ML20137B969

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Proposed Tech Specs 3/4.4.5 & 3.4.6.2 Re SG Tube Repair Using Laser Welded Sleeves
ML20137B969
Person / Time
Site: South Texas  STP Nuclear Operating Company icon.png
Issue date: 03/17/1997
From:
HOUSTON LIGHTING & POWER CO.
To:
Shared Package
ML20137B951 List:
References
NUDOCS 9703240113
Download: ML20137B969 (20)
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Text

{{#Wiki_filter:' 3 ATTACHMENT 2 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS 3/4.4.5 AND 3.4.6.2 Revised Pages Attached 3/4 4-12 3/4 4-16a 3/4 4-20 3/4 4-13 3/4 4-16b B 3/4 4-2a 3/4 4-13a* 3/4 4-17* B 3/4 4-3* 3/4 4-14* 3/4 4-18 B 3/4 4-3a 3/4 4-15 3/4 4-18a B 3/4 4-4 3/4 4-16 3/4 4-19* B 3/4 4-5 

                                             *No changes on this page, provided for continuity.

The attached pages replace all of the revised pages submitted on May 17,1996 (ST-HL-AE-5362). Amendments that have been incorporated on the pages previously submitted have necessitated resubmission of all of the pages. Additions are indicated by redlining and deletions by strikeout. IER*A855R3E8S$?98 P PDR E\WFRWRC-WK\TSC 9?JS50 mp $TI 30158506 

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4' REACTOR ' COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. l ACTION: With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing t,y above 200*F. SURVEILLANCE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance j of the following augmented inservice inspection program and the requirements , of Specification 4.0.5. i 4.4.5.1 Steam Generator Samole Selection and Insoection - Each steam generator shPfal be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1. 1 4.4.5.2 Steam Generator Tube Samole Selection and Insoection - The steam generator tube minimum sample size, inspection result classification, and the correspondin action required shall be as specified in Table 4.4-2 and Tab 1si4 W 3%g$The inservice inspection of steam generator tubes shall"be beff6rsed*aB*bhe frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of 

                                                                       ^

S ecification 4.4.5.4. WhsrRs#1pihsithWexcept1bssNf?4047572Ts?through 

                   ' p% 572WMprevlodsidefsctsf or41mperf actfionsiinsthetarssyrepaired!bpTsleeving 4

Wreinodeonsideredi Anisreairequiringkreinspestiord"The*6ubus~sslebted"for ~ Eai:n'isssevibe~inspectibrf*shal1~Tnclude~at ^1enstl"34 of the total number of nonrepaired tubes in all steam. generators Es#720f[ogthaltiotagnumberlo( yep,airedjd@gsddis11?[ steam; generators; th's EGbes selected for tihese inspectior.s shill be selected on a random basis except:

a. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas;
b. The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:
1) All nonplugged tubes that previously had detectable wall penetrations (greater than 20%),
2) Tubes in those areas where experience has indicated potential problems, and SOUTH TEXAS - UNITS 1 & 2 3/4 4-12
                                          .- ..       . _ ~ .      __ - -          -    -.   .  - - . _

l REACTOR COOLANT SYSTEM 

,          STEAM GENERATORS J

SURVEILLANCE REQUIREMENTS (Continued)

3) A tube inspection (pursuant to Specification 4.4.5.47a.e))

shall be performed on each selected tube. If any selectsd tube does not permit the passage of the eddy current probe 4 i for a tube inspection, this shall be recorded and an - adjacent tube shall be selected and subjected to a tube inspection.

4) Indications left in service as a result of application of ]

1 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 orETsbis during each inservice inspection may be subjected ~EB"a"p!474?3)srEl's1 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. i
d. For Unit 1, any tube allowed to remain in service per Acceptance criterion ts114(of Technical Specification 4.4.5.4a) shall be inspected vil"lhe rotating pancake coil (RPC) eddy" current method over the F* distance. Such tubes are exempt from eddy current 1

inspection over the portion of the tube below the F* distance which is not structurally relevant, j

e. For Unit 1, 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 tue 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 i of tubes inspected over their full length.  ! The results of each sample inspection shall be classified into one of the 4 following three categories: ' Cateaorv IneDection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective. SOUTH TEXAS - UNITS 1 f 2 3/4 4-13 

     , a REACTOR' COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued)

C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes. C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective. Note: In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations. M SOUTH TEXAS - UNITS 1 & 2 3/4 4-13a l 

