in accordance with the In accordance Steam Generator Program. with the Steam Generator Program SR 3.4.17.2 Verify that each inspected SG tube that satisfies the Prior to entering tube repair criteria is plugged in accordance with the MODE 4 following Steam Generator Program. a SG tube inspection Cook Nuclear Plant Unit 1 3.4.17-2 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained.

In addition, the Steam Generator Program shall include the following provisions:

a. Provisions for condition monitoring assessments.

Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes.Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.b. Performance criteria for SG tube integrity.

SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.1. Structural integrity performance criterion:

All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents.

This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials.

Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse.

In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.2. Accident induced leakage performance criterion:

The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed I gpm for all SGs.Cook Nuclear Plant Unit I 5.5-5 Amendment No. 2-97, Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Program (continued)

3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE." c. Provisions for SG tube repair criteria.

Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.d. Provisions for SG tube inspections.

Periodic SG tube inspections shall be performed.

The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria.

The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection.

An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2. Inspect 100% of the tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.e. Provisions for monitoring operational primary to secondary LEAKAGE.Cook Nuclear Plant Unit I 5.5-6 Amendment No. 28-7, Cook Nuclear Plant Unit 1 5.5-6 Amendment No. 2-97, Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation.

The program shall include: a. Identification of a sampling schedule for the critical variables and control points for these variables;

b. Identification of the procedures used to measure the values of the critical variables;

c. Identification of process sampling points;d. Procedures for the recording and management of data;e. Procedures defining corrective actions for all off control point chemistry conditions; and f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.5.5.9 Ventilation Filter Testing Program (VFTP)The VFTP shall establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems. Tests described in Specifications 5.5.9.a and 5.5.9.b shall be performed once per 24 months; after each complete or partial replacement of the HEPA filter bank or charcoal adsorber bank; after any structural maintenance on the HEPA filter bank or charcoal adsorber bank housing; and, following painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation that could adversely affect the filter bank or charcoal adsorber capability.

Tests described in Specification 5.5.9.c shall be performed once per 24 months;after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of adsorber operation; after any structural maintenance on the HEPA filter bank or charcoal adsorber bank housing; and, following painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation that could adversely affect the charcoal adsorber capability.

Tests described in Specification 5.5.9.d shall be performed once per 24 months.The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test Frequencies.

Cook Nuclear Plant Unit I 5.5-7 Amendment No. 2-97-,

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Ventilation Filter Testing Program (VFTP) (continued)

a. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a removal efficiency of > 99%of the dioctyl phthalate (DOP) when tested in accordance with the standard and at the system flowrate specified below: ESF Ventilation System CREV System ESF Ventilation System FHAEV System ANSI Standard N510-1975 N510-1980 N510-1980 Flowrate (cfm)> 5,400 and 5 6,600 a 22,500 and 5 27,500? 27,000 and 5 33,000 b. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a removal efficiency of > 99% of a halogenated hydrocarbon refrigerant test gas when tested in accordance with the standard and at the system flowrate specified below: ESF Ventilation System CREV System ESF Ventilation System FHAEV System ANSI Standard N510-1975 N510-1980 N510-1980 Flowrate (cfm) 5,400 and 5 6,60022,500 and :5 27,500 a27,000 and 5 33,000 c. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained from either at least one test canister or at least two carbon samples removed from one of the charcoal adsorbers, shows the methyl iodide penetration less than or equal to the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86 0 F) and the relative humidity (RH) specified below: Cook Nuclear Plant Unit I 5.5-8 Amendment No. 28-~, Cook Nuclear Plant Unit 1 5.5-8 Amendment No. 2-97, Programs and Manuals 5.5 I 5.5 Programs and Manuals 5.5.9 Ventilation Filter Testing Program (VFTP) (continued)

ESF Ventilation System CREV System ESF Ventilation System FHAEV System Face Velocity (fpm) Penetration

(%) RH (%)NA 1 45.5 46.8 5 5 95 95 95 In addition, the carbon samples not obtained from test canisters shall be prepared by either: 1. Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed; or 2. Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.d. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters and the charcoal adsorbers is less than the value specified below when tested at the system flowrate specified below: Delta P ESF Ventilation System (inches water gauge)CREV System ESF Ventilation System FHAEV System 6 6 6 Flowrate (cfm)> 5,400 and : 6,600> 22,500 and 5 27,500> 27,000 and < 33,000 Cook Nuclear Plant Unit 1 5.5-9 Amendment No. 297, Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Waste Gas Holdup System, the quantity of radioactivity contained in gas storage tanks and the quantity of radioactivity contained in unprotected outdoor temporary liquid storage tanks.The program shall include: a. The limits for concentrations of hydrogen and oxygen in the Waste Gas Holdup System and a Surveillance program to ensure the limits are maintained.

Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion);

b. A Surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than the amount that would result in a whole body exposure of > 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents; and c. A Surveillance program to ensure that the quantity of radioactivity contained in all outdoor temporary liquid storage tanks that are not surrounded by liners, dikes, or walls capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program Surveillance Frequencies.

5.5.11 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established.

The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards.

The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has: 1. An API gravity, an absolute specific gravity, or a specific gravity within limits;Cook Nuclear Plant Unit I 5.5-10 Amendment No. 2&~, Cook Nuclear Plant Unit I 5.5-10 Amendment No. 2-97-,

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Diesel Fuel Oil Testing Pro-gram (continued)

2. A flash point within limits and, if the gravity was not determined by comparison with the supplier's certification, a kinematic or saybolt viscosity within limits; and 3. A clear and bright appearance with proper color;b. Within 31 days following addition of the new fuel oil to storage tanks, verify that the properties of the new fuel oil, other than those addressed in Specification 5.5.11 .a above, are within limits; and c. Total particulate concentration of the fuel oil is < 10 mg/I when tested every 31 days in accordance with ASTM D-2276, Method A.The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test Frequencies.

