ML070860765
| ML070860765 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 03/14/2007 |
| From: | Jensen J Indiana Michigan Power Co |
| To: | Document Control Desk, NRC/NRR/ADRO |
| References | |
| AEP:NRC:7449 | |
| Download: ML070860765 (14) | |
Text
Indiana Michigan Power INDIANA Cook Nuclear Plant MICHIGAN One Cook Place Bridgman, MI 49106 POWERO AEPcom A unit of American Electric Power March 14, 2007 AEP:NRC:7449 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 I and 2 Docket Nos. 50-315 and 50-316 Corrections to Technical Specification Table of Contents
References:
- 1. Letter from J. Donohew, Nuclear Regulatory Commission (NRC), to M. K. Nazar, Indiana Michigan Power Company (I&M), "D. C. Cook Nuclear Plant, Units I and 2 - Issuance of Amendments for the Conversion to the Improved Technical Specifications with Beyond Scope Issues (TAC Nos. MC2629, MC2630, MC2653 through MC2687, MC2690 through MC2695, MC3152 through MC3157, MC3432 through MC3453)," dated June 1, 2005 (ML050620034).
- 2. Letter from P. Tam, NRC, to M. K. Nazar, I&M, "D. C. Cook Nuclear Plant, Units 1 and 2 - Issuance of Amendments Re: Elimination of Requirements for Monthly Operating Reports and Occupational Radiation Exposure Reports (TAC Nos. MC8166 and MC8167)," dated January 12, 2006 (ML053570406).
- 3. Letter from J. N. Jensen, I&M, to NRC Document Control Desk, "Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity," AEP:NRC:6449, dated May 26, 2006 (ML061570157).
- 4. Memorandum from R. P. Zimmerman, NRC, to S. A. Varga et. al., NRC, "License Amendment Corrections of Technical Specifications," dated January 16, 1997.
Dear Sir or Madam:
By Reference 1, the Nuclear Regulatory Commission (NRC) issued Amendment Numbers 287 and 269 to Donald C. Cook Nuclear Plant (CNP) Units 1 and 2 Facility Operating Licenses (FOL)
DPR-58 and DPR-74. These amendments approved the conversion of the CNP Unit I and Unit 2 Technical Specifications (TS) consistent with Improved Standard Technical Specifications (ISTS) as described in NUREG-1431, "Standard Technical Specifications - Westinghouse Plants," Revision 2.
By Reference 2, the NRC issued Amendment Numbers 292 and 274 to CNP Units 1 and 2
U. S. Nuclear Regulatory Commission AEP:NRC:7449 Page 2 FOL approving the deletion of the TS requirements to submit the annual occupational radiation exposure report and the monthly operating report.
Indiana Michigan Power Company (I&M) staff recently identified administrative errors in the CNP Unit I and Unit 2 TS Table of Contents (TOC). Specifically, in both CNP Unit 1 and Unit 2 TS TOC Page 2 of 5, the word "System" was left off of the title of Table 3.3.7-1, "Control Room Emergency Ventilation System Actuation Instrumentation," and Specification 3.3.8, "Boron Dilution Monitoring Instrumentation (BDMI)," was inadvertently left off of the contents listing during the conversion to ISTS approved by Reference 1. Also, in both CNP Unit 1 and Unit 2 TS TOC Page 5 of 5, the license amendment (Reference 2) deleting TS 5.6.1, Occupational Radiation Exposure Report, and TS 5.6.4, Monthly Operating Report, inadvertently left the titles in the TS TOC listing versus annotating them as "Deleted." These administrative errors constitute erroneous changes that were not addressed in the notice to the public nor reviewed by the staff. NRC memorandum dated January 16, 1997 (Reference 4), states that such administrative errors may be corrected by a letter from the licensee to the staff instead of an amendment to the license.
By Reference 3, I&M proposed to modify TS requirements related to steam generator tube integrity consistent with TS Task Force (TSTF) generic change traveler TSTF-449, "Steam Generator Tube Integrity." The Reference 3 license amendment request, currently under review by the NRC, proposed changes which modified the Unit 1 and Unit 2 TS TOC Page 2 of 5 and Page 5 of 5.
Enclosed are revised TS pages that reflect the proposed amendment request submitted by Reference 3 with the administrative corrections to the TOC listing discussed above. I&M requests that the NRC issue the revised pages when issuing the amendment requested by Reference 3. provides an affirmation statement pertaining to this letter. Attachment I provides the CNP Unit 1 and Unit 2 marked-up TS pages to replace the corresponding pages submitted in Attachments IA and IB to Reference 3. Attachment 2 provides the TS pages, with the changes incorporated, to replace the corresponding pages submitted in Attachments 2A and 2B to Reference 3.