     .,6

... n 

 .                                                                                       1 REACTOR' COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.3   InsDection Frecuencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
a. The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality Subsequent inservice inspections shall be 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 preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months;
b. If the results of the inservice inspection of a steam generator conducted in accordance with 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 I
4) A main steam line or feedwater line break.

t i l i SOUTH TEXAS - UNITS 1 & 2 3/4 4-14 l l 1 i

REACTOR' COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Accertance Criteria

a. As used in this specification:

1)}' 

                           ~ w@^ hich s forms;thelprimans)yste,mj tof,s.econdaryjsystemjpre
                      ~

bo@daryg 2)" Imoerfection means an exception to the dimensions, finish, or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections; 3) 

                      "      Deoradation means a service-induced cracking, wastage, wear, or general corrosion occurring on either inside or outside of a tube; f)     Deoraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation; 5)
                      ^~
                             % Decradation means the percentage of the tube wall thickness affected or removed by degradation; 6)
  • Defect means an imperfection of such severity that it exceeds the plugging A tube containing a defect is defective; ;otirepsif3l limit.

l 7) 

                      ~      Plucoino Limit F6r7Reesir' Limit means the imperfection depth    !

at or beyond which the tube shall be removed from service i and ic cgual tc 005 cf the ncmincl tube .cIl thichnccc. by 7 pl6ggingyorfrepairedibp2sleevihgiinitheV'affectedfarea "' , because31timaysbecomeiunserviceable prioratosthe?next inspectionb.7Thelpluggingsor?repairslimi.tsimperfection depthsfarelspecifiedgnipercentagelofhthejnominalswalf thicknessq.asifpilowsg sP% rig'inslWsbeTwall? - 1.. .f . . . 140% bQj(Westinghousei laser @eldedssleeveNall? ,540% For Unit 1, this definition does not apply to the tube support plate intersections for which voltage-based repair criteria are being applied. Refer to 4.4.5.4.a.4+12 for the repair limit applicable to these intersections.

8) Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in Specification 4.4.5.3c., above; 9)
                      ~

Tube Insoection 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; and SOUTH TEXAS - UNIT 1 &2 3/4 4-15

.o REACTOR COOLANT SYSTEM i STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) 10) 

                     ~~~   Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed prior to initial POWER OPERATION using the equipment and             l 1

techniques expected to be used during subsequent inservice i inspections. 11) 

                      ~~

F* criteria IFor Unit 1 on1vl Tube degradation below a specified distance from the hard roll contact point at or i near the top of tubesheet (the F* distance) can be excluded i from consideration to the acceptance criteria stated in this section (i.e., plugging of such tubes is not required) . The methodology for determination of the F* distance as well as the list of tubes to which the F* criteria is not applicable is described in detail in Topical Report - BAW 10203P, Revision O. 12) 

                     ~~

For Unit 1, Tube Support Plate Plugging 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 thickness of the tube support plates. At tube support plate intersections, the plugging (repair) limit is i based on maintaining steam generator tube serviceability as described below: a) Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within 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 voltage greater than the lower voltage repair limit (Note 1), will be repaired or plugged, except as noted 4 . 4 . 5. 4 . a .M-12. c below. c) Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube 4 support with a bobbin voltage greater than the lower voltage limit (Note 1) but less than or equal to the upper repair voltage limit (Note 2), may remain inservice if a rotating pancake coil inspection does not detect degradation. Steam generator tubes, with indications outside diameter stress corrosion cracking degradation with bobbin voltage greater than the upper voltage repair limit (Note 2) will be plugged or repaired. SOUTH TEXAS - UNITS 1 & 2 3/4 4-16 

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4 

               ,    (

g + t REACTOR

  • COOLANT SYSTEM
f. t STEA)( GENERATORS 4- SURVEILLANCE REQUIREMENTS (Continued) t
d) Certain intersections as identified in Framatome
'                                                                                         Technologies, Inc. Topical Report BAW-10240P, " South                                                   <

