Text

Amends 150 & 130 to Licenses DPR-70 & DPR-75,respectively, Incorporating Guidance of NRC GL 90-06 That Addresses power-operated Relief Valve & Block Valve Reliabililty & Addl low-temperature Overpressure Protection for LWR
	ML20065J641
Person / Time
Site:	Salem
Issue date:	04/07/1994
From:	Chris Miller; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20065J643	List: ML20065J641; ML20065J668;
References
	GL-90-06, GL-90-6, NUDOCS 9404180346
	Download: ML20065J641 (20)
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NUCLEAR REGULATORY COMMISSION

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PUBLIC SERVICE ELECTRIC & GAS COMPANY PHILADELPHIA ELECTRIC COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-272 t

SALEM NVCLEAR GENERATING STATION. UNIT NO. 1 AMENDMc.4T TO FACILITY OPERATING LICENSE Amendment No. 150 License No. DPR-70 1.

The Nuclear Regulatory Commission (the Commission or the NRC) has found that:

A.

The application for amendment filed by the Public Service Electric &

Gas Company, Philadelphia Electric Company, Delmarva Power and Light Company and Atlantic City Electric Company (the licensees) daxed December 8, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commr - i ' s rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-70 is hereby amended to read as follows:

9404180346 940407 PDR ADOCK 05000272 P

PDR

, (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendices A and B,'as revised through Amendment No. 150, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION cc 1

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Charles L. Miller, Director Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

April 7, 1994

ATTACHMENT TO LICENSE AMENDMENT NO. 150 EACILITY OPERATING LICENSE NO. DPR-70 DOCKET NO. 50-272 Revise Appendix A as follows:

Remove Paaes Insert Paaes 3/4 4-5 3/4 4-5 3/4 4-Sa 3/4 4-30 3/4 4-30 3/4 4-31 3/4 4-31 3/4 5-6 3/4 5-6 3/4 5-6a B 3/4 4-la B 3/4 4-la

REACTOR C001 ANT SYSTEM 3/4.4.3 RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.3 Two power relief valves (PORVs) and their' associated block valves shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a.

With one or both PORVs inoperable and capable of being manually cycled, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV(s) to OPERABLE status or close the associated block valve (s) with power maintained to the block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With one PORV inoperable and not capable of being manually cycled, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV to OPERABLE status or close its associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

With both PORVs inoperable and not capable of.being manually cycled, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either restore at least one PORV to OPERABLE status or close the associated block valves and remove power from the block valves and be in HOT STANDBY within the'next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Restore the remaining PORV to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from failure of that valve to meet the Limiting Condition for Operation.

d.

With one block valve inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore the block valve to OPERABLE status or place the associated PORV in manual' control; restore the block valve to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With both block valves inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore the block e.

valves to OPERABLE status or place the associated PORVs in manual control; restore at least one block valve to OPERABLE-status within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Restore the remaining block valve to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from failure of that valve to meet the Limiting Condition for Operation.

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SALEM - UNIT 1 3/4 4-5 Amendment No. 150

REACTOR COOLANT SYSTEM

- 3/4.4. 3 RELIEF VALVES SURVEILIANCE REQUIREMENTS 4.4.3.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE at least once per 18 months by:

Operating the PORV through one complete cycle of full travel during a.

MODES 3 or 4, and b.

Operating solenoid valves, air control valves, and check valves on associated air accumulators in PORV control systems through one complete cycle of full travel, and c.

Performing 4 CHANNEL CALIBRATION of the actuation instrumentation.

4.4.3.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed in order to meet the requirements of ACTION b, or-c in Specification 3.4.3.

5 SALEM - UNIT 1 3/4 4-Sa Amendment No. 150

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REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following overpressure protection systems shall be OPERABLE:

a.

Two Pressurizer Overpressure Protection System relief valves (POPS) with a lift setting of less than or equal to 375 psig, or i

b.

A reactor coolant system vent of greater than or equal to 3.14 square inches.

APPLICABILITY: When the temperature of one or more of the RCS cold legs is l

less than or equal to 312*F, except when the reactor vessel head is removed.

ACTION:

With one POPS inoperable in MODE 4 and the temperature of one or a.

more of the RCS cold legs is less than or equal to 312*F, either restore the inoperable POPS to OPERABLE status within 7 days or depressurize and vent the RCS through a 3.14 square inch vent (s) within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; maintain the RCS in a vented condition until both POPSs have been restored to OPERABLE status, b.

With one POPS inoperable in MODES 5 or 6 with the Reactor Vessel Head installed, restore the inoperable POPS to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or complete depressurization and venting of the RCS through at least a 3.14 square inch vent (s) within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; maintain the RCS in a vented condition until both POPSs have been restored to OPERABLE status.

c.