There are no commitments made in this letter. Should you have any questions, please contact Ms. Susan D. Simpson, Regulatory Affairs Manager, at (269) 466-2428.
Sincerely, ZN.Jensen Site Vice President KAS/rdw
U. S. Nuclear Regulatory Commission AEP:NRC:7449 Page 3
Enclosure:
Affirmation Attachments: 1. Donald C. Cook Nuclear Plant Unit I and Unit 2 Technical Specification Pages Marked To Show Changes
- 2. Donald C. Cook Nuclear Plant Unit 1 and Unit 2 Technical Specification Pages With the Proposed Changes Incorporated c: J. L. Caldwell, NRC Region III K. D. Curry, Ft. Wayne AEP, w/o enclosure/attachments J. T. King, MPSC MDEQ - WHMD/RPMWS NRC Resident Inspector P. S. Tam, NRC Washington, DC
Enclosure to AEP:NRC:7449 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 ite1i T N. Jensen SiteVice President SWORN TO AND SUBSCRIBED BEFORE ME THIS ____DAY OF iAQ f-CV 2007
+* Notary Public fEGAN D.WENDZEL R
My Commission Expires Notary Public. BRrripn o.p,, &q MY Commission Expires Jan. 21, 2009
Attachment 1 to AEP:NRC:7449 DONALD C. COOK NUCLEAR PLANT UNIT I AND UNIT 2 TECHNICAL SPECIFICATION PAGES MARKED TO SHOW CHANGES Unit 1: Table of Contents, Page 2 of 5 Unit 1: Table of Contents, Page 5 of 5 Unit 2: Table of Contents, Page 2 of 5 Unit 2: Table of Contents, Page 5 of 5
UNIT 1 APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Chapter/Specification Paqe 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 Instru m e ntatio n ......................................................................................................... 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 Sytem Actuation Instrumentation .............................................. 3.3.7-3 i Boron DilutionMont entation (BDMI) ....... ... ..... ............... 3.8-11 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 F P Y ) ........................................................................................................ 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 RC S Loops - MO D ES 1 and 2 ......................................................................................... 3.4.4-1 3.4.5 R C S Loops - MO D E 3 ..................................................................................................... 3.4.5-1 3.4.6 R C S Loops - MO D E 4 .................................................................................................... 3.4.6-1 3.4.7 RCS Loops - MO DE 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 re ssu rize r ..................................................................................................................... 3 .4 .9-1 3.4.10 Pressurizer Safety V alves .............................................................................................. 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 RC S O perational LEA KAG E .......................................................................................... 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
- .4.u17 Steam Generator (SG) Tube Integrity. ... ........... 3.41-1...............
Cook Nuclear Plant Unit 1 Page 2 of 5 Amendment No. 287
UNIT 1 APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Chapter/Specification Paqe 5.0 ADMINISTRATIVE CONTROLS 5 .1 R e s p o n s ib ility ..................................................................................................................... 5 .1-1 5 .2 O rg a n izatio n ...................................................................... ................................................ 5 .2-1 5.2.1 O nsite and O ffsite O rganizations .................................................................................... 5.2-1 5.2.2 Unit Staff ....................................................................... 5.2-1 5 .3 Unit S taff Q ualifications ..................................................................................................... 5 .3-1 5 .4 P ro ce d u re s ........................................................................................................................ 5 .4 -1 5 .5 P rogram s and Manuals ..................................................................................................... 5 .5-1 5.5.1 Offsite Dose Calculation Manual (O DCM ) ...................................................................... 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 Com ponent Cyclic or Transient Lim its ............................................................................ 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 G enerator (SG ) Program ..................................................................................... 5.5-5 Table 5.5.7 1, MiRnimum. Number o-f Steam Generators o bt b nspected During Inervpi, Inpction............................................. 5.5* 9 Table 5.5.72, Steam G-nerartor G Tube Inspec.ion ........................ 5.5 10 5.5.8 Secondary W ater Chem istry Program ......................................................................... 5.5-744-5.5.9 Ventilation Filter Testing Program .(VFTP) ...................................................................... 5.5- ,
5.5.10 Explosive Gas and Storage TankRadioactivity Monitoring Program ............................... 5.5-I-1-4 5.5.11 Diesel Fuel Oil Testing Program ................................ ....... 5.5-1 4 5.5.12 Technical Specifications (TS) Bases Control Program ........................ 5.5-1-1-5 5.5.13 Safety Function Determination Program (SFDP) ............................................................ 5.5-12*-6 5.5.14 Containment Leakage Rate Testing Program ................................................................. 5.5-1341-7 5.5.15 Battery Monitoring and Maintenance Program ................................................................ 5.5-14 4-8 5.6 R eporting R equirem ents ... I............................................................................................... 5.6-1 5.6.1 eletedOccupationA RadiatAon 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 te 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 R adiation A rea ........................................................................................................ 5 .7-1 Cook Nuclear Plant Unit 1 Page 5 of 5 Amendment No. 287
UNIT 2 APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Chapter/Specification Paqe 3.3 INSTRUMENTATION 3.3.1 Reactor Trip System (RTS) Instrum entation ................................................................... 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 Instru m e ntatio n ......................................................................................................... 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 System Actuation Instrumentation .............................................. 3.3.7-3 3A8 -Boron Dilution Monitorin Instrumentation (BDMl)............. w....... .......8-.1j 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)
