Text

Amends 194 & 177 to Licenses DPR-70 & DPR-75,respectively, Changing TSs 3.4.3 & 3.4.5, Relief Valves, to Ensure That Automatic Capability of PORVs to Relieve Pressure Is Maintained When Valves Are Isolated
	ML20148K084
Person / Time
Site:	Salem
Issue date:	06/04/1997
From:	Stolz J; NRC (Affiliation Not Assigned)
To:	;
Shared Package
ML20148K062	List: ML20148K055; ML20148K084; ML20148K107; ML20148K151;
References
	NUDOCS 9706170353
	Download: ML20148K084 (15)
	v • d • e

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p 8 UNITED STATES

$ NUCLEAR REGULATORY COMMISSION S " WASHINGTON, D.C. 20666-0001

- W PUBLIC SERVICE ELECTRIC & GAS COMPANY PHILADELPHIA ELECTRIC COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-272 SALEM NUCLEAR GENERATING STATION. UNIT fi0. I affMDtfEtT TO FACILITY OPERATING LICENSE Amendment No.194 License No. DPR-70 h leer Regulatory Commission (the Commission or the NRC) has found OI lication for amendment filed by the Public Service Electric &

E any, Philadelphia Electric Company, Delmarva Power and Light L and Atlantic City Electric Company (the licensees) dated dipury 31, 1997, as supplemented by letters dated March 14, April 8, j 11 28, 1997, complies with the standards and requirements of !

ic Energy Act of 1954, as amended (the Act), and the E ion's rules and regulations set forth in 10 CFR Chapter I; L Ohcility will operate in conformity with the application, the n ions of the Act, and the rules and regulations of the

h. ion; i 'Jk%Eisreasoaableassurance: (1) that the activities authorized by

4% amendment can be conducted without endangering the health and

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sf of the public, and (ii) that such activities will be conducted v liance with the Commission's regulations set forth in 10 CFR h I;

'ei e of this amendment will not be inimical to the common

de and security or to the health and safety of the public; and i

M b{ ra of this amendment is in accordance with 10 CFR Part 51 of f *f

Fhumission's regulations and all applicable requirements have been 2t. m ed.

hi , the Ticense is amended by changes to the Technical Specifica-

. :t as icated in the attachment to this license amendment, and 11 rap .C.(2) of Facility Operating License No. DPR-70 is hereby

Meo ' read as follows

9706170353 970604 PDR ADOCK 05000272 p PDR

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l (2) Technical Soecifications and Environmental Protection Plan l

The Technical Specifications contained in Ap)endices A and B, as l l revised through Amendment No.194 , are here)y incorporated in the ;

license. The licensee shall o the Technical Specifications. perate the facility in accordance with In addition, the license is amended by changes to Appendix C as indicated l in the attachment to this license amendment, and paragraph 2.C.(10) of the Facility Operating License No. DPR-70 is amended to read as follows: i l (10) Additional Conditions The Additional Conditions contained in Appendix C, as revised through Amendment No.194 . are hereby incorporated into this license. Public Service Electric and Gas Company shall operate the facility in accordance with the Additional Conditions.

3. This license amendment is effective as of its date of issuance and the l change to the facility shall be implemented prior to entry into Mode 3

! from the current outage for Salem Unit 1. Implementation of this l amendment shall include upgrading the initiation circuitry for the power i operated relief valves as described in the licensee's a) plication dated

! January 31. 1997, as supplemented by letters dated Marc 1 14, April 8 and A)ril 28, 1997, and evaluated in the staff's safety evaluation attached to t11s amendment.

FOR THE NUCLEAR REGULATORY COMMISSION

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L F. Stolz, Directo I o

ject Directorate I-Division of Reactor Projects - I/II i Office of Nuclear Reactor Regulation Attachments: 1. Page 1 to Ap)endix C of License

DPR-70

2. Changes to t1e Technical Specifications Date of Issuance: June 4, 1997 1

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Page 1 of Apaendix C is attached, for convenience, for the composite license i to reflect t1is change.

._ . _ _ . . _ - _ - . - . _ _ = . _ _ _ . _ _ _

ATTACHMENT TO LICENSE AMENDMENT NO.194 FACILITY OPERATING LICENSE NO. DPR-70 DOCKET NO. 50-272 j 1. Remove Insert Appendix C, page 1 Appendix C, page 1

2. Revise Appendix A as follows:

Remove Paaes Insert Paaes 3/4 4-5 3/4 4-5 B 3/4 4-la B 3/4 4-la I 3

B 3/4 4-lb I

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l j APPENDIX C i ADDITIONAL CONDITIONS

OPERATING LICENSE NO. DPR-70 3

Public Service Electric and Gas Company Philadelphia Electric Company.

Delmarva Power and Light Company, and Atlantic City Electric Company shall comply with the following conditions on the schedules noted below

I Amendment Additional Condition Implementation Number Date 192 The licensee is authorized to relocate certain The amendment

Technical Specification requirements to shall be j licensee-controlled documents. Implementation implemented of this amendment shall include the relocation within 60 days

)

of these technical specification requirements from March 21.

to the appropriate documents, as described in 1997.

the licensee's application dated January 11.

i 1996, as supplemented by letters dated February 26. May 22. June 27. July 12.

December 23, 1996, and March 17. 1997, and evaluated in the staff's safety evaluation attached to this amendment.

194 The licensee is authorized to upgrade the The amendment initiation circuitry for the power operated shall be relief valves, as described in the licensee's implemented application dated January 31,1997, as prior to entry supplemented by letters dated March 14 into Mode 3 A)ril 8. and April 28, 1997, and evaluated in from the tie staff's safety evaluation attached to this current outage ,

amendment. for Salem. '

Unit 1.

i 1 Amendment No. 493 194

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1 REACTOR COOLANT SYSTEM i 3/4.4.3 RELIEF VALVES i

i LIMITING CONDITION FOR OPERATION

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! 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 because of excessive seat leakage, 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 inoper0. ole due to causes other than excessive seat leakage, 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 due to causes other than excessive seat leakage, 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 ,

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.

aALEM - UNIT 1 3/4 4-5 Amendment No.194

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l REACTOR COOLANT SYSTEM BASES

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3/4.4.2 SAFETY VALVES The prersurizer 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 '

i 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 I 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 umximum 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 trjp on the loss of load) and also assuming no operation of the power operated reilrf 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 1

The OPERABILITY of the PORVs and block valves is determined on the basis of i their being capable of performing the following functions A. Manual control of PORVs to control reactor coolant system pressure.

This is a function that is used for the steam generator tube rupture j accident and for plant shutdown.

B. Automatic control of PORVs to control reactor coolant system pressure.

This is a function that reduces challenges to the code safety valves for overpressurization events, including an inadvertent actuation of the Safety Injection System.

C. Maintaining the integrity of the reactor coolant pressure boundary. !

This is a function that is related to controlling identified leakage and ensuring the ability to detect unidentified reactor coolant pressure boundary leakage.

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SALEM - UNIT 1 B 3/4 4-la Amendment No.190

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l REACTOR COOLANT SYSTEM BASES .

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3/4.4.3 RELIEF VALVES (continued) ;

D. Manual control of the block valve to : (1) unblock an isolated PORV to allow it to be used for manual and automatic control of Reactor Coolant i i System pressure (Items A & B), and (2) isolate a PORV with excessive !

seat leakage (Item C).

E. Manual control of a block valve to isolate a stuck-open PORV.

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! SALEM - UNIT 1 B 3/4 4-lb Amendment No.194 3

e atto p t UNITED STATES s* y NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20665 0001

\...../

PUBLIC SERVICE ELECTRIC & GAS COMPANY PHILADELPHIA ELECTRIC COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-311 SALEM NUCLEAR GENERATING STATION. UNIT N0. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.177 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 January 31, 1997, as supplemented by letters dated March 14, April 8, and April 28, 1997, 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 i 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:

(2) Technical Soecifications and Environmental Protection Plan The Technical Specificat' ions contained in Ap)endices A and B. as revised through Amendment No.177 . are here)y incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

In addition. the license is amended by changes to Appendix C as indicated in the attachment to this license amendment, and paragraph 2.C.(26) to Facility Operating License No. DPR-75 is amended to read as follows:

(26) Additional Conditions The Additional Conditions contained in Appendix C as revised through Amendment No.177 . are hereby incorporated into this license. Public Service Electric and Gas Company shall operate the facility in accordance with the Additional Conditions.

3. This amendment is effective as of its date of issuance and the change to the facility shall be implemented prior to entry into Mode 3 from the l current outage for Salem Unit 2. Implementation of this amendment shall l include upgrading the initiation circuitry for the power operated relief valves as described in the licensee's application dated January 31, 1997.

as supplemented by letters dated March 14. April 8. and April 28. 1997. ,

and evaluated in the staff's safety evaluation attached to this amendment. j FOR THE NUCLEAR REGULATORY COMMISSION Joh S

F. Stolz. Dir. tor Pr ject Directora I-2 vision of Reactor Projects - I/II ;

Office of Nuclear Reactor Regulation !

