ML070720537

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Improved Technical Specifications, Amendment 233
ML070720537
Person / Time
Site: Indian Point Entergy icon.png
Issue date: 02/26/2007
From:
Entergy Nuclear Operations
To:
Document Control Desk, NRC/NRR/ADRO
References
Download: ML070720537 (22)
v • d • e


Text

eFebruary 26, 2007 DISTRIBUTION CONTROL LIST Document Name: IP3 ITS / BASES CC# NAME DEPARTMENT LOCATION 1 OPS PROC GP SUPERVISOR OPS PROCEDURE GROUP iP2 5 CONTROL ROOM OPERATIONS iP3 11 RES DEPARTMENT MGR RES 45-4-A 19 CONROY, PAT LICENSING / ROOM 205 GSB-2D 20 CHEMISTRY SUPERVISOR CHEMISTRY DEPARTMENT 45-4-A 21 TSC (IP3) 53' ELEVATION 1P2 22 SHIFT MANAGER OPERATIONS IP3 23 LIS LICENSING & INFO SERVICE OFFSITE 25 SIMULATOR TRAINING 48-2-A 28 RESIDENT INSPETOR US NRC 88' ELEVATION IP2 32 EOF E-PLAN EOF 47 CHAPMAN, N BECHTWL OFFSITE 50 SCOZZARO, ANTHONY WESTINGHOUSE ELECTRIC OFFSITE 61 SIMULATOR TRAINING 48-2-A 69 CONROY, PAT LICENSING / ROOM 205 GSB/2D 99 BARANSKI J ST. EMERGENCY MGMT. OFFICE OFFSITE 106 SIM INSTRUCT AREA TRAINING #48 164 CONTROL ROOM OPERATIONS IP3 273 FAISON, CHARLENE NUCLEAR LICENSING WPO-12 319 GRANT, LEAH LRQ TRAINING #48 354 GRANT, LEAH LRQ TRAINING #48 357 GRANT, LEAH ILO TRAINING 48-2-A 424 GRANT, LEAH TRAINING #48 474 OUELLETTE P ENG, PLAN & MGMT INC OFFSITE 489 CLOUGHNESSY, PAT PLANT SUPPORT TEAM GSB/3B 492 WORK CONTROL OPERATION 72' ELEVATION IP2 493 OPERATIONS FIN TEAM 33 TURBIN DECK 45-1-A 494 AOEF / A. GROSSJEAN E-PLAN WPO/12-D 496 GRANT, LEAH LRQ TRAINING #48 497 GRANT, LEAH LRQ TRAINING #48 500 GRANT, LEAH LRQ TRAINING #48 501 GRANT, LEAH LRQ TRAINING #48 512 GRANT, LEAH LRQ TRAINING #48 513 GRANT, LEAH LRQ TRAINING #48 518 DOC CONTROL DESK NRC OFFSITE ko (

IPEC SITE IP-SMM-AD-103 Revision 0

-Enter-y' MANAGEMENT QUALITY RELATED ADMINISTRATIVE PROCEDURE MANUAL INFORMATIONAL USE Page 13 of 21 ATTACHMENT 10.1 SMM CONTROLLED DOCUMENT TRANSMITTAL FORM SITE MANAGEMENT MANUAL CONTROLLED DOCUMENT TRANSMITTAL FORM - PROCEDURES Page 1 of 1

-E terogy CONTROLLED DOCUMENT TRANSMITTAL FORM - PROCEDURES TO: DISTRIBUTION DATE: 3/2/07 PHONE NUMBER: 271-7057 FROM: IPEC DOCUMENT CONTROL The Document(s) identified below are forwarded for use. In accordance with IP-SMM-AD-1 03, please review to verify receipt, incorporate the document(s) into your controlled document file, properly disposition superseded, void, or inactive document(s). Sign and return the receipt acknowledgement below within fifteen (15) working days.

AFFECTED DOCUMENT: IP3 ITSIBASES/TRM DOC# REV# TITLE INSTRUCTIONS

                                • FOLLOW THE ATTACHED INSTRUCTIONS****************************
                      • PLEASE NOTE EFFECTIVE DATE***********

RECEIPT OF THE ABOVE LISTED DOCUMENT(S) IS HEREBY ACKNOWLEDGED. I CERTIFY THAT ALL SUPERSEDED, VOID, OR INACTIVE COPIES OF THE ABOVE LISTED DOCUMENT(S) IN MY POSSESSION HAVE BEEN REMOVED FROM USE AND ALL UPDATES HAVE BEEN PERFORMED IN ACCORDANCE WITH EFFECTIVE DATE(S) (IF APPLICABLE) AS SHOWN ON THE DOCUMENT(S).