5.5.12 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical 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 provided the changes do not require either of the following:

1. A change in the TS incorporated in the license; or 2. A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.d. Proposed changes that meet the criteria of Specification 5.5.12.b above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71 (e).Cook Nuclear Plant Unit I 5.5-11 Amendment No. 28-7, Cook Nuclear Plant Unit 1 5.5-11 Amendment No. 2-97, Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Safety Function Determination Program (SFDP)This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate actions may be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6.a. The SFDP shall contain the following:

1. Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected;

2. Provisions for ensuring the unit is maintained in a safe condition if a loss of function condition exists;3. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and 4. Other appropriate limitations and remedial or compensatory actions.b. A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power, or no concurrent loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed.

For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and: 1. A required system redundant to the system(s) supported by the inoperable support system is also inoperable;

2. A required system redundant to the system(s) in turn supported by the inoperable supported system Is also inoperable; or 3. A required system redundant to the support system(s) for the supported systems described in Specifications 5.5.13.b.1 and 5.5.13.b.2 above is also inoperable.

c. The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.Cook Nuclear Plant Unit I 5.5-12 Amendment No. 2-97, Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 Containment Leakage Rate Testing Program a. A program shall establish the leakage rate testing of the containment 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 in accordance with the guidelines contained in Regulatory Guide 1.163,"Performance-Based Containment Leak-Test Program," dated September, 1995, as modified by the following exceptions:

1. The Type A testing Frequency specified in NEI 94-01, Revision 0, Paragraph 9.2.3, as "at least once per 10 years based on acceptable performance history" is modified to be "at least once per 15 years based on acceptable performance history." This change applies only to the interval following the Type A test performed in October 1992.2. A one-time exception to the requirement to perform post-modification Type A testing is allowed for the steam generators and associated piping, as components of the containment barrier. For this case, ASME Section XI leak testing will be used to verify the leak tightness of the repaired or modified portions of the containment barrier. Entry into MODES 3 and 4 following the extended outage that commenced in 1997 may be made to perform this testing.b. The calculated peak containment internal pressure for the design basis loss of coolant accident, P., is 12 psig.c. The maximum allowable containment leakage rate, L 8 , at Pa, shall be 0.25%of containment air weight per day.d. Leakage rate acceptance criteria are: 1. Containment leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are : 0.60 La for the Type B and C tests and 5 0.75 La for Type A tests.2. Air lock testing acceptance criterion is overall air lock leakage rate is< 0.05 La when tested at > P,.e. The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.Cook Nuclear Plant Unit I 5.5-13 Amendment No. 28-~, Cook Nuclear Plant Unit I 5.5-13 Amendment No. 2-87-,

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 Battery Monitoring and Maintenance Program This program provides for battery restoration and maintenance, based on the recommendations of IEEE Standard 450-1995, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," or of the battery manufacturer including the following:

a. Actions to restore battery cells with float voltage < 2.13 V; and b. Actions to equalize and test battery cells that had been discovered with electrolyte level below the minimum established design limit.Cook Nuclear Plant Unit I 5.5-14 Amendment No. 2-94ý,

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.5.6.6 Post Accident Monitoring Report When a report is required by Condition B or H of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.5.6.7 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.7, Steam Generator (SG) Program. The report shall include: a. The scope of inspections performed on each SG, b. Active degradation mechanisms found, c. Nondestructive examination techniques utilized for each degradation mechanism, d. Location, orientation (if linear), and measured sizes (if available) of service induced indications, e. Number of tubes plugged during the inspection outage for each active degradation mechanism, f. Total number and percentage of tubes plugged to date, and g. The results of condition monitoring, including the results of tube pulls and in-situ testing.Cook Nuclear Plant Unit 1 5.6-4 Amendment No. 2-7-, 28, Attachment 2B to AEP:NRC:6449 DONALD C. COOK NUCLEAR PLANT UNIT 2 TECHNICAL SPECIFICATION PAGES WITH THE PROPOSED CHANGES INCORPORATED Table of Contents, Page 2 of 5 Table of Contents, Page 5 of 5 1.1-3 3.4.13-1 3.4.13-2 3.4.17-1 (new)3.4.17-2 (new)5.5-5 5.5-6 5.5-7 5.5-8 5.5-9 5.5-10 5.5-11 5.5-12 5.5-13 5.6-4 UNIT 2 APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Chapter/Specification Paae 3.3 INSTRUMENTATION 3.3.1 Reactor Trip System (RTS) Instrumentation

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3.3.1-1 Table 3.3.1-1, Reactor Trip System Instrumentation

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3.3.1-11 3.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation

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3.3.2-1 Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation

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3.3.2-7 3.3.3 Post Accident Monitoring (PAM) Instrumentation

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3.3.3-1 Table 3.3.3-1, Post Accident Monitoring Instrumentation

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3.3.3-4 3.3.4 Remote Shutdown Monitoring Instrumentation

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3.3.4-1 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation

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3.3.5-1 3.3.6 Containment Purge Supply and Exhaust System Isolation Instrumentation

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3.3.6-1 Table 3.3.6-1, Containment Purge Supply and Exhaust System Isolation Instrumentation

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3.3.6-4 3.3.7 Control Room Emergency Ventilation (CREV) System Actuation Instrumentation

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3.3.7-1 Table 3.3.7-1, CREV Actuation Instrumentation

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3.3.7-3 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)Lim its .............................................................................................................................

3 .4 .1-1 3.4.2 RCS Minimum Temperature for Criticality

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3.4.2-1 3.4.3 RCS Pressure and Temperature (P/T) Limits ..................................................................

3.4.3-1 Figure 3.4.3-1, Reactor Coolant System Pressure versus Temperature Limits -Heatup Limit, Criticality Limit, and Leak Test Limit (Applicable for service period up to 32 EFPY) ...........................................................................................................