Texas Project Tube Repair Criteria for ODSCC At Tube Support Plates" will be excluded from application of , the voltage-base repair criteria as it is determined  ; that these intersections may collapse or deform  ; {r following a postulated LOCA + SSE event.  ! f l' e) If an unscheduled mid-cycle inspection is performed, i i the mid-cycle repair limits apply instead of the i limits identified in 4.4.5.4.a. R12.a,  ! i- 4 . 4 . 5. 4 . a . M12. b, and 4 . 4 . 5. 4 . a . -1412. c . The mid-cycle i ! repair limitiiI'will be determined fE6m the equations  ! l for mid-cycle repair limits of NRC Generic Letter ' L 95-05, Attachment 2, page 3 of 7. Implementation of j these mid-cycle repair limits should follow the same

approach as in TS 4.4.5.4.a. W 12.a, ~~

4.4.5.4.a. M12.b, ~ { and 4.4.5.4.a M Q.c. l Note 1: The lower voltage repair limit is 1.0 volt for 3/4-inch diameter tubing or 2.0 volts for 7/8-inch diameter tubing. i Note 2: The upper voltage repair limit (V ,u.) o is calculated according to

the methodology in Generic Letter _95-05 as supplemented. Vuiu. may t

r differ at the TSPs and flow distribution baffle. d 1 

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{ ~serviceabilitpijC;;Accohtabl41tubiisM84WillibeJ6rformediis kccordancefwitM theinethods Assesifed M iuestF N F ~- - " Rep 3/4 %InchiaiametsslTube'Psodsorts ^ j fand3Westinghouss ucARF135983Reirisidial2 "~ WIaseib?We ! PreheatsiMateestGineratovelff si9951addjjWCAPs146537 l l PSpec;ific?Applidatine?6fi%aserdeeldsdis14eynopfcW8outh Texasifreject1PowerXPlanMSteamiGeneratorsM9une219967 j subjecgtojphejllmi}st.ionslagfjfes@{cpionslasgoteggypM 8!n c u e t f a a j TubeffMEIldsludeiii!EEE!fdisoWI$Edf7p1UiiWRhhtWeid

revious1 j p$

A tiaboli!*Winstalled U tion issiaEcorrectivelorsprevehti+d?insaeGei^NJ EN74I5741laT9 Eishaqdiradipsiorito' " '"~ ~ M urn @ fkreyi g g j{$piep ) M [Q {sep icej ~ ~ ~ } i

b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug 6Hrep~sTfiall tubes exceeding the i j

plugging cracks) required pr}tepaiR11mit 

                                                                                                 ~by Table 4.4-2        and--#.-.-

all[ind3Tablsjy ~ N V3.E!ibes e6Etainin , 4.4.5.5 Reoorts 1 i a. Within 15 days following the completion of each inservice , inspection of steam generator tubes, the number of tubes plugged bQi(AiilisdRin each steam generator shall be reported to the j Commis5To^ii"In a Special Report pursuant to Specification 6.9.2; i 1 i i SOUTH TEXAS - UNIT 1 & 2 3/4 4-16a i , 4 i d

s

  • o REACTOR' COOLANT SYSTEM EIEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued)
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:
1) Number and extent of tubes inspected,
2) Location and percent of wall-thickness penetration for each ,

indication of an imperfection, and i

3) Identificationoftubespluggedbj{rspyired.
c. Results of steam generator tube inspections which fall into l 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 plant operation. This report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
d. For Unit 1, implementation of the voltage-based criteria to tube support plate intersections, notify the Staff prior to returning the steam generators to service should any of the following conditions arise:

17 

                        "     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.

2T" 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. 57 

                        ~

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-a, notify the NRC and provide an assessment of the safety of the occurrence. SOUTH TEXAS - UNITS 1 & 2 3/4 4-16b

TABLE 4.4-1 , NINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION Preservice Inspection No Yes No. of Steam Generators per Unit Two Three Four Two Three Four First Inservice Inspection All One Two Two Second & Subsequent Inservice Inspections one' One' One* One* TABLE NOTATIONS

1. The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 3 N % of the tubes (where N is the number of steam generators in the plant) if the results of the first or previous inspections indicate that all steam generators are performing in a like manner.

Note that under some circumstances, the operating conditions in one or more steam generators may be found to be more severe than those in other steam generators. Under such circumstances the sample sequence shall be modified to inspect the most severe conditions.

2. The other steam generator not inspected during the first inservice inspection shall be inspected.

The third and subsequent inspections should follow the instructions described in 1.above.

3. Each of the other two steam generators not inspected during the first inservice inspections shall be inspected during the second and third inspections. The fourth and subsequent inspections shall follow the instructions described in 1 above.

SOUTH TEXAS - UNITS 1 & 2 3/4 4-17

O 

                                                                                                                                                              . s Table 4.4-2                                                          .