NAME (PRINT) SIGNATURE DATE CC#

U

Distribution of 1P3 Technical Specification Amendment 233 (Approved by NRC February 13, 2007)

Pages are to be inserted into your controlled copy of the IP3 Improved Technical Specifications following the instructions listed below. The TAB notation indicates which section the pages are located.

TAB - Facility Operating License FOL Page 3, Amd 232 1 FOL Page 3, Amd 233 TAB - List of Effective Pages Pages 1 through 3, (Amd 232) 1 Pages 1 through 3, (Amd 233)

TAB - List of Amendments Page 14 [ Page 14 TAB - Table of Contents Page I, Amd 227 _ Page I, Amd 233 Page II, Amd 228 Page II, Amd 233 Page III, Amd 227 Page III, Amd 233 Page IV, Amd 227 Page IV, Amd 233 TAB 1.0 - Use and Application 1.1-4, Amd 205 1.1-2, Amd 233 TAB 3.4 - Reactor Coolant System (RCS) 3.4.13-1, Amd 205 3.4.13-1, Amd 233 3.4.13-2, Amd 205 3.4.13-2, Amd 233

________' _3.4.17-1, Amd 233 (new)

  • __3.4.17-2, Amd 233 (new)

Page 1 of 2

Distribution of IP3 Technical Specification Amendment 233 (Approved by NRC February 13, 2007)

Pages are to be inserted into your controlled copy of the IP3 Improved Technical Specifications following the instructions listed below. The TAB notation indicates which section the pages are located.

TAB 5.0 - Administrative Controls 5.0-13, Amd 205 5.0-13, Amd 233 5.0-14, Amd 205 5.0-14, Amd 233 5.0-15, Amd 205 5.0-15, Amd 233 5.0-16, Amd 205 5.0-17, Amd 205 Deleti ng pgs. 5.0-16 'thru,5.0-1 9" 5.0-18, Amd 205 5.0-19, Amd 205 5.0-36, Amd 205 5.0-36, Amd 233 Page 2 of 2

(4) ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, Amdt. 203 to receive, possess, and use in amounts as required any 11/27/00 byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5) ENO pursuant to the Act and 10 CFR Parts 30 and 70, to Amdt. 203 possess, but not separate, such byproduct and special 11/27/00 nuclear materials as may be produced by the operation of the facility.

C. This amended license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level ENO is authorized to operate the facility at steady state reactor core power levels not in excess of 3216 megawatts thermal (100% of rated power).

(2) Technical SDecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 233 are hereby incorporated in the License. ENO shall operate the facility in accordance with the Technical Specifications.

(3) (DELETED) Amdt. 205 2-27-01 (4) (DELETED) Amdt. 205 2-27-01 D. (DELETED) Amdt.46 2-16-83 E. (DELETED) Amdt.37 5-14-81 F. This amended license is also subject to appropriate conditions by the New York State Department of Environmental Conservation in its letter of May 2, 1975, to Consolidated Edison Company of New York, Inc., granting a Section 401 certification under the Federal Water Pollution Control Act Amendments of 1972.

Amendment No. 233

INDIAN POINT 3 TECHNICAL SPECIFICATIONS - APPENDIX A LIST OF EFFECTIVE PAGES Page 1 of 3 Page Amend Section 3.0 2 205 1 229 3 205 Table of Contents 2 226 Section 3.2.3 i 233 3 229 1 205 ii 233 4 212 2 205 iii 233 5 226 3 205 iv 233 Section 3.1.1 4 205 Section 1.1 1 I205 Section 3.2.4 1 205 Section 3.1.2 1 205 2 205 1 205 2 205 3 224 2 205 3 205 4 233 Section 3.1.3 4 205 5 225 1 205 Section 3.3.1 6 205 2 205 1 205 7 205 Section 3.1.4 2 205 8 205 1 205 3 205 Section 1.2 2 205 4 205 1 205 3 205 5 205 2 205 4 205 6 205 3 205 5 205 7 205 Section 1.3 Section 3.1.5 8 205 1 205 1 205 9 205 2 205 2 205 10 205 3 205 Section 3.1.6 11 205 4 205 1 205 12 205 5 205 2 205 13 225 6 205 3 205 14 205 7 205 Section 3.1.7 15 225 8 205 1 205 16 205 9 205 2 205 17 205 10 205 3 205 18 205 11 205 4 205 19 225 12 205 Section 3.1.8 20 225 13 205 1 205 Section 3.3.2 Section 1.4 2 205 1 205 1 205 Section 3.2.1 2 205 2 205 1 205 3 205 3 205 2 205 4 205 4 205 3 205 5 205 Section 2.0 Section 3.2.2 6 205 1 7I 225 1 205 7 205 The latest amendment reflected in this list is: Amendment 233

