ML20154C005

Proposed Tech Specs Reflecting Replacement SG Tube Insp Surveillance Differences & Deleting Provision Allowing Use of F* Alternate Repair Criteria

ML20154C005
Person / Time
Site:	South Texas
Issue date:	09/30/1998
From:	HOUSTON LIGHTING & POWER CO.
To:
Shared Package
ML20154C003	List: ML20154C005 NOC-AE-0162, Application for Amends to Licenses NPF-76 & NPF-80, Reflecting Replacement SG Tube Insp Surveillance Differences & Deleting Provision Allowing Use of F* Alternate Repair Criteria
References
NUDOCS 9810060111
Download: ML20154C005 (17)

Text

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NOC-AE-0162 Attachment 4 Page 1 of16 1

1 ATTACHMENT 4 1

1 TECHNICAL SPECIFICATION MARKED-UP PAGES  !

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l 9810060111 980930 PDR ADOCK 05000498 P PM ,

NOC-AE-0162 Attnchm3nt 4 Page 2 of16 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS The below listed Technical Specification pages are marked-up to identify the proposed changes associated with this submittal and follow this page. Listed pages numbers with "No changes" following the page number indicate that those pages contain no proposed changes, but are i included for continuity of review. In addition, one markup on Page 3/4 4-13, one markup on Page 3/4 4-15, two markups noted on Page 3/4 4-16, two markups noted on Page 3/4 4-16a, one markup on Page 3/4 4-16b, one markup on Page B 3/4 4-3, and two markups on Page B 3/4 4-4 are not part of this submittal's proposed changes, but are included for clarity since they are part of proposed changes from Reference 1 (See Attachment 2, Page 9). These ten Reference 1 markups are identified by adjacent asterisks (*).

P_ ages 3/4 4-12 3/4 4-13 3/4 4-13a (No changes) 3/4 4-14 3/4 4-15 3/4 4-16 3/4 4-16a 3/4 4-16b 3/4 4-17 (No changes) 3/4 4-18 (No changes) 3/4 4-18a B 3/4 4-2a B 3/4 4-3 B 3/4 4-3a

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B 3/4 44 l

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REACTOR COOLANT $YSTEM 3/4.4.5 STEAM GENERATOR LIMITING _ CONDITION FOR OPERATION N n l

l 3.4.5 Each steam generator shall be OPERABLE. /

I l

APPLICABILITY: MODES 1, 2, 3, and 4. (f0R. N o e[ E Sfeerw gerJe.rafors o#h ACTION:

With one or more steam generators inoperable, restore the inoperable generator (s) to CPERABLE status prior to increasing T above 200'F.

SURVEILLANCE REOUTREMENTS - 1 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by perform nce of the following augmented inservice inspection program and the requirem ts of Specification 4.0.5.

4.4.5.1 Steam Generator Samole Selecti_on and Inspection - Each s am generator shall be determined GPERABLE during shutdown by select g and inspecting at least the minimum number of steam generators spe ' ied in Table 4.4-1. '

4.4.5.2 Steam Generator Tube SLmole Selection and Insoect n - The steam  ;

generator tube minimum sample size, inspection result cl sification, and the corresponding action required shall be as specified in able 4.4-2 and Table 4.4-3. The inservice inspection of steam gener or tubes shall be performed at the frequencies specified in Specifica on 4.4.5.3 and the  !

inspected tubes shall be verified acceptable per t acceptance criteria of Specification 4.4.5.4. When applying the excepti s of 4.4.5.2.a through 4.4.5.2.c, previous defects or imperfections in he area repaired by sleeving are not considered an area requiring reinspect n. The tubes selected for each inservice inspection shall include at le t 3% of the total number of nonrepaired tubes in all steam generators anda 0% of the total number of repaired tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:

a. Where experience in similar plants with similar water chemistry indicates critical areas-to be inspected, then at least 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 (greaterthan20%),

2) Tubes in those areas where experience has indicated potential problems, and SOUTH TEXAS - UNITS 1 & 2 3/4 4-12 Unit 1 - Amendment No.90 Unit 2 - Amendment No.77