Attachments: 1. Page 1 to Apaendix C of License

DPR-75

2. Changes to tie Technical Specifications Date of Issuance: June 4, 1997 l

Page 1 of Ap)endix C is attached, for convenience, for the composite license to reflect tais change.

i 1

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i ATTACHMENT TO LICENSE AMENDMENT N0.177 1 I

i FACILITY OPERATING LICENSE NO. DPR-75 l DOCKET NO. 50-311 l

1. Remove Insert 1

Appendix C,~page 1 Appendix C, page 1 -

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2. Revise Appendix A as follows: l I

Remove Paaes Insert Paaes i t

3/4 4-8 3/4 4-8 B 3/4 4-2 B 3/4 4-2 8 3/4 4-3 B 3/4 4-3 B 3/4 4-3a l

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APPENDIX C j ADDITIONAL CONDITIONS OPERATING LICENSE NO. DPR-75 4

Public Service Electric and Gas Company, Philadelphia Electric Company,

Delmarva Power and Light Company, and Atlantic City Electric Company shall

. comply with the following conditions on the schedules noted below:

Amendment Additional Condition Implementation

, Number -Date

, 175 The licensee is authorized to relocate certain The amendment

! Technical Specification requirements to ' shall be licensee-controlled documents. Implementation implemented of this amendment shall include the relocation within 60 days of these technical specification requirements from March 21, ;

to the appropriate documents. as described in 1997. l the licensee s application dated January 11, !

i 1996. as supplemented by letters dated ;

] February 26. May 22. June 27 July 12, '

December 23, 1996, and March 17, 1997, and j evaluated in the staff's safety evaluation !

4, attached to this amendment. 1 177 The licensee is authorized to upgrade the The amendment '

! initiation circuitry for the power operated shall be i relief valves. as described in the licensee's implemented i application dated January 31. 1997, as prior to entry l supplemented by letters dated March 14, into Mode 3

A3ril 8. and April 28, 1997, and evaluated in from the

t7e staff's safety evaluation attached to this current outage i amendment. for Salem, j Unit 2.

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REACTOR COOLANT SYSTEM 3/4.4.5 RELIEF VALVES l

LIMITING CONDITION FOR OPERATION I

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l 3.4.5 Two power relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

l ACTION:

a. With one or both PORVs inoperable because of excessive seat leakage, 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 . j I

b. With one PORV inoperable due to causes other than excessive seat leakage, 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 due to causes other than excessive seat ,

leakage, 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 va] e to meet the Limiting Condition for Operation. l

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 l control; restore the block valve to operable status within 72 !

hours 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 SHUTDCWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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

, 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 l

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 i

. for Operation. '

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l SALEM - UNIT 2 3/4 4-8 Amendment No. 177

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REACTOR COOLANT SYSTEM BASES

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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 which could occur du:ing shutdown. In the event that no safety valves are OPERABLE, an operrting RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization. 3n addition, the Overpressure Protection System provides a diverse means of protecticn 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 setpcint 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 nmximum water volume in the pressurizer assures that the parameter is maintained within the normal steady-state envelope of operation assumed in the SAR. The limit 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 OPERABILITY of the PORVs and block valves is determined on the Lasis of their being capable of performing the following functions:

A. Manual control of PORVs to control reactor coolant system pressure.

This is a function that is used for the steam generator tube rupture accident and for plant shutdown.

SALEM - UNIT 2 B 3/4 4-2 Amendment No.177 l

REACTOR COOLANT SYSTEM l BASES a l

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I 3/4.4.5 RELIEF VALVES (continued)

I B. Automatic control of PORVs to control reactor coolant system pressure. !

This is a function that reduces challenges to the code safety valves for l overpressurization events, including an inadvertent actuation of the l

Safety Injection System. l l

C. Maintaining the integrity of the reactor coolant pressure boundary. l This is a function that is related to controlling identified leakage and j ensuring the ability to detect unidentified reactor coolant pressure boundary leakage. l

.D. Manual control of the block valve to : (1) unblock an isolated PORV to 1 allow it to be used for nanual and automatic control of Reactor Coolant System pressure (Items A & B), and (2) isolate a PORV with excessive seat leakage (Item C).

E. Manual control of a block valve to isolate a stuck-open PORV.

3/4.4.6 STEAM GENERATORS i

The Surveillance Requirements for inspection of the steam generator tubes l 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 lindts 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 1r 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 (prinary-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 loads )

imposed during normal operation and by postulated accidents. Operating plants l have demonstrated that primary-to-secondary leakage of 500 gallons per day per , l 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.

i SALEM - UNIT 2 B 3/4 4-3 Amendment No.177 )i

REACTOR COOLANT SYSTEM BASES

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l 3/4.4.6 STEAM GENERATORS (continued)

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.

I SALEM - UNIT 2 B 3/4 4-3a Amendment No. 177

ML20148K084

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