INDIAN POINT 3 TECHNICAL SPECIFICATIONS - APPENDIX A LIST OF EFFECTIVE PAGES Page 2 of 3 5 220I Section 3.4.13 1 205 Section 3.4.4 1 1233 2 205 1 052 ]233 3 205 Section 3.4.5 Section 3.4.14 4 205 1 205 1 205 5 205 2 205 2 205 Section 3.6.3 3 205 3 205 1 205 Section 3.4.6 4 205 2 205 1 205 5 205 3 205 2 205 Section 3.4.15 4 205 3 205 1 226 5 205 Section 3.4.7 2 205 6 205 1 205 3 205 Section 3.6.4 2 205 4 205 1 205 3 205 Section 3.4.16 Section 3.6.5 Section 3.4.8 1 226 1 2-05 1 205 2 205 Section 3.6.6 2 205 3 205 1 205 Section 3.4.9 4 205 2 205 1 225 Section 3.4.17 3 205 2 225 1 234 205 Section 3.4.10 2 233 Section 3.6.7 1 205 Section 3.5.1 1 2-05 2 205 1 222 2 205 Section 3.4.11 2 205 Section 3.6.8 1 226 3 205128 2 205 Section 3.5.2 Section 3,6.9 3 205 1 205 1 =205 Section 3.4.12 2 205 2 205 1 205 3 230 Section 3.6. 10 2 226 4 205 1 T205 3 205 Section 3.5.3 2 205 4 205 1 226 3 205 5 205 2 205 4 205 6 205 Section 3.5.4 Section 3.7.1 7 205 1 205 1 205 8 205 2 205 2 205 9 220 Section 3.6.1 3 225 10 220 1 205 4 205 11 220 2 J205 12 220 Section 3.6.2 Section 3.7.2 The latest amendment reflected in this list is: Amendment 233

INDIAN POINT 3 TECHNICAL SPECIFICATIONS - APPENDIX A LIST OF EFFECTIVE PAGES Page 3 of 3 3 232 4 232 5 232 6 205 7 205 8 205 9 210 10 232 11 232 12 205 13 233 14 233 15 233 16 Deleted 17 Deleted 18 Deleted 19 Deleted 20 205 21 224 22 224 23 224 24 224 25 224 26 232 27 205 28 205 29 205 30 206 31 225 32 227 33 227 34 225 35 225 36 233 37 232 38 232 39 232 40 232 41 232 The latest amendment reflected in this list is: Amendment 233

Entergy Nuclear Operations, Inc Indian Point 3 Nuclear Power Plant License Amendments Page 14 AMENDMENT SUBJECT LETTER DATE 232 Adoption of TSTF-258, TSTF-308, and Related 12/13/2006 Administrative Control Changes Based on NUREG-1431 233 Adoption of TSTF-449 - Steam Generator Tube 02/13/2007 Integrity

Facility Operating License No. DPR-64 Appendix A - Technical Specifications TABLE OF CONTENTS 1.0 USE AND APPLICATION 1.1 Definitions 1.2 Logical Connectors 1.3 Completion Times 1.4 Frequency 2.0 SAFETY LIMITS (SLs) 2.1 Safety Limits 2.2 Safety Limit Violations 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY SURVEILLANCE REQUIREMENT (SR) APPLICABILITY 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 Shutdown Margin (SDM) 3.1.2 Core Reactivity 3.1.3 Moderator Temperature Coefficient (MTC) 3.1.4 Rod Group Alignment Limits 3.1.5 Shutdown Bank Insertion Limits 3.1.6 Control Bank Insertion Limits 3.1.7 Rod Position Indication 3.1.8 PHYSICS TESTS Exceptions-MODE 2 3.2 POWER DISTRIBUTION LIMITS 3.2.1 Heat Flux Hot Channel Factor (FQ(Z))