REACTOR COOLANT SYSTEM

/

STEAM GENERATORS For Mode / E sbrngenemws ou p ,

SURVEILLAN REOUIREMENTS (Continued) h

3) A tube inspection (pursuant to S)ecification 4.4. 4a shall  !

be performed on each selected tu>e. If any select d does l not permit the passage of the eddy current probe ube i inspection, this shall be recorded and an adjacent tube shall i be selected and subjected to' a tube inspection. I

4) hndications 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 or Table 4.4-3) 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, and

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

d. for Unit I, ;ny tak ell =:d t r a;in in : rvi:: per A ::pt;nce Criterien !! (cf T;;hnical Spuf fic:tien 4.4.5.4 ) :h:11 bc l in:p::ted vi: the ret; ting p;ne;h ::ii (TO) addy current aethod

-over the f' distence. Such tubes ere eu +t frea addy current in:pecti:n =:r th: p:rti:n f the t be b:l= the P di:t;nc; which it net structurally rel e rt.

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h. (:n r - plementation of the steam. generator tube / tube support

'.'a i t p aQrYepair 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 (00 SCC) 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.

Cateoory Insoection Results C-1 Less than Sk of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

SOUTH TEXAS - UNITS 1 & 2 3/4 4-13 Unit 1 - Amendment No. 82r83,90 Unit 2 - Amendment No. 77

l .

REACTOR COOLANT SYSTEM A/O C bA MCE3

. STEAM GENERATORS l

SURVEILLANCE RE0VIREMENTS (Continued)

C-2 One or more tubes, but not more than 1% of the total i tubes inspected are defective, or between 5% and 10%

of the total tubes inspected are degraded tubes.

l 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 must. exhibit significant. (greater than 10%) further wall penetrations to be included in the above percentage calculations.

I l SOUTH TEXAS - UNITS 1 & 2 3/4 4-13a Unit 1 - Amendment No. 83 l-i

l Q))pg 0 6 t" 0t TCh

• REACTOR COOLANT SYSTEM gyg $4derCor r~eglAcetheN f STEAM GENERATORS L $*N*#8 l yg p)ACthef l _ SURVEILLANCE REOUJREMENTS (Continued) 4.4.5.3 Inspection Frequencies - The abov required inservice inwpections of steam generator tubes shall be performed the following frequencies:

a. The first inservice inspection,shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticalitg.

Subsequent 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, not including the preser-vice 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 Table 4.4-2 at 40-month 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 4.4.5.3a.; the interval may then be extended to a maximum of once per 40 months; and

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

1) Primary-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2, or

2) A seismic occurrence greater than the Operating Basis Earthquake, or

3) A loss-of-coolant accident requiring actuation of the Engineered Safety Features, or

4) A main steam line or feedwater line break.

Note: T NSEMlcl b)Spechos Is Nob re pirec.

dwing %e skr> gese<& redace nec O WF Ag e .

SOUTH TEXAS - UNITS 1 & 2 3/4 4-14

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

REACTOR COOLANT SYSTEM l

l STEAM GENERATORS SURVEILLANCE REOUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria ([or Mojel E dgm

a. As used in this specification: pedrS _

oh

1) Tubina or Tube means that ortion of the tube or sleeve which I forms the primary system o secondary system pressure boundary; l

2) Imperfection means an e eption to the dimensions, finish, or contour of a tube from at required by fabrication drawings or l s)ecifications. Eddy- rrent testing indications below 20% of tie nominal tube wall hickness, if detectable, may be considered as imperfe tions;

3) Deoradation means a ervice'inducedcracking, wastage, wear,orl general corrosion o curring on either inside or outside of a tube;

4) Dearaded Tube sea s a tube containing imperfections greater l than or equal to 0% of the nominal wall thickness caused by degradation;

5) X_Dearadation ans the percentage of the tube wall thickness j affected or re oved by degradation;

6) Defect means n imperfection of such severity that it exceeds the plugging 3r repair limit. A tube containing a defect is defective;