3.4.3-3 Figure 3.4.3-2, Reactor Coolant System Pressure versus Temperature Limits -Various Cooldown Rates Limits (Applicable for service period up to 32 EFPY) ........ 3.4.3-4 3.4.4 RCS Loops -MODES 1 and 2 .........................................................................................

3.4.4-1 3.4.5 RCS Loops -MODE 3 ......................................................................................................

3.4.5-1 3.4.6 RCS Loops -MODE 4 ......................................................................................................

3.4.6-1 3.4.7 RCS Loops -MODE 5, Loops Filled ................................................................................

3.4.7-1 3.4.8 RCS Loops -MODE 5, Loops Not Filled ..........................................................................

3.4.8-1 3.4.9 Pressurizer

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3.4.9-1 3.4.10 Pressurizer Safety Valves ................................................................................................

3.4.10-1 3.4.11 Pressurizer Power Operated Relief Valves (PORVs) ......................................................

3.4.11-1 3.4.12 Low Temperature Overpressure Protection (LTOP) System ...........................................

3.4.12-1 3.4.13 RCS Operational LEAKAGE ............................................................................................

3.4.13-1 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage ..................................................................

3.4.14-1 3.4.15 RCS Leakage Detection Instrumentation

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3.4.15-1 3.4.16 RCS Specific Activity ...................................................................................................

3.4.16-1 Figure 3.4.16-1, Reactor Coolant DOSE EQUIVALENT 1-131 Specific Activity Limit Versus Percent of RATED THERMAL POWER .........................................................

3.4.16-3 3.4.17 Steam Generator (SG) Tube Integrity

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3.4.17-1 Cook Nuclear Plant Unit 2 Page 2 of 5 Amendment No. 26~, Cook Nuclear Plant Unit 2 Page 2 of 5 Amendment No. 264, UNIT 2 APPENDIX A TECHNICAL SPECIFICATIONS TAI0I r- nC f'"KlTrEMlTC Definitions 1.1 1.1 Definitions ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable.

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

In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.LEAKAGE LEAKAGE shall be: a. Identified LEAKAGE 1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank, 2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE, or 3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE; and c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE)through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.I I Cook Nuclear Plant Unit 2 1.1-3 Amendment No. 264, RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to: a. No pressure boundary LEAKAGE;b. 1 gpm unidentified LEAKAGE;c. 10 gpm identified LEAKAGE; and d. 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).APPLICABILITY:

MODES 1, 2, 3, and 4.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RCS operational A.1 Reduce LEAKAGE to within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months LEAKAGE not within limits.limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not AND met.B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR Pressure boundary LEAKAGE exists.OR Primary to secondary LEAKAGE not within limit.Cook Nuclear Plant Unit 2 3.4.13-1 Amendment No. 249, RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1 NOTES 1. Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

2. Not applicable to primary to secondary LEAKAGE.Verify RCS operational LEAKAGE is within limits by 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months performance of RCS water inventory balance.SR 3.4.13.2 ---------NOTES----Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Verify primary to secondary LEAKAGE is s 150 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months gallons per day through any one SG.Cook Nuclear Plant Unit 2 3.4.13-2 Amendment No. 269, SG Tube Integrity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.17 Steam Generator (SG) Tube Integrity LCO 3.4.17 SG tube integrity shall be maintained.

AND All SG tubes satisfying the tube repair criteria shall be plugged in accordance with the Steam Generator Program.APPLICABILITY:

MODES 1, 2, 3, and 4.ACTIONS Separate Condition entry is allowed for each SG tube.CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SG tubes A.1 Verify tube integrity of the 7 days satisfying the tube repair affected tube(s) is criteria and not plugged maintained until the next in accordance with the refueling outage or SG Steam Generator tube inspection.

Program.AND A.2 Plug the affected tube(s) in Prior to entering accordance with the Steam MODE 4 following the Generator Program. next refueling outage or SG tube inspection B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not AND met.B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR SG tube integrity not maintained.

Cook Nuclear Plant Unit 2 3.4.17-1 Amendment No.

SG Tube Integrity 3.4.17 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify SG tube integrity in accordance with the In accordance Steam Generator Program. with the Steam Generator Program SR 3.4.17.2 Verify that each inspected SG tube that satisfies the Prior to entering tube repair criteria is plugged in accordance with the MODE 4 following Steam Generator Program. a SG tube inspection Cook Nuclear Plant Unit 2 3.4.17-2 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained.

In addition, the Steam Generator Program shall include the following provisions:

a. Provisions for condition monitoring assessments.

Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes.Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.b. Performance criteria for SG tube integrity.

SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.1. Structural integrity performance criterion:

All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents.

This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials.

Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse.

In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.2. Accident induced leakage performance criterion:

The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 1 gpm for all SGs.Cook Nuclear Plant Unit 2 5.5-5 Amendment No. 2",

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Program (continued)

3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE." c. Provisions for SG tube repair criteria.

Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.d. Provisions for SG tube inspections.

Periodic SG tube inspections shall be performed.

The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria.

The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection.

An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2. Inspect 100% of the tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.e. Provisions for monitoring operational primary to secondary LEAKAGE.Cook Nuclear Plant Unit 2 5.5-6 Amendment No. 26w, Cook Nuclear Plant Unit 2 5.5-6 Amendment No. 269, Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation.

The program shall include: a. Identification of a sampling schedule for the critical variables and control points for these variables;

b. Identification of the procedures used to measure the values of the critical variables;

c. Identification of process sampling points;d. Procedures for the recording and management of data;e. Procedures defining corrective actions for all off control point chemistry conditions; and f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.5.5.9 Ventilation Filter Testing Program (VFTP)The VFTP shall establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems. Tests described in Specifications 5.5.9.a and 5.5.9.b shall be performed once per 24 months; after each complete or partial replacement of the HEPA filter bank or charcoal adsorber bank; after any structural maintenance on the HEPA filter bank or charcoal adsorber bank housing; and, following painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation that could adversely affect the filter bank or charcoal adsorber capability.