3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FNH) 3.2.3 AXIAL FLUX DIFFERENCE (AFD) 3.2.4 QUADRANT POWER TILT RATIO (QPTR) 3.3 INSTRUMENTATION 3.3.1 Reactor Protection System (RPS) Instrumentation 3.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation 3.3.3 Post Accident Monitoring (PAM) Instrumentation 3.3.4 Remote Shutdown 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation 3.3.6 Containment Purge System and Pressure Relief Line Isolation Instrumentation 3.3.7 Control Room Ventilation (CRVS) Actuation Instrumentation 3.3.8 Fuel Storage Building Emergency Ventilation System (FSBEVS)

Actuation Instrumentation (continued)

INDIAN POINT 3 i Amendment 233

Facility Operating License No. DPR-64 Appendix A - Technical Specifications TABLE OF CONTENTS 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits 3.4.2 RCS Minimum Temperature for Criticality 3.4.3 RCS Pressure and Temperature (P/T) Limits 3.4.4 RCS Loops-MODES 1 and 2 3.4.5 RCS Loops-MODE 3 3.4.6 RCS Loops-MODE 4 3.4.7 RCS Loops-MODE 5, Loops Filled 3.4.8 RCS Loops-MODE 5, Loops Not Filled 3.4.9 Pressurizer 3.4.10 Pressurizer Safety Valves 3.4.11 Pressurizer Power Operated Relief Valves (PORVs) 3.4.12 Low Temperature Overpressure Protection (LTOP) 3.4.13 RCS Operational LEAKAGE 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage 3.4.15 RCS Leakage Detection Instrumentation 3.4.16 RCS Specific Activity 3.4.17 Steam Generator (SG) Tube Integrity 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1 Accumulators 3.5.2 ECCS-Operating 3.5.3 ECCS-Shutdown 3.5.4 Refueling Water Storage Tank (RWST) 3.6 CONTAINMENT SYSTEMS 3.6.1 Containment 3.6.2 Containment Air Locks 3.6.3 Containment Isolation Valves 3.6.4 Containment Pressure 3.6.5 Containment Air Temperature 3.6.6 Containment Spray System and Containment Fan Cooler System 3.6.7 Spray Additive System 3.6.8 Not Used 3.6.9 Isolation Valve Seal Water (IVSW) System 3.6.10 Weld Channel and Penetration Pressurization System (WC & PPS)

(continued)

INDIAN POINT 3 ii Amendment 233

Facility Operating License No. DPR-64 Appendix A - Technical Specifications TABLE OF CONTENTS 3.7 PLANT SYSTEMS 3.7.1 Main Steam Safety Valves (MSSVs) 3.7.2 Main Steam Isolation Valves (MSIVs) and Main Steam Check Valves (MSCVs) 3.7.3 Main Boiler Feedpump Discharge Valves (MBFPDVs), Main Feedwater Regulation Valves (MFRVs), Main Feedwater Inlet Isolation Valves (MFIIVs) and Main Feedwater (MF) Low Flow Bypass Valves 3.7.4 Atmospheric Dump Valves (ADVs) 3.7.5 Auxiliary Feedwater (AFW) System 3.7.6 Condensate Storage Tank (CST) 3.7.7 City Water (CW) 3.7.8 Component Cooling Water (CCW) System 3.7.9 Service Water (SW) System 3.7.10 Ultimate Heat Sink (UHS) 3.7.11 Control Room Ventilation System (CRVS) 3.7.12 Control Room Air Conditioning System (CRACS) 3.7.13 Fuel Storage Building Emergency Ventilation System (FSBEVS) 3.7.14 Spent Fuel Pit Water Level 3.7.15 Spent Fuel Pit Boron Concentration 3.7.16 Spent Fuel Assembly Storage 3.7.17 Secondary Specific Activity 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources-Operating 3.8.2 AC Sources-Shutdown 3.8.3 Diesel Fuel Oil and Starting Air 3.8.4 DC Sources-Operating 3.8.5 DC Sources-Shutdown 3.8.6 Battery Cell Parameters 3.8.7 Inverters-Operating 3.8.8 Inverters-Shutdown 3.8.9 Distribution Systems-Operating 3.8.10 Distribution Systems -Shutdown 3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration 3.9.2 Nuclear Instrumentation 3.9.3 Containment Penetrations 3.9.4 Residual Heat Removal (RHR) and Coolant Circulation-High Water Level 3.9.5 Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level 3.9.6 Refueling Cavity Water Level (continued)

