ML061720131

Technical Specification Pages Steam Generatory Tube Integrity

ML061720131
Person / Time
Site:	Kewaunee
Issue date:	07/17/2006
From:	Jaffe D NRC/NRR/ADRO/DORL/LPLIII-1
To:	Christian D Dominion, Dominion Energy Kewaunee
D. Jaffe LPL3-1
Shared Package
ML061700091	List: ML061700208 ML061720131
References
TAC MC9581
Download: ML061720131 (10)

Text

TABLE OF CONTENTS TECHNICAL SPECIFICATIONS APPENDIX A Section Title Page 1.0 Definitions ............................................................................................................... 1.0-1 1.0.a Quadrant-to-Average Power Tilt Ratio ........................................................ 1.0-1 1.0.b Safety limits ................................................................................................ 1.0-1 1.0.c Limiting Safety System Settings .................................................................. 1.0-1 1.0.d Limiting Conditions for Operation ................................................................. 1.0-1 1.0.e Operable - Operability ................................................................................ 1.0-1 1.0.f Operating .................................................................................................. 1.0-1 1.0.g Containment System Integrity ...................................................................... 1.0-2 1.0.h Protective Instrumentation Logic ................................................................. 1.0-2 1.0.i Instrumentation Surveillance ....................................................................... 1.0-3 1.0.j Modes ....................................................................................................... 1.0-4 1.0.k Reactor Critical ........................................................................................... 1.0-4 1.0.1 Refueling Operation ................................................................................... 1.0-4 1.0.m Rated Power .............................................................................................. 1.0-4 1.0.n Reportable Event ....................................................................................... 1.0-4 1.0.0 Radiological Effluents ................................................................................. 1.0-5 1.0.p Dose Equivalent 1-131 ................................................................................ 1.0-6 1.0.q Core Operating Limits Report ...................................................................... 1.0-6 1.0.r Shutdown Margin ....................................................................................... 1.0-6 1.0.s Immediately ................................................................................................ 1.0-6 1.0.t Leakage .................................................................................................... 1.0-7 2.0 Safety Limits and Limiting Safety System Settings ..................................................... 2.1-1 2.1 Safety Limits, Reactor Core ......................................................................... 2.1-1 2.2 Safety Limit, Reactor Coolant System Pressure ........................................... 2.2-1 2.3 Limiting Safety System Settings, Protective Instrumentation .......................................................................................... 2.3-1 2.3.a Reactor Trip Settings ................................................................ 2.3-1 2.3.a.1 Nuclear Flux ........................................................ 2.3-1 2.3.a.2 Pressurizer .......................................................... 2.3-1 2.3.a.3 Reactor Coolant Temperature ............................. 2.3-2 2.3.a.4 Reactor Coolant Flow .......................................... 2.3-3 2.3.a.5 Steam Generators ............................................... 2.3-3 2.3.a.6 Reactor Trip Interlocks ........................................ 2.3-4 2.3.a.7 Other Trips .......................................................... 2.3-4 3.0 Limiting Conditions for Operation .............................................................................. 3.0-1 3.1 Reactor Coolant System ............................................................................. 3.1-1 3.1.a Operational Components ........................................................... 3.1-1 3.1.a.1 Reactor Coolant Pumps ...................................... 3.1-1 3.1.a.2 Decay Heat Removal Capability ........................... 3.1-1 3.1.a.3 Pressurizer Safety Valves .................................... 3.1-3 3.1.a.4 Pressure Isolation Valves .................................... 3.1-4 3.1.a.5 Pressurizer PORV and PORV Block Valves ......... 3.1-4 3.1.a.6 Pressurizer Heaters ............................................. 3.1-5 3.1 .a.7 Reactor Coolant Vent System .............................. 3.1-5 3.1.b Heatup & Cooldown Limit Curves for Normal Operation ............. 3.1-6 3.1.c Maximum Coolant Activity ......................................................... 3.1-7 3.1.d RCS Operational Leakage .......................... 3.1-8 3.1.e Maximum Reactor Coolant Oxygen, Chloride and Fluoride Concentration .............................................................. 3.1-9 3.1.f Minimum Conditions for Criticality ............................................ 3.1-10 3.1 .g Steam Generator (SG) Tube Integrity ...................................... 3.1-11 TS i