Tests described in Specification 5.5.9.c shall be performed once per 24 months;after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of adsorber operation; after any structural maintenance on the HEPA filter bank or charcoal adsorber bank housing; and, following painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation that could adversely affect the charcoal adsorber capability.

Tests described in Specification 5.5.9.d shall be performed once per 24 months.The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test Frequencies.

Cook Nuclear Plant Unit 2 5.5-7 Amendment No. 26~, Cook Nuclear Plant Unit 2 5.5-7 Amendment No. 269, Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Ventilation Filter Testing Program (VFTP) (continued)

a. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a removal efficiency of > 99%of the dioctyl phthalate (DOP) when tested in accordance with the standard and at the system flowrate specified below: ESF Ventilation System CREV System ESF Ventilation System FHAEV System ANSI Standard N510-1975 N510-1980 N510-1980 Flowrate (cfm)5,400 and s 6,600 22,500 and : 27,500 a 27,000 and : 33,000 b. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a removal efficiency of > 99% of a halogenated hydrocarbon refrigerant test gas when tested in accordance with the standard and at the system flowrate specified below: ESF Ventilation System CREV System ESF Ventilation System FHAEV System ANSI Standard N510-1975 N510-1980 N510-1980 Flowrate (cfm)2 5,400 and : 6,600 a 22,500 and : 27,500 a 27,000 and : 33,000 c. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained from either at least one test canister or at least two carbon samples removed from one of the charcoal adsorbers, shows the methyl iodide penetration less than or equal to the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 300C (86°F) and the relative humidity (RH) specified below: Cook Nuclear Plant Unit 2 5.5-8 Amendment No. 26~, Cook Nuclear Plant Unit 2 5.5-8 Amendment No. 269, Programs and Manuals 5.5 I 5.5 Programs and Manuals 5.5.9 Ventilation Filter Testing Program (VFTP) (continued)

ESF Ventilation System Face Velocity (fpm) Penetration

(%) RH (%)CREV System ESF Ventilation System FHAEV System NA 1 45.5 46.8 5 5 95 95 95 In addition, the carbon samples not obtained from test canisters shall be prepared by either: 1. Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed; or 2. Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.d. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters and the charcoal adsorbers is less than the value specified below when tested at the system flowrate specified below: Delta P ESF Ventilation System (inches water aauae)CREV System ESF Ventilation System FHAEV System 6 6 6 Flowrate (cfm)> 5,400 and < 6,600> 22,500 and < 27,500> 27,000 and < 33,000 Cook Nuclear Plant Unit 2 5.5-9 Amendment No. 26g, Cook Nuclear Plant Unit 2 5.5-9 Amendment No. 2.6-9, Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Explosive Gas and Storage Tank Radioactivity Monitorinq Program This program provides controls for potentially explosive gas mixtures contained in the Waste Gas Holdup System, the quantity of radioactivity contained in gas storage tanks and the quantity of radioactivity contained in unprotected outdoor temporary liquid storage tanks.The program shall include: a. The limits for concentrations of hydrogen and oxygen in the Waste Gas Holdup System and a Surveillance program to ensure the limits are maintained.

Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion);

b. A Surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than the amount that would result in a whole body exposure of a 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents; and c. A Surveillance program to ensure that the quantity of radioactivity contained in all outdoor temporary liquid storage tanks that are not surrounded by liners, dikes, or walls capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program Surveillance Frequencies.

5.5.11 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established.

The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards.

The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has: 1. An API gravity, an absolute specific gravity, or a specific gravity within limits;Cook Nuclear Plant Unit 2 5.5-10 Amendment No. 2.64, Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Diesel Fuel Oil Testing Program (continued)

2. A flash point within limits and, if the gravity was not determined by comparison with the supplier's certification, a kinematic or saybolt viscosity within limits; and 3. A clear and bright appearance with proper color;b. Within 31 days following addition of the new fuel oil to storage tanks, verify that the properties of the new fuel oil, other than those addressed in Specification 5.5.11 .a above, are within limits; and c. Total particulate concentration of the fuel oil is < 10 mg/I when tested every 31 days in accordance with ASTM D-2276, Method A.The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test Frequencies.

5.5.12 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical 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 provided the changes do not require either of the following:

1. A change in the TS incorporated in the license; or 2. A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.d. Proposed changes that meet the criteria of Specification 5.5.12.b above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).Cook Nuclear Plant Unit 2 5.5-11 Amendment No. 26~, Cook Nuclear Plant Unit 2 5.5-11 Amendment No. 249, Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Safety Function Determination Program (SFDP)This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate actions may be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6.a. The SFDP shall contain the following:

1. Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected;

2. Provisions for ensuring the unit is maintained in a safe condition if a loss of function condition exists;3. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and 4. Other appropriate limitations and remedial or compensatory actions.b. A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power, or no concurrent loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed.

For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and: 1. A required system redundant to the system(s) supported by the inoperable support system is also inoperable;

2. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or 3. A required system redundant to the support system(s) for the supported systems described in Specifications 5.5.13.b.1 and 5.5.13.b.2 above is also inoperable.

c. The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.Cook Nuclear Plant Unit 2 5.5-12 Amendment No. 2464, Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 Containment Leakage Rate Testing Program a. A program shall establish the leakage rate testing of the containment 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 in accordance with the guidelines contained in Regulatory Guide 1.163,"Performance-Based Containment Leak-Test Program," dated September, 1995, as modified by the following exceptions:

1. The Type A testing Frequency specified in NEI 94-01, Revision 0, Paragraph 9.2.3, as "at least once per 10 years based on acceptable performance history" is modified to be "at least once per 15 years based on acceptable performance history." This change applies only to the interval following the Type A test performed in May 1992.b. The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 12 psig.c. The maximum allowable containment leakage rate, La, at Pa, shall be 0.25%of containment air weight per day.d. Leakage rate acceptance criteria are: 1. Containment leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C tests and < 0.75 La for Type A tests.2. Air lock testing acceptance criterion is overall air lock leakage rate is 5 0.05 La when tested at ? Pa.e. The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.5.5.15 Battery Monitoring and Maintenance Program This program provides for battery restoration and maintenance, based on the recommendations of IEEE Standard 450-1995, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," or of the battery manufacturer including the following:

a. Actions to restore battery cells with float voltage < 2.13 V; and b. Actions to equalize and test battery cells that had been discovered with electrolyte level below the minimum established design limit.Cook Nuclear Plant Unit 2 5.5-13 Amendment No. 26.9, Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.5.6.6 Post Accident Monitoring Report When a report is required by Condition B or H of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.5.6.7 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.7, Steam Generator (SG) Program. The report shall include: a. The scope of inspections performed on each SG, b. Active degradation mechanisms found, c. Nondestructive examination techniques utilized for each degradation mechanism, d. Location, orientation (if linear), and measured sizes (if available) of service induced indications, e. Number of tubes plugged during the inspection outage for each active degradation mechanism, f. Total number and percentage of tubes plugged to date, and g. The results of condition monitoring, including the results of tube pulls and in-situ testing.Cook Nuclear Plant Unit 2 5.6-4 Amendment No. 269, 27-,

Attachment 3 to AEP:NRC:6449 DONALD C. COOK NUCLEAR PLANT UNIT 1 TECHNICAL SPECIFICATION BASES PAGES MARKED TO SHOW CHANGES B 3.4.4-2 B 3.4.5-3 B 3.4.6-2 B 3.4.7-3 B 3.4.13-1 B 3.4.13-2 B 3.4.13-3 B 3.4.13-4 B 3.4.13-5 B 3.4.13-6 B 3.4.17-1 (new)B 3.4.17-2 (new)B 3.4.17-3 (new)B 3.4.17-4 (new)B 3.4.17-5 (new)B 3.4.17-6 (new)B 3.4.17-7 (new)

RCS Loops -MODES 1 and 2 B 3.4.4 BASES APPLICABLE SAFETY ANALYSES (continued) four pump coastdown, single pump locked rotor, single pump coastdown, and rod withdrawal events (Ref. 1).Steady state DNB analyses have been performed for the four RCS loop operation.

These analyses establish allowable RCS loop average temperature and AT for the minimum measured flow and power distribution as a function of RCS pressure.

These analyses also establish a locus of power, pressure, and temperature conditions for which the departure from nucleate boiling ratio (DNBR) is equal to its Safety Limit value. The area of permissible operation is bounded by the combination of assumed reactor trips for Power Range Neutron Flux -High, Overtemperature AT, Overpower AT, Pressurizer Pressure -Low, and Pressurizer Pressure -High Functions.

The difference between the reactor trip values assumed in the safety analyses and the nominal reactor trip setpoints provides an allowance for instrumentation channel error and setpoint error.The unit is designed to operate with all RCS loops in operation to maintain DNBR above the SL, during all normal operations and anticipated transients.

By ensuring heat transfer in the nucleate boiling region, adequate heat transfer is provided between the fuel cladding and the reactor coolant.RCS Loops -MODES I and 2 satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The purpose of this LCO is to require an adequate forced flow rate for core heat removal. Flow is represented by the number of RCPs in operation for removal of heat by the SGs. To meet safety analysis acceptance criteria for DNBR, four pumps are required at rated power.An OPERABLE RCS loop consists of an OPERABLE RCP in operation providing forced flow for heat transport and an OPERABLE SG-iR accor~dance with the Steam Generator Proegram.APPLICABILITY In MODES 1 and 2, the reactor is critical and thus has the potential to produce maximum THERMAL POWER. Thus, to ensure that the assumptions of the accident analyses remain valid, all RCS loops are required to be OPERABLE and in operation in these MODES to prevent DNB and core damage.The decay heat production rate is much lower than the full power heat rate. As such, the forced circulation flow and heat sink requirements are reduced for lower, noncritical MODES as indicated by the LCOs for MODES 3,4, and 5.Cook Nuclear Plant Unit 1 B 3.4.4-2 Revision No. 0 RCS Loops -MODE 3 B 3.4.5 BASES LCO (continued)

An OPERABLE RCS loop consists of one OPERABLE RCP and one OPERABLE SGI .A SG is OPERABLE if it mFets the rFequireneMts of the Steam Program and has the minimum water level specified in SR 3.4.5.2. An RCP is OPERABLE if it is capable of being powered and is able to provide forced flow if required.APPLICABILITY In MODE 3, this LCO ensures forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing.The most stringent condition of the LCO, that is, two RCS loops OPERABLE and two RCS loops in operation, applies to MODE 3 with the Rod Control System capable of rod withdrawal.

The least stringent condition, that is, two RCS loops OPERABLE and one RCS loop in operation, applies to MODE 3 with the Rod Control System not capable of rod withdrawal.

Operation in other MODES is covered by: LCO 3.4.4, "RCS Loops -MODES I and 2";LCO 3.4.6, "RCS Loops -MODE 4";LCO 3.4.7, "RCS Loops -MODE 5, Loops Filled";LCO 3.4.8, "RCS Loops -MODE 5, Loops Not Filled";LCO 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation

-High Water Level"; and LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation

-Low Water Level." ACTIONS A.1 If one required RCS loop is inoperable, redundancy for heat removal is lost. The Required Action is restoration of the required RCS loop to OPERABLE status within the Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . This time allowance is a justified period to be without the redundant, nonoperating loop because a single loop in operation has a heat transfer capability greater than that needed to remove the decay heat produced in the reactor core and because of the low probability of a failure in the remaining loop occurring during this period.B.1 If restoration for Required Action A.1 is not possible within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , the unit must be brought to MODE 4. In MODE 4, the unit may be placed on the Residual Heat Removal System. The additional Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is compatible with required operations to achieve cooldown and depressurization from the existing unit conditions in an orderly manner and without challenging unit systems.Cook Nuclear Plant Unit 1 B 3.4.5-3 Revision No. 0 RCS Loops -MODE 4 B 3.4.6 BASES LCO (continued)

Utilization of the Note is permitted provided the following conditions are met: a. No operations are permitted that would dilute the RCS boron concentration with coolant with boron concentrations less than required to meet the requirements of LCO 3.1.1, "SHUTDOWN MARGIN (SDM)," therefore maintaining the margin to criticality.

