ML062420133

Tech Specs, Correction Amendment 188

ML062420133
Person / Time
Site:	Kewaunee
Issue date:	08/29/2006
From:	NRC/NRR/ADRO/DORL/LPLIII-1
To:
Shared Package
ML062430179	List: ML062260326 ML062420133
References
TAC MC9581
Download: ML062420133 (8)

Text

ATTACHMENT TO LICENSE AMENDMENT NO. 188 FACILITY OPERATING LICENSE NO. DPR-43 DOCKET NO. 50-305 Replace the following page of Facility Operating License No. DPR-43 with the attached revised page. The changed area is identified by a marginal line.

REMOVE INSERT Page 3 Page 3 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT TS i TS i TS ii TS ii TS iii TS iii TS iv TS iv TS v TS v TS 1.0-7 TS 3.1-8 TS 3.1-8 TS 3.1-11 TS 4.18-1 TS 4.19-1 II TS 4.2-2 TS 4.2-2 TS 4.2-3 thru TS 4.2-6 TABLE TS 4.2-2 TABLE TS 4.2-2 TS 6.9-6 TS 6.9-6 TS 6.9-7 TS 6.22-1 TS 6.22-2 Revised by letter dated August 29, 2006

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 1.0.1 Instrumentation Surveillance ........................................................................ 1.0-3 1.0.j Modes ............................................ 1.0-4 1.0.k Reactor Crtical ........................................................................................... 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.o 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 Safetj 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 Setn s ................................................................ 2.3-1 2.34a.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 irmit 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 TSi Amnendment No. 188

Section Title Page 3.2 Chemical and Volume Control System ............................................................ 3.2-1 3.3 Engineered Safety Features and Auxiliaiy Systems ....................................... 3.3-1 3.3.a Accumulators ........................................................................... 3.3-1 3.3.b Emergency Core Cooling System .............................................. 3.3-2 3.3.c Containment Cooling Systems ............................................... 3.3-4 3.3.d Component 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 Instrumentation System .............................................................................. 3.5-1 3.6 Containment 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 3.1 0.a Shutdown Reactivity ................................................................ 3.10-1 3.10.b Power Distribution Limits ......................................................... 3.10-1 3.10.c Quadrant Power Tilt Limits ...................................................... 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 Anendment No. 188 1 Revised by letter dated August 29 , 2006

Section Title Paqe 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 I 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 Amendment No. 188 I Revised by letter dated August 29 , 2006

Section Title Page 6.0 Administrative Controls ............................................................................................ 6.1-1 6.1 Responsibility ............................................................................................. 6.1-1 6.2 Organization ............. ... ............................................................................. 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 Limit Violation ................................................................................. 6.7-1 6.8 Procedures ................................................................................................ 6.8-1 6.9 Reporting Requirements ........................................................................... 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 Requirements .......................... 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 Requirements ................................................ 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 Chemistry .................... ..... . ................................. 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 Waste Treatment Systems ......................................................... 6.19-1 6.20 Containment 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 Amendment No. 188 I Revised by letter dated August .29 , 2006

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 Amendment No. 188 1 Revised by letter dated' August 29 , 2006

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 Amendment No. 188 Revised by letter dated August 29 , 2006

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 oleration 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 Amendment No. 188 I Revised by letter dated August 29 , 2006

g. Steam Generator (SG) Tube Integrity

1. When the average reactor coolant system temperature is > 2001F 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 verify-tube 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 Amendment No. 188 Revised by letter dated August 29 , 2006

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 ROS 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 Amendment No. 188 Revised by letter dated August 29, 2006

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 Amendment No. 188 I Revised by letter dated August 29 2006

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 Amendment No. 188 I Revised by letter dated August "29, 2006

TABLE TS 4.2-2 STEAM GENERATOR TUBE INSPECTION TS Table 4.2-2 has been deleted Paae 1 of 1 Amendment No. 188 Revised by letter dated August'.29 , 2006

b. Unique Reporting Requirements

1. Annual Radiological Environmental Monitoring Report A. Routine Radiological Environmental Monitoring Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the OFF-SITE DOSE CALCULATION MANUAL (ODCM) and Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

2. Radioactive Effluent Release Report Routine. Radioactive Effluent Release Reports covering the operation of the unit for the previous calendar year shall be submitted by May 1 of each year. The report shall Include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and the PCP, and in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.

3. Special Reports A. Special reports may be required covering inspections, test and maintenance activities. These special reports are determined on an individual basis for each unit and their preparation and submittal are designated in the Technical Specifications.

(1) Special reports shall be submitted to the Director of the NRC Regional Office listed in Appendix D, 10 CFR Part 20, with a copy to the Director, Office of Inspection and Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555 within the time period specified for each report.

4. Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the Initial entry into INTERMEDIATE SHUTDOWN following completion of an inspection performed in accordance with the Specification 6.22, 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, TS 6.9-6 Amendment No. 188 I Revised by letter dated August. 2 q , 2006

d. Location, orientation (ifflinear), 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,

9. The results of condition monitoring, including the results of tube pulls and in-situ testing,

h. The effective plugging percentage for all plugging in each SG.

TS 6.9-7 Amendment No. 188 Revised by letter dated August 29 , 2006

6.22 STEAM GENERATOR (SG) PROGRAM I A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage.

during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met..

b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.

Leakage is not to exceed 150 gpd per SG.

3. The operational LEAKAGE performance criterion is specified in TS 3.1.d, "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.

TS 6.22-1 Amrendment No. 188 Revised by letter dated August 29 2006

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed.

The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

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

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

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e. Provisions for monitoring operational primary to secondary LEAKAGE.

TS 6.22-2 AmendmentNo. 188 Revised by letter dated August. 29 2006