Section Title Page 3.2 Chem ical and Volume Control System ............................................................ 3.2-1 3.3 Engineered Safety Features and Auxiliary Systems ........................................ 3.3-1 3.3.a Accum ulators ........................................................................... 3.3-1 3.3.b Em ergency Core Cooling System .............................................. 3.3-2 3.3.c Containment Cooling Systems ................................................... 3.3-4 3.3.d Com ponent Cooling System ...................................................... 3.3-6 3.3.e Service W ater System ............................................................... 3.3-7 3.4 Steam and Power Conversion System ......................................................... 3.4-1 3.4.a Main Steam Safety Valves ........................................................ 3.4-1 3.4.b Auxiliary Feedwater System ...................................................... 3.4-1 3.4.c Condensate Storage Tank ......................................................... 3.4-3 3.4.d Secondary Activity Limits ........................................................... 3.4-3 3.5 Instrum entation System .............................................................................. 3.5-1 3.6 Containm ent System .................................................................................. 3.6-1 3.7 Auxiliary Electrical Systems ......................................................................... 3.7-1 3.8 Refueling Operations .................................................................................. 3.8-1 3.9 Deleted 3.10 Control Rod and Power Distribution Limits ................................................. 3.10-1 3.10.a Shutdown Reactivity ................................................................ 3.10-1 3.10.b Power Distribution Limits ......................................................... 3.10-1 3.10.c Quadrant Power Tilt Lim its ...................................................... 3.10-4 3.10.d Rod Insertion Limits ................................................................ 3.10-4 3.10.e Rod Misalignment Limitations .................................................. 3.10-5 3.10.f Inoperable Rod Position Indicator Channels ............................ 3.10-5 3.10.g Inoperable Rod Limitations ...................................................... 3.10-7 3.10.h Rod Drop Time ........................................................................ 3.10-7 3.10.i Rod Position Deviation Monitor ............................................... 3.10-7 3.10.j Quadrant Power Tilt Monitor .................................................... 3.10-7 3.10.k Core Average Temperature ..................................................... 3.10-7 3.10.1 Reactor Coolant System Pressure ........................................... 3.10-7 3.10.m Reactor Coolant Flow ............................................................. 3.10-8 3.10.n DNBR Parameters ................................................................... 3.10-8 3.11 Core Surveillance Instrumentation ............................................................. 3.11-1 3.12 Control Room Post-Accident Recirculation System .................................... 3.12-1 3.14 Shock Suppressors (Snubbers) ................................................................. 3.14-1 4.0 Surveillance Requirements ....................................................................................... 4.0-1 4.1 Operational Safety Review .......................................................................... 4.1-1 4.2 ASME Code Class In-service Inspection and Testing .................................. 4.2-1 4.2.a ASME Code Class 1, 2, 3, and MC Components and Supports .................................................................................. 4.2-1 4.2.b Deleted ................................................................................... 4.2-2 4.3 Deleted TS ii