Boron reduction with coolant at boron concentrations less than required to assure SDM is maintained is prohibited because a uniform concentration distribution throughout the RCS cannot be ensured when in natural circulation; and b. Core outlet temperature is maintained at least 1 0°F below saturation temperature, so that no vapor bubble may form and possibly cause a natural circulation flow obstruction.

Note 2 requires that the secondary side water temperature of each SG be< 50°F above each of the RCS cold leg temperatures or the pressurizer water level be < 62% before the start of an RCP with any RCS cold leg temperature

< 152 0 F. This restraint is to prevent a low temperature overpressure event due to a thermal transient when an RCP is started.An OPERABLE RCS loop comprises an OPERABLE RCP and an OPERABLE SG .A S ic OPERABL i it meots th- .... ;....t.of the Steam Generator Program and has the minimum water level specified in SR 3.4.6.2.Similarly for the RHR System, an OPERABLE RHR loop comprises an OPERABLE RHR pump (either the east or west) capable of providing forced flow to an OPERABLE RHR heat exchanger.

RCPs and RHR pumps are OPERABLE if they are capable of being powered and are able to provide forced flow if required.APPLICABILITY In MODE 4, this LCO ensures forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing.One loop of either RCS or RHR provides sufficient circulation for these purposes.

However, two loops consisting of any combination of RCS and RHR loops are required to be OPERABLE to meet single failure considerations.

Operation in other MODES is covered by: LCO 3.4.4, "RCS Loops -MODES I and 2";LCO 3.4.5, "RCS Loops -MODE 3";LCO 3.4.7, "RCS Loops -MODE 5, Loops Filled";LCO 3.4.8, "RCS Loops -MODE 5, Loops Not Filled";Cook Nuclear Plant Unit 1 B 3.4.6-2 Revision No. 0 RCS Loops -MODE 5, Loops Filled B 3.4.7 BASES LCO (continued) is to prevent a low temperature overpressure event due to a thermal transient when an RCP is started.Note 4 provides for an orderly transition from MODE 5 to MODE 4 during a planned heatup by permitting removal of RHR loops from operation when at least one RCS loop is in operation.

This Note provides for the transition to MODE 4 where an RCS loop is permitted to be in operation and replaces the RCS circulation function provided by the RHR loops.RHR pumps are OPERABLE if they are capable of being powered and are able to provide flow. A;; OPERAK-E SG can perform as a heat sink via natural circulation when it has an adequate water level and is OPERABLE in acor.dan.e with the Steam Generator Program.APPLICABILITY In MODE 5 with RCS loops filled, this LCO requires forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing. One loop of RHR provides sufficient circulation for these purposes.

However, one additional RHR loop is required to be OPERABLE, or the secondary side water level of at least two SGs is required to be above the lower tap of the SG wide range water level instrumentation by > 420 inches.Operation in other MODES is covered by: LCO 3.4.4, "RCS Loops -MODES 1 and 2";LCO 3.4.5, "RCS Loops -MODE 3";LCO 3.4.6, "RCS Loops -MODE 4";LCO 3.4.8, "RCS Loops -MODE 5, Loops Not Filled";LCO 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation

-High Water Level"; and LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation

-Low Water Level." ACTIONS A.1, A.2, B.1 and B.2 If one RHR loop is OPERABLE and either the required SGs do not have secondary side water levels above the lower tap of the SG wide range level instrumentation by > 420 inches or one required RHR loop is inoperable, redundancy for heat removal is lost. Action must be initiated immediately to restore a second RHR loop to OPERABLE status or to restore the secondary side water levels to within limit for the required SGs. Either Required Action will restore redundant heat removal paths.The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal.Cook Nuclear Plant Unit 1 B 3.4.7-3 Revision No. 0 RCS Operational LEAKAGE B 3.4.13 B 3.4 REACTOR COOLANT SYSTEM (RCS)B 3.4.13 RCS Operational LEAKAGE BASES BACKGROUND Components that contain or transport the coolant.to or from the reactor core make up the RCS. Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systems from the RCS.During unit life, the joint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration.

The purpose of the RCS Operational LEAKAGE LCO is to limit system operation in the presence of LEAKAGE from these sources to amounts that do not compromise safety. This LCO specifies the types and amounts of LEAKAGE.Plant Specific Design Criterion 16 (Ref. 1), requires means for detecting and, to the extent practical, identifying the source of reactor coolant LEAKAGE. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems.The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration.

Therefore, detecting and monitoring reactor coolant LEAKAGE into the containment area is necessary.

Quickly separating the identified LEAKAGE from the unidentified LEAKAGE is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility and the public.A limited amount of leakage inside containment is expected from auxiliary systems that cannot be made 100% leaktight.

Leakage from these systems should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with RCS leakage detection.

This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded.

The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA).APPLICABLE SAFETY ANALYSES Except for primary to secondary LEAKAGE, the safety analyses do not address operational LEAKAGE. However, other operational LEAKAGE is related to the safety analyses for LOCA; the amount of leakage can affect the probability of such an event. The safety analysis for an event resulting in steam discharg to the atmosphere assumes f-hi p-rinma'ryt-o Cook Nuclear Plant Unit I B 3.4.13-1 Revision No. 0 RCS Operational LEAKAGE B 3.4.13 BASES APPLICABLE SAFETY ANALYSES (continued)

Primary to secondary LEAKAGE is a factor in the dose releases outside containment resulting from a steam line break (SLB) accident and other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR). The leakage contaminates the secondary fluid.The UFSAR (Ref. 3) analysis for SGTR assumes the contaminated secondary fluid is released via the steam generator power operated relief valves (and safety valves if their setpoint is reached) if offsite power is not available or if the condenser steam dump system fails to operate. The safety analysis for the SLB accident assumes the amount of primary to secondary LEAKAGE in the three intact SGs is 1 gpm minus a faulted SG tube LEAKAGE of 500 gallons per day as an initial condition.