STEAM GENERATOR TUBE INSPECTION 

                                                                                                                                                        ~

1ST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Size Result Action Required Result . Action Required Result Action Required A minimum of C-1 None N.A. N.A. N.A. N.A. S Tubes per S.G. C-2 Plug IIr[rijisilt C-1 None N.A. N.A. defective tubes and inspect additional 2S C-2 Plug ji]t M defective C-1 None tubes in this S.G. t a 9n S bes in C-2 Plug [M defective this S.G. tubes C-3 Perform action for C-3 result of first sample C-3 Perform action for C-3 N.A. N.A. result of first sample C-3 Inspect all tubes in All other this S.G., plug y S.G.s are None N.A. N.A. M defective C-1 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 to NRC additional pursuant to 50.72 S.G. are (b)(2) of 10 CFR Part C-3 50 Additional Inspect all tubes in each S.G. is C-3 S.G. and plug {M defective tubes. Notification to NRC N.A. N.A. pursuant to 50.72 (b)(2) of 10 CFR Part 50 S=3E% where N is the number of steam generators in the unit, and n is the number of steam generators inspected during an 12 inspection. SOUTH TEXAS - UNITS 1 & 2 3/4 4-18 - _ _ _ _ _ _ - . - .- -- -. - ----.__--_.-___x___--.-_.-_____-_-_____..._..___-_-____._-__--___-_--.

m._at _ .. .. 4_. -.Y nip 4: :3 . STCAM GENERATOR REPAIRED TUBE INSPECTION 1stsAMPLE INSPECTION 2ND'SAMPEINSPECTION semple'Siid Result NctioriRoduired Result AcWiEMoquir6d E.mminim.G.in"st;20%. C4 Nods N. :Ai. 

                                                                                                                                       -                                       N. X
o. f repaired ~ tubes..,P3 y g gg gy y tube. end _inspeettioowa .

seu repared,tubesEtiniF~ CM M_._*'*c** '***!'# S. ;G..;. tubes c:a._ 

                                                                                                                                  -                     _ Peam 6H..E. ~.ser_ios. ..ts.r. c.:.a reouit: t est sampie Cd3                   Insps6tiall..~W 16beiM                       All W jig;Q @s                                           M                                    '

this;S;Gfplug;&feche Cil  ; tW uba:end)nspect  : 20Wof the;; repaired, tubes Soniir$',GIC12 Perforrvisctisdisi'C?2 d e M W iS[G5 nUtlufedditional M6f firksempee~ S.GFere C4 Notificatiori;tsNRC Wyso@6)@~M WJAM W;S.K@d 10...,.C. F.~R...Pa. r. t.,50 C4- escli Ggan plugWM l Noun onon:to nRc M tg 507db82[hi 

                                                                                                                                                      -10 CFR P. art':.60 T
               !"1Eacl.Qe..pa..ir.v'.;'. method Isiso. nside.ared ,Aisiiissatipopulatisd..w.6..
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                                                                              -v                 -        - fF..#<d. etermination';6...fls.
                                                                                                                             .+           - sw cope ~sipenelo61         ev I

g i L 3/4 4-18a SOUTH TEXAS - UNITS 1 & 2 i t I [ 

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I REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System Leakage Detection Instrumentation shall be OPERABLE:

a. One Containment Atmosphere Radioactivity Monitor (gaseous or particulate), and
b. The Containment Normal Sump Level and Flow Monitoring System. -{

APPLICABILITY: MODES 1, 2, 3, and 4.  ; 1 ACTION: ' I

a. With the required containment atmosphere radioactivity monitor inoperable perform the following actions or be in at least HOT l STANDBY within the next 6 hours and in COLD SHUTDOWN within the ,

following 30 hours: i

1) Restore one containment atmosphere radioactivity monitoring system to OPERABLE within 30 days and
2) Obtain and analyze a grab sample of the containment atmosphere for gaseous and particulate radioactivity at i least once per 24 hours, or l
3) Perform a Reactor Coolant System water inventory balance at least once per 24 hours.
b. With the required containment normal sump level and flow monitoring system inoperable perform the following actions or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours:
1) Restore the containment normal sump level and flow monitoring system to OPERABLE status within 30 days and
2) Perform a Reactor Coolant System water inventory balance at least once per 24 hours,
c. With both a. and b. inoperable, enter 3.0.3.

SURVEILLANCE REQUIREMENTS 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by:

a. Containment Atmosphere Gaseous and Particulate Monitoring Systems performance of CHANNEL CHECK, CHANNEL CALIBRATION, and DIGITAL CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3, and
b. Containment Normal Sump Level and Flow Monitoring System performance of CHANNEL CALIBRATION at least once per 18 months.