7) Pluaaina Lin. t or Repair limit means the imp ^erfection depth at or beyond which the tube shal1 be removed from service by pluggingor[repairedbysleevingintheaffectedareabecause it may become unserviceable prior to the'next inspection. The fo6. inode g plugging or repair limit imperfection depths are specified in fgj percentage of the nominal wall thickness as follows:

qBd* a. ' original tube wall - 40%

d b. West house laser welded sleeve wall 40%

'".,'tAIsdefinitiondoesnotapfsiy4otubesupport p ate in er ctionsforwhichthevopage-b ed repair criteria are being applied. Refer to 4.4.5. . atte f r the repair limit applicable to these intersections, 11

8) Unserviceable describes the condition of a tube if it leaks or l 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-line or-feedwater line break as specified in Specification 4.4.5.3c., above;

9) Tube Insoection means an inspection of e st a generatortubel l from the )oint of entry (hot leg sid complete 1 ' around the U-bend to t1e top support of the cold 'eg; *nd-1 SOUTH TEXAS - UNITS 1 & 2 3/4 4-15 Unit 1 ment No. 83,90 Unit 2 - Amendment No. 77

d REACTOR COOLANT SYSTEM g /V)ohe, [ gg gey p) ggS op/ p STEAM GENERATORS SURVEILLANC E0VIREMENTS (Continued) i

10) Preservice Inspection means an inspection of the full length of l

, 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 arior to initial POWER OPERATION using the equipment and teciniques expected to be used during subsequent inservice inspections.

i

41) r* criteria fFer LSit ! er'"! Tebe degrad:tien bele'c: : l 2 3. 0$..$$.2U.

.. $.5[33233.%...$'U. .N. 5....!!2, ..$..Y_ E_5N.'$_$.

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fdcr:ti: t: th: ::::pttr. : criteri: :t:ted in thi: ::: tion (i.e., g;ing Of :::S tri;; i: r.;t required). The

th:d:j':;y fer d;t:r; ir.; tier, fer the T' distence es well u

. I the' list ;f tubes te t.ich the i' criterie is net eipliceble 13 l 3 de: rit:d in det3 il k J.7 helJ.egat z OL' ;0202 , Revision G.

7 F mill A wa r4 Le> g

,hr " nit h Tube Uuooort Plate Pluant limit is used for the 4

lj 4&)A disposition-of a vall yTOWelim generator tube for continued l service that'to riencing predominately axially oriented

. outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the plugging (repair) limit is based on maintaining steam generator tube serv.ceability as described below:

a) Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltage less than or equal to the lower voltage repair limit (Note 1),

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 voltagpG greater than the lower voltage repair limit (Note 1)/ wi 1 be repaired or plugged, except as noted in 4.4.5.4.p.tt.

below, l

il c) Steam generator tubes,.with indications of potentia 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 (Note 1) but less than or equal to the upper repair voltage limit (Note 2), may reinain in service if a rotating pancake coil inspection does not detect #

degradation. Steam generator tubes, with indications of outside diameter stress corrosion cracking degradation with bobbin voltage greater than the upper voltage repa.ir limit (Note 2) will be plugged or repaired.

SOUTH TEXAS - UNITS 1 & 2 3/4 4-16 Unit 1 - Amendment No. Bh83,90 Unit 2 - Amendment No. 77

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SEP. 4.1997 . 2: 24Pr1 ,

NO.668 P.15/36 l REACTOR COOLANT SYSTEM .

g- #

STEAM GENERATORS < "" ' 7 SVRVEILLANCE RE0VIREMENTS (Con ued) m ,. . a _ b _ __ _ _ _ u _ _ _ __ u__ m _; ,_ r___ i___

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d)/ If an unscheduled mid-cycle inspection is performed, the l

mid-cycle identifiednin repair limits app 4.4.5.4.a M . %y insteadM

, 4.4.5.4.a ef,,the

.b, limits and c- J 4.4.5.4.a. X c. The mid-cycle repair limits will be TO' b 8*llb determined from the equations for mid-cycle repair limits steg m{,,.g of NRC Generic letter 95-05, Attachment.2, page 3 of 7.

