ML20137E289

Amends 125 & 119 to Licenses NPF-2 & NPF-8,respectively, Revising TS to Incorporate Latest Revised Topical Repts Governing Installation of Laser Welded SG Tube Sleeves

ML20137E289
Person / Time
Site:	Farley
Issue date:	03/24/1997
From:	Berkow H NRC (Affiliation Not Assigned)
To:
Shared Package
ML20137E281	List: ML20137E276 ML20137E289 ML20137E295
References
NUDOCS 9703270145
Download: ML20137E289 (26)

Text

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1 UNITED STATES i s* j 2

NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30006 4001 i

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*%*****/  ;

! SOUTHERN NUCLEAR OPERATING COMPANY. INC.  :

1 ALABAMA POWER COMPANY )

DOCKET NO. 50-348 ,

i

JOSEPH M. FARLEY NUCLEAR PLANT. UNIT 1

[

l AMENDMENT TO FACILITY OPERATING LICENSE  :

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{ Amendment No.125

License No. NPF-2 i  !

1. The Nuclear Regulatory Commission (the Commission) has found that: 1

! A. The application for amendment by Southern Nuclear Operating I j Company, Inc. (Southern Nuclear), dated January 10, 1997, as i i

supplemented by letter dated February 24, 1997, complies with the 1

standards and requirements of the Atomic Energy Act of 1954, as l l amended (the Act), and the Commission's rules and regulations set l forth in 10 CFR Chapter I; 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; i

?

C. There is reasonable assurance (1) that the activities authorized i 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;

(- D. The issuance of this license amendment will not be inimical to the F common defense and security or to the health and safety of the j- public; and -

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

j- .2. Accordingly, the license is amended by changes to the Technical

Specifications, as indicated in the attachment to this license 4

amendment; and paragraph 2.C.(2) of Facility Operating License No.

I NPF-2 is hereby amended to read as follows:

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) 9703270145 970324 DR ADOCK 050003 8 1

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(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as j revised through Amendment No. 125 , are hereby incorporated in the 4

license. Southern Nuclear 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 30 days of issuance.

I FOR THE NUCLEAR REGULATORY COMMISSION l kN +)< [kw hr i Herbert N. Berkow, Director

., Project Directorate Il-2 Division of Reactor Projects - I/II l Office of Nuclear Reactor Regulation l

Attachment:

i Changes to the Technical

Specifications Date of Issuance: March 24, 1997 i

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ATTACHMENT TO LICENSE AMENDMENT NO.125 TO FACILITY OPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.

Remove Paaes Insert Paaes 3/4 4-9 3/4 4-9 3/4 4-10 3/4 4-10 3/4 4-11 3/4 4-11 3/4 4-12 3/4 4-12 3/4 4-12a 3/4 4-12a 3/4 4-13 3/4 4-13 3/4 4-15a B 3/4 4-3a B 3/4 4-3a l

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REACTOR COOLANT SYSTEM 3/4.4.6 STEAM GENERATORS LIMITING CONDITION FOR OPERATION 3.4.6 Each steam generator shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

j ACTION:

With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing Tavg above 200*F.

SURVEILLANCE REQUIREMENTS , 4.4.6.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.

4.4.6.1 Steam Generator Sample Selection and Inspection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the udnimum number of steam generators specified in Table 4.4-1.

4.4.6.2 Steam Generator Tube # Sample Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Tables 4.4-2 and 4.4-3. The inservice inspection of steam generator tubes shall be I

performed at the frequencies specified in Specification 4.4.6.3 and the inspected tubes shall be verified acceptable per the acceptance cr'iteria of Specification 4.4.6.4. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam

, generators. When applying the exceptions of 4.4.6.2.a through 4.4.6.2.c, previous defects or imperfections in the area repaired by sleeving are not considered an area requiring reinspection. 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 50% of the tubes inspected shall be from these critical areas.

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

1. When referring to a steam generator tube, the sleeve shall be considered a part of the tube if the tube has been repaired per Specification 4.4.6.4.a.9.

FARLEY-UNIT 1 3/4 4-9 AMENDMENT NO.. 125

0 e REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

1. All nonplugged tubes that previously had detectable wall penetrations greater than 20%.

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

3. A tube inspection (pursuant to Specification 4.4.6.4.a.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 or sleeve inspection, this shall .

be recorded and an adjacent tube shall be selected and subjected to a tube inspection.

4. Indications lef t in service as a result of application of l the tube support plate voltage-based repair criteria shall be inspected by bobbin coil probe during all future refueling outages.

c. The tubes selected as the second and third samples (if required by Tables 4.4-2 and 4.4-3) during each inservice inspection may l be subjected to a partial tube inspection provided:

1. The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found.

2. The inspections include those portions of the tubes where imperfections were previously found.

d. Implementation of the steam generator tube / tube support plate repair criteria requires a 100 percent bobbin coil inspection for hot-leg and cold-leg tube support plate intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODscc) indications. The determination of the lowest cold leg tube .

support plate intersections having ODSCC indications shall be based on the performance of at least a 20 percent random sampling of tubes inspected over their full length.

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

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i FARLEY-UNIT 1 3/4 4-10 AMENDMENT NO. 125 l

i REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

Category ,

Inspection 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 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.

C-3 More 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 or sleeves must exhibit 1 signifi cant (greater than 10%) further wall penetrations to be included in the above percentage calculations.

4.4.6.3 Inspection Frequencies - The above required inservice .  ;

inspections of steam generator tubes shall be performed at the following l frequencies:

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a. The first inservice inspection shall be performed after 6 '

Effective Full Power Months but within 24 calendar months of initial criticality. Subsequent inservice inspections shall be l performed at intervals of not less than 12 nor more than 24 i 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 1 occurred, the inspection interval may be extended to a maximum of once per 40 months.

b. If the results of the inservice inspection of an steam generator conducted in accordance with Tables 4.4-2 and 4.4-3 l at 40 month intervals fall in Category C-3, the inspection fr ; sney shall be increased to at least once per 20 months.

The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.6.3.a; the interval may then be extended to a maxihum of  ;

once per 40 months.

c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Tables 4.4-2 and 4.4-3 during l the shutdown subsequent to any of the following conditions:

1. Primary-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.7.2.

2. A seismic occurrence greater than the operating Basis Earthquake.

3. A loss-of-coolant accident requiring actuation of the engineered safeguards.

4. A nain steam line or feedwater line break.

FARLEY-UNIT 1 3/4 4-11 AMENDMENT NO. 125 9

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. . 1 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) l 4.4.6.4 Acceptance Criteria

a. As used in this Specification:

1. Imperfection means an exception to the dimensions, finish or contour of a tube or sleeve from that required by j fabricat.sn drawings or specifications. Eddy-current testing indications below 20% of the nominal wall thickness, if detectable, may be considered as imperfections.

2. D3 gradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube or sleeve.

3. Degraded Tube means a tube, including the sleeve if the tube has been repaired, that contains imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation.

4.  % Degradation means the percentage of the tube or sleeve wall thickness affected or removed by degradation.

5. Defect means an imperfection of such severity that it exceeds the plugging or repair limit. A tube or sleeve containing a defect is defective.

6. Plugging or Repair Limit means the imperfection depth at or beyond which the tube shall be repaired (i.e.,

sleeved) or removed from service by plugging and is greater than or equal to 40% of the nominal tube wall thickness. For a tube that has been sleeved with a mechanical joint sleeve, through wall penetration of greater than or equal to 31% of sleeve nominal wall thickness in the sleeve requires the tube to be removed from service by plugging. For a tube that has been sleeved with a welded joint sleeve, through wall penetration greater than or equal to 24% of sleeve l nominal wall thickness in the sleeve between the weld joints requires the tube to be removed from service by plugging. This definition does not apply to tube support l plate intersections for which the voltage-based repair criteria are being applied. Refer to 4.4.6.4.a.11 for ,

the repair limit applicable to these intersections.  !

7. Unserviceable describes the condition of a tube or sleeve if it leaks or contains a defect large enough to affect i its structural integrity in the event of an operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.6.3.c, above.

EARLEY-UNIT 1 3/4 4-12 AMENDMENT No.125

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REACTOR COOLANT SYSTEM i

SURVEILLANCE REQUIREMENTS (Continued)

8. Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg. For a tube that has been repaired by sleeving, ]

the tube inspection should include the sleeved portion of I the tube.

9. Tube Repair refers to mechanical sleeving, as described  ;

by Westinghouse report WCAP-lll?8, Rev. 1, or laser  !

welded sleeving, as described by Westinghouse reports WCAP-13088, Revision 4, and WCAP-14740 dated January 1997, which is used to maintain a tube in service or return a tube to service. This includes the removal of plugs that were installed as a corrective or preventive measure.

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10. Preservice Inspection 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. This inspection shall be performed after the field hydrostatic test and prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent I inservice inspections. '

11. Tube Support Plate Repair Limit is used for the disposition of an alloy 600 steam generator tube for continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the repair limit is based on maintaining steam generator tube serviceability as described below:

a. Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within Ae bounds of the tube support plate with bobba voltage less than or equal to the lower voltage repair limit [2.0 volts), will be allowed to remain in service.

b. Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair limit [2.0 volts), will be repaired or plugged except as noted in 4.4.6.4.a.ll.c below.

EARLEY-UNIT 1 3/4 4-12a AMENDMENT NO.125

REACTOR COOLANT SYSTD4 SURVEILLANCE REQUIREMENTS (Continued)

b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair of all .

tubes exceeding the plugging or repair limit) required by Tables 4.4-2 and 4.4-3. l 4.4.6.5 Reports

a. Following each inservice inspection of steam generator tubes, the number of tubes plugged or repaired in each steam generator shall be reported to the Commission within 15 days of the completion of the plugging or repair effort,

b. The complete results of the steam generator tube and sleeve inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This Special Report shall include:

1. Number and extent of tubes and sleeves inspected.

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

3. Identification of tubes plugged or repaired.

c. Results of steam generator tube inspections which fall into Category C-3 shall be considered a REPORTABLE EVENT and shall be reported pursuant to 10CFR50.73 prior to resumption of plant operation. The written report shall provide a description of -

investigations conducted to determine the cause of the tube degradation and corrective measures taken to prevent recurrence.

d. For implementation of the voltage-based repair criteria to tube support plate intersections, notify the NRC staff prior to returning the steam generators to service (Mode 4) should any of the following conditions arise:

1. If estimated leakage based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds the leak limit (determined from the licensing basis dose calculation for the postulated main steam line break) for the next operating cycle.

2. If circumferential crack-like indications are detected at the tube support plate intersections.

3. If indications are identified that extend beyond the confines of the tube support plate.

4. If indications are identified at the tube support plate elevations that are attributable to primary water stress corrosion cracking.

5. If the calculated conditional burst probability based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds 1 x 10-2, notify the NRC and provide an assessment of the safety significance of the occurrence.

FARLEY-UNIT 1 3/4 4-13 AMENDMENT No. 125

TABLE 4.4-3  ;

i STEAM GENERATOR REPAIRED TUBE INSPECTION . ,

"1 IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION Sample Size Result Actions Required Result Action Required Q A minimum of C-1 None NA NA -

20% of ,

h repaired tubes  ;

H (1) (2)

>o C-2 Plug or repair defective C-1 None '

repaired tubes and inspect ,

100% of the repaired tubes in C-2 Plug or repair defective i this steam generator. repaired tubes.  :

C-3 Perform action for C-3 result of first sample.

C-3 Inspect all repaired tubes in All other steam None s this steam generator, plug or generators are i

, repair defective tubes and C-1. j

} inspect 20% of the repaired Some steam Perform action for C-2 l tubes in each steam generator generators C-2 result of first sample. '

i but no additional U Notification to NRC pursuant steam generators

• to 10 CFR 50.72 (b) (2) . are C-3. -

Additional steam Inspect all repaired tubes generator is C-3. in each steam generator and j plug or repair defective tubes. Not;fication to NRC pursuant to 10 CFR 50.72 (b) (2) . I (1) Each repair method is considered a separate population for determination of scope expansion.

g (2) The inspection of repaired tubes may be performed on tubes from 1 to 3 steam generators based on outage g plans. j 2

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REACTOR COOLANT SYSTEM BASES The voltaga structural limit is the voltage from the burst pressure / bobbin voltage correlation at the 95-percent prediction interval curve reduced to account for the lower 95/95-percent tolerance for tubing material  ;

properties at 650 *F (i.e., the 95-percent LTL curve). The voltage structural limit must be adjusted downward to account for potential flaw growth during an operating interval and to account for NDE uncertainty. .

The upper voltage repair limit, Vau, is determined from the structural l voltage limit by applying the following equation:

1 Vmu, = Vsi - Vor - Vnu

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where Vor represents the allowance for flaw growth between inspections and Vna represents the allowance for potential sources of error in the measurement of the bobbin coil voltage. Further discussion of the assumptions necessary to determine the voltage repair limit is contained in GL 95-05.

The mid-cycle equation in 4.4.6.4.a.11.d should only be used during unplanned inspections in which eddy current data is acquired for indications at the tube support plates.  ;

4.4.6.5 implements several reporting requirements recommended by GL 95-05 for situations in which the NRC wants to be notified prior to returning the SGs to service. For the purposes of this reporting requirement, leakage and conditional burst probability can be calculated based on the as-found voltage distribution rather than the projected end-of-cycle voltage distribution (refer to GL 95-05 for more information) when it is not practical to complete these calculations using the projected EOC voltage ,

distributions prior to returning the SGs to service. Note that if leakage and conditional burst probability were calculated using the measured EOC voltage distribution for the purposes of addressing the GL section 6.a.1 and 6.a.3 reporting criteria, then the results of the projected EOC voltage distribution should be provided per the GL section 6.b(c) criteria.

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 or repair will be required for all tubes with imperfections exceeding 40% of the tube nominal wall thickness. If a sleeved tube is found to have thrrugh wall penetration of greater than or equal to 31% for the mechanical sleeve and 24% for the laser welded sleeve of sleeve nominal wall thickness in the sleeve, it must be plugged. The 31% and 24% limits are derived from R.G. 1.121 calculations with 20% added for conservatism.

The portion of the tube and the sleeve for which indicati'ns o of wall degradation must be evaluated can be summarized as follows:

EARLEY-UNIT 1 B 3/4 4-3a AMENDMENT NO.125

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j Mp ter%.

p t UNITED STATES l

g j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 300 SHOO 1 i

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4

SOUTHERN NUCLEAR OPERATING COMPANY. INC.

ALABAMA POWER COMPANY i

, DOCKET NO. 50-364 l

l JOSEPH M. FARLEY NVCLEAR PLANT. UNIT 2 l- AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.119

[ License No. NPF-8 l

1. The Nuclear Regulatory Comission (the Comission) has found that:

} A. The application for amendment by Southern Nuclear Operating Company, Inc. (Southern Nuclear), dated January 10, 1997, as j supplemented by letter dated February 24, 1997, complies with the i standards and requirements of the Atomic Energy Act of 1954, as l amended (the Act), and the Comission's rules and regulations set

i. forth in 10 CFR Chapter I; i

j B. The facility will operate in conformity with the application, the i provisions of the Act, and the rules and regulations of the i Comission; i

l C. There is reasonable assurance (1) 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 Comission's regulations; D. The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the i public; and

! E. The issuance of this amendment is in accordance with 10 CFR Part

51 of the Comission's regulations and all applicable requirements j have been satisfied.

I 2. Accordingly, the license is amended by changes to the Technical i Specifications, as indicated in the attachment to this license

amendment; and paragraph 2.C.(2) of Facility Operating License No. NPF-8

, is hereby amended to read as follows:

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e (2) Technical Soecifications  ;

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! The Technical Specifications contained in Appendices A and B, as l revised through Amendment No.119, are hereby incorporated in the

, license. Southern Nuclear shall operate the facility in i accordance with the Technical Specifications.

i 3. This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION 4

(& N. h /

Herbert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects - I/II l Office of Nuclear Reactor Regulation I

Attachment:

Changes to the Technical ,

Specifications i 1

Date of Issuance: March 24, 1997 1

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ATTACHMENT TO LICENSE AMENDMENT NO.119 TO FACILITY OPERATING LICENSE NO. NPF-8 i

DOCKET NO. 50-364 I

Replace the following pages of the Appendix A Technical Specifications with j the enclosed pages. The revised areas are indicated by marginal lines.

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i Remove Insert 3/4 4-9 3/4 4-9 l 3/4 4-10 3/4 4-10 i

! 3/4 4-11 3/4 4-11 )

3/4 4-12a 3/4 4-12a 3/4 4-12b 3/4 4-12b ,

! 3/4 4-13 3/4 4-13 I i

3/4 4-13a 3/4 4-13a i 3/4 4-13b 3/4 4-13b 3/4 4-15a i B 3/4 4-3 B 3/4 4-3

{' B 3/4 4-3a B 3/4 4-3a l B 3/4 4-3b B 3/4 4-3b g B 3/4 4-3c ---

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• . o REACTOR COOLANT SYSTEM 3/4.4.6 STEAM GENERATORS .

LIMITING CONDITION FOR OPERATION 3.4.6 Each steam generator shall be OPERABLE.

APPLICABILITY: HODES 1, 2, 3 and 4.

ACTION-With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing Tavg above 200*F.

SURVEILLANCE REQUIREMENTS .

i 4.4.6.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.

4.4.6.1 Steam Generator Sample Selection and Inspection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.

4.4.6.2.1 Steam Generator Tube # Sample Selection and Inspection'- The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Tables 4.4-2 and 4.4-3. The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.4.6.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.6.4. The tubes selected for each inservice inspection l shall include at least 3% of the total number of tubes in all steam generators. Selection of tubes to be inspected is not affected by the F* l '

designation. When applying the exceptions of 4.4.6.2.1.a through 4.4.6.2.1.c, previous defects or imperfections in the area repaired by sleeving are not considered an area requiring re-inspection. 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 50% of the tubes inspected shall be from these critical areas.

b. The first 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%.

1. When referring to a steam generator tube, the sleeve shall be considered a part of the tube if the tube has been repaired per Specification 4.4.6.4.a.9.

EARLEY-UNIT 2 3/4 4-9 AMENDMENT NO.119

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REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) l

2. Tubes in those areas where experience has indicated l potential problene.

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3. A tube inspection (pursuant to Specification 4.4. 6.4.a.8) I shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube or sleeve inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection. ,  ;

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4. Indications left in service as a result of application of l '

the tube support plate voltage-based repair criteria shall be inspected by bobbin coil probe during AA1 future ,

refueling outages.

c. The tubes selected as the second and third samples (if required by Tables 4.4-2 and 4.4-3) during each inservice inspection may l be subjected to a partial tube inspection provided:

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1. The tubes selected for these samples include the tubes i from those areas of the tube sheet array where tubes with j imperfections were previously found.

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2. The inspections include those portions of the tubes where '

imperfections were previously found.

d. Implementation of the steam generator tube / tube support plate repair criteria requires a 100 percent bobbin coil inspection for hot-leg and cold-leg tube support plate intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of the lowest cold-leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20 percent random sampling of tubes inspected over their full length.

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

EARLEY-UNIT 2 3/4 4-10 AMENDMENT No.119 -

. o REACTOR COOLANT 3YSTEM  !

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SURVEILLANCE REQUIREMENTS (Continued)

Category Inspection 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 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.

I C-3 More 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 or sleeves must exhibit i significant (greater than 10%) further wall penetrations to be included in the above percentage calculations.

{

4.4.6.2.2 Steam Generator F* Tube Inspection - In addition to the minimum sample size as determined by specification 4.4.6.2.1, all F* tubes will be inspected within the tubesheet region. The results of this inspection will not be a cause for additional inspections per Tables 4.4-2 and 4.4-3. l 4.4.6.3 Inspection Frequencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:

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a. The first inservice inspection shall be performed after 6 Ef fective Full Power Months but within 24 calendar months of initial criticality fubsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections following service under AVT cdnditions, 4

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 40 months.

b. If the results of the inservice inspection of a steam generator conducted in accordance with Tables 4.4-2 and 4.4-3 at 40 month l

intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in ,

inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.6.3.a; the interval may then be extended to a umximum of once per 40 months.

c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Tables 4.4-2 and 4.4-3 during l

, the shutdown subsequent to any of the following conditions:

FARLEY-UNIT 2 3/4 4-11 AMENDMENT NO.119

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,RF. ACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) i

6. Plugging or Repair Limit means the imperfection depth at

] or beyond which the tube shall be repaired (i.e.,

sleeved) or removed from service by plugging and is greater than or equal to 40% of the nominal tube wall thickness. This definition does not apply for tubes that l- ' meet the F* criteria. For a tube that has been sleeved l j with a mechanical joint sleeve, through wall penetration of greater than or equal to 31% of sleeve nominal wall l thickness in the sleeve requires the tube to be removed i

from service by plugging. For a tube that has been 4

sleeved with a welded joint sleeve, through wall 1 penetration greater than or equal to 24% of sleeve l l nominal wall thickness in the sleeve between the weld  !

i joints requires the tube to be removed from service by plugging. This definition does not apply to tube support j plate intersections for which the voltage-based repair i criteria are being applied. Refer to 4.4.6.4.a.16 for i

the repair limit applicable to these intersections. For I a tube with an imperfection or flaw in the tubesheet

below the lower joint of an installed elevated laser welded sleeve, no repair or plugging is required provided the installed sleeve meets all sleeved tube inspection i requirements.

7. Unserviceable describes the condition of a tube or sleeve i if it leaks or contains a defect large enough to affect its structural integrity in the event of an operating Basis Earthquake, a loss-of-coolant accident, or a steam j line or feedwater line break as specified in 4.4.6.3.c, above.

8. Tube Inspection means an inspection of the steam

! generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg. For a tube with a tube sheet sleeve installed, the point of entry is the bottom of the tube sheet sleeve below the lower sleeve joint. For a tube that has been repaired by sleeving, the tube inspection should include the sleeved portion of the tube.

9. Tube Repair refers to mechanical sleeving, as described by Westinghouse report WCAP-ll178, Rev. 1, or laser welded sleeving as described by Westinghouse reports WCAP-13088, Revision 4, and WCAP-14740 dated January 1997, which is used to maintain a tube in service or return a tube to service. This includes the removal of plugs that were installed as a corrective or preventive measure.

FARLEY-UNIT 2 3/4 4-12a AMENDMENT NO.119

4 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

10. Preservice Inspection 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. This inspection shall be performed after the field hydrostatic test and prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.

11. F* Distance is the distance of the expanded port, ion of a tube which provides a sufficient length of undegraded tube expansion to resist pullout of the tube from the tubesheet. The F* distance is equal to 1.54 inches plus allowance for eddy current uncertainty measurement and is measured down from the top of the tube sheet or the bottom of the roll transition, whichever is lower in elevation. The allowance for eddy current uncertainty is documented in the steam generator eddy current inspection procedure.

12. F* Tube is a tube:

a) with degradation equal to or greater than 40% below .

the F* distance, and b) which has no indication of imperfections greater than or equal to 20% of nominal wall thickness within the F* distance, and c) that remains inservice.

i J

13. Tube Expansion is that portion of a tube which has been l increased in diameter by a rolling process such that no crevice exists between the outside diameter of the tube and the hole in the tubesheet. Tube expansion a' Iso refers to that portion of a sleeve which has been increased in diameter by a rolling process such that no crevice exists between the outside diameter of the sleeve and the parent steam generator tube.

14. Tube Support Plate Repair Limit is used for the l disposition of an alloy 600 steam generator tube for continued service that is experiencing predominately axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the repair limit is based on maintaining steam generator tube serviceability as described below:

FARLEY-UNIT 2 3/4 4-12b AMENDMENT NO.119

l l

REACTOR COOLANT SYSTEM I SURVEILLANCE REQUIREMENTS (Continued)

a. Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltages less than or equal to the lower voltage repair limit [2.0 volts), will be allowed to remain in service.

b. Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion '

cracking within the bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair limit [2.0 volts), will be repaired or plugged except as noted in 4.4.6.4.a.14.c below. l

c. Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair limit

[2.0 volts) but less than or equal to the upper voltage repair limit *, may remain in service if a rotating probe inspection does not detect degradation. Steam generator tubes, with indications of outside diameter stress corrosion cracking degradation with a bobbin voltage greater [

than the upper voltage repair lindt*, will be  ;

plugged or repaired. .

d. If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of the lindts identified in 4.4.6.4.a.14.a, 4. 4. 6. 4. a .14.b, and 4. 4. 6. 4. a .14. c. l The mid-cycle repair limits are determined from the following equations:

i Yst Vnma= 1.0 + NDE + Gr [ CL-At ]

CL

. I Van =Vanc-[Vun-V L n] [ CL-At ]

CL )

)

The upper voltage repair limit is calculated according to the methodology in Generic Letter 95-05 as supplemented.

FARLEY-UNIT 2 3/4 4-13 AMENDMENT No. 119

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1

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REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) where:

Vma = upper voltage repair limit Vua = lower voltage repair limit VWat = mid-cycle upper voltage repair limit based on l time into cycle

• l Vxua = mid-cycle lower voltage repair limit based on Vana and time into cycle at = length of time since last scheduled I inspection during which Vma and Vua were  !

implemented l CL = cycle length (the time between two scheduled 1 steam generator inspections)

V,6 = structural limit voltage

)

Gr =

average growth rate per cycle length NDE =

95-percent cumulative probability allowance for nondestructive examination uncer,tainty  !

(i.e., a value of 20-percent has been approved by NRC)

Implementation of these mid-cycle repair limits should follow the same approach as in TS 4.4.6.4.a.14.a, 4.4.6.4.a.14.b, and 4.4.6.4.a.14.c.

b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair of all tubes exceeding the plugging or repair limit) required by l Tables 4.4-2 and 4.4-3. l 4.4.6.5 Reports *

a. Following each inservice inspection of steam generator tubes, the number of tubes plugged, repaired or designated F* in each l steam generator shall be reported to the Commission within 15 days of the completion of the inspection, plugging or repair effort.

b. The complete results of the steam generator tube and sleeve inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within ,12 months following the completion of the inspection. This Special )

Report shall include:

l. Number and extent of tubes and sleeves inspected. l

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

I

3. Identification of tubes plugged or repaired.

l FARLEY-UNIT 2 3/4 4-13a AMENDMENT NO.119

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) t

c. Results of steam generator tube inspections which fall into Cntegory C-3 shall be considered a REPORTABLE EVENT and shall be reported pursuant to 10CFR50.73 prior to resumption of plant ,

operation. The written report shall provide a description of  !

investigations conducted to determine the cause of the tube degradation and corrective measures taken to prevent recurrence. l

d. For implementation of the voltage-based repair criteri*a to tube l support plate intersections, notify the staff prior to )

returning the steam generator to service (Mode 4) should any of l the following conditions arise: j i

1. If estimated leakage based on the projected end-of-cycle '

(or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds the leak limit (determined from the licensing basis dose calculation for the postulated main steam line break) for the next operating cycle.

2. If circumferential crack-like indications are detected at l the tube support plate intersections. -

3. If indications are identified that extend beyond the confines of the tube support plate.

4. If indications are identified at the tube support plate I elevations that are attributable to primary water stress corrosion cracking.

5. If the calculated conditional burst probability based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds 1 x 10'#, notify the NRC and provide an dasessment of the safety significance of the occurrence.

EARLEY-UNIT 2 3/4 4-13b AMENDMENT NO. jjg

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TABLE 4.4-3 r P

STEAM GENERATOR REPAIRED TUBE INSPECTION IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION t

" Sample Size Result Actions Required Result Action Required k A minimum of C-1 None NA NA E 20% of N repaired tubes (1) (2)

C-2 Plug or repair defective C-1 None repaired tubes and inspect '

100% of the repaired tubes in C-2 Plug or repair defective this steam generator. repaired tubes. ,

C-3 Perform action for C-3 result of first sample.

w g C-3 Inspect all repaired tubes in All other steam None

, this steam generator, plug or generators are i 4 repair defective tubes and C-1.

p inspect 20% of the repaired some steam Perform action for C-2  ;

tubes in each steam generator generators C-2 result of first sample.

but no additional i Notification to NRC pursuant steam generators [

to 10 CFR 50.72 (b) (2) . are C-3. .

t Additional steam Inspect all repaired tubes i

generator is C-3. in each steam generator and plug or repair defective tubes. Notification to NRC  !

pursuant to 10 CFR

• 50.72 (b) (2) .

a '

2 (1) Each repair method is considered a separate population for determination of scope expansion.

g (2) The inspection of repaired tubes may be performed on tubes from 1 to 3 steam generators based on outage G plans.

I

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a )

I REACTOR COOLANT SYSTEM BASES 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 j 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. l Inservice inspection of steam generator tubing also provides a means or I

! characterizing the nature and cause of any tube degradation so that corrective measures can be taken. l The plant is expected to be operated in a manner such that the sec*ondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary.

1 coolant chemistry is not maintained within these limits, localized '

corrosion may likely result in stress corrosion cracking. The extent of l cracking during plant operation would be limited by the limitation of steam l i

generator tube leakage between the primary coolant system and the secondary I coolant system (primary-to-secondary leakage = 150 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. Operational leakage of this magnitude can be readily detected

• by existing Farley Unit 2 radiation monitors. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged or repaired.

The voltage-based repair lindts of 4.4.6.4.a.14 implement the guidance in l GL 95-05 and are applicable only to Westinghouse-designed steam generators (SGs) with outside diameter stress corrosion cracking (ODSCC) located at the tube-to-tube support plate intersections. The voltage-based repair l limits are not applicable to other forms of SG tube degradation nor are they applicable to ODSCC that occurs at other locations within the SG.

Additionally, the repair criteria apply only to indications where.the degradation mechanism is dominantly axial ODSCC with no significant cracks i extending outside the thickness of the support plate. Refer to GL 95-05 for additional description of the degradation morphology.