SOUTH TEXAS - UNITS 1 & 2 3/4 4-19

j REACTOR' COOLANT SYSTEM QPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION j 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE, j l
b. 1 gpm UNIDENTIFIED I2AKAGE, i
c. Fcr Unit 1, 150 gallons per day of primary-to-secondary leakage 1 through any one sceam generator, and fcr Unit 2, 1 3pr total {

reactor to accendary 1;; hag; through all etcar gcncratcrc and 500 i gallcnc per day thrcugF any cnc ctcar generater,

d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, and
e. 0.5 gpm leakage per nominal inch of valve size up to a maximum of )

5 gpm at a Reactor Coolant System pressure of 2235 1 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in  ! Table 3.4-1.* I APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 1 hours.
b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage 4

from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the ne?:t 6 hours and in COLD SHUTDOWN within the following 30 hours.
  • Test pressures less than 2235 psig but greater than 150 psig are allowed.

Observed leakage shall be adjusted for the actual test preeaure up to 2235 psio assuming the leakage to be directly proportional to pressure differential to the one-half power. SOUTH TEXAS - UNITS 1 & 2 3/4 4-20

s ' t Q s REACTOR COOLANT SYSTEM BASES RELIEF VALVES (Continued) C. Manual control of the block valve to: (1) unblock an isolated PORV to allow it to be used for manual control of reactor coolant system pressure (Item A), and (2) isolated 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 i The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken. The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to minimize corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the 3.4.6.2.c limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System. Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage as low as 150 gallons per day per steam generator can readily be detected. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged of&spaired.7 DefectiVeitubss % f;bsTrepsiredtby?a Westisghouseflaser@ eldedisleeve M Theftechnical sa's.es/ forts.leevingsrepsifars, describediinsWestinghousefReports;WCAPs136983 Revision 42 M Laser!Weldedisleeves for13/ Cinch) Diameter 3TubelFeedring-TypeiandiWestlinghouseiPreheatersSteaf ~ ~~ GeneratorspfApril31995(andiWCAP-14653 M SpecificfApplicationtofiLaseriWelded SleevesfforiSouthhTexasiProjectsPower PlantlSteam1Generatorsg Qune(1996) 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. Except as discussed below, plugging b((repair will be required for all tubes with imperfections exceeding the plugging"or[ repair limit of 40% of the original tube nominal wall thickness. IfialtubeGcontainsyalWestinghouse? laser!weldedtsleeve ki th ? imper fecti on$ ed:eedingi40 % iof 2 nominall wall i thicknes s Mit[mus t : be l plugged ! Thelbasisyforsthefsleeveipluggingilimitjisibased(on(RegulatorylGuidet14121 analysis R and1isidescribedtinnthelWestinghouse?sleevin abovek "She'am^gsnerator tube"insp' set' ions of"opskating^gitechnicalireportsolisted plants"haVe~demonstfated" 

                                                                                                                   ~

the Espability to reliably detect degradation that has penetrated 20% of the orig]inal tube wall thickness. Repaired?tubss7srs[siso7 incl 6deTiWtheyinsefVics tube inspe;ct16n[prbgfam] 

                                               ^ ~ ~~' ~ '            ~'~ "~~ ~~~"' ~ ' " '

SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-2a 

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

l l , 

        'O i

> i.  ; l REACTOR COOLANT SYSTEM l BASES 4 STEAM GENERATORS (Continued) Exclusion of certain areas of Unit 1 tubes from consideration has been , analyzed using an F* criteria. The criteria allows service induced degradation i deep within the tubesheet to remain in service. The analysis methodology > determines the length of sound fully rolled expanded tubing required in the  ! i uppermost area within the tubesheet to preserve needed structural margins for  ! all service conditions. The remainder of the tube, below the F* distance, is i considered not structurally relevant and is excluded from consideration to the , customary plugging criteria of 40% throughwall. i The amount of primary to secondary leakage from tubes left in service by '. application of the F* criterion has been determined by verification testing. . This leakage has been considered in the calculation of postulated primary to  ; 5 secondary leakage under accident conditions. Primary to secondary leakage {

during accident conditions is limited such that the associated radiological  ;

consequences as a result of this leakage is less than the 10 CFR 100 limits. ] For Unit 1, the voltage-based repair limits of SR 4.4.5 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 j 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, j 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 I the degradation morphology. Implementation of SR 4.4.5 requires a derivation of the voltage structural limit from the burst versus voltage empirical correlation and then the j subsequent derivation of the voltage repair limit from the structural limit (which is then implemented by this surveillance). ] 4 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; V,v is determined from the structural voltage limit by applying the follouing equation

V 3, = V,i, - Va, - V, where V , represent the allowance for flaw growth between inspections and V , 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. SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-3 1 t f

_ __ _ . __ _ _ _ _ _ .. _ ... - _ _ . . _ - _ -_-__m. _ __m . . . _ , _ . - - _ , o a 4 

!        , e t

REACTOR COOLANT SYSTEM i BASES ! STEAM GENERATORS (Continued) l l The mid-cycle equation in SR 4.4.5.4.a.+t12.e should only be used during unplanned inspections in which eddy current data"is acquired for indications at the tube support plates. 