Implementation of these mid-cycle repair limits should Odly follow the same approach as in TS 4.4.5.4.a.4r.a, 4.4.5.4.a. X.b, and 4.4.5.4.a. W.c. Il 88 at ing .'m Note 2: The upper voltage repair limit (V is calculated according to the i methodology in Generic Letter 95-5)as supplemented. Vun may iffer at the TSPs and flow distribution baffle.

h Iube Reoair rehrs to a process that reestablishes tube serviceabilityY Acceptable tube repair will be oerformed in accordance wib the methods described in Westingiouse Reports WCAP-13698, Revision 2, " Laser Welded Sleeves for 3/4 Inch Diameter Tube Feedring-Type and Westinghouse Preheater Steam Generators," April 1995 and WCAP-14653, " Specific Application ,

of Laser Welded Sleeves for South Texas Project Power Plant  !

Steam Generators," June 1996, including post-weld stress relief; Tube repair includes the removal of, plugs that were previously installed as a corrective or preventive measure. A tube inspection per 4.4.5.4.a.9 is required prior to returning previously plugged tubes to service,

b. The steam generator shall be determined OPERABLE after completing the corresponding actions [ plug orVrepair all tubes exceeding the plugging or repair limit and all Jbes containing through-wall cracks required by Table 4.4-2 nd Table 4.4-3.

4.4.5.5 Reports $o Me el S e es tHerAYor$ e

a. Within 15 days following the completion of each inservice inspection ofsteamgeneratortubes,thenumberoftubespluggedorrepairedinj each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2; SOUTH TEXAS - UNITS 1 & 2 3/4 4-16a Unit 1 - Amendment No. 83,90 Unit 2 - Amendment No. 77

SEP. 4.1997 2:25PM NO.668 P.16/35 REACTOR COOLANT SYSTEM i STEAM GENERATORS SURVEJLLANCE REOUIREMENTS (Cont _inued) -

1

b. The complete results of the steam generator tube 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: I ok 1) Number and extent of tubes inspected, S

eM hob 2) location and percent of wall-thickness penetration for each indication of an imperfection, and

3) Identification of tubes plugged or repaired. l l

c. Results of steam generator tube inspections which fall into Category C-3 shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days and prior to resumption of 1 plant operation. This report shall provide a description of l investigations conducted to determine cause of the tube degradation 1 and corrective measures taken to prevent recurrence.

d.

For tubEMortbh implementation plate intersections,ofnotif thethe voltahe-based Staff prior to repair criteri!

returning the steam generators to service should any of the '

following conditions arise:

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

line break) for the next operating cycle. l

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

3) If indications are identified that extend beyond the confines l 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 l projected end-of-cycle (or if not practical, using the actgal measured end-of-cycle) voltage distribution exceeds 1 x 10' ,

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

1 SOUTH TEXAS - UNITS 1 & 2 3/4 4-16b Unit 1 - Amendment No. 81,90 Unit 2 - Amendment No. 77

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g TABt.E 4.4-.1 g  !

Z

[ MINIMUM NUMBER OF STEAM GENERATORS TO BE d- .

e. INSPECTED DURING INSERVfCE INSPECTION O i to U) r l;

P.u.1k.E inspection No Yes }

w .

1 No. of Steam Generators per Unit Two Three Four Two Tfwee Four

= i g First inservice Inspection , ,

All One Two Two Second & S.Lma..; inservice impections One' One' One2 Ord TABt.E NOTATIONS 1, The Inservice Impection may be flmited to one steam generator on a rotating schedule encompassing 3 N % of the tutes I (where N is the enember of steam generators in tfra plant) If the results of the first or previous inspections indicate that  :

sit steem generators are performing in a like menner. Note that under some d w....;saces, the operating conditions in '

one or more steem generators may be found to be more severe than those in other steam generators. Under such circum.

stances the semple sequence shall be modified to inspect the most severe conditions. .

