Text

Technical Specifications Pages Re Steam Generator Tube Integrity
	ML070530178
Person / Time
Site:	Indian Point
Issue date:	02/13/2007
From:	; NRC/NRR/ADRO/DORL/LPLI-1
To:	;
References
	TAC MD2178, TAC MD2179, TAC MD0083
	Download: ML070530178 (29)
	v • d • e

instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, Amdt. 42 to receive, possess, and use in amounts as required any 10-17-78 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. 220 possess, but not separate, such byproduct and special 09-06-01 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; 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 Amdt. 241 reactor core power levels not in excess of 3216 10-27-04 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 251, are hereby incorporated in the license. ENO shall operate the facility in accordance with the Technical Specifications.

(3) The following conditions relate to the amendment approving the conversion to Improved Standard Technical Specifications:

1. This amendment authorizes the relocation of certain Technical Specification requirements and detailed information to licensee-controlled documents as described in Table R, "Relocated Technical Specifications from the CTS," and Table LA, "Removed Details and Less Restrictive Administrative Changes to the CTS" attached to the NRC staff's Safety Evaluation enclosed with this amendment. The relocation of requirements and detailed information shall be completed on or before the implementation of this amendment.

Amendment No. 251

Facility Operating License No. DPR-26 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 SLs 2.1.1 Reactor Core SLs 2.1.2 Reactor Coolant System Pressure SL 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 (Fo(Z))

3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FAH) 3.2.3 AXIAL FLUX DIFFERENCE (AFD) (Constant Axial Offset Control (CAOC) Methodology) 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 System (CRVS) Actuation Instrumentation Indian Point 2 Amendment No. 251

Facility Operating License No. DPR-26 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 SYSTEM (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 Unit (FCU)

System 3.6.7 Recirculation pH Control 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)

Point 22 .iiii Amendment No. 251 Indian Point Amendment No. 251

Facility Operating License No. DPR-26 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 Feedwater Isolation 3.7.4 Atmospheric Dump Valves (ADVs) 3.7.5 Auxiliary Feedwater (AFW) System 3.7.6 Condensate Storage Tank (CST) 3.7.7 Component Cooling Water (CCW) System 3.7.8 Service Water System (SWS) 3.7.9 Ultimate Heat Sink (UHS) 3.7.10 Control Room Ventilation System (CRVS) 3.7.11 Spent Fuel Pit Water Level 3.7.12 Spent Fuel Pit Boron Concentration 3.7.13 Spent Fuel Pit Storage 3.7.14 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 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 4.0 DESIGN FEATURES 4.1 Site Location 4.2 Reactor Core 4.3 Fuel Storage Indian Pont iii Amendment No. 251

Facility Operating License No. DPR-26 Appendix A - Technical Specifications TABLE OF CONTENTS 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.2 Organization 5.2.1 Onsite and Offsite Organizations 5.2.2 Unit Staff 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 Radioactive Effluent Controls Program 5.5,4 Component Cyclic or Transient Limit 5.5.5 Reactor Coolant Pump Flywheel Inspection Program 5.5.6 Inservice Testing Program 5.5.7 Steam Generator (SG) Program 5.5.8 Secondary Water Chemistry Program 5.5.9 Ventilation Filter Testing Program (VFTP) 5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Program 5.5.11 Diesel Fuel Oil Testing Program 5.5.12 Technical Specification (TS) Bases Control Program 5.5.13 Safety Function Determination Program (SFDP) 5.5.14 Containment Leakage Rate Testing Program 5.5.15 Battery Monitoring-and Maintenance 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 Post Accident Monitoring Report 5.6.7 Steam Generator Tube Inspection Report Indian Point 2 iv Amendment No. 251

Definitions 1.1 1.1 Definitions 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.

MASTER RELAY TEST A MASTER RELAY TEST shall consist of energizing each required master relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required master relay. The MASTER RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.

MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-i with fuel in the reactor vessel.

INDIAN POINT 2 1.1 -3 Amendment No. 251

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 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months LEAKAGE not within 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

'LEAKA AGE exists.

OR Primary to secondary LEAKAGE not within limit.

INDIAN POINT 2 3.4.13- 1 Amendment No. 251

RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1

- NOTES -

1. Not required to be performed in MODE 3 or 4 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of steady state operation.

2. Not applicable to primary to secondary LEAKAGE.

Verify RCS Operational LEAKAGE is within limits by performance of RCS water inventory balance. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months SR 3.4.13.2

- NOTE -

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 gallons per day through any one SG. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months INDIAN POINT 2 3.4.13 -2 Amendment No. 251

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 MODE 4 following the Steam Generator next refueling outage Program. 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.

OR SG tube integrity not 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months maintained.

INDIAN POINT 2 3.4.17 - 1 Amendment No. 251

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 INDIAN POINT 2 3.4.17 -2 Amendment No. 251

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.6 Inservice Testing Program (continued)

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities,

c. The provisions of SR 3.0.3 are applicable to inservice testing activities, and

d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.

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 INDIAN POINT 2 5.5 -6 Amendment No. 251

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Program (continued) 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 150 gpd per SG.

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 120, 90, 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 INDIAN POINT 2 5.5 -7 Amendment No. 251

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Program (continued) 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 48 effective full power months or two 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). Ifdefinitive 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.

INDIAN POINT 2 5.5 -8 Amendment No. 251

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

8. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report", April 1995;

9. WCAP-10079-P-A, "NOTRUMP, A Nodal Transient Small Break and General Network Code", August 1985;

10. WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code", August 1985; and

11. WCAP-10054-P-A, Addendum 2, Revision 1, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection Into the Broken Loop and Cosi Condensation Model", July 1997.

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 to the NRC upon issuance for each reload cycle.

5.6.6 Post Accident Monitoring Report When a report is required by Condition B or F 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, INDIAN POINT 2 5.6 -4 Amendment No. 251

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.7 Steam Generator Tube Inspection Report (continued)

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 2 5.6 -5 Amendment No. 251

(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 Specifications 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

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 (FN,)

3.2.3 AXIAL FLUX DIFFERENCE (AFD) 3.32.4 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 9233

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)

(coF t nueud) iNDiAN POINT 3 11 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)

T IKnT AtNf' T NIT D i-ii I I nil ruu kjN, J Amiendmient 233 A

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 T kr.r T A 1INu I AN fi-lTLI rvi Nii J

-r 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. 1igpm 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 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.

INDIAN POINT 3 3.4,13-1 Amendment 233

RCS Operational Leakage 3.4.13 SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 0.02 hours 1.190476e-4 weeks 2.7396e-5 months by performance of RCS water inventory balance.

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

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.

INDIAN POINT 3 3.4.13 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not met. AND OR B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SG tube integrity not maintai ned.

INDIAN POINT 3 3.4.17 - I 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, PINT POT 3 3.4.117 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 asofrom 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.

(continued)

INDIAN POINT 3 5.0 - 15 Amendment 2133

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

v t e Letter Sequence Other
TAC:MD0083, Steam Generator Tube Integrity, Delete Fire Protection Program Implementation from Administrative Controls, TSTF-44 (Approved, Closed) TAC:MD2178, Steam Generator Tube Integrity, Delete Fire Protection Program Implementation from Administrative Controls, TSTF-44 (Approved, Closed) TAC:MD2179, Steam Generator Tube Integrity, Delete Fire Protection Program Implementation from Administrative Controls, TSTF-44 (Approved, Closed)
Initiation Request Acceptance... Administration Questions 12 July 2006 Answers 30 August 2006 Results Approval, Approval Other: ML070530178
MONTHYEAR2006-08-30 [Table View]

ML070530178

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