ML17346A248

Proposed Tech Specs Revising Inservice Insp & Inservice Test Program

ML17346A248
Person / Time
Site:	Turkey Point
Issue date:	04/02/1984
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17346A247	List: ML17346A246 ML17346A248
References
NUDOCS 8404050197
Download: ML17346A248 (58)

Text

TABLE OF CONTENTS Section Title ~Pa e TECHNICAL SPECIFICATIONS 1.0 DEFINITIONS l-l 1.1 Safety Limits l-l 1.2 Limiting Safety System Settings 1-1 1.3 . Limiting Conditions for Operation 1-1 lA Operable l-l 1.5 Containment Integrity 1-2 1.6 Protective Instrumentation Logic 1-2 1.7 Instrumentation Surveillance 1-3 1.8 Shutdown 1-3 1.9 Power Operation 1-0 1.10 Refueling Operation 1-0 1.11 Rated Power 1-0 1.12 Thermal Power 1-0 1.13 Design Power 1-0 1.10 Dose Equivalent I-131 1-5 1.15 Power Tilt 1-5 1.16 Interim Limits 1-6 1.17 Low Power Physics Tests 1-6 1.18 Engineered Safety Features 1-6 1.19 Reactor Protection System 1-6 1.20 Safety Related Systems and Components 1-6 1.21 Per Annum 1-6 1.22 Reactor Coolant System Pressure Boundary Integrity 1-6 1.23 Coolant Loop 1-7 1.20 E-Average Disintegration Energy 1-7 2.0 SAFETY LIMITS AND LIMITINGSAFETY SYSTEM SETTINGS 2.1-1 2.1 Safety Limit, Reactor Core 2.1-1 2.2 Safety Limit, Reactor Coolant System Pressure 2.2-1 2.3 Limiting Safety Sy te- Setting, Protective Instrumentation 2.3-1 3.0 LIMITING CONDITIONS FOR OPERATION 3.0-1 3.1 Reactor Coolant System 3.1-1 Operational Components 3.1-1 Pressure-Temperature Limits 3.1-2 Leakage 3.1%

Maximum Reactor Coolant Activity 3.1-5 Reactor Coolant Chemistry 3.1-6 DNB Parameters 3.1-7 3.2 Control Rod and Pow'er Distribution Limits 3.2-1 Control Rod Insertion Limits 3.2-1 Misaligned Control Rod 3.2-2 Rod Drop Time 3.2-2 Inoperable Control Rods 3.2-2 Control Rod Position Indication , 3.2-3 Power Distribution Limits 3.2-3 In-Core Instrumentation 3.2-7 Axial Offset Alarms 3.2-8 3.3 Containment 3.3-1

~4050+97 840402 PDR ADQCK 05000250 P PDR S9' Amendment Nos. 95 and

TABLE OF CONTENTS (Continued)

Section Title ~Pa e Engineering Safety Features 3.0-1 Safety Injection and RHR Systems 3.0-1 Emergency Containment Cooling Sy -"ms 3.0-3 Emergency Containment Filtering Sy"tern 3.0-0 Component Cooling System 3.0A htake Cooling Water System 3.0-5 Post Accident Containment Vent System 3.0-6 Control Room Ventilation 3.0-6 3.5 Instrumentation 3.5-1 3.6 Chemical and Volume Control System 3.6-1 3.7 Electrical Systems 3.7-1 3.8 Steam Power Conversion Systems 3.8-1 3.9 Radioactive Materials Release 3.9-1 Liquid Wastes 3.9-1 Gaseous Wastes 3.9-3 Containerized Wastes 3.9-5 3.10 Refueling 3.10-1 3.11 Miscellaneous Radioactive Materials Sources 3.11-1 3.12 Cask Handling 3.12-1 3.13 Snubbers 3.13-1 3.10 Fire Protection Systems 3.10-1 3.15 Overpressure 'Mitigating System 3.15-1 l.

0.0 SURVEILLANCE REQUIREMENTS 0.0-1 0.1 Operational Safety Review 0.1-1 c.2 Reactor Coolant System In Service Inspection 0.2-1 Reactor Coolant System Integrity 0.3-1 Containment Tests 0.0-1 Integrated Leakage Rate Test - Post Operational 0.0-1 Local Penetration Tests 4.0-2 Report of Test Results 0.0-2 Isolation Valves 0.0-3 Residual Heat Removal Sy 'em OA>>3 Tendon Surveillance 0.0-0 End Anchorage Concrete Surveillance 4.0-6 Liner Surveillance 0.0-7 0.5 Safety Injection 4.5-1 0.6 Emergency Containment Cooling Systems 0.6-1 0.7, Emergency Containment Filtering and Post Accident Containment Vent Systems 0.7-1 0.8 Emergency Power System Periodic Tests 0.8-1 0.9 Main Steam Isolation Valves 0.9-1 Amendment tVos. and

TABLE OF CONTENTS (Continued)

Section Title ~Pa e 0.10 Auxiliary Feedwater System 0.10-1 0.11 Reactivity Anomalies O.l l-l 0.12 Environmental Radiation Survey 0.12-1 4.13 Radioactive Materials Sources Surveillance 0.13-1 0.10 Shock Suppressors (Snubbers) 0.10-1 0.15 Fire Protection Systems 0.15-1 0.16 Overpressure Mitigating System 0.16-1 5.0 DESIGN FEATURES 5.1-1 5.1 Site 5.l-l 5.2 Reactor 5.2-1 5.3 Containment 5.3-1 5.0 Fuel Storage 5A-I 6.0 ADMINISTRATIVECONTROLS 6-1 6.1 Responsibility 6-1 6.2 Organization 6-1 6.3 Facility Staff Qualifications 6-5 6A Training 6-5 6.5 Review and Audit 6-6 6.6 Reportable Occurrence Action 6-10 6.7 Safety Limit Violation 6-10

'.8 Procedures 6-10 ~

6.9 Reporting Requirements 6-16 6.10 Record Retention 6-27 6.11 Radiation Protection Program 6-29 6.12 High Radiation Area 6-29 6.13 Environmental Qualif ication 6-30 B2.1 Bases for Safety Limit, Reactor Core B2.1-1 B2.2 Bases for Safety Limit, Reactor Coolant System Pressure B2.2-1 B2.3 Bases for Limiting Safety System Settings, Protective Instrumentation B2.3-1 B3.1 Bases for Limiting Conditions for Operation, Reactor Coolant System B3.1-1 B3.2 Bases for Limiting Conditions for Operation, Control and Power Distribution Limits B3.2-1 B3.3 Bases for Limiting Conditions for Operation, Containmen't B3.3-1 B3.0 Ba~es for Limiting Conditions for Operation, Engineered Safety Features B3.0-1 90 Amendment Nos. 62 and 53 12/19/80

TABLE OF'ONTENTS (Continued)

Section Title ~Pa e Bases for Limiting Conditions for Operation, Instrumentation B3.5-1 B3.6 Bases for Limiting Conditions for Operation, Chemical and Volume Control Sy +e- B3.6-1 B3.7 Bases for Limiting Conditions for Operation, Electrical Sy-gems B3.7-1 B3.8 Bases for Limiting Conditions for Operation, Steam and Power Conversion Systems B3.8-1 B3.9 Bases for Limiting Conditions for Operation, Radioactive Materials Release B3.9-1 B3.10 Bases for Limiting Conditions for Operation, Refueling B3.10-1 B3.11 Bases for Limiting Conditions for Operation, Miscellaneous Radioactive Material Sources B3.11-1 B3.12 Bases for Limiting Conditions for Operation, Cask Handling B3.12-1 B3.13 Bases for Limiting Conditions for Operation, Snubbers B3.13-1 B3.10 Bases for Fire Protection System B3.10-1 B3.15 Bases for Limiting Conditions of Operation, Over pressure Mitigating System B3.15-1 B0.0 Bases for Surveillance Requirements B0.0-1 B0.1 Bases for Operational Safety Review BO.l-i~

B0.2 Bases for Reactor Coolant System In-Service Inspection B0.2-1 B0.3 Bases for Reactor Coolant System Integrity B0.3-1 Bases for Containment Tests B0.0-1 B4.5 Bases for Safety Injection Tests B0.5-1 B0.6 Bases for Emergency Containment Cooling System Tests B0.6-1 B0.7 Bases for Emergency Containment Filtering and Post Accident Containment Venting Systems Tests B0.7-1 B0.8 Bases for Emergency Power System Periodic Tests B0.8-1 B4.9 Bases for 'Main Steam Isolation Valve Tests B0.9-1 B0.10 Bases for Auxiliary Feedwater System Tests B0.10-1 B0.11 Bases for Reactivity Anomalies B0.11-1 B0.12 Bases for Environmental Radiation Survey B0.12-1 B0.13 Bases for Fire Protection Systems B0.13-1 B0.10 Bases for Snubbers B4.10-1 B0.15 Bases for Surveillance Requirements, Overpressure Mitigating System B0.15-1 iv Amendment Nos. and

LIST OF TABLES Table Title 3.5-1 Instrument Operating Conditions for Reactor Trip 3.5-2 Engineering Safety Features Actuation 3.5-3 Instrument Operating Conditions for isolation Functions 3.5-0 Engineered Safety Feature Set Points 3.13>>1 Safety Related Snubbers 3.10-1 Fire Detection System O. 1-1 Minimum Frequencies for Checks, Calibrations and Test of Instrument Channels 0.1-2 Minimum Frequencies for Equipment and Sampling Tests 0.2-1 Deleted 0.2-2 Minimum Number of Steam Generators to be Inspected During Inservice Inspection 0.2-3 Steam Generator Tube Inspection 0.12-1 Operational Environmental Radiological Surveillance Program 0.12-2 Operational Environmental Radiological Surveillance Program Types of Analysis 6.2-1 Operating Personnel Amendment Nos. and

LIST OF FIGURES

~F1 ure Title Reactor Core Thermal and Hydraulic Safety Limits, Three Loop Operation 2.l-la Reactor Cor e Thermal and Hydraulic Safety Limits, Three Loop Operation 2.1-1b Reactor Core Thermal and Hydraulic Safety Limits, Three Loop Operation 2.1-2 Reactor Core Thermal and Hydraulic Safety Limits, Two Loop Operation 3.1-1 DOSE EQUIVALENTI-131 Primary Coolant Specific Activity. Limit Versus Percent of RATED POWER with the Primary Coolant Specific Activity > 1.0 p Ci/gram Dose Equivalent I-131 3.l-la Reactor Coolant System Heatup and Cooldown Pressure Limits 3.1-lb Reactor Coolant System Heatup and Cooldown Pressure Limits 3.1-lc Reactor Coolant System Heatup and Cooldown Pressure Limits 3.1-1d Reactor Coolant System Heatup and Cooldown Pressure Limits 3.1-2 Radiation Induced Increase in Transition Temperature for A302-B Steel 3.1-2c Radiation Induced Increase in Transition Temperature for A302-B Steel 3.1-2d Radiation Induced Increase in Transition Temperature for A302-B Steel 3.2-1 Control Group Insertion Limits for Unit 0, Three Loop Operation 3.2-1a Control Group Insertion Limits for Unit 0, Two Loop Operation 3.2-1b Control Group Insertion Limits for Unit 3, Three Loop Operation 3.2-lc Control Group Insertion Limits for Unit 3, Two Loop Operation 3.2-2 Required Shutdown Margin 302-3 Hot Channel Factor Normalized Operating Envelope 3.24 Maximum Allowable Local KW/FT 0.12-1 Sampling Locations 6.2-1 Offsite Organization Chart 6.2-2 Plant Organization Chart B3.1-1 Effect of Fluence and Copper Content on Shift of RTNDT for Reactor Vessel Steels Exposed to 550 F Temperature B3.1-2 Fast Neutron Fluence (E>lMEV) as a function of Effective Full Power Years B3.2-1 Target Band on Indicated Flux Difference as a Function of Operating Power Level B3.2-2 Permissible Operating Band on Indicated Flux Difference as a Function of Burnup (Typical) vi Amendment Nos. 95 and 39

t 3.3 CONTAINMENT

~Ob'ective:

To define the operating status of the containment.

1. CONTAINMENT INTEGRITY

a. The containment integrity (as defined in 1.5) shall riot be violated unless the reactor is in the cold shutdown condition. Specification 3.0.1 applies to 3.3.1.

b. The containment integrity shall not be violated when the reactor vessel head is removed unless the reactor is in the refueling shutdown condition.

2. INTERNAL PRESSURE If the internal pressure exceeds 3 psig or the internal vacuum exceeds 2 psig, the condition shall be corrected within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or the reactor shall be brought to hot shutdown.

3. CONTAINMENT ISOLATION VALVES With Keff >0.99. % thermal power excluding decay heat >0, and a*

average coolant temperature Tavg >200 F, the following conditions shall be met:

The containment isolation valves shall be operable in accordance with the requirements of Specification 0.0.3., or the valve is closed.

3.3-1 Amendment Nos. and

5. TWO residual heat removal pumps shall be operable.

6. TWO residual heat exchangers shall be operable.

7. All valves, interlocks and piping associated with the above components and required for post accident operation, shall be operable except'valves that are positioned and locked. Valves 860-A, B, 862-A, B, 865-A, B, C; 866-A, B shall have power removed from their motor operators by locking open the circuit breakers at the Motor Control Centers. The air supply to valve 758 shall be shut off to the valve operator.

b. During power operation, the requirements of 3.0.1a may be modified to allow one of the following components to be inoperable (including associated valves and piping) at any one time except for the cases stated in 3.0.1.b.2. If the system is not restored to meet the requirements of 3.0.1a within the time period specified, the reactor shall be placed in the hot shutdown condition. If the requirements of 3.0.1a are not satisfied within an additional 08 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the reactor shall be placed in the cold shutdown condition. Specification 3.0.1 applies to 3.0.l.b.

l. ONE accumulator may be out of service for a period of up to 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.

2. ONE of FOUR safety injection pumps may be out of service for 30 days. A second safety injection pump may be out of service, provided the pump is restored to operable status within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. Compliance with the requirements of Specification 0.0.3 shall be verified for the remaining 2 pumps.

3. ONE channel of heat tracing on the flow path may be out of service for 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.~

ONE residual heat removal pump may be out of service, provided the pump is restored to operable status within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. Compliance with the requirements of Specification 0.0.3 shall be verified for the other pump.

+ See r eference (11) on page B.3.0-'2 3.0-2 Amendment Nos. and

5. ONE residual heat exchanger may be out of service for a period of 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. Compliance with the requirements of Specification 0.0.3 shall be verified for the remaining heat exchanger.

6. Any valve in the system may be inoperable provided repairs are completed within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. Compliance with the requirements of Specification 4.0.3 shall be verified for all valves that provide the duplicate function.

7. To permit temporary operation of the valve, e.g., for surveillance of valve operability, for the purpose of valve maintenance, etc., the valves specified in 3Ã.l.a.7 may be unlocked and may have supplied air or electric power restored for a period not to exceed 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

c. During power operation three Reactor Coolant Loops shall be in oper ation.

1. With less than three Reactor Coolant Loops in operation, the reactor must be in hot shutdown within one hour.

d. In hot shutdown at least two Reactor Coolant Loops shall be operable and at least one Reactor Coolant Loop shall be in operation.+

1. With less than two Reactor Coolant Loops operable, restore the required Coolant Loops to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or reduce Tavg to less than or equal to 350 F within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2. With no Reactor Coolant Loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required Coolant Loop to operation.

e. With average coolant temperature less than 350 F, at least two Coolant Loops shall be operable or immediate corrective action must be taken to return two Coolant Loops to operable as soon as possible. One of these Coolant Loops shall be in operation.+

1. With no Coolant Loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required Coolant Loop to operation.

All reactor coolant pumps and residual heat removal pumps may be de-energized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided 1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and 2) core outlet temperature is maintained as last 10 F below saturation temperature.

3.0-2a Amendment Nos. and

In refueling shutdown, at least one residual heat removal Coolant Loop shall be in operation or all operations involving an increase in the reactor decay heat load or a reduction in boron concentration in the Reactor Coolant System must be suspended, and all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere must be closed in four hours. As an exception, the single residual heat removal Coolant Loop may be removed from operation during the performance of core alterations in the vicinity of the reactor pressure vessel hot legs, provided core outlet temperature is maintained below 160 F.

g. In refueling shutdown, when the water level above the top of the pressure vessel flange is less than 23 feet, two residual heat removal Coolant Loops shall'be operable or action to return two residual heat removal Coolant Loops to operable shall be taken as soon as possible.

2. EVlERGENCY CONTAINMENT COOLING SYSTEMS

a. The reactor shall not be made critical, except for low power physics tests unless the following conditions are met:

1. Three emergency containment cooling units are operable.

2. Two containment spray pumps are operable.

3. All valves and piping associated with the above components, and required for post accident operation, are operable.

b. During power operation, the requirements of 3.0.2a may be modified to allow one of the following components to be inoperable (including associated valves and piping) at any one time. If the system is not restored to meet the requirements of 3.0.2a within the time period specified, the reactor shall be placed in the hot shutdown condition. If the requirements of 3.0.2a are not satisfied within an additional 08 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the reactor shall be placed in the cold shutdown condition. Specification 3.0.1 applies to 3.0.2.b.

3A-3 Amendment Nos. 87 R 81

l. ONE emergency containment cooling unit may be out of service for a period of 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. Prior to initiating maintenance the other TWO units shall be tested to demonstrate operability.

2. ONE containment spray pump may be out of service provided it is restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Compliance with the requirements of Specification 0.0.3 shall be verified for the other pump.

3. Any valve in the system may be inoperable provided repairs are completed within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. Compliance with the requirements of Specification 0.0.3 shall be verified for all other valves that provide the duplicate function.

3. EMERGENCY CONTAINMENT FILTERING SYSTEM

a. The reactor shall not be made critical, except for low power physics tests unless:

1. THREE emergency containment filtering units are operable.

2. All valves, interlocks and piping associated with the above components and required for post-accident operation, are operable.

b. During power operation:

l. ONE unit may be inoperable for a period of 7 days if the other TWO are operable.

2. Any valve in the system may be inoperable provided repairs are completed within 7 days. Compliance with the requirements of Specification 0.0.3 shall be verified for all other valves that provide the duplicate function.

3. If after 7 days the unit is still inoperable, Specification 3.0.1 applies to 3.0.3.b.

0. COMPONENT COOLING SYSTEM

a. The reactor shall not be made critical, except for low power physics tests unless the following conditions are met:

l. THREE component cooling pumps are operable.

2. THREE component cooling heat exchangers are operable.

3. All valves, interlocks and piping associated with the above components are operable.

3.0-0 Amendment Nos. and

ID

b. During power operation, the requirements of 3AJk.a may be modified as stated below. If the sytem is not restored to meet the conditions of 3JtA.a within the time period specified, the reactor shall be placed in the hot shutdown condition. If the requirements of 3.0A.a are not satisfied within an additional 08 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the reactor shall be placed in the cold shutdown condition.

Specification 3.0.l applies to 3.0.0.b.

l. ONE pump may be out of service for 7 days.

2. a) A second pump may be out of service for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Compliance with the requirements of Specification 0.0.3 shall be verified for the third pump.

b) ONE heat exchanger may be out of service for a period of 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. Compliance with the requirements of Specification 0.0.3 shall be verified for the other heat exchangers.

5. INTAKE COOLING WATER SYSTEM

a. The reactor shall n'ot be made critical unless the following conditions are met:

1. THREE intake cooling water pumps and TWO headers are operable.

2. All valves, interlocks and piping associated with the operation of these pumps, and required for post accident operation, are operable.

b. During power operation, the requirements of 3.0.5.a., above, may be modified to allow any one of the following components to be inoperable provided the remaining systems are in continuous operation. If the system is not restored to meet the requirements of 38.5.a. within the time period specified, the reactor shall be placed in the hot shutdown condition. If the requirements of 38.5.a are not satisfied within an additional 08 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the reactor shall be placed in the cold shutdown condition. Specification 3.0.1 applies to 3.4.5.b.

l. One of the two headers may be out of service for a period of 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. Compliance with the requirements of Specification 4.0.3 shall be verified for the remaining header.

2. One intake cooling water pump may be out of service for a period of 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. Compliance with the requirements of Specification 0.0.3 shall be verified for the other two pumps.

3.0-5 Amendment Nos. and

6. POST ACCIDENT CONTAINMENTVENT SYSTEM

a. The reactor shall not be made critical, except for low power physics tests unless:

1. The post accident containment vent system is operable.

2. All valves, interlocks, and piping associated with the above components and required for post-accident operation are operable.

b. During power operation:

1. The unit may be inoperable for a period of 7 days.

2. Any valve in the system may be inoperable provided repairs are completed within 7 days. Compliance with the requirements of Specification 0.0.3 shall be verified for all valves that provide the duplicate function.

3. If after 7 days the unit is still inoperable, Specification 3.0.1 applies to 3.0.6.b.

7. CONTROL ROOM VENTILATION

a. The reactor shall not be made critical, except for low power physics tests unless:

l. The control room ventilation system is operable.

2. All valves, interlocks, and piping associated with the above

'components and required for post-accident operation are operable.

b. During power operation:

1. The unit may be inoperable for a period of 3 1/2 days.

2. Any valve in the system may be inoperable provided repairs are completed within 3 1/2 days. Prior to initiating maintenance, all valves that provide the duplicate function shall be tested to demonstrate operability.

3. If after 3 1/2 days the unit. is still inoperable, Specification 3.0.1 applies to 3.0.7.b.

A'.0>>6 Amendment Nos. and

3.8

~ STEAM AND POWER CONVERSION SYSTEMS

~ ~ ~

~ *>>"'~

l E'ystems.

~Ob'ective: To define conditions of the steam-relieving capacity and auxiliary feedwater system.

1. When the reactor coolant of a nuclear unit is heated above 350oF, the following conditions must be met:

a. TWELVE (12) of its steam generator safety valves shall be operable (except for testing).

b. System piping, interlocks and valves directly associated with the related components shall be operable.

c. Its condensate storage tank shall contain a minimum of 185,000 gallons of water.

d. Its main steam stop valves shall be operable in accordance with the requirements of Specification 0.0.3.

2. The iodine-131 activity on the secondary side of a steam generator shall not exceed 0.67 pCi/cc.

3. During power operation, if any of the conditions of 3.8.1 or 3.8.2 cannot be met within 08 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the reactor shall be shutdown and the reactor coolant temperature reduced below 350oF.

The following number of independent steam generator auxiliary feedwater pumps and associated flow path shall be operable in accordance with the requirements of Specification 0.10 when the reactor coolant is heated above 350oF:

a. Sin le Nuclear Unit 0 eration Two auxiliary feedwater pumps capable of being powered from an operable steam supply.

b. Dual Nuclear Unit 0 eration Three auxiliary feedwater pumps capable of being powered from an operable steam supply.

5. During power operation, if any of the conditions of 3.8.0 cannot be met, the reactor shall be shutdown and the reactor coolant temperature reduced below 350oF, unless one of the following conditions can be met:

3.8-1 Amendment Nos. and

a. For single unit operation with one of the two required auxiliary feedwater pumps inoperable, restore the inoperable pump to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the reactor shall be shutdown and the reactor coolant temperature reduced below 350 F within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. For dual unit operation with one of the three required auxiliary feedwater pumps inoperable, restore the inoperable pump to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or a reactor shall be shutdown and its reactor coolant temperature 'reduced below 350 F within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

3.8-2 Amendment Nos. and

0.0 SURVEILLANCE RE UIREMENTS 0.0.1 Specified intervals may be adjusted plus or minus 2596'to accommodate normal test schedules.

0.0.2 When the reactor is in a shutdown condition, some of the surveillance requirements discussed in this section are not required to be satisfied provided that the safety limits or limiting conditions for operation for the shutdown status are satisfied. When a surveillance activity is not completed because the reactor is shutdown and the surveillance is not required, the surveillance requirement shall be met prior to the time indicated in the applicable footnote.

0.0.3 Surveillance ReqUirements for inservice inspection and testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows:

a) Inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.5'5a(g)(6)(i).

S.

b) Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:

ASME Boiler and Pressure Vessel Required frequencies for Code and applicable Addenda performing inservice terminology for inservice ~

inspection and testing ins ection and testin activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 180 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days 0.0-1 Amendment Nos. and

~ ~

c) The provisions of Specification 0.0.1 are applicable to the above required frequencies for performing inservice inspection and testing activities.

d) Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements.

e) Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.

0.0-2 Amendment Nos. and

0.1 OPERATIONAL SAFETY REVIEW Applies to items directly related to safety limits and limiting conditions for operation.

~Ob'ective: To specify the minimum frequency and type of surveillance to be applied to equipment and conditions.

Calibration, testing, and checking of analog channels and testing of logic channels shall be performed as specified in Table O.l-l.

Equipment and sampling tests shall be conducted as specified in Table 0.1-2.

0 Amendment Nos. and

TABLE 0.1-2 (Sheet 2 of 3)

/

MINIMUMFRE UENCIES FOR E UIPMENT AND SAMPLING TESTS Max. Time Check ~recruencn Between Tests (Days)

5. Control Rods (cont'd) Partial movement of Biweekly while 20 full length rods critical

6. Pressurizer Safety Valves Tests performed in compliance

7. Main Steam Safety Valves with Specification 0.0.3

8. Containment Isolation Trip Functioning Each refueling shutdown

9. Refueling System Interlocks Functioning Prior to each NA refueling

10. Accumulator Boron Concentration At least once per 31 days and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after each t

solution volume increase of . 196 of tank volume.t

11. Reactor Coolant System Evaluate Daily Leakage

12. Diesel Fuel Supply Fuel inventory Weekly 10

13. Spent Fuel Pit Boron Concentration Prior to refueling

10. Secondary Coolant I-131 Concentration Weekly+0 10

15. Vent Gas and Particulates I-131 and Particulate Activity Weekly+ 10

16. Fire Protection Pump and Operable Monthly Power Supply

17. Turbine Stop and Control Closure 'monthly+++

Valves, Reheater Stop and Intercept Valves

18. LP Turbine Rotor Inspector V, MT, PT Every 5 years 6 years (w/o rotor disassembly)

19. Spent Fuel Cask Crane Functioning Within 7 days 7 days when Inter locks crane is being used to maneuver spen fuel cask.

Amendment Nos. and

l 0.2 REACTOR COOLANT SYSTEM IN-SERVICE INSPECTION

~Ob'ective:

ecif ication:

reactor coolant system boundary.

To assure the continued integrity of the reactor coolant system boundary.

0.2.1 a. Except as listed below, there are no additional surveillance requirements other than those required by Specification 0.0.3.