Limits ........................................................ 3.4.1-1 3.4.2 RCS Minimum Temperature for Criticality ................................ 3.4.2-11 3.4.3 RCS Pressure and Temperature (P/T) Limit s ................................................................. 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 F P Y ) ......................................................................................................... 31 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 RC S Loops - MO D ES 1 and 2 ....................................................................................... 3.4.4-1 3.4 .5 R C S Loops - MO D E 3 .................................................................................................... 3.4.5-1 3.4.6 R C S Loops - MO D E 4 .................................................................................................... 3.4.6-1 3.4.7 RCS Loops - MO DE 5, Loops Filled ............................................................................... 3.4.7-1 3.4.8 RCS Loops - MO DE 5, Loops Not Filled ......................................................................... 3.4.8-1 3 .4 .9 P ressu rize r ..................................................................................................................... 3 .4 .9-1 3.4.10 Pressurizer S afety V alves .............................................................................................. 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 RC S O perational LEA KAG E ........................................................................................... 3.4.13-1 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage ................................................................. 3.4.14-1 3.4.15 RCS Leakage Detection Instrum entation ........................................................................ 3.4.15-1 3 .4 .16 R C S S pecific 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 Steam G..enerator (SG)TubeIiei.t. 1.4.17 ..... . . 3..-**..,..................4 l
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 R e s p o n s ib ility .................................................................................................................... 5 .1-1 5 .2 O rg a n iza tio n ...................................................................................................................... 5 .2 -1 5.2.1 O nsite and O ffsite O rganizations ................................................................... ......... 5.2-1 5 .2 .2 Un it S ta ff ........................................................................................................................ 5 .2-1 5.3 Unit S taff Q ualifications ..................................................................................................... 5 .3-1 5 .4 P ro ce d u re s ........................................................................................................................ 5 .4 -1 5.5 Programs and Manuals .............................................. 5.5-1 5.5.1 Offsite Dose Calculation Manual (O DCM ) ...................................................................... 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 Com ponent Cyclic or Transient Lim its ............................................................................ 5.5-3 5.5.5 Reactor Coolant Pump Flywheel Inspection Program ..................................................... 5.5-4 5.5.6 Inservice T esting P rogram .............................................................................................. 5.5-4 5.5.7 Steam G enerator (SG ) Program ..................................................................................... 5 .5-5 Tabie 5.5.7-1, Minimum Number of Steam Generators to be Inspected During lnscVG, , * ,,,cct, , .................. . .... ................... . ..................................................... 5. 9 Table 5.5.7 2, Ste aG e. tnerr (9G) Tube In.pcctin........................ 5.510 5.5.8 Secondary Water Chemistry Program ................................... 5.5-L7,4-5.5.9 Ventilation Filter Testing Program (VFTP) ................................ 5.5-Z14 5.5.10 Explosive Gas and StorageTank Radioactivity Monitoring Program .............................. 5.5-L71-4 5.5.11 D iesel Fuel O il Testing Program ......... ......................................................................... 5.5-1 44 5.5.12 Technical Specifications.(TS) Bases Control Program ....................... 5 5.5.13 Safety Function Determination Prog.ram (SFDP) ............................................................. 5.5-E4-6 5.5.14 Containment Leakage Rate Testing Program ............................................................. 5.5-- 34 5.5.15 Battery Monitoring and Maintenance Program ................................................................ 5.5-E34-8 5.6 R eporting R equirem ents ................................................................................................... 5.6-1 5.6.1 Rd e Repo ........................................................... 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 DeletedMonthly Operatin g, R............................... ............................... ..... 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 R adiation A rea .......................................................................................................... 5 .7-1 Cook Nuclear Plant Unit 2 Page 5 of 5 Amendment No. 269
Attachment 2 to AEP:NRC:7449 DONALD C. COOK NUCLEAR PLANT UNIT I AND UNIT 2 TECHNICAL SPECIFICATION PAGES WITH THE PROPOSED CHANGES INCORPORATED Unit 1: Table of Contents, Page 2 of 5 Unit 1: Table of Contents, Page 5 of 5 Unit 2: Table of Contents, Page 2 of 5 Unit 2: Table of Contents, Page 5 of 5
UNIT 1 APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Chapter/Specification Page 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 Instru m e ntatio n ......................................................................................................... 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 System Actuation Instrumentation .............................................. 3.3.7-3 3.3.8 Boron Dilution Monitoring Instrumentation (BDMI) .......................................................... 3.3.8-1 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 ............ I.......................................................... 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 F P Y) ........................................................................................................ 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 R C S Loops - MO D E 3 .................................................................................................... 3.4.5-1 3.4.6 R C S Loops - MO D E 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 re ssu rize r ..................................................................................................................... 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 R C S S pecific A ctivity ...................................................................................................... 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. 2-97,