2. The other steem generster not Impected during the first inservice lenpection shaft be Impected. The third and b,Lm_..;

impections shouhl follow the Imtructions described in 1 above. ,

3. Each of the other two steem generators not impected during the first inservice inspections shaft be Impetted during the second and third Impections. The fourth and subia..m.; impections shall fotfow the imtructions descrew in i sbove. i i

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Tr. bra 4.4-2 f STEAM GENERATOR TUBE INSPECTION p 1 i 1ST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION y Iji 2 '

Sample Size Result Action Required Result Action Required Result Action Required g  !

A minimum of C-1 None N.A. N.A. N.A. N.A.

S Tubes per S.G. C-2 Pfug or repair C.1 None N.A. N.A. l ]lI defective tubes and inspect additional 2S C-2 Pfuo or repair defective C-1 None g  :

tubes in this S.G* tubes and inspect ,  !

C-2 Pfug or repair defective '

additional 4S tubes in g this S.G. tubes i C-3 Perform action for C-3 result of first sample l I

C-3 Perform action for C-3 N.A. N.A.

result of first sampfe [

t

, C-3 Inspect all tubes In All other l this S.G., plug or

. S.G.s are None N.A. N.A.

repair defective C-1 l  ;

tubes and inspect 2S  ;

tubes in each other Some Perform action for C-2 '

N.A. N.A.  !

S.G. S.G.s C-2 result of second sample <

but no .

Notification ta NRC additional pursuant to 50.72 S.G. are j (bH2) of 10 CFR Part C-3 i 50 Additional Insnect all tubes in each S.G. is C-3 S.G. and plug or repair l  ;

defective tubes.  !

Notification to NRC N.A. N.A. h!

gj pursuant to 50.72 (bH2) of 10 CFR Part 50 m, where N is the number of steam generators in the unit, and n is the number of steam Generators inspected during

. S=3Et n an inspection. y{

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SOUTH TEXAS - UNITS 1 8. 2 3/4 4-18 Unit 1 - Amendment No.90 Unit 2 - Amendment No.77 t

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Table 4.4-3 s:

he STEAM GENERATOR REPAIRED TUBE INSPECTION 8' A

1ST SAMPLE INSE dCTION 2ND SAMPLE INSPECTION -

Sampfe Size Result Action Required Result Action Required 3i A minimum of 20% C-1 None N.A. N.A. .

of. repaired tubes "'

C-2 Pfug defective repaired C-1 None epa snh C-2 Plug defective repaired t

S.G. tubes C-3 Perform action for C-3 resuft of first sample i

C-3 Inspect all repaired tubes in All other S.G.s are None this S.G., plus defective C-1 repaired tubes and inspect 20% of the repaired tubes Some S.G.s C-2 Perform action for C-2 i

in each other S.G. but no additional result of first sample S.G. are C-3 Notification to NRC pursuant to 50.72 (bH2) of Additional S.G. is Inspect all repaired tubes in  ;

10 CFR Part 50 , C-3 each S.G. and plug defective repaired tubes.

Notification to NRC pursuant to 50.72 (b)(2} of 10 CFR Part 50 "3 Each repair method is considered a separate population for determination of scope expansion. g m,

8:

r TI o-SOUTH TEXAS - UNITS 1 & 2 3/4 4-18a Unit 1 - Amendment No. 90 $

Unit 2- Amendment No. 77 j t

SEP. 4.1997 2:26PM , NO.668 P.22/36 REACTORCOOLANT_SY$IIM BASES _

l RELIEF VALVES (Continued)

C. Manual control of the block valve to:

allow it to be used for manual control o(1) unblock f reactor an isolated coolant system pressurePORV to (Item A), and (2) isolate the PORV with excessive seat leakage (Item B).

D. Manual control allows a block valve to isolate a stuck-open PORV.

3/4,4,5 STEAM GENERATORS l 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' l.83, Revision 1. Inservice i inspection of steam generator tubing is essential in order to maintain I surveillance of the conditions of the tubes in the event that there is l i evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing also prevides a means of l characterizing the nature and cause o tuha danendn+ ins er t" t e ~ ecti measures can be taken.

( Q . y njej g g fg g p e m aogh)

The plant is expected to be o erate in manner MUITidihne secoridary

~

coolant will be maintained within ose chemistry limits found to minimize corrosion of the steam generator ub . If the secondary coolant chemistry is not maintained within these limi s, lo lized corrosion may likely result in stress corrosion cracking. The xtent cracking during plant operation would be limited by the 3.4.6.2 e limita 'on of steam generator tube leakage between the Reactor Coolant Sysuem and th Secondary Coolant System. Cracks having a primary-to-secondary 1 !akage less han this limit during operation will have an adequate margin of safety to w thstand the loads imposed during 8g y normal operation and by postula:ed accidents Operating plants have n E steam demonstrated that primary-to-sei ondary leaka dayi ege,ghs l per steam generator can readily e detected. geLeakage as low as 150 gallons in excess of thisper[limi l will require plant shutdown and unschedul the '

leaking tubes will be located and luggedor/!d inspection, repaired. during Defective whic(may be tubes repaired by a Westinghouse laser w ded sleeve. The technical bases fo,r

! sleeving repair are described in Wes inghouse Reports WCAP-13698, Revision 2,

" Laser Welded Sleeves for 3/4 Inch Di eter Tube Feedring-Type and l Westinghouse Preheater Steam Generator " April 1995 and WCAP-14653, " Specific i Application of Laser Welded Sleeves for uth Texas Pro et Powe,r_ Plant steam Generators," June 1996. .por Mode} E deaf 1

l m o ewAfors Wastage-type defects are unlikely th proper chemi ry treatment bf the s -)

i secondary coolant. However, even if a efect should deve 3 in service, it will be found during scheduled inservi e steam generator t ae examinations.

Except as discussed below, plugging or repair will be requd red for all tubes with imperfections exceeding the plugging or repair limit ilf 40% of the original tube nominal wall thickness. If a tube contains i Westinghouse laser welded sleeve with imperfection exceeding 40% of nominal ill thickness, it must be plugged. The basis for the sleeve plugging limit ,s based on Regulatory Guide 1.121 analysis, and is described in the Westinghouse sleeving technical reports listed above. Steam generator tube inspections of operating SOUTH TEXAS - UNITS 1 & 2 8 3/4 4-2a Unit 1 - Amendment No. 55,827e3,90 Unit 2 - Amendment No. 44,77

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_-.c . _ _ _ _ _ . ._ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ . _ _

l REACTOR COOLANT SYSTEM 3 BASES N:t STEAM GENERATORS (Continued) t0vi pignts have demonstrated the capability to reliably detect degradation that l L has penetrated 20% of the original tube wall thickness. Repaired tubes are ,

% also included in the inservice tube inspection program. - I E

o _xclusion of certain areas of Unit I tubes from consideration has I naly sing an F* criteria. The criteria allows n rvice.induc  !

%(degradation within the tubesheet to remain in service. analysis methodology det Qes the length of sound fully rolled nded tubing j f' required in the uppemog area within the tubeshee preserve needed 5 structural margins for at ice conditio'ns e remainder of the tube, y below the F* distance, is cons from consideration to the customar d not ing cturally criteria relevant of 40%

and is excluded throughwall. .'

i i

l

The amount of primar secondary leaka om tubes left in service by e application of the

• terion has been determine verification testing.

O This leakage h en considered in the calculation of p lated primary to E secondar age under accident conditions. Primary to seco leakage e duri ccident conditions is limited such that the associated ra ical o sequences as a result of this leakage is less than the 10 CFR 1001 w& "-i'. ' Wlhe voltage-based repair limits of SR 4.4.5 implement the guidance in 95-05 and are applicable only to Westinghouse-designed steam generators (SGs) with outside diameter stress corrosion cracking (0DSCC) located at the tybe-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 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 For model morpholoav-G sten ReNe"I&

Implementation of SR 4.4.5 requ res a &rivation 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 re air limit; V is determined from the structural voltage limit by app ying the fo1Nw,ing equation:

Vunt ."_ Vst . -Vn-Vuos c ._.

%V represent the allowance for flaw growth between inspections and Vuor

i. -On s the allowance for potential sources of error in the measurement oT the bobbin coil voltage. Further discussion of the assumptions necessary to determine the voltage repair limit are discussed in GL 95-05.

SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-3 Unit 1 - Amendment No. 82T83,90 Unit 2 - Amendment No. 77

REACTOR COOLANT SYSTEM g g N

4 BASES STEAM GENERATORS (Conti ued)

The mid-cycle eq tion in SR 4.4.5.4. a. e hould only be used during l unplanned inspections n which eddy curren dat is acquirad for_ indica + ions

at the tube support p ates. on Model E sb D  !

^

x J eAieraro rs SR 4.4.5.5 implements several reporting requirement recommendedhyCLN

. 95-05 for situations which the NRC wants to be notified p ior'to returning the

SGs to service. For the purpose of this re' porting 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 i probability were calculated using the E0C 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.

! Whenever the results of any steam generator tubing inservice inspection i fall into Category C-3, these results will be promptly reported to the

, Commission in a Special Report pursuant to Specification 6.9.2 within 30 days. 1

. and prior to resumption of plant operation. Such cases will be considered by 2 the Commission on a case-by-case basis and may result in a requirement for

analysis, laboratory examinations, tests, additional eddy-current inspection, l

and revision of the Technical Specifications, if necessary.

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION _ SYSTEMS

! The RCS Leakage Detection Systems required by this specification are

, provided to monitor and detect leakage from the reactor coolant pressure boundary. These Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.

3/4.4.6.2 OPERATIONAL LEAKAGE PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.

Industry experience has shown that while a limited amount .of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than I gpm. This threshold vale is sufficiently low to ensure early detection of additional leakage.  ;

SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-3a Unit 1 - Amendment No. 83,90 Unit 2 - Amendment No.77

e l REACTOR COOLANT SYSTEM fEWCe [# WM C Applicdr'on of vohge-based Repaia caifer,n BASES ^ad Issen-welded sleeving to Model E sk OPERATIONAL LFAKAAF (Continued) 8 eNerators. [*** The] -- " '

$ fe,. U..R 4 leaka .6 are more restrtetive tL. t- stand!erdlimits incor!

operating orated 1 akage into limits andSR 4 intended to i provide an additional margin to accommodate a crack which ight grow at a greater than expected rate or unexpectedly extend outside the thickness of the tube support plate. Hence, the reduced leakage limit, when combined with an effective leak rate monitoring program, provides additional assurance that should a significant leak' be experienced in service, it will be detected, and the plant shut down in_a timely manner.

M r tit; ; =f steam generator tube leakage limit of 150 gpd fo each steam generator nit Tsolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of 10 CFR Part 100 dose guideline valves in the event of either a steam generator 1

tube rupture or steam line break. The 150 gpd limit per steam generator is f conservative compared to the assumptions used in the analysis of these accidents. The 150 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.

The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems.

The specified allowed leakage from any RCS pressure isolation valve is sufficiently low to ensure early detection of possible in-series check valve failure. It is apparent that when pressure isolation is provided by two in-series check valves and when failure of one valve in the pair can go  ;

undetected for a substantial length of'tims, verification of valve integrity l is required. Since these valves are imporr. ant in preventing overpressurization and rupture of the ECCS low pressure pi)ing which could result in a LOCA that bypasses containment, these valves siould be tested periodically to ensure low probability of gross failure, j The Surveillance Requirements for RCS pressure isolation valves provide I added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS pressure .

isolation valve is IDENTIFIED LEAKAGE and will be considered as a portion of .

l the allowed limit'.  !

i 3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining The 8]diNounf mer pcovi L #e h>cacec Ienhage limib w il b e reMaj w; g +k , m y s}e w g enera k s.

SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-4 Unit 1 - Amendment No. 83,90 Unit 2 - Amendment No.77

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