0.2.2 The inspection interval shall be 10 years.

0.2.3 In addition to the requirements of Specification 0.0.3 each reactor coolant pump flywheel shall be inspected per the recommendations of Regulatory Position CA.b of Regulatory guide 1.10, Revision 1, August 1975.

Irradiation specimens shall be examined by Tensile and Charpy V Notch methods. Capsule 1 shall be removed and examined at the first region replacement. Capsule 2 shall be removed and examined at the fourth region replacement. Capsule 3 shall be

'removed and examined after twenty years of operation. Capsule 0 shall be removed and examined after thirty years of operation.

Capsule 5 shall be removed and examined after forty years of operation.

0.2-1 Amendment Nos. and

0.2.5 STEAM GENERATOR INSPECTION 0.2.5.1 Steam Generator Sam le Selection and Ins ection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 0.2-2.

0.2.5.2 Steam Generator Tube Sam le Selection and Ins ection-The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 0.2-3. The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 0.2.5.3 and the inspected tubes shall be ver ified acceptable per the acceptance criteria of Specification 0.2.5.0. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:

a. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 5096 of the tubes inspected shall be from these critical areas.

b. The sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:

1. All nonplugged tubes that previously had detectable wall penetrations (greater than 20'6),

and

2. Tubes in those areas where experience has indicated potential problems.

3. A tube inspection (pursuant to Specification 0.2.5.0.8) shall be performed on each selected tube.

If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.

0.2-2 Amendment Nos. 81 R 75

c. The tubes selected as the second and third samples in the inservice inspection may be less than a full tube inspection by concentrating (selecting at least 5096 of the tubes to be inspected) the inspection on those areas of the tube sheet array and on those portions of tubes where tubes with imperfections were previously found.

The results of each sample inspection shall be classified into one of the following three categories:

~cate or Ins ection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

C-2 One or more tubes but not more than 16 of the total tubes inspected are defective, or between 5%

and 10% of the total tubes inspected are degraded tubes.

C-3 Vore than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.

NOTE: In all inspections, previously degraded tubes must exhibit significant (greater than 10;6) further wall penetrations to be included in the above percentage calculations.

0.2.5.3 Inspection Fre uencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:

a. The first inservice inspection shall be performed after six effective full power months of operation but within 20 calendar months following replacement of steam generators. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 20 calendar months after the previous inspection.

If two consecutive inspections following service under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 00 months.

0.2-3 Amendment Nos. 81 R 75

b. If the inservice inspection of a steam generator conducted in accordance with Table 0.2-3 requires a third sample inspection whose results fall in Category C-3, the inspection frequency shall be reduced to at least once per 20 months. The reduction in inspection frequency shall apply until a subsequent inspection demonstrates that a third sample inspection is not required.

c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 0.2-3 during the shutdown subsequent to any of the following conditions:

1. Primary-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.1.3.g.

2. A seismic occurrence greater than the operating Basis Earthquake (OBE).

3. A loss-of-coolant accident resulting in rapid depressurization of the primary system, or L A main steam line or feedwater line break resulting in rapid depressurization of the affected steam generator.

0.2.5.0 Acce tance Criteria

a. As used in this Specification:

dimensions, finish or contour of a tube from that required by fabrication drawings or specifications.

Eddy-current testing, indications below 20% of the nominal tube wall thickness, if detectable may be considered as imperfections.

2.

wastage, wear or general corrosion occurring on either inside or outside of a tube.

d De raded Tube means a tube containing imperfections greater than or equal to 2096 of the nominal wall thickness'caused by degradation.

0.2-0 Amendment Nos. 81 k 75

wall thickness affected or removed by degradation.

5. Defect means an imperfection or such severity that it exceeds the plugging limit . A tube containing a defect is defective. Any tube which does not permit the passage of the eddy-current inspection probe shall be deemed a defective tube.

beyond which the tube shall be removed from service because it may become unserviceable prior to the next inspection and is equal to 0096 of the nominal tube wall thickness.

7. Unserviceable described the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of OBE, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 0.2.5.3.c, above.

generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.

9. Preservice Ins ection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing.

b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 0.2-3.

0.2-5 Amendment Nos. Sl 2 75

~ ~

I

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rt.2.5.5 ~Re orts

a. Following each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission within 15 days.

b. The complete results of the steam generator tube inservice inspection shall be included in the Annual Changes, Tests and Experiment Reports for the period in which this inspection was completed. This report shall include:

1. Number and extent of tubes inspected.

2. Location and percent of wall-thickness penetration for each indication of an imperfection.

3. Identification of tubes plugged.

c. Results of steam generator tube inspections which fall

'nto Category C-3 and require prompt notification of the Commission shall be reported pursuant to Specification 6.9.2.a prior to resumption of plant operation. The written followup of this report shaf1 provide a description of investigations conducted to determine the cause of the tube degradation and corrective measures taken to prevent recurrence.

0.2-6 Amendment Nos. 81 2 75

i l TABLE 0.2-1 VAS DELETED IN ITS ENTIRETY

0.0.2 LOCAL PENETRATiON TESTS Test Procedure and Fre uenc Local leak detection tests of the following components shall be performed at a pressure not less than 50 psig using pressure decay, soap bubble, halogen detection or equivalent methods at the frequency listed, unless otherwise noted:

1. Containment purge valves (pressure applied in connecting duct) - each refueling.

2. Personnel and Emergency Airlocks

a. +V/ithin 3 days of every first of a series of openings when containment integrity is required, verify that door seals have not been damaged or seated improperly by vacuum testing the volume between the door seals in accordance with approved plant procedures.

b. At least once per 6 months, conduct an overall airlock leakage test to verify that the overall airlock leakage rate is within its limit.

3. Equipment access opening (pressure applied between gaskets)-

annually and after use.

Fuel transfer tube flange (pressure applied between gaskets)-

each ref uleing.

5. Electrical penetrations (pressure applied to canister) - each refueling.

Acce tance Criteria:

Repairs and tests shall be made whenever the sum of the local leak rate tests, including the isolation valves discussed in 0.4.3, exceeds sixty percent of the total containment allowable leak rate.

0.0-2 Amendment Nos. 73 2 67

0Ã.3 ISOLATION VALVES Containment isolation valves shall be tested in accordance with 10 CFR 50, Appendix 3.

No additional surveillance requirements other than those required by Specification 0.0.3.

4.0.4 RESIDUAL HEAT REMOVAL SYSTE'M No additional surveillance requirements other than those required by Specification 0.0.3.

V-3 Amendment Nos. and

0.5 SAFETY IN3ECTION

~

• "'""""'" 5'v

~

~Ob'ective: To verify that the subject systems will respond promptly and perform their design functions.

ecif ication: 1. SYSTEM TESTS

a. System tests shall be performed at each:refueling shutdown. The test shall be performed in accordance with the following procedure:

With the Reactor Coolant System pressure equal to or less than 350 psig and temperature equal to or less than 350 F, a test safety injection signal will be applied to initiate operation of the system.

The breakers for the residual heat removal pump motors will be tested either in the test position or by actual residual heat removal pump motor operation resulting from the test safety injection signal.

b. The test will be considered satisfactory if control panel indication and visual observations indicate that all components have received the safety injection signal in the proper sequence and

'timing, appropriate breakers shall open and close, and all automatic valves shall complete their travel.

2. COMPONENT TESTS

a. ~Pum s No additional surveillance requirements other than those required by Specification 0.0.3.

b. Valves No additional surveillance requirements other than those required by Specification 0.0.3.

0.5-L Amendment Nos. and

2. COMPONENT TESTS

a. ~Pum s For the containment spray pumps, there are no additional surveillance requirements other than those required by Specif ication 0.0.3.

b. Fans Emergency Containment Cooling fans shall be started. at intervals not greater than one (1) month.tt Acceptable levels of performance shall be that the fan motors reach their nominal operating current for the containment atmosphere during the test, and operate for at least fifteen minutes.

c. Valves No additional surveillance requirements other than those required by Specification 0.0.3.

l td - N.A. during cold or refueling shutdowns. The specified tests, however, shall be performed within one surveillance interval prior to heatup above 200 F.

0.6-2 Amendment Nos. and

2. POST ACCIDENT CONTAINMENT VENT SYSTEM ae Operating tests shall be performed during refueling but not longer than 18 months. The tests shall consist of visual inspection of the system and pressure drop and air flow measurements. Visual inspection shall include a search for any foreign materials and gasket deterioration in the HEPA filters and charcoal adsorbers. Less than 6" of water pressure drop at 55 cfm flow shall constitute acceptable performance.

b. Valves No additional surveillance requirements. other than those required by Specification 0.0.3.

2. Performance Tests

a. A visual inspection of the sytem shall be made before each DOP test and halogenated hydrocarbon leak test. At least once per 18 months or after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation, in-place DOP and halogenated hydrocarbon tests at design flow (55 cfm + 1096) and carbon analysis, or carbon replacement, for the Post Accident Containment Vent filters shall be

, performed. In addition, carbon analysis (or carbon replacement), DOP and halogenated hydrocarbon tests at design flow (55 cfm + 1096) shall be performed after (1) any structural maintenance on system housings which might have affected filter bank efficiency, (2) after complete or partial replacement of a filter bank or (3) after exposure of the filters to effluents from painting, fire or chemical release. Removal "of > 9996 DOP and > 99% halogenated hydrocarbon shall constitute acceptable performance.

b. Laboratory carbon sample analysis shall show >90% methyl radio-iodine removal or the charcoal shall be replaced with charcoal that meets or exceeds the criteria of position C.6.a of Regulatory Guide 1.52 (Revision 2). The sample shall be taken in accordance with position C.6.b of Regulatory Guide 1.52. Carbon analysis shall be performed in accordance with ANSI N510-1975. Analysis shall verify the above removal efficiency for radio-iodine within 05 days after removal of the sample.

C. The hydrogen concentration measurement instrument shall be calibrated with proper consideration for humidity.

0.7-3 Amendment Nos. and

3. CONTROL ROOM VENTILATION EMERGENCY INTERNAL CLEANUP) SYSTEM

l. A visual inspection shall be made before each in-place DOP test, halogenated hydrocarbon leak test, and airflow distribution test.

The Control Room Ventilation System shall be'operated monthly for at least 15 minutes to demonstrate operability. Auto initiation of the systems operations shall be checked during refueling, but not longer than 18 months. Pressure drop measurements across the filter bank shall be made annually. Less than 6" of water pressure drop at designed flow (1,000 cfm + 10%)

across the combined HEPA filter and charcoal adsorbers shall constitute acceptable performance. A visual inspection shall include a search for any foreign materials and gasket deterioration in the HEPA filters and charcoal adsorbers.

2. Performance Tests

a. A visual inspection shall be made before each in-place DOP test, halogenated hydrocarbon leak test and airflow distribution test. At least once per 18 months or after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation, in-place DOP and halogenated hydrocarbon tests at design flow (1,000 cfm + 1096) and carbon analysis shall be performed after (1) any structural maintenance on system housings, which might have bank efficiency, (2) after complete or partial affected'ilter replacement of a filter bank, or (3) after operational exposure 0.7-0 Amendment NNos. and

0.9 MAIN STEAM ISOLATION VALVES Applies to periodic testing of the main steam isolation valves.

~Ob'ective: To verify the ability of the main steam isolation valves to close upon signal.

Each main steam isolation valve shall be demonstrated OPERABLE by verifying full closure within 5 seconds when tested pursuant to Specif ication 0.0.3.

0.9-1 Amendment Nos. and

~ I

~ q t 0.10 AUXILIARYFEEDVATER SYSTEM

~Ob'ective:

• >>i 'i '* i 8'ystem.

To verify the operability of the auxiliary feedwater system and its ability to respond properly when required.//+

accordance with the requirements of Specification 0.0.3 and a flow rate of 373 gpm established to the steam generators.

2. Part of the tests acceptance criteria shall be considered satisfactory if control panel indication and visual observation of the equipment demonstrate that all components have operated properly.

3. At least once per 18 months:

(a) Verify that each automatic valve in the flow path actuates to its correct position upon r eceipt of each auxiliary feedwater actuation test signal.

(b) Verify that each auxiliary feedwater pump receives a start signal t

as designed automatically upon receipt of each auxiliary" feedwater actuation test signal.

+ N.A. during cold or refueling shutdowns (only for the unit at cold or refueling shutdown).

The specified tests, however, shall be performed within one surveillance interval prior to starting the turbine.

k/ NOTE: If any local manual realignment of valves is required when operating the Auxiliary Feedwater pumps,"a dedicated individual, who is in communication with the control room, shall'be stationed at the auxiliary feedwater pump area. Upon instructions from the control room, this operator would realign the valves in the AFW system train to its normal operational alignment.

0.10-1 Amendment Nos. and

10.0 BASES FOR SURVEILLANCE RE UIREMENTS 0.0.1 This specification provides a grace period so that intervals may be adjusted to accommodate normal test schedules.

0.0.2 This specification provides guidance for surveillance requirements when the reactor is in a shutdown condition.

0.0.3 This specification ensures that inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves will be performed in accordance with a periodically updated version of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.55a. Relief from any of the above requirements has been provided in writing by the Commission and is not a part of these Technical Speci fications.

This specification includes a clarification of the frequencies for performing the inservice inspection and testing activities required by Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda. This clarification is provided to ensure consistency in surveillance intervals throughout these Technical Specifications and to remove any ambiguities relative to the frequencies for performing the required inservice inspection and testing activities.

l-Under the terms of this specification, the more restrictive requirements of the Technical Specifications take precedence over the ASME Boiler and Pressure Code and applicable Addenda.

B0.0-1 Amendment Nos. and

BASES FOR REACTOR COOLANT SYSTEM IN-SERVICE INSPECTION This specification ensures that inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves will be performed in accordance with a periodically updated version of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.55a. This is accomplished by referencing the inspections required by Specification 0.0.3.

MISCELLANEOUS INSPECTIONS Reactor Coolant Pum Fl wheels The flywheels shall be visually examined per the recommendations of Regulatory Position CA.b of Regulatory Guide 1.10 Revision 1, August 1975.

Materials Irradiation Surveillance S ecimens The reactor vessel surveillance program includes eight specimen capsules to ev'aluate radiation damage based on pre-irradiation and post-irradiation testing of specimens.

The specimens are located about 3" from the vessel wall at the axial midplane.

Capsule No. 1 is scheduled to be removed at the first region replacement. The exposure of this capsule leads the vessel maximum exposure by a factor of 2.1. Thus~

this capsule provides information for approximately a four-year exposure to the vessel.

Capsule No. 2 is scheduled to be removed at the fourth region replacement. This capsule leads the vessel maximum exposure by a factor of O.S and thus will provide data for a four-year exposure to the vessel. This sample also contains weld metal which is not present in Capsule No. l.

Capsule No. 3 leads the vessel maximum exposure by a factor of 2.2 and is scheduled to be removed after twenty years. Thus, sample No. 3 will provide data for an exposure to the vessel of approximately forty years.

Capsules No. 0 and 5 lead the maximum vessel exposure by factors of 0.7 and 0.5, respectively. Thus, Capsule No. 0, which is scheduled to be removed after thirty years, provides data for a vessel exposure of twenty-one years and Capsule No. 5, which is scheduled to be removed at forty years, provides data for a vessel exposure of twenty years.

In addition to the capsules discussed above, there are three spares.

Steam Generator Tube Ins ection The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision l. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. In service inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

B0.2-1 Amendment Nos. and

~ ~

t I

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those parameter limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these parameter limits, localized corrosion may likely result in stress corrosion racking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = I gallon per minute, total). Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of I gallon per minute can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.

Wastage-type defects are unlikely with the all volatile treatment (AVT) of secondary coolant. However, even if a defect of similar type should develop in service, it will be found during scheduled inservice steam generator tube examinations. Plugging will be required of all tubes with imperfections exceeding the plugging limit which, by the definition of Specification 0.2.5.0.a is 0096 of the tube nominal wall thickness.

Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 2096 of the original tube wall thickness.

Whenever the results of any steam generator tubing in-service inspection fall into Category C-3, these results will be promptly reported to the Commission pursuant tb Specification 6.9.2.a prior to resumption of plant operation. Such cases will be 4 considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

B0.2-2 Amendment Nos. and