With both POPSs inoperable, depressurize and vent the RCS through I

a 3.14 square inch vent (s) within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; maintain the RCS in a vented condition until both POPSs have been restored to OPERABLE

status, d.

In the event either the POPS or the RCS vent (s) are used to I

mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the POPS or vent (s) on the transient and any corrective action necessary to prevent recurrence.

SALEM - UNIT 1 3/4 4-30 Amendment No. 150

l REACTOR COOI).NT SYSTEM OVERPRESSURE PROTECTION SYSTEMS SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each POPS shall be demonstrated OPERABLE by:

a.

Performance of a CHANNEL FUNCTIONAL TEST on the POPS' actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the POPS is required OPERABLE.

b.

Performance of a CHANNEL CALIBRATION on the POPS actuation channel at least once per 18 months.

c.

Verifying the POPS isolation valve is open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when the POPS is being used for overpressure protection.

d.

Testing pursuant to Specification 4.0.5.

4.4.9.3.2 The RCS vent (s) shall be verified to be open at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

when the vents (s) is being used for overpressure protection.

Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.

SALEM - UNIT 1 3/4 4-31 Amendment No.

150 l

9 EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - Tave <350*F LIMITING CONDITION'FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be l

OPERABLE:

One OPERABLE centrifugal charging pump # and associated flow path

l a.

capable of taking suction from the refueling water storage tank and transferring suction to the residual heat removal pump discharge piping and;

1. Discharging into each Reactor Coolant System (RCS) cold leg.

b.

One OPERABLE residual heat removal pump and associated residual heat removal heat exchanger and flow path capable of taking suction from the refueling water storage tank on a safety injection signal and transferring suction to the containment sump during the recirculation phase of operation and; 1.

Discharging into each RCS cold leg, and; upon manual initiation, 2.

Discharging into two RCS hot legs.

APPLICABILITY: MODE 4 ACTION:

With no ECCS subsystem OPERABLE because of the inoperability of a.

either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months , b.

Within no ECCS subsystem OPERABLE because of the inoperability of-either the residual heat removal heat exchanger or residual heat removal pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reactor Coolant System T less than 350*F by use of alternate heat removal methods.

""8 c.

In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

A maximum of one safety injection pump or one centrifugal charging pump shall be OPERABLE in MODE 4 when the temperature of one or more of the RCS cold legs is less than or equal to 312*F, Mode 5, or Mode 6 when the head is on the reactor vessel.

SALEM - UNIT 1 3/4 5-6 Amendment No. 150 l,

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EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - Tave < 350*F SURVE111ANCE REQUIREMENTS-i 4.-5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per applicable Surveillance Requirements of 4.5.2.

4.5.3.2-All safety injection pumps and centrifugal charging pumps,' except the above required OPERABLE pump, shall be demonstrated to be inonerable at-least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months while in MODE 4 and the temperature of one or more of the RCS cold legs is less than or equal to 312*F, MODE 5, Lor MODE 6 when the head is on the reactor vessel by either of the.following methods:

By verifying that the motor circuit breakers have been removed from

- r a.

their electrical power supply circuits or, l

b.

For testing purposes, by verifying that the pump is in a recirculation flow path and that the manual discharge valve is'

closed, i

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SALEM - UNIT 1 3/4 5-6a Amendment'No. 150 f

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REACTOR C001 ANT SYSTEM BASES 3/4.4.2 -SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed-to relieve 420,000 pounds per hour of saturated steam at the valve setpoint.

The relief capacity of a single safety valve is adequate to relieve any-overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization. In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperature.

While in Mode 5 the safety valve requirement may be met by establishing a vent path of equivalent relieving capacity when no code safety valves are OPERABLE.

During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig.

The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip setpoint is reached (i.e.,

no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump valves.

Demonstration of the safety valves lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.

3/4.4.3 RELIEF VALVES The power operated relief valves (PORV) and steam bubble function to

'l relieve RCS pressure during all design transients up to and including the design step load decrease with steam ducp. Operation of the power operated relief valves minimizes the undesirable opening of the spring-loaded pressurizer code safety valves. Each power operated relief valve has a remotely operated block valve to provide positive shutoff capability should a relief valve become inoperable. Although a PORV may be inoperable, it may be able to be manually opened or closed, and therefore, able to perform its intended design function.

PORV inoperability may be due to excessive seat leakage, instrumentation problems, automatic control problems, or other causes that do not prevent manual use, and do not create a possibility for a small break LOCA.

The Action Statement requires that the block valve be closed, and power maintained to the valve in those circumstances when the PORV is still capable of being manually cycled.

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SALEM - UNIT 1 B 3/4 4-la Amendment No. 150

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UNITED STATES i')T*/'g 3

NUCLEAR REGULATORY COMMISSION gi 7 1!

