ML20100B823

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Proposed Tech Specs 3.4.5 & 3.4.6.2 Including Associated Bases 3/4.4.5 & 3/4.4.6.2,allowing Implementation of Alternate SG Tube Plugging Criteria for Tube Support Plate/ Tube Intersections
ML20100B823
Person / Time
Site: South Texas STP Nuclear Operating Company icon.png
Issue date: 01/22/1996
From:
HOUSTON LIGHTING & POWER CO.
To:
Shared Package
ML19311B930 List:
References
NUDOCS 9601260235
Download: ML20100B823 (10)
v • d • e


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ATTACHMENT 3 1

PROPOSED TECHNICAL SPECIFICATION CHANGES 1

4 9601260235 960122 ADOCK 05000498 PDR PDR P

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ST-IIL-AE-5269 Page 1 of 8 REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) 3)

A tube inspection (pursuant to Specification 4.4.5.4a.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 inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.

4)~1 LTubes left in"servics'afa ieselt'of applicsti6ii'of the inbe~sOpp6rt' plats

' alternate l plugging criteriaishall.be~inspectedg' bobbinyoi.lfprobe'.ddririg 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, and 2)

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

df For Unit:1;: implemsntstion'of the tubs support plateLatternate'pitigging briteria

" limit requires a 100 percent bobbin coil'. inspection.for'all hot leg tube support plats intersections and all coki leg intersectionsLdown to;the lowest" cold le~g' tube support plateLwith outer diameter stress borrosion~crac_ ing'(ODSCC) indications?l;The '

k determination of theftubeisupport plate intersections having'ODSCC indications shall be based.on the performance of at least a _20% random samplingLof tubes inspected over their full le'ngth.

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

Catecory Insnection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

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

C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.

Note: In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage i

calculations SOUTil TEXAS - UNITS 1 & 2 3/4 4-13 i

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ST-HL-AE-5269 Page 2 of 8 REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Accentance Criteria a.

As used in this specification:

1)

Jmnerfection 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; 2)

Decradation means a service-induced cracking, wastage, wear, or general corrosion occurring on either inside or outside of a tube; 3)

Decraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation; 4)

% Deuradation means the percentage of the tube wall thickness affected or removed by degradation; 5)

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

Pluccine Limit means the imperfection depth at or beyond which the tube shall be removed from service and is equal to 40% of the nominal tube wall thickness; (For Unit'1, thisitermitioridoes not aipl9Lt(the regionif the

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tube subject to.the tube support plate alternate p ugging criteria limit;i.e.1 the tube support plate intersections. Specification;4.4.5.4.a.10 describes.the plugging limit for use within the tube support plate intersection ~of the tube.)

7)

Unserviceable describes the condition of a tube ifit 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; 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

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support of the cold leg; and 9)

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 techniques expected to be used during subsequent inservice inspections.

SOUTH TEXAS - UNIT 1 & 2 3/4 4-15 1semme.

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Page 3 of 8 i

REACTOR COOI? ANT SYSTEM i

STEAM OENERATORS SURVEILL:ANCE: REQUIREMENTS l(Continued).,

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i 4.~4;5;41, r^ccentance' Criteria l

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10)MForjUsit':1[thsT6be Sssn6rt^ Plats ^AlternesPlushi6ECriidriiLimitTis ussd fof the disposition ^of a steam' generator tube for continued service that is ~

t sxperiending' outer diameter stress corrosion cracking confined lwithin the thickness lof the tube support plateMAt; tube' support plats intersectionsithe

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e desc.ribed b,elow:

j repair limit is based on maintaining [steamigeneratottubsjsesvicsabilityjs a_)1 * '.? Degrsdatiojiattrib6ted 16'6uisidi'dismeter stress"66rrosi6h~6facking

.m thas~or eqdal.to;:1.0 volt wil,1 bs allowed to remain in service.: ~ ~

b)],^ Degriidstio6~attiibuied 16"ouisidsdiameter7sMc6rrosion~ cracking i

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~ ithin the boundsiof;the tube'suppett plitelwith bobbihkoltage i

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4 greater (than11.0 volt lwill lie plugged;except as n6ted. id"

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4..,4,.5,.4;a 1. 0..c..

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7 Indications'of potential;~degradatioh'~attributsd 16 outside^ diameter

~~ ^ ~~" stress corrosion crackingLwithin the bounds'of the tube support piste with;a bobbin voltage" greater than 1.0 volt butlless than~or: equal t6

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2.85Lvolts may remain in servics ifialrotsting pancake coll linspectioh 4

does notl detect degradationhilndications of outsidel diameter stress corrosion' cracking dsgradation:with~ bobbin voltage greater than:2.85 yollsfwil.l.be;;plugge,d: ^

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iTub6 inisisecