ML20101L917

Proposed TS 3/4.4.6, Steam Generators & Associated Bases

ML20101L917
Person / Time
Site:	Farley
Issue date:	03/29/1996
From:	SOUTHERN NUCLEAR OPERATING CO.
To:
Shared Package
ML20101L907	List: ML20101L904 ML20101L917
References
NUDOCS 9604050049
Download: ML20101L917 (27)

Text

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Attachment 2 Unit 2 Technical Specification Pages i

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9604050049 960329 PDR ADOCK 05000364 P PDR

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Unit 2 Technical Specification Page Markups t

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:.:uce at:u:amms :snunnai

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2 Tubes in those areas where experience has indicated '

potential problems, i

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At least 36 of the total number of sleeved tubes in all three stsam generators or all of the sleeved tubes in i the generator chosen for the inspection program, j whichever is less. These inspections will include both the tube and the sleeve.

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"" 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

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i eddy current probe for a tube or sleeve inspection, i

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

VO Nvf C- I S. L ";;:: 12tddications 1.ft in service as a

!' I @ ig gjh( % %resul4ofapplicationofthetubesupportplate

.a; ::15::12- shall le i "

probe during a "ll 5 tor:e.i2,gi aspected by bobbin coil 8'l'-- Icefueling outages.

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

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

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

i Isplementation of the see generator tube / tube support plate

[ep & h $ 1r;;in; criteria require 00 percent bobbigeoil inspection N C' for hot-leg 22: cr;;:n ;_::: irt:;;;nia .ad cold-leg g(-- " JcQ W g intersections down to the lowest cold-leg tube support plate eja/O with known outside diameter stress porresta= craewn= (oDscc) F indications. The determination of Mube support plate m intersections having OD8cc indications shall be based on the performance of at least a 20 percent random sampling of tubes e,,

inspected over their full length.

The results of esob sagle inspection shall be classified into one of the j following three categories: g/d,jlg t__

&};b'$

FARLEY-UNIT 2 3/4 4-10 AMDHBIENT N0. 106

. _ _ . - - . - -. ~ .- .._. - _- . - - -- __-. .~.. . -- . - _ - _

SEA:70s ::

• ANT 3V57D1 1

. --S'JAVEILLANCE RECUIPD'ENTS scontinued)

J j 6. F19eging or Roosir Limit means the imperfection depth i

at er beyond which the tube shall be repaired (i.e.,  :

l sleeved) or removed from service by plugging and is greater than er equal to 404 of the acminal tube wall i

thickness. This definition does not apply to the area l of the tubesheet region below' the F* distance in the F*

tubes. For a tube that has been sleeved with e

( mschanical joiht sleeve, through wall penetration of j "

greater than or equal to 31% of sleeve nominal wall f 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 well penetration greater than er equal to 374 et sleeve  ;

j naminal well thickness la the sleeve between the weld '

joints requires the tube to be removed from service by j plugging. This definition does not apply to tube

! support plate intersections for which the voltage-based d

l ;h;;i.q criteria are being a lied. Refer te imit applicable to these l fy (- 4.4.6.4.a.14 for latersections. th)e ' u;;i__,

i 1 j 7. Unserviceab1_e,, describes the condittom of a tube or  ;

sleeve if it leaks er contains a defect large enough to l

, affect its structural integrity in the event of an j operating Basis Earthquake, a loss-of-coolant accident, j or a steam line or feedwater line break as specified in j 4.4.6.3.c, above.

! 8. Tube Inspection means an inspection of the steen

{ generator tube from the point of entry thot leg sidel cospletely around the U-bood to the top support of the l cold leg. For a tube that has been repaired by

. eleeving, the tube inspection should include the i sleeved portion of the tube.

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j g. Tube Repair refers to mechanical sleeving, as described j by Westinghouse report WCAP-11178, Rev.1, or laser welded sleeving as described by Westinghouse report i WCAP-12672, which is used to maintain a tube in service l or return a tube to service. This includes the removal l of plugs that were installed as a corrective or

preventive measure.

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1 f ( FARLEY-UNIT 2 3/4 4-12a AMEN!MNT NO.105 l 4

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, PEAb:2 ::C:. ANT SYSTEM

• SURVEILLANCE REQUIRENENTS I:"entinuedi h

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 I using the equipment and techniques espected to be used

.o during subsequent inservice inspections.

11. F* Distance is the distance of the expanded portion of a tube which provides a sufEicient length of undegraded l tube expansion to resist pullout of the tube from the  ;

tubesheet. The F* distance is equal to 1.19 inches and is measured down from the top of the tubesheet or the bottom of the roll transition, whichever is lower in elevation.