WASHINGTON, D.C. 20555-0001

%. ~... f PUBLIC SERVICE ELECTRIC & GAS COMPANY PHILADELPHIA ELECTRIC COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY l

DOCKET N0. 50-311 SALEM NUCLEAR GENERATING STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 130 License No. DPR-75 1.

The Nuclear Regulatory Commission (the Commission or the NRC) has found that:

A.

.The application for amendment filed by the Public Service Electric &

Gas Company, Philadelphia Electric Company, Delmarva Power and Light Company and Atlantic City Electric Company (the licensees) dated December 8, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have_ been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-75 is hereby amended to read as follows:

j (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendices A and B, as revised through Amendment No.

130, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION l l.

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(up Charles L. Miller, Director Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation.

Attachment:

Changes to the Technical Specifications Date of Issuance:

April 7, 1994 I

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' ATTACHMENT TO LICENSE AMENDMENT NO. 130 FACILITY OPERATING LICENSE NO. DPR-75 DOCKET NO. 50-311 Revise Appendix A as follows:

Remove Paaes Insert Paaes 3/4 4-8 3/4 4-8 3/4 4-8a 3/4 4-31 3/4 4-31 3/4 5-7 3/4 5-7 3/4 5-8 3/4 5-8 8 3/4 4-2 8 3/4 4-2 8 3/4 4-3 8 3/4 4-3 s

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REACTOR COOLANT SYSTEM 3/4.4.5 RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.5 Two power relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With one or both PORVs inoperable and capable of being manually a.

cycled, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV(s) to OPERABLE status or close the associated block valve (s) with power maintained to the block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , b.

With one PORV inoperable and not capable of being manually cycled, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV to OPERABLE status or close its associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With both PORVs inoperable and not capable of being manually cycled, c.

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either restore at least one PORV to OPERABLE status or close the associated block valves and remove power from the block valves and be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Restore the remaining PORV to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from failure of that valve to meet the Limiting Condition for Operation, d.

With one block valve inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore the block valve to OPERABLE status or place the associated PORV in manual control; restore the block valve to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6. hours and in HOT SHUTDOWN within the following 6' hours.

e.

With both block valves inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore the block valves to OPERABLE status or place the associated PORVs in manual control; restore at least one block valve to OPERABLE status within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Restore the remaining block valve to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from failure of that valve to meet the Limiting Condition for Operation.

SALEM - UNIT 2 3/4 4-8 Amendment No. 130

- REACTOR C001 ANT SYSTEM 3/4.4.5 RELIEF VALVES SURVEILIANCE REQUIREMENTS l

4.4.5.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE at least once per 18 months by:

Operating the PORV through one complete cycle of full travel during a.

MODES 3 or 4, and b.

Operating solenoid valves, air control valves, and check valves on associated air accumulators in PORV control systems through one complete cycle of full travel, and c.

Performing a CHANNEL CALIBRATION of the actuation instrumentation.

4.4.5.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed in order to meet the requirements of ACTION b, or e in Specification 3.4.5.

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l SALEM - UNIT 2 3/4 4-8a Amendment No. 130

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8 REACTOR C001 ANT SYSTEM DVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 1

I 3.4.10.3 At least one of the following overpressure protection systems shall be OPERABLE:

a.

Two Pressurizer Overpressure Protection System relief valves (POPS) with a lift setting of less than or equal to 375 psig, or b.

The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 3.14 square inches.

APPLICABILITY: When the temperature of one or more of the RCS cold legs is less than or equal to 312'F, except when the reactor vessel head is removed.

ACTION:

a.

With one POPS inoperable in MODE 4 and the temperature of one or more of the RCS cold legs is less than or equal to 312*F, restore the inoperable POPS to OPERABLE status within 7 days or depressurize and vent the RCS through a 3.14 square inch vent (s) within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; maintain the RCS in a vented condition until both POPSs l

have been restored to OPERABLE status.

I b.

With one POPS inoperable in MODES 5 or 6 with the Reactor Vessel Head installed, restore the inoperable POPS to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or complete depressurization and venting of the RCS through at least a 3.14 square inch vent (s) within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; maintain the RCS in a vented condition until both POPSs have been restored to OPERABLE status, c.

With both POPSs inoperable, depressurize and vont the RCS through a 3.14 square inch vent (s) within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; maintain the RCS in a vented condition until both POPSs have been restored to OPERABLE

status, d.

In the event either the POPS or the RCS vent (s) are used to l

mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days.

The report shall describe the circumstances-initiating the transient, the effect of the POPS or vent (s) on the transient and any corrective ' action necessary to prevent recurrence.