UNIT 1 APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Chapter/Specification Paqe 5.0 ADMINISTRATIVE CONTROLS 5 .1 R e sp o ns ib ility .................................................................................................................... 5 .1-1 5 .2 O rg a n iza tio n ....................................................................................................................... 5 .2 -1 5.2.1 O nsite and O ffsite O rganizations ......................................... ........................................... 5.2-1 5 .2 .2 Un it S ta ff ........................................................................................................................ 5 .2 -1 5 .3 Unit S taff Q ua lifications ..................................................................................................... 5 .3-1 5 .4 P ro c e d u re s ........................................................................................................................ 5 .4 -1 5 .5 P rogram s 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 T esting P rogram .............................................................................................. 5.5-4 5.5.7 Steam Generator (SG) Program ..................................................................................... 5.5-5 5.5.8 Secondary Water Chemistry 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 Oil Testing Program ............... ....................... 5.5-10 5.5.12 Technical Specifications (TS) Bases Control Program... ... *...... ................. 5.5-11 5.5.13 Safety Function Determination Program (SFDP) ... ........................................... ..... 5.5-12 5.5.14 Containment Leakage Rate Testing Program .............................. 5.5-13 5.5.15 Battery Monitoring and Maintenance Program ............................................................ 5.5-14 5.6 R eporting R equirem ents ................................................................................................... 5.6-1 5 .6 .1 De le te d ........................................................................................................................... 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 D e le te d ........................................................................................................................... 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 Hig h R ad iatio n A rea .......................................................................................................... 5 .7 -1 Cook Nuclear Plant Unit 1 Page 5 of 5 Amendment No. 2-P7,
UNIT 2 APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Chapter/Specification Paqe 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 Instrum e ntatio n ......................................................................................................... 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 System Actuation Instrumentation .............................................. 3.3.7-3 3.3.8 Boron Dilution Monitoring Instrumentation (BDMI) ......................................................... 3.3.8-1 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 F PY ) .......................................... ............................................................. 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 R C S Loops - MO D E 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 re ssurize r ..................................................................................................................... 3 .4 .9-1 3.4.10 Pressurizer Safety V alves .............................................................................................. 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 RC S O perational LEA KAG E ........................................................................................... 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 R CS S pecific A ctivity ...................................................................................................... 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 2 Page 2 of 5 Amendment No. 2-69,
UNIT 2 APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Chapter/Specification Paqe 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility .................................................................................................................... 5.1-1 5.2 O rganization ...................................................................................................................... 5.2-1 5.2.1 O nsite and Offsite O rganizations ................................................................................... 5.2-1 5.2.2 Unit Staff ........................................................................................................................ 5.2-1 5.3 Unit Staff Q ualifications ................................................................................................... 5.3-1 5.4 Procedures ........................................................................................................................ 5.4-1 5.5 Program s and Manuals ..................................................................................................... 5.5-1 5.5.1 Offsite Dose Calculation Manual (O DCM ) ...................................................................... 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 Com ponent Cyclic or Transient Lim its ............................................................................ 5.5-3 5.5.5 Reactor Coolant Pum p Flywheel 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 G as 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 Determination Program (SFDP) ......................... ... 5.5-12 5.5.14 Containm ent Leakage Rate Testing Program ................................................................. 5.5-13 5.5.15 Battery Monitoring and Maintenance Program ................................. ............................. 5.5-13 5.6 Reporting Requirem ents .................................................................................................. 5.6-1 5.6.1 Deleted ........................................................................................................................... 5.6-1 5.6.2 Annual Radiological Environm ental O perating Report .................................................... 5.6-1 5.6.3 Radioactive Effluent Release Report ............................................................................. 5.6-2 5.6.4 Deleted ........................................................................................................................... 5.6-2 5.6.5 CO RE O PERATING LIM ITS REPO RT (CO LR) .............................................................. 5.6-2 5.6.6 Post Accident Monitoring 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 2 Page 5 of 5 Amendment No. 2-69,