Implementation of 4.4.6.4.a.14 requires a derivation of the voltage l structural limit from the burst versus voltage empirical correlation and then the subsequent derivation of the voltage repair limit from the structural limit (which is then implemented by this surveillance).

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FARLEY-UNIT 2 B 3/4 4-3 AMENDMENT NO. I19

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4

• 1 i

REACTOR COOLANT SYSTEM BASES I

_-_ - - g - _ _ - _-_- _-_ - .. - - - - - _-

l The voltage structural limit is the voltage from the burst pressure / bobbin '

voltage correlation, at the 95-percent prediction interval curve reduced to account for the lower 95/95-percent tolerance bound for tubing material properties at 650 *F (i.e., the 95-percent LTL curve). The voltage structural limit must be adjusted downward to account for potential flaw growth during an operating interval and to account for NDE uncertainty.

The upper voltage repair limit: Vyn, is determined from the structural i voltage limit by applying the following equation:

vm=Vn-V=-Vm

where Va represents the allowanc<e for flaw growth between inspections and I Vms represents the allowance for potential sources of error in the measurement of the bobbin coil voltage. Further discussion of the assumptions necessary to determine the voltage repair limit are discussed in GL 95-05.

The mid-cycle equation in 4.4.6.4.a.14.d should only be used during l ,

unplanned inspections in which oddy current data is acquired for j

, indications at the tube support plates.

4.4.6.5 implements several reporting requirements recommended by GL 95-05 i for situations which the NRC wants to be notified prior to returning the SGs to service. For the purposes of this reporting requirement, leakage i

and conditional burst probability can be calculated based on the as-found voltage distribution rather tha.n the projected end-of-cycle volte.ge distribution (refer to GL 95-05 for more information) when it is not practical to ecmplete these calculations using the projected EOC voltage distributions prior to returning the SGs to service. Note that if, leakage and conditional burst probabilaty were calculated using the measured EOC voltage distribution for the p2rposes of addressing the GL section 6.a.1 i and 6.a.3 reporting criteria, then the results of the projected EOC voltage distribution should be provided per the GL section 6.b(c) criteria.

l Wastage-type defects are unlikely with preper 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 or repair will be required for all tubes with imperfections exceeding 40% of the tube nondnal wall thickness. If a sleeved tube is found to have through wall penetration of greater than or equal to 31% for the mechanical sleeve and 24% for the laser welded sleeve of sleeve nominal ,

wall thickness in the sleeve, it must be plugged. The 31% and 24% limits are derived from R.G. 1.121 calculations with 20% added for conservatism.

The portion of the tube and the sleeve for which indications of wall degradation must be evaluate:d can be summarized as follows:

a. Mechanical

1. Indications of degradation in the entire length of the sleeve must be evaluated against the sleeve plugging limit.

FARLEY-UNIT 2 B 3/4 4-3a AMENDMENT NO.119

_ . .~. .- .

I

REACTOR COOLANT SYSTEM .

l BASES l . . . .

i 2. Indication of tube degradation of any type including a complete

guillotine break in the tube between the bottom of the upper l joint and the top of the lower roll expansion does not require

( that the tube be removed from service.

3. The tube plugging limit continues to apply to the portion of l the tube in the entire upper joint region and in the lower roll expansion. As noted above, the sleeve plugging lindt applies to these areas also.

4. The tube plugging limit continues to apply to that portion of l the tube above the top of the upper joint. ,

b. Laser Welded

1. Indications of degradation in the length of the sleeve between the weld joints must be evaluated against the sleeve plugging limit.

2. Indication of tube degradation of any type including a complete break in the tube between the upper weld joint and the lower weld joint does not require that the tube be removed from service.

3. At the weld joint, degradation must be evaluated in both the sleeve and tube.

4. In a joint with more than one weld, the weld closest to the end of the sleeve represents the joint to be inspected and the limit of the sleeve inspection.

5. The tube plugging limit continues to apply to the portion of the tube above the upper weld joint and below the lower weld joint.

F* tubes do not have to be plugged or repaired provided the remainder of the tube within the tubesheet that is above the F* distance is not degraded. The F* distance.is equal to 1.54 inches plus allowance for eddy current uncertainty measurement and is measured down from the top of the -

tubesheet or the bottom of the roll transition, whichever is lower in elevation.

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

Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pursuant to 10 CFR 50.73 prior.to resumption of plant operation. Such casts will be 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 to the Technical Specifications, if necessary.

FARLEY-UNIT 2 B 3/4 4-3b AMENDMENT NO.119

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