SURVEILIANCE REQUIREMENTS 4.4.10.3.1 Each POPS shall be demonstrated OPERABLE by:

SALEM - UNIT 2 3/4 4-31 Amendment No. 130

9 EMERGENCY CORE C00LINC SYSTEMS ECCS SUBSYSTEMS - Tave <350*F LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be l

OPERABLE:

One OPERABLE centrifugal charging pump # and associated flow path l

a.

capable of taking suction from the refueling water storage tank and transferring suction to the residual heat removal pump discharge piping and;

1. Discharging into each Reactor Coolant System (RCS) cold leg.

b.

One OPERABLE residual heat removal pump and associated residual heat removal heat exchanger and flow path capable of taking suction from the refueling water storage tank on a safety injection signal and transferring suction to the containment sump during the recirculation phase of operation and; 1.

Discharging into each RCS cold leg, and; upon manual initiation, 2.

Discharging into two RCS hot legs.

APPLICABILITY: MODE 4.

ACTION:

With no ECCS subsystem OPERABLE because of the inoperability of a.

either the centrifugal chargin5 pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months , b.

Within no ECCS subsystem OPERABLE because of the inoperability of either the residual heat removal heat exchanger or residual heat removal pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reactor Coolant System T less than 350*F by use of alternate heat removal methods.

""3 c.

In the event the ECCS. is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to dr-The current value of the usage factor for each affected safety oje tion nozzle shall be provided in this Special Report whenever ica alue exceeds 0.70.

A maximum of one safety injection pump or une centrifugal charging pump shall be OPERABLE in MODE 4 when the temperature of one or more of the RCS cold legs is less than or equal to 312*F, Mode 5, or Mode 6 when the head is on the reactor vessel.

SALEM - UNIT 2 3/4 5-7 Amendment No.130 l

EMERCENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - Tave < 350'F SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per applicable Surveillance Requirements of 4. 5.2.

4.5.3.2 All safety injection pumps and centrifugal charging pumps, except the above required OPERABLE pump, shall be demonstrated to be inocerable at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months while in MODE 4 and the temperature of one or more of the RCS cold legs is less than or equal to 312*F, MODE 5, or MODE 6 when the head is on the reactor vessel by either of the following methods:

a.

By verifying that the motor circuit breakers have been removed from their electrical power supply circuits or, b.

For testing purposes, by verifying that the pump is in a recirculation flow path and that the manual discharge valve is closed.

SALEM - UNIT 2 3/4 5-8 Amendment No.130

REACTOR COOLANT SYSIE5

?

BASES 3/4.4.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.

Each safety valve is designed to relieve 420,000 pounds per hour of saturated steam at the valve setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition whi:h could occur during shutdown.

In the event that no safety valves are OPEPABLE, an operating RHR loop, connected to the RCS, provides overpressure rei.ief capability and will prevent RCS ove rpressurization.

In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperature. While in Mode 5 the safety valve requirement may be met by establishing a vent path of equivalent relieving capacity when no code safety valves are OPERABLI.

During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig.

The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip setpoint is reached (i.e.,

no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump valves.

Demonstration of the safety valves lift settings will occur only during shutdown and will be performed in accordance with.the provisions of Section XI of the ASME Boiler and Pressure Code.

3/4.4.4 PRESSURIZER The limit on the maximum water volume in the pressurizer assures that the parameter is maintained within the normal steady-state envelope of operation assumed in the SAR.

The l!mit is consistent with the initial SAR assumptions.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation. The maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control RCS pressure and establish natural circulation.

3/4.4.5 RELIEF VALVES The power operated relief valves (PORV) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step. load decrease with steam dump. Operation of the power operated-relief valves minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.

Each power operated relief. valve has a remotely operated block valve to provide positive shutoff capability should a relief valve become inoperable. Although a PORV may be inoperable, it may be able to be manually opened or closed, and therefore, able to perform its SALEK - UNIT 2 B 3/4 4-2 Amendment No. 130 l

REACTOR COOLANT SYSTEM BASES 3/4.4.5 RELIEF VALVES (continued) intended design function.

PORV inoperability may be due to excassive seat leakage, instrumentation problems, automatic control problems, or other causes that do not prevent manual use, and do not create a possibility for a small break LOCA.

The Action Statement requires that the block valve be closed, and power maintained to the valve in those circumstances when the PORV is still capable of being manually cycled.

3/4.4.6 STEAM GENERATORS 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 1.

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.

Inservice 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.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.

If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. 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 - 500 gallons per day per steam generator).

Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the leads imposed during normal operation and by postulated accidents.

Operating plants have demonstrated that primary-to secondary leakage of 500 gallons per day per steam generator 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 proper chemistry treatment of the.

secondary coolant. However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations.

Plugging will be required for all tubes with imperfections exceeding the plugging limit of 40% of the tube nominal wall thickness.

Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.

SALEM - UNIT 2 B 3/4 4-3 Amendment No.130

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