Section Title r-ae 4.4 Containment Tests 4.4-1 4.4.a Integrated Leak Rate Tests (Type A) ......................................... 4.4-1 4.4.b Local Leak Rate Tests (Type B and C) ...................................... 4.4-1 4.4.c Shield Building Ventilation System ............................................. 4.4-1 4.4.d Auxiliary Building Special Ventilation System ............................. 4.4-3 4.4.e Containment Vacuum Breaker System ...................................... 4.4-3 4.4.f Containment Isolation Device Position Verification ............................................................................... 4.4-3 4.5 Emergency Core Cooling System and Containment Air Cooling System Tests ................................................................................ 4.5-1 4.5.a System Tests ........................................................................... 4.5-1 4.5.a.1 Safety Injection System ....................................... 4.5-1 4.5.a.2 Containment Vessel Internal Spray System ................................................................ 4.5-1 4.5.a.3 Containment Fan Coil Units ................................. 4.5-2 4.5.b Component Tests ...................................................................... 4.5-2 4.5.b.1 Pumps ................................................................ 4.5-2 4.5.b.2 Valves ................................................................. 4.5-2 4.6 Periodic Testing of Emergency Power System ............................................. 4.6-1 4.6.a Diesel Generators ..................................................................... 4.6-1 4.6.b Station Batteries ........................................................................ 4.6-2 4.7 Main Steam Isolation Valves ....................................................................... 4.7-1 4.8 Auxiliary Feedwater System ........................................................................ 4.8-1 4.9 Reactivity Anomalies .................................................................................. 4.9-1 4.10 Deleted 4.11 Deleted 4.12 Spent Fuel Pool Sweep System ................................................................ 4.12-1 4.13 Radioactive Materials Sources .................................................................. 4.13-1 4.14 Testing and Surveillance of Shock Suppressors (Snubbers) ...................... 4.14-1 4.15 Deleted 4.16 Reactor Coolant Vent System Tests .......................................................... 4.16-1 4.17 Control Room Postaccident Recirculation System ..................................... 4.17-1 4.18 RCS Operational Leakage ......................................................................... 4.18-1 4.19 Steam Generator (SG) Tube Integrity ........................................................ 4.19-1 5.0 Design Features ...................................................................................................... 5.1-1 5.1 Site ........................................................................................................... 5.1-1 5.2 Containment ............................................................................................... 5.2-1 5.2.a Containment System ................................................................. 5.2-1 5.2.b Reactor Containment Vessel .................................................... 5.2-2 5.2.c Shield Building .......................................................................... 5.2-2 5.2.d Shield Building Ventilation System ............................................. 5.2-2 5.2.e Auxiliary Building Special Ventilation Zone and Special Ventilation System ........................................................ 5.2-2 5.3 Reactor Core .............................................................................................. 5.3-1 5.3.a Fuel Assemblies ........................................................................ 5.3-1 5.3.b Control Rod Assemblies ............................................................ 5.3-1 5.4 Fuel Storage .............................................................................................. 5.4-1 5.4.a Criticality .................................................................................. 5.4-1 5.4.b Capacity ................................................................................... 5.4-1 5.4.c Canal Rack Storage .................................................................. 5.4-1 TS iii

Section Title Paqe 6.0 Adm inistrative Controls ............................................................................................ 6.1-1 6.1 Responsibility ............................................................................................. 6.1-1 6.2 O rganization ............................................................................................... 6.2-1 6.2.a Off-Site Staff ............................................................................ 6.2-1 6.2.b Facility Staff ............................................................................. 6.2-1 6.2.c Organizational Changes ............................................................ 6.2-1 6.3 Plant Staff Qualifications ............................................................................ 6.3-1 6.4 Training ..................................................................................................... 6.4-1 6.5 Deleted .......................................................................................... 6.5 6.5-6 6.6 Deleted ..................................................................................................... 6.6-1 6.7 Safety Lim it Violation .................................................................................. 6.7-1 6.8 Procedures .................................................. ....................................... 6.8-1 6.9 Reporting Requirem ents ............................................................................. 6.9-1 6.9.a Routine Reports ........................................................................ 6.9-1 6.9.a.1 Startup Report ..................................................... 6.9-1 6.9.a.2 Annual Reporting Requirem ents .......................... 6.9-1 6.9.a.3 Monthly Operating Report .................................... 6.9-3 6.9.a.4 Core Operating Limits Report ............................. 6.9-3 6.9.b Unique Reporting Requirem ents ................................................ 6.9-6 6.9.b.1 Annual Radiological Environmental Monitoring Report ................................................ 6.9-6 6.9.b.2 Radioactive Effluent Release Report ................... 6.9-6 6.9.b.3 Special Reports ................................................... 6.9-6 6.9.b.4 Steam Generator Tube Inspection Report ........... 6.9-6 6.10 Record Retention ..................................................................................... 6.10-1 6.11 Radiation Protection Program .................................................................... 6.11-1 6.12 System Integrity ........................................................................................ 6.12-1 6.13 High Radiation Area ................................................................................. 6.13-1 6.14 Deleted ................................................................................................... 6.14-1 6.15 Secondary W ater Chem istry ...................................................................... 6.15-1 6.16 Radiological Effluents ............................................................................... 6.16-1 6.17 Process Control Program (PCP) ................................................................ 6.17-1 6.18 Offsite Dose Calculation Manual (ODCM) .................................................. 6.18-1 6.19 Major Changes to Radioactive Liquid, Gaseous and Solid W aste Treatm ent Systems ......................................................... 6.19-1 6.20 Containm ent Leakage Rate Testing Program ............................................ 6.20-1 6.21 Technical Specifications (TS) Bases Control Program ............................... 6.21-1 6.22 Steam Generator (SG) Program ................................................................ 6.22-1 7/8.0 Deleted TS iv