                                                                                                                                       ]

SR 4.4.5.5 implements several reporting requirements recommended by ' 1 GL 95-05 for situations which the NRC wants to be notified prior to returning { the SGs to service. For the purpose 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 l to GL 95-05 for more information) when it is not practical to complete thesa l calculations using the projected EOC voltage distributions prior to returning . the SGs to service. Note that if leakage and conditional burst probability were  ! j 

'                   calculated using the 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. Whenever the results of any steam generator tubing inservice inspection 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 and 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 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 j indicative of an impending gross failure of the pressure boundary. Therefore, 1 the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly I 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 4 1 to a threshold value of less than 1 gpm. This threshold value is sufficiently j low to ensure early detection of additional leakage. i SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-3a

I e, ? is REACTOR CODLANT SYSTEM BASES l OPERATIONAL LEAKAGE (Continued) For Unit 1, the leakage limits incorporated into SR 4.4.6 are more restrictive than the standard operating leakage limits and are intended to J provide an addition margin to accommodate a crack which might grow at a greater I than expected rate or unexpectedly extend outside the thickness of the tube  ! support plate. Hence, the reduced leakage limit, when combined with an j 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. shd 2, thegeneratore beba4- steam generator tube leakage limit of 4-gpm For Unithy 150Tgpdjfor e44"g sachisteam not isolated from the RCS ensures that the d6sige contribut'lo6^'from the tube leakage will be limited to a small fraction of  ; 10 CFR Part 100 dose guideline values in the event of either a steam generator ' tube rupture or steam line break. The 4-gpm 150  ; is :: :ictant with I:fonservative@caparodito the~Jgpd limit pdW#taanGgineratiorass'Umptionsi of these accidents.""Ths~s44"150'5pd~15akiye limit per steam generator ensures j that steam generator tube inth5fity 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 f 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 time, verification of valve integrity is required. Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA that bypasses containment, these valves should be tested periodically to ensure low probability of gross failure. The Surveillance Requirements for RCS pressure isolation valves provide 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 the allowed limit. 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 SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-4

g 4/ se - REAC70R COOLANT SYSTEM BASES GRilS_TRX (Continued) the chemistry within the Steady-State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant conc?ntration l levels in excess of the Steady-State Limits, up to the Transient Limits, for the  ! specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady-State Limits. The Surveillance Requirements provide adequate assurance that l concentrations in excess of the limits will be detected in sufficient time to ' take corrective action. 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure that the resulting 2-hour doses at the SITE BOUNDARY will not exceed an appropriately small fraction of 10 CFR Part 100 dose guidelines values following a steam generator tube rupture accident in conjunction with an assumed steady-state reactor-to-secondary steam generator leakage rate of 4-gpm 150 % d!p"er steamigenefator. The values for the limits on specific activity Fepresent ~ l'i'mit's based ~upon a parametric evaluation by the NRC of typical site locations. 4 These values are conservative in that specific site parameters of the STPEGS I site, such as SITE BOUNDARY location and meteorological conditions, were not f considered in this evaluation. The ACTION statement permitting POWER OPERATION to continue for limited time periods with the reactor coolant's specific activity greater than 1  ; microcurie / gram DOSE EQUIVALENT I-131, but within the allowable limit shown on i Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. The sample analysis for determining the gross specific activity and E can exclude the radioiodines because of the low reactor coolant limit of 1 microcurie / gram DOSE EQUIVALENT I-131 and because, if the limit is exceeded, the i radiciodine level is to be determined every 4 hours. If the gross specific activity level and radiciodine level in the reactor coolant were at their limits, the radioiodine contribution would be approximately 1%. In a release of reactor coolant with a typical mixture of radioactivity, the actual radioiodine contribution would probably be about 20%. The exclusion of radionuclides with half-lives less than 15 minutes from these determinations has SOUTH TEXAS - UNITS 1 & 2 B 3/4 4-5}}

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