INDIAN POINT 3 ii i Amendment 233

Facility Operating License No. DPR-64 Appendix A - Technical Specifications TABLE OF CONTENTS 4.0 DESIGN FEATURES 4.1 Site Location 4.2 Reactor Core 4.3 Fuel Storage 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.2 Organization 5.3 Unit Staff Qualifications 5.4 Procedures 5.5 Programs and Manuals 5.5.1 Offsite Dose Calculation Manual (ODCM) 5.5.2 Primary Coolant Sources Outside Containment 5.5.3 NOT USED 5.5.4 Radioactive Effluent Controls Program 5.5.5 Component Cyclic or Transient Limit 5.5.6 Reactor Coolant Pump Flywheel Inspection Program 5.5.7 Inservice Testing Program 5.5.8 Steam Generator (SG) Program 5.5.9 Secondary Water Chemistry Program 5.5.10 Ventilation Filter Testing Program (VFTP) 5.5.11 Explosive Gas and Storage Tank Radioactivity Monitoring Program 5.5.12 Diesel Fuel Oil Testing Program 5.5.13 Technical Specification (TS) Bases Control Program 5.5.14 Safety Function Determination Program (SFDP) 5.5.15 Containment Leakage Rate Testing Program 5.6 Reporting Requirements 5.6.1 NOT USED 5.6.2 Annual Radiological Environmental Operating Report 5.6.3 Radioactive Effluent Release Report 5.6.4 NOT USED 5.6.5 CORE OPERATING LIMITS REPORT (COLR) 5.6.6 NOT USED 5.6.7 Post Accident Monitoring Instrumentation (PAM) Report 5.6.8 Steam Generator Tube Inspection Report 5.7 High Radiation Area INDIAN POINT 3 iv Amendment 233

Definitions 1.1 1.1 Definitions LEAKAGE (continued) system not directly connected to the atmosphere.

Leakage past the pressurizer safety valve seats and leakage past the safety injection pressure isolation valves are examples of reactor coolant system leakage into closed systems.)

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 for leakage into closed systems and RCP seal water injection or leakoff) that is not identified LEAKAGE;
c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

MASTER RELAY TEST A MASTER RELAY TEST shall consist of energizing each master relay and verifying the OPERABILITY of each relay.

The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant loop temperature, and reactor (continued)

INDIAN POINT 3 1.1- 4 Amendment 233

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 LEAKAGE A.1 Reduce LEAKAGE to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> not within limits for within limits.

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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR Pressure boundary LEAKAGE exists.

OR Primary to secondary LEAKAGE not within limit.

INDIAN POINT 3 3.4.13-1 Amendment 233

RCS Operational Leakage 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4

SR 3.4.13.1 ------------------- NOTES------------------ ----- NOTE----

1. Not required to be performed in MODE 3 or 4 Only required to until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of steady state operation. be performed during steady
2. Not applicable to primary to secondary state operation LEAKAGE.

Verify RCS Operational LEAKAGE is within limits 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> by performance of RCS water inventory balance.

SR 3.4.13.2 ------------------ NOTE-------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.

Verify primary to secondary LEAKAGE is < 150 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> gallons per day through any one SG.

INDIAN POINT 3 3.4.13-2 Amendment 233

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

NOTE ---------------------------------

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 7 days satisfying the tube the affected tube(s) is repair criteria and maintained until the not plugged in next refueling outage or accordance with the SG tube inspection.

Steam Generator AND Program.

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

INDIAN POINT 3 3.4.17 - 1 Amendment No. 233

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 Prior to satisfies the tube repair criteria is plugged entering MODE 4 in accordance with the Steam Generator Program. following a SG tube inspection INDIAN POINT 3 3.4.17 - 2 Amendment No. 233

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 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 (continued)

INDIAN POINT 3 5.0 - 13 Amendment 233

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program (continued) 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 0.3 gpm per SG and 1 gpm through all SGs.

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.

(continued)

INDIAN POINT 3 5.0 - 14 Amendment 233

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program (continued)

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-line 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.

NOTE---------------

Pages 5.0-16 through 5.0-19 are deleted.

Next page is 5.0-20.

(conti nued)

INDIAN POINT 3 5.0 - 15 Amendment 233

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.7 Post Accident Monitoring Instrumentation (PAM) Report When a report is required by LCO 3.3.3, "Post Accident Monitoring (PAM)

Instrumentation," a report shall be submitted within the next 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.

5.6.8 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 5.5.8 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,
g. The results of condition monitoring, including the results of tube pulls and in-situ testing, and
h. The effective plugging percentage for all plugging in each SG.

INDIAN POINT 3 5.0 - 36 Amendment 233

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