LIST OF TABLES TABLE TITLE 1.0-1 ................. Frequency Notations 3.1-1 ................. Deleted 3.1-2 ................. Reactor Coolant System Pressure Isolation Valves 3.5-1 ................. Engineered Safety Features Initiation Instrument Setting Limits 3.5-2 ................. Instrument Operation Conditions for Reactor Trip 3.5-3 ................. Emergency Cooling 3.5-4 ................. Instrument Operating Conditions for Isolation Functions 3.5-5 ................. Instrument Operation Conditions for Safeguards Bus Power Supply Functions 3.5-6 ................. Accident Monitoring Instrumentation Operating Conditions for Indication 4.1-1 ................. Minimum Frequencies for Checks, Calibrations and Test of Instrument Channels 4.1-2 ................ Minimum Frequencies for Sampling Tests 4.1-3 ................. Minimum Frequencies for Equipment Tests 4.2-1 ................. Deleted 4.2-2 ................. Deleted 4.2-3 ................. Deleted TS v

t. 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.

TS 1.0-7

d. RCS Operational LEAKAGE

1. When the average RCS temperature is > 200'F, 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).

2. If the limits contained in TS 3.1.d.1 are exceeded for reasons other than pressure boundary LEAKAGE or primary-to-secondary LEAKAGE, then reduce the LEAKAGE to within their limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

3. If the limits contained in TS 3.1.d.1 for pressure boundary or primary to secondary LEAKAGE are exceeded, or the time limit contained in TS 3.1.d.2 is exceeded, then initiate action to:

- Achieve HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and

- Achieve COLD SHUTDOWN within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

4. When the reactor is critical and above 2% power, two reactor coolant leak detection systems of different operating principles shall be in operation with one of the two systems sensitive to radioactivity. Either system may be out of operation for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> provided at least one system is OPERABLE.

TS 3.1-8

g. Steam Generator (SG) Tube Integrity

1. When the average reactor coolant system temperature is > 200OF the following shall be maintained:

A. SG Tube integrity shall be maintained, and B. All SG tubes satisfying the tube repair criteria shall be plugged in accordance with the Steam Generator Program.

Note: Separate entry condition is allowed for each SG tube.

2. If the requirements of TS 3.1.g.1..B are not met, then:

A. Within 7 days verifytube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection, and B. Plug the affected tube(s) in accordance with the Steam Generator Program prior to entering INTERMEDIATE SHUTDOWN following the next refueling outage or SG tube inspection.

3. If the requirements of TS 3.1.g.2.A or TS 3.1.g.1.A are not met, then initiate action to:

- Achieve HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

- Achieve COLD SHUTDOWN within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

TS 3.1-11

b. Whenever integrity of a pressure isolation valve listed in Table TS 3.1-2 cannot be demonstrated, the integrity of the remaining pressure isolation valve in each high pressure line having a leaking valve shall be determined and recorded daily.

In addition, the position of the other closed valve located in the high pressure piping shall be recorded daily.

b. Deleted TS 4.2-2

TABLE TS 4.2-2 STEAM GENERATOR TUBE INSPECTION TS Table 4.2-2 has been deleted Page 1 of 1

4.18 RCS Operational LEAKAGE APPLICABILITY Applies to the surveillance requirements for RCS operational LEAKAGE in TS 3.1 .d.

OBJECTIVE To assure that the RCS operational LEAKAGE requirements are verified in a sufficient periodicity.

SPECIFICATION Note 1: LEAKAGE surveillances are 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.

Note 2: TS 4.18.a is not applicable to primary to secondary LEAKAGE

a. Verify RCS operational LEAKAGE, except for primary to secondary LEAKAGE, is within limits by performance of RCS water inventory balance each 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

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

TS 4.18-1

4.19 Steam Generator (SG) Tube Integrity APPLICABILITY Applies to the surveillance requirements for Steam Generator (SG) Tube Integrity in TS 3.1.g.

OBJECTIVE To assure that the Steam Generator Tube Integrity requirements are verified in a sufficient periodicity.

SPECIFICATION

a. Verify SG tube integrity in accordance with the Steam Generator Program.

b. Verify that each inspected SG tube that satisfies the tube repair criteria is plugged in accordance with the Steam Generator Program prior to entering INTERMEDIATE SHUTDOWN following a SG tube inspection.

TS 4.19-1