12. F* Tube is a tubes al with degradation equal te or greater than 406 below the F* distance, and b) which has no indication of imperfections greater thaa er equal to 20% of nominal well thickness within the F* distance, and c) that remains inservice.

13. Tube Eicoansion is that portion of a tubs which has been increased in diameter by a rolling process such that no crevice exists between the outside diameter of the t .

T and the hole in the tubesheet, predom.)ttc),[ % [

LL6. Q.NOf 14. bh Tube supporthatefrptLit' O fid*-

th;_' y_ Limit is for the disposition of4d-steam generet for continued service that is experiencing utside diameter stress corrosies cracking confined within the thickness of the y A ubin's plate intersections, the repair limit is based en maintaining steam generater tube serviceability as described below:

6 yallegradation attributed to outside diameter stress Lg g # gho corrosion cracking within the bounds of the tube support ate with bobbia voltage less than or equal .0 voltgwill be allowed to reunia is service ,

b. ttributed to outside diameter stress y /s //%f0 j splogradationhacking corrosion c within the bounds of the tube upport late with a hobbia voltage greater thaa

[ p '/" // /M/. .0 vol 11 be repaired or plugged ascept as p noted in . 4. 6. 4. a .14. c below.

FARLEY-UNIT 2 3/4 4-12h AManBENT 110. 105 y -p p al+ 4 -13

ptA: :s :::'.w. 3YST m C kWer VO 6

, _3'.;RVE*i,* AN:E PE;'.':RmENT5 8 continued -

J g ggMgg pdicationsofpotentialdegr tion attributed to outside diameter stress co osion cracking within the bounds of the tub upport ste with a hobbia voltage greater p .0 volt but less than or equal ta.0.0 ...i. may r'emain a service tf a rotating p&acake coil inspection does not detect degradationdMcations of outside 7

fg /d[ .

c- i diameter stress cerrosion cracking degradation with a b'obbin voltage greater tha 0. 0 ..." $111 f ///M #by be plugged er repaired,

b. The steam generator shall be determined CPEAASLE after completing the corresponding actions (plug or repair of all tubes exceeding the plugging or repair limiti required by Table 4.4-2.

p- 7% VPP * 'fg in %ario

-j a- por as sepwa' FARLEY-UNIT 2 3/4 4-13 ANENDMENT NO. 106

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Insert A ,

d. If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of ,

the limits identified in 4.4.6.4.a.14.a,  !

4.4.6.4.a.14.b, and 4.4.6.4.a.14.c. The mid-cycle repair limits are determined from the following equations:

Vst Vmn= 1.0 + NDE + Gr [ CL-At ]

CL Vm,n=Vmn-[Vun-V t e] [ CL-At ]

CL where:

= upper voltage repair limit I Von Vnt

= lower voltage repair limit

=

Vmn mid-cycle upper voltage repair limit based on time into cycle Vm.n = mid-cycle lower voltage repair limit based on Ven and time into cycle At = length of time since last scheduled i inspection during which Von and V t n were implemented '

CL = cycle length (the time between two scheduled steam generator inspections) .

Vn = structural limit voltage Gr = average growth rate per cycle length NDE = 95-percent cumulative probability allowance for nondestructive examination >

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

-Implementation of these mid-cycle repair limits should I 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.

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PEA *T A **C'E7 S Y5704

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$*JRVE!*.:.ANCE PEQUIPD4ENTS tConnnuedI 4.4.6.5 Reports  ;

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

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b. .The complete results of the steam generator tube and sleeve inservice inspection shall be submitted to the Ceemission 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 idtich fall into Category C-3 shall be considered a REPORTABLE EVENT and shall be reported pursuant to 10CFRSO.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' staff prior to returning the steam generator to service (Mode 4) should any of the following conditions arises _

(If estimated leakage based on the actual end-of-c voltage distribution would have exceeded the leak limit 7 g,(- h h 7 (for the postulated mala steau line break utilising licensing basis assumptions) during the previous ,

( operating cycle. l

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

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

f the calculated conditional burst probability exceeds

{ 1 a 10, notify the N3tc and provide as assessment of

.,/Lf 6 (the safety significance of the occurrence.

rARLEY-UwIT 2 3/4 4-13a AMINEMENT NO. 106

46) 63, (of-p4-o matly p.'>l4 t-i3

Insert B

, 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 j (determined from the licensing basis dose calculation for l the postulated main steam line break) for the next operating l cycle.

Insert c

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, notify the NRC and provide an assessment of the safety significance of the occurrence.

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. . su: :: ::: _u: : : t" 3A3E3 1 3/4.4.6 STEAM GENERATORS i

4 The surveillance pequirements for inspection of the steam generator tubes j ensure that the structural integrity of this portion of the RCS will be I maintained. The program for inservice inspection of steam generator tubes

is based on a modification of Regulatory Guide 1.03, Revision 1. Inservice

] inspection of steam generator tubing is essential in order to maintain j surveillance of the conditions oft the tubes in the event that there is j- evidence of mechanical damage or progressive degradation due to design, j manufacturing errors, or inservice conditions that lead to corrosion.

l Inservice inspection of steam generator tubing also provides a means of i characterising the nature and cause of any tube degradation so that j corrective measures can be taken.

) The plant is expected to be operated in manner such that the secondary coolant will be maintained withis 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 plaat operation would be limited by the limitation of steam; generator tube (primary-to-secondary coolant system leakage between the primary leakage coolant systemper

= 150 gallons and the day secondary l per

. . steam generator). Cracks having a primary-to-secondary leakage less than j this limit during operaties will have an adequate margin of safety to j withstand the loads imposed during normal operation and ky postulated 1 accidents. Operational leakage of this magnitude can be readily detected j by existing Farley Unit 2 radiation monitors. Leakage in excess of ghis

limit will require plant shutdown and an unscheduled inspection, during

} which the leaking tubes will be located and plugged or repaired.

j The repair limit for ODSCC at tube support plate intersections is based on )

the analysis contained in NCAF-12071, Revision 2, "J. M. Farley Units 1 and l r ( 2 SG Tube Flugging Criter!.a for ODSCC at Tube Support Plates," and j _,M documentation contained in EPRI Report TR-100407, Revision 1, *FWR Steam M

Generator Tube Repair Limits - Technical Support Document for Outside l Diameter Stress corrosion Cracking at Tube support Plates." The 3

, application of this criteria is based on liatting primary-to-secondary leakage during a steam line break to ensure the applicable Part 100 limits I 5

i Qenotescoeded.

j wastage-type defects are unlikely with proper chemistry treatment of the l

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

Flugging or repair will be required for all tubes with isperfections exceeding 40% of the tube nominal wall thickness. If a sleeved tube is

found to have through wall penetration of greater than or equal to 314 for

! the mechanical sleeve and 376 for the laser welded sleeve of sleeve nominal l

wall thickness in the sleeve, it must be plugged. The 31% and 37% limits are derived f rom R. G. 1.121 calculations with 204 added for conservatism.

The portion of the tube and the sleeve for which indications of wall

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degradation must be evaluated can be summarized as follows:

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FARLEY-UNIT 2 3 3/4 4-3 AMEN!Beff No. 106 i ( -

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. Insert D The voltage-based repair limits of 4.4.6 implement the guidance in 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 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 I dominantly axial ODSCC with no significant cracks 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 requires a derivation of the voltage 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).

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; Vun, is determined from the structural voltage limit by applying the following equation:

Vum. = Vn - Vor - Vms where Va, represents the allowance for flaw growth between inspections and 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.

l The mid-cycle equation in 4.4.6.a.14.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 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.

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l Unit 2 Technical Specification Pages Replacement Pages Page 3/4 4-10 Replace Page 3/4 4-12a Replace Page 3/4 4-12b Replace l Page 3/4 4-13 Replace Page 3/4 4-13a Replace l

Page B 3/4 4-3 Replace l Page B 3/4 4-3a Replace l Page B 3/4 4-3b Replace l I l

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, REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

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

3. At least 3% of the total number of sleeved tubes in all three steam generators or all of the sleeved tubes in the generator chosen for the inspection program, whichever is less. These inspections will include both the tube and the sleeve.

4. 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.

5. Indications left in service as a result of application of 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 Table 4.4-2) during each inservice inspection may 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:

FARLEY-UNIT 2 3/4 4-10 AMENDMENT NO.

f REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

6. Plugging or Repair Limi,1 means the imperfection depth at or l beyond which the tube shall be repaired (i.e., sleeved) or I

removed from service by plugging and is greater than or equal to 40% of the nominal tube wall thickness. This definition does not apply to the area of the tubesheet region below the F* distance in the F* tubes. 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 l 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 37% of sleeve nominal wall thickness in the sleeve between the weld joints requires the tube to be removed from service by plugging.

l This definition does not apply to tube support plate intersections for which the voltage-based repair criteria are being applied. Refer to 4.4.6.4.a.14 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 its structural integrity in the event of an Operating Basis l Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.6.3.c, above.

l 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 the tube.

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9. Tube Repair refers to mechanical sleeving, as described by Westinghouse report WCAP-11178, Rev. 1, or laser welded l

sleeving as described by Westinghouse report WCAP-12672, 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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l 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. )

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

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• REACTOR COOLANT SYSTEM a

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! SURVEILLANCE REQUIREMENTS (Cc-ntinued)

11. F* Distance is the distance of the expanded portion of a l tube which provid,es a sufficient length of undegraded tube l expansion to resist pullout of the tube from the tubesheet.

The F* distance is equal to 1.79 inches and is measured down from the top of the tubesheet or the bottom of the roll transition, whichever is lower in elevation.

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12. F* Tube is a tube:

a) with degradation equal to or greater than 40% below the F* dintance, and b) which has no indication of imperfections l

greater than or equal to 20% of nominal wall thickness within the F* distance, and c) that remains inservice.

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 l the hole in the tubesheet.

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14. Tube Support Plate Repair Limit is used for the 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 I thickness of the tube support plates. At tube support plate l intersections, the repair limit is based on maintaining l steam generator tube serviceability as described below I l l a. Steam generator tubes, whose degradation is attributed l l l to outside diameter stress corrosion cracking within l the bounds of the tube support plate with bobbin l voltages less than or equal to the lower voltage l repair limit (2.0 volts), will be allowed to remain in i service.

b. Steam generator tubes, whose degradation is attributed l I 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, i

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

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REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

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 l indications of outside diameter stress corrosion l cracking degradation with a bobbin voltage greater than the upper voltage rcpair limit *, will be plugged l or repaired.

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

! Vst Vmn= 1.0 + NDE + Gr [ CL-At )

i CL t

VM on=Vmn-[Vun-Vt n] [ CL-At )

CL where:

l Vun = upper voltage repair limit

= lower voltage repair limit l Vua

=

l Vmn mid-cycle upper voltage repair limit based on time into cycle Vka = mid-cycle lower voltage repair limit based on Vmn and time into cycle l At = length of time since last scheduled l inspection during which Vun and Vua were implemented CL = cycle length (the time between two scheduled steam generator inspections)

Vst = structural limit voltage Gr = average growth rate per cycle length NDE = 95-percent cumulative probability allowance for nondestructive examination l uncertainty (i.e., a value of 20-percent has been approved by NRC) l Implementation of these mid-cycle repair limits should follow the same approach as in TS 4.4.6.4.a.14.a, j 4.4.6.4.a.14.b, and 4.4.6.4.a.14.c.

l l b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair of all tubes exceeding i the plugging or repair limit) required by Table 4.4-2.

. 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.

REACTOR COOLANT SYSTEM

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SURVEILLANCE REQUIREMENTS (Continued) 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 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:

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 j degradation and corrective measures taken to prevent recurrence.

d. For implementation of the voltage-based repair criteria to tube  !

support plate intersections, notify the staff prior to returning i the steam generator to service (Mode 4) should any of the j following conditions arise: }'

l. 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.

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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 i' 1 x 10-2, notify the NRC and provide an assessment of the safety significance of the occurrence. ,

FARLEY-UNIT 2 3/4 4-13a AMENDMENT NO. j i

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, 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 l surveillance of the conditions of the tubes in the event that there is

( evidence of mechanical damage or progressive degradation due to design, I 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 i measures can be taken.

l 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 j likely result in stress corrosion cracking. The extent of cracking during l plant operation would be limited by the limitation of steam generator tube l leakage between the primary coolant system and the secondary 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 I i 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

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shutdown and an unscheduled inspection, during which the leaking tubes will be l located and plugged or repaired.

l The voltage-based repair limits of 4.4.6 implement the guidance in GL 95-05

and are applicable only to Westinghouse-designed steam generators (SGs) with l outside diameter stress corrosion cracking (ODSCC) located at the tube-to-tube support plate intersections. The voltage-based repair limits are not l applicable to other forms of SG tube degradation nor are they applicable to l

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 extending outside the thickness of the support plate. Refer to GL 95-05 for additional description of the degradation morphology.

l Implementation of 4.4.6 requires a derivation of the voltage 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  ;

i then implemented by this surveillance). I i

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

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REACTOR COOLANT SYSTEM BASES l The voltage structural limit is the voltage from the burst pressure / bobbin I

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 I 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; Vuu, is determined from the structural voltage limit by applying the following equation

Vun = Vn - Var - Vnz where Va, represents the allowance for flaw growth between inspections and Vos 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.a.14.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 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 baseo on the as-found voltage j 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 through wall penetration of greater than or equal to 31% for the mechanical sleeve and 37% for the laser welded sleeve of sleeve nominal wall thickness in the sleeve, it must be plugged. The 31% and 37% limits are derived from R. G.

1.121 calculations with 20% added for conse.rvatism. The portion of the tube and the sleeve for which indications of wall degradation must be evaluated can be summarized as follows:

a. Mechanical l

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

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

] guillotine break in the tube between the bottom of the upper joint l

and the top of the lower roll expansion does not require that the tube be removed from service.

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

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• REACTOR COOLANT SYSTEM BASES

3. The tube plugging limit continues to apply to the portion of 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 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 l 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. 1 1

r* 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 1.79 inches and is measured down from the top of the tubesheet or the bottom of the roll transition, whichever is lower in elevation. Included in this distance is an allowance of 0.25 inch for eddy current elevation measurement uncertainty.

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 cases 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 B3/4 4-3b AMENDMENT NO. l

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l Attachment 3 1

Significant Hazards Evaluation Voltage-Based Repair Criteria i

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i t Joseph M. Farley Nuclear Plant - Unit 2 Voltage-Based Repair Criteria for the Repair of Westinghouse Steams Generator Tubes ,

Affected by Outside Diameter Stress Corrosion Cracking j 1

Sinnificant Hazards Consideration Analysis  !

DESCRIPTION OF CHANGES As required by 10 CFR 50.91(a)(1), an analysis is provided to i...ai.une that the proposed license amendment to implement the voltage-based repair criteria for tube support plate elevations in accordance with Generic letter 95-05, " Voltage-Based Repair Criteria for Westmghouse Steam Generator Tubes l Affected by Outside Diameter Stress Corrosion Crackmg," involves no significant hazards. He voltage- i l based repair criteria involve a correlation between oddy current bobbin probe signal amplitudes (voltage) l and the tube burst and leakage capabilities. i Specifically, crack indications with bobbin probe voltages less than or equal to 2.0 volts, regardless of j indicated depth, do not require remedial action if postulated steam line break leakage can be shown to be acceptable. Inspections will also be performed to ensure other forms of degradation are not occurring at the tube support plates and that cracks are not being masked at tube support plates by other factors. l The proposed an=aadment would modify Technical Specification 3/4.4.6 " Steam Generators" and its 4 l

associated bases. He steam generator repair limit will be modified to clarify that the appropriate method l for determining serviceability for tubes with outside diameter stress corrosion cracking at the tube support i plate is by a methodology that more reliably assesses structural integrity. For Unit 2, the operational leakage requirement has previously been modified to reduce the total allowable primary-to-=amadary  :

leakage for any one steam generator from 500 gallons per day to 150 gallons per day. In addition, the Technical Specification limit for specific activity of dose equivalent 1" and its transient dose equivalent p I"' reactor coolant specific activity has previously been reduced N a factor of 2 in order to increase the allowable leakage in the event of a steam line break.

EV_ALUATION Steam Generator Tube Intenrity in the development of the voltage-based repair criteria, R.G.1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes," and R.G.1.83, "Insenice inspection of PWR Steam Generator Tubes," were used as the bases for determimng that steam generator tube integrity considerations are maintained within acceptable limits. R.G.1.121 describes a method acceptable to the NRC staff for meeting General Design Criteria (GDC) 2,14,15, 31, and 32 by reducing the probability and consequences of steam generator tube

rupture through determming the limiting safe conditions of tube wall degradation beyond which tubes with unacceptable crackmg, as established by inservice inspection, should be removed from service by plugging or repair. His regulatory guide uses safety factors on loads for tube burst that are consistent with the

requiremonts of Section 111 of the ASME Code. For the tube support plate elevation degradation occurring 2

in the Farley steam generators, tube burst criteria are inherently satisfied during normal operating l

conditions by the presence of the tube support plate. He presence of the tube support plate enhances the integrity of the degraded tubes in that region by precluding tube deformation beyond the diameter of the

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i Significant Hazards Evaluation Page 2 i

, Voltage-Based Repair Criteria drilled hole. Analyses in WCAP-12871 show that for open crevices with as-designed gaps, the tube support plate may not function to provide a similar constraining effect during accident condition loadings.

'Ihe WCAP-12871 analyses for Farley Unit I with corroded and packed crevices, as confirmed by bobbin coil inspection, show that the tube support plates would not be significantly displaced even under steam line break loadmg conditions. For conservatism in the voltage-based repair criteria, no credit is taken in the development of the repair criteria for the presence of the tube support plate during accident condition l loadmgs Conservatively, based on the existing data base, burst testmg shows that the safety requirements for tube burst margins during accident condition loadmgs can be satisfied with bobbin coil signal amplitudes several times larger than the proposed 2.0 voit voltage-based repair criteria, regardless of the depth of tube wall penetration of the crackmg R.G.1.83 describes a method acceptable to the NRC staff for implementag GDC 14,15,31, and 32 through periodic inservice inspection for the Waa of significant tube wall degradation Upon implementation of the voltage-based repair criteria in accordance with Generic leter 95-05, tube leakage considerations must also be addressed. It must be deternuned that the cracks will not leak excessively during all plant conditions. For the voltage-based tube repair criteria developed for the steam generator tubes, no leakage is expected during normal operating conditions even with the presence of through-wall cracks. 'Ihis is the case as the stress corrosion cracking occurring in the tubes at the support plate elevations in the Farley steam generators is short, tight, axially oriented micro cracks often separated by ligaments of material. No leakage during normal operatmg conditions has been observed in the field for l crack indications with signal amplitudes less than 7.7 volts in a 3/4 inch tube. Voltage correlation to 7/8 l inch tubing size would result in an expected voltage of about 10 volts. Relative to the expected leakage l during accident condition loadmgs, the limiting event with respect to primary-to-secondary leakage is a j postulated steam line break event. For 7/8 inch tubing, the data supports no leakage up to 2.8 volts and a low probability ofleakage between 2.8 and 6.0 volts. The threshold of significant leakage (20.3 liter / hour 4

or 10 gpm) in a 7/8 inch tube diameter is about 6 volts.

Additional Considerations I The proposed amendment would preclude occupational radiation exposuie that would otherwise be incurred by plant workers involved in tube plugging or repair operations. 'Ibc proposed amendment would mmunize the loss of margm in the reactor coolant flow through the steam generator by keeping structurally sound tubes in service and not unnecessarily plugging or sleevmg them and, therefore, assist in d-aanratmg that muumum flow rates are maintained in excess of that required for operation at full power. Reduction in the amount of tube pluggmg and sleeving can reduce the length of plant outages and reduce the time that the steam generator is open to the containment environment during an outage.

ANALYSIS In accordance with the three factor test of 10 CFR 50.92(c), implementation of the proposed license amendment is analyzed using the following standards and found not to: 1) involve a significant increase in the probability or consequences for an accident previously evaluated; or 2) cream the possibility of a new or different kind of accident from any accident previously evaluated; or 3) involve a significant reduction in a margin of safety.

Conformance of the proposed aws.dwc.t to the standards for a determination of no significant hazard as defmed in 10 CFR 50.92 (three factor test) is shown in the following:

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• Significant Hazards Evaluaten Page 3 Voltage-Based Repair Criteria

1) Operation of Farley units in accordance with the proposed license amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Testmg of model boiler specimens for free standmg tubes at room temperature conditions shows burst pressures as high as approxunately 5000 psi for indications of outer diameter stress corrosion crackmg with voltage measurements as high as 26.5 volts. Burst testmg performed on pulled tubes, including tubes pulled from Farley Unit 2, with up to 7.5 volt indications show burst pressures in excess of 5300 psi at room temperature. As stated earlier, tube burst criteria are inherently satisfied during normal operating conditions by the presence of the tube support plate.

Furthermore, correctmg for the effects of temperature on material properties and minimum strength  ;

levels (as the burst testing was done at room temperature), tube burst capability significantly exceeds the R.G.1.121 criterion requiring the maintenance of a margin of 1.43 times the steam line break pressure differential on tube burst if through-wall cracks are present without regard to the presence of the tube support plate. Considering the existing data base, this criterion is satisfied with bobbin coil indications with signal amplitudes over twice the 2.0 volt voltage-based repair criteria, regardless of the indb=wl depth measurement. This structural limit is based on a lower 95% confidence level limit of the data at operatmg temperatures. The 2.0 volt criterion provides a conservative margin of safety to the structural limit considering expected growth rates of outside diameter stress corrosion crackmg at Farley, Alternate crack morphologies can correspond to a voltage so that a unique crack length is not defined by a burst pressure to voltage correlation. i However, relative to expected leakage during normal operatmg conditions, no field leakage has been reported from tubes with indications with a voltage level of under 7.7 volts for a 3/4 inch tube 4 with a 10 volt correlation to 7/8 inch tubing (as compared to the 2.0 volt proposed voltage-based tube repair limit). Thus, the proposed amendment does not involve a significant increase in the probability or consequences of an accident.

Relative to the expected leakage during accident condition loadmgs, the accidents that are affected by primary-to-mnadary leakage and steam release to the environment are Loss of External ,

Electrical Load and/or Turbine Trip, Loss of All AC Power to Station Auxiliaries, Major Secondary System Pipe Failure, Steam Generator Tube Rupture, Reactor Coolant Pump Locked Rotor, and Rupture of a Control Rod Drive Mechanism Housing. Of these, the Major Secondary System Pipe Failure is the most limiting for Farley in considering the potential for off-site doses.

The offsite dose analyses for the other events which model primary-to-mandary leakage and steam releases from the mandery side to the environment assume that the enad=ry side remains intact.

The steam generator tubes are not subjected to a sustamed increase in differential pressure, as is the case following a steam line break event. This increase in differential pressure is responsible for the postulated increase in leakage and associated offsite doses following a steam line break event.

In addition, the steam line break event results in a bypass of containment for steam generator leakage. Upon implementation of the voltage-based repair criteria, it must be verified that the expected distributions of cracking indications at the tube support plate intersections are such that primary-to-secondary leakage would result in site boundary dose within the current licensing basis.

Data indicate that a threshold voltage of 2.8 volts could result in through-wall cracks long enough to leak at steam line break conditions. Application of the proposed repair criteria requires that the current distribution of a number ofindications versus voltage be obtamed during the refueling outages. The current voltage is then combined with the rate of change in voltage measurement and a voltage measurement uncertainty to establish an end of cycle voltage distribution and, thus, leak rate during steam line break pressure differential. The leak rate during a steam line break is further increased by a factor related to the probability of detection of the flaws. Ifit is found that the i

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• Significant Hazards Evaluation Page 4

, Voltage-Based Repair Criteria potential steam line break leakage for degraded intersections planned to be left in service coupled j with the reduced allowable specific activity levels result in radiological consequences outside the current licensing basis, then additional tubes will be plugged or repaired to reduce steam line break leakage potential to within the acceptance limit. Hus, the consequences of the most limiting 4

design basis accident are constramed to present licensing basis limits.

l 2) The proposed license amendment does not create the possibility of a new or different kind of

! accident from any accklent previously evaluated.

J l Implementation of the proposed voltage-based tube support plate elevation steam generator tube )

repair criteria does not introduce any significant changes to the plant design basis. Use of the j criteria does not provide a mechanism that could result in an accident outside of the region of the tube support plate elevations Neither a single or multiple tube rupture event would be expected in a steam generator in wiuch the repair criteria have been applied during all plant conditions. He bobbin probe signal amplitude repair criteria are established such that operational leakage or j excessive leakage during a postulated steam line break condition is not anticipated. Southern I i

Nuclear has previously implemented a maximum leakage limit of 150 gpd per steam generator. i De R.G.1.121 criterion for establishing operational leakage limits that require plant shutdown are I based upon leak-before-break considerations to detect a free span crack before potential tube rupture. The 150 gpd limit provides for leakage detection and plant shutdown in the event of the occurrence of an unexpected single crack resulting in leakage that is associated with the longest permissible crack length. R.G.1.121 acceptance criteria for establishing operating leakage limits are based on leak-before-break considerations such that plant shutdown is initiated if the leakage associated with the longest permissible crack is exc-W The longest permissible crack is the

{ length that provides a factor of safety of 1,43 aganst burstmg at steam line break pressure j differential. A voltage amplitude of approximately 9 volts for typical outside diameter stress corrosion crackmg corresponds to meetag this tube burst requirement at the 95% prediction interval on the burst correlation. Alternate crack morphologies can correspond to a voltage so that a unique crack length is not defined by the burst pressure versus voltage correlation.

Consequently, a typical burst pressure versus through-wall crack length correlation is used below to define the " longest permissible crack" for evaluatmg operatmg leakage limits.

The single through-wall crack lengths that result in tubo burst at 1.43 times steam line break pressure differential and steam line break conditions are about 0.54 inch and 0.84 inch, respectively. Normal leakage for these crack lengths would range from about 0.4 gallons per minute to 4.5 gallons per minute, respectively, while lower 95% c-Manca level leak rates would l

. range from about 0.06 gallons per minute to 0.6 gallons per minute, respectively.

An operatmg leak rate of 150 gpd per steam generator has been implemented. This leakage limit 4 provides for detection of 0.4 inch long cracks at nonunal leak rates and 0.6 inch long cracks at the lower 95% ennMance level leak rates. Thus, the 150 gpd limit provides for plant shutdown prior to reaching critical crack lengths for steam line break conditions at leak rates less than a lower 95% confidence level and for three times normal operatmg pressure differential at less than 1

nonunal leak rates.

Considering the above, the implementation of voltage-based plugging criteria will not create the possibility of a new or different kind of accident from any previously evaluated.  ;

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,, Significant Hazards Evduation Page 5

, Voltage-Based Repak Criteria

3) The proposed license amendment does not involve a significant reduction in margin of safety.

The use of the voltage-based tube support plate elevation repair criteria is demonstratal to

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maintain steam generator tube integrity commensurate with the requirements of Generic Letter 95-

{

05 and R.G.1.121. R.G.1.121 describes a method acceptable to the NRC staff for meeting GDC 'j' 2,14,15,31, and 32 by reducing the probability of the consequences of steam generator tube rupture. His is accomplished by determining the limiting conditions of degradation of steam

]

generator tubing, as established by inservice inspection, for which tubes with unacceptable cracking should be removed from service. Upon impleraenw+iori of the criteria, even under the worst case conditions, the occurrence of outside diameter stress cuirosion cracking at the tube support plate elevations is not expected to lead to a steam generator tube rupture event during normal or faulted plant conditions. He most limiting effect would be a possible increase in leakage during a steam line break event. Excessive leakage during a steam line break event, however, is precluded by verifying that, once the criteria are applied, the expected end of cycle distribution of crack indications at the tube support plate clevations would result in minimal, and acceptable primary to =~-lary leakage during the event and, hence, help to demonstrate l radiological conditions are less than an appropriate fraction of the 10 CFR 100 guideline.

The margin to burst for the tubes using the voltage-based repair criteria is comparable to that currently provided by existing Technical Specifications. I In addressing the combined effects of LOCA + SSE on the steam generator component (as required by GDC 2), it has been determmed that tube collapse may occur in the steam generators at some plants. This is the case as the tube support plates may become deformed as a result oflateral loads at the wedge supports at the periphery of the plate due to either the LOCA rarefaction wave and/or SSE loadings. Den, the resulting pressure differential on the deformed tubes may cause some of the tubes to collapse.

)

There are two issues associated with steam generator tube collapse. First, the collapse of steam generator tubing reduces the RCS flow area through the tubes. The reduction in flow area increases the resistance to flow of steam from the core during a LOCA which, in tum, may potentially increase Peak Clad Temperature (PCT). Second, there is a potential the partial through-wall cracks in tubes could progress to through-wall cracks during tube deformation or collapse or that short through-wall indications would leak at significantly higher leak rates than included in the leak rate assessments.

1 Consequently, a detailed leak-before-break analysis was performed and it was concluded that the leak-before-break methodology (as permitted by GDC 4) is applicable to the Farley reactor coolant system primary loops and, thus, the probability of breaks in the primary loop piping is sufficiently low that they need not be considered in the structural design basis of the plant. Excluding breaks in the RCS primary loops, the LOCA loads from the large branch line breaks were analyzed at Farley and were found to be ofinsufficient magnitude to result in steam generator tube collapse or significant deformation.

1 Regardless of whether or not leak-before-break is applied to the primary loop piping at Farley, any flow area reduction is expected to be minimal (much less than 1%) and PCT margin is available to account for this potential effect. Based on analyses' results, no tubes near wedge locations are expected to collapse or deform to the degree that secondary to primary in-leakage would be l

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Voltage-Based Repair Criteria

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increased over current expected levels. For all other steam generator tubes, the possibility of secondary-to-primary leakage in the event of a LOCA + SSE event is not significant. In actuality, the amount of secondary-to-primary leakage in the event of a LOCA + SSE is expected to be less than that originally allowed, i.e.,500 gpd per steam generator. Furthermore, anmaary-to-primary in-leakage would be less than primary-to-secondary leakage for the same pressure differential since the cracks would tend to tighten under a secondary-to-primary pressure differential. Also, the presence of the tube support plate is expected to reduce the amount ofin-leakage.

Addressing the R.G.1.83 consideratians, implementation of the tube repair criteria is supplemented by 100% inspection requirements at C tube support plate elevations having outside diameter v.ress corrosion crackmg indications, reduced operatmg leakage limits, eddy current inspection guidelines to provide consistency in voltage normalization, and rotatmg probe inspection requirements for the larger indications left in service to characterize the principle degradation E

mechamsm as outside diameter stress corrosion crackmg As noted previously, implementation of the tube support plate elevation repair criteria will decrease the number of tubes that must be taken out of service with tube plugs or repaired. 'Ihe installation of steam generator tube plugs or tube sleeves would reduce the RCS flow margin, thus implementation of the voltage-based repair criteria will maintain the margin of flow that would otherwise be reduced through increased tube pluggmg or sleeving.

Considering the above, it is concluded that the proposed change does not result in a significant reduction in margin with respect to plant safety as defined in the Final Safety Analysis Report or any bases of the plant Technical Specifications CONCLUSION Based on the precedung analysis, it is concluded that using the voltage-based steam generator tube repair criteria in accordance with Generic Letter 95-05, " Voltage-Based Repair Criteria for Westmghouse Steam Generator Tubes Affected by Outside Diart:ter Stress Corrosion Crackmg," for removing tubes from service or repairing tubes at Farley is acceptable and the proposed license amendment does not involve a Significant Hazards Consideration as defmed in 10 CFR 50.92.

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