The dose consequences resulting from events resulting in a steam discharge to the atmosphere are within a small fraction of the limits defined in 10 CFR 100 and within GDC-19.The RCS Operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO RCS operational LEAKAGE shall be limited to: a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration.

LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.b. Unidentified LEAKAGE The 0.8 gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air particulate monitoring equipment can detect within a reasonable time period. The limit is established for the pressurizer surge line in the leak before break methodology.

Violation of this LCO could result in Cook Nuclear Plant Unit I B 3.4.13-2 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.4.13-2 Revision No. 0 RCS Operational LEAKAGE B 3.4.13 continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary.Cook Nuclear Plant Unit 1 B 3.4.13-3 Revision No. 0 RCS Operational LEAKAGE B 3.4.13 BASES LCO (continued)

c. Identified LEAKAGE Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of unidentified LEAKAGE and is well within the capability of the RCS Makeup System. Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE).

Violation of this LCO could result in continued degradation of a component or system.Ls~~~~~~~~~~

Onear tSG~n1rt EkG t~i~ n~elimit o-5.aln porrday pe S.G is based opthatio.Pperfora criterion in NEI .. .06 Sa nen rateo r Pro.gram Guidelines 4). -Th St+eamq Geerator Programr_, paiol LEAKAGE performantce critUon 1in E 97-06 stae ,riheRCS operational primary to secondary leakage through any one~iG shall be limited to 150 gallons pedy.' The l~imi is based onj Deratin exeinewith SG tuedgaainmcaim thal'eS~Ulint~ube leakae ~Theoperational leakage ate~ criterion inI onjuntion with teipmnaioofthe SteamGnerto Progra I n feciemeasure fo rm -nr-iingwn th rq~-yo ~g benerator tube.ru rs d. P~~s~ tq Se a LEAKAGE thgq All Sta Geea~To9tal priar to:eodary LEAKAGE amounting to 1 gpmn thr~oug the ;Intac Sý roues accGeptable effseite and coentrol room. doses in the S-GTR accGiden;t analysis.

Fmor t-he RSLB accnident, the amount of primar', to secoendary LEKG nthe three wintact SGs is assumed to be 1 gpm min*us a faulted SG tube LEAKA.GE of 500 gallons per day. Thc LCOQ limit of 600 gallons pcr day is more conservative than the I gpm; value assumed in the offeite dose calculations.R This lFimt is imposed to help m~inimize the potential for cxcessiv.e leakage or tube burst in the ev.ent of a M8LB Or LOCA conssietnt with the LCO limit on primar; to secondary LEAKAGE fthrugh any; one SG. in addition, the conservative limit is approepriate due to the increased steam release as a result of the replacement SGs. Violation of this LCOQ could exceed the offsite dose limitS for these accidents Prim~ary to secondary LEAKAGE must be icueinthc totall allowable lim~it fo identified LEAKAGE.e..,. P4,.~ to Seena LEAKAGE fk..-..- Any~ On SG~Cook Nuclear Plant Unit 1 B 3.4.13-4 Revision No. 0 RCS Operational LEAKAGE B 3.4.13 The 150 gallons pcr day limit on one SG is based on the assumption that a single crack leaking this amount would not propagate to a tube burst under the stress conditions of a LOC A. or a main stcam 'one rupture. if leaked through m~any c~acks, the crackes are Ye~' small, and the above assumption is Gonservative.

APPLICABILITY In MODES 1, 2, 3, and 4, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.

In MODES 5 and 6, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.Cook Nuclear Plant Unit 1 B 3.4.13-5 Revision No. 0 RCS Operational LEAKAGE B 3.4.13 BASES APPLICABILITY (continued)

LCO 3.4.14, "RCS Pressure Isolation Valve (PIV) Leakage," measures leakage through each individual PIV and can impact this LCO. Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS LEAKAGE when the other is leak tight. If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable identified LEAKAGE.ACTIONS A.1 and A.2 With unidentified LEAKAGE > 0.8 gpm, the pressurizer surge line must be verified not to be the source of unidentified LEAKAGE or the unidentified LEAKAGE must be reduced to within limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . These Required Actions are necessary to satisfy the requirements for the application of Leak-Before-Break methodology to the pressurizer surge line as documented in Reference 4 and approved by the NRC as documented in Reference 5, and are necessary to prevent further deterioration of the RCPB associated with the pressurizer surge line. The Completion Time allows time to verify leakage rates and either identify the unidentified LEAKAGE or reduce LEAKAGE to within limit before the reactor must be shut down.B.1 Unidentified LEAKAGE > 1.0 gpm must be reduced to < 1.0 gpm within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This Completion Time allows time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down. This action is necessary to prevent further deterioration of the RCPB.C.1 Identified LEAKAGE or primary to .e.ondry LE.' GE, in excess of the LCO limits must be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This Completion Time allows time to verify leakage rates and reduce LEAKAGE to within limits before the reactor must be shut down. This action is necessary to prevent further deterioration of the RCPB.D.1 and D.2 If any Required Action and associated Completion Time of Condition A, B, or C is not met, if any pressure boundary LEAKAGE exists, or if Uria6vtoscdj A:AG-ý th SR 3.4.13.2 is not I met, the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences.

It should be noted that LEAKAGE past seals and gaskets is not pressure boundary Cook Nuclear Plant Unit 1 B 3.4.13-6 Revision No. 0 RCS Operational LEAKAGE B 3.4.13 BASES ACTIONS (continued)

LEAKAGE. The reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This action reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary.The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. In MODE 5, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.SURVEILLANCE SR 3.4.13.1 REQUIREMENTS Verifying RCS LEAKAGE to be within the LCO limits ensures the integrity of the RCPB is maintained.

Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be positively identified by inspection.

It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. Unidentified LEAKAGE and identified LEAKAGE are determined by performance of an RCS water inventory balance. Pim.'a.; to sendary LEAKA^GE is also meaSUred by perform9ance f an RCS water iry.nt..r; in conjunction

'ith effluont moneitoring within the seconidar,'

steam and fedaerssems.

The RCS water inventory balance must be performed with the reactor at steady state o eratin conditions (stable RCS pressure, temperature, and power level, rhe ance is mciied- bW Note. The-efer-ea Note LJ !ýd a-t- mg that this SR is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady state operation.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance provides sufficient time to collect and process all necessary data after stable unit conditions are established.

Steady state operation is required to perform a proper inventory balance since calculations during maneuvering are not useful. For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, and power level.An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment sump level. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. These leakage detection systems are specified in LCO 3.4.15, "RCS Leakage Detection Instrumentation." r o e s a te t a is S Is not applicble to primary to secondary EA Ebecaus LAKAGE of 150 gallons per day cannoj b eaue cuaeyba CSwjrivnoyblne Cook Nuclear Plant Unit I B 3.4.13-7 Revision No. 0 Cook Nuclear Plant Unit I B 3.4.13-7 Revision No. 0 RCS Operational LEAKAGE B 3.4.13 The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents.

Cook Nuclear Plant Unit 1 B 3.4.13-8 Revision No. 0 RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.13.2* A ...........

provides tMe mneans niecessar';

to deter-mine R-- UI-PRRAU R11 ITY I in an operation;al MODE=. The raguiremen~t to dcmonstrate SG fub integrity inaccordance with, the Steam GceReator Program emphasizes the impoace of SG tube intcgrit,', even though this Suweillanre cannot be performed at normal operating con~ditions; REFERENCES

1. UFSAR, Section 1.4.3.2. Regulatory Guide 1.45, May 1973.3. UFSAR, Section 14.2.4.4. Letter from Indiana Michigan Power Company (M. W. Rencheck) to the NRC dated October 26, 2000 (Letter C1000-20).

5. Letter from NRC (John F. Stang) to Indiana Michigan Power Company (Robert P. Powers), dated November 8, 2000..0, "E 7OJSteami Geeao rgamGieie.

Cook Nuclear Plant Unit 1 B 3.4.13-9 Revision No. 0 RCS Operational LEAKAGE B 3.4.13 Cook Nuclear Plant Unit I B 3.4.13-10 Revision No. 0 SG Tube Integrity B 3.4.17 P' -t b 'nfegr-f- -en t-'f he--tub-,s----e-ca"pa bI ,f -r 6r-n--i " -iri ntended ~ ~ RCBsft ucincnitn1 ihtelcnjgbss e se rused to met te SG perfra nce cieaare deine e y 1h SernGeeatr r rm.ud~ie -,e -Cook Nuclear Plant Unit I B 3.4.17-1 Rev. X.X Cook Nuclear Plant Unit I B 3.4.17-1 Rev. X.X SG Tube Integrity B 3.4.17 ea gene r -integrity, satisfirs'Criterion-2 of 10 C*FR t-'i-n...

rato" iuU 0. 3-6 (c) (2) (i i): Lco The LCO requires that SG tube integrity-bemaintained.

The LCO also fequires that all SG tubes that satisfy the repair criteriýjýIuElqýýd in pcýýordance

'with the Steam Generator Program.Cook Nuclear Plant Unit I B 3.4.17-2 Rev. X.X Cook Nuclear Plant Unit 1 B 3.4.17-2 Rev. X.X SG Tube Integrity B 3.4.17 Cook Nuclear Plant Unit I B 3.4.17-3 Rev. X.X Cook Nuclear Plant Unit 1 B 3.4.17-3 Rev. X.X SG Tube Integrity B 3.4.17 ftýj -ý-n&A.2 Cook Nuclear Plant Unit I B 3.4.17-4 Rev. X.X Cook Nuclear Plant Unit 1 B 3.4.17-4 Rev. X.X SG Tube Integrity B 3.4.17 F-tions (cpntinued) que ons a as ia so in e o odt rh 6 1e 'i '" _ 6 fd 6 m el i -i C '- -Zý-U M V L: I LLI-MNU rZ ZD f_ý .3. Z+. I IýEQUIREMENTS Cook Nuclear Plant Unit 1 B 3.4.17-5 Rev. X.X SG Tube Integrity B 3.4.17ý-ORVEILCMTcý I~h Feq'-ehtr-y of pri'toenterg-MODE'4fll'w ing-nSG-inspecto nsrs thtteSrelac as bencmltedan al tu es etn h repair cieri are puged proo sujctn ta Gtb týigniicap piayt eodr ps edferental Cook Nuclear Plant Unit 1 B 3.4.17-6 Rev. X.X SG Tube Integrity B 3.4.17 FE-rEýN'CES

-1"-. "fý7-NIHI 97-bý Steam Generator Program Guýidýlings.

2. -10 CFR 50 Appendix A, GDC 19,'3 ------ 1'O.CFR 10P, 4ý AS EBole ndPesur ese Cde ecio II ubecin B-6. -'r1 r -St ea .m --Ge .nera I to I r .Exa -minat -ion Cook Nuclear Plant Unit 1 B 3.4.17-7 Rev. X.X

ML061570157

Text

SUBJECT:

Dear Sir or Madam:

Enclosures:

Subject:

1.0 INTRODUCTION

2.0 DESCRIPTION

3.0 BACKGROUND

10.0 REFERENCES

1.0 INTRODUCTION

2.0 DESCRIPTION

3.0 BACKGROUND

10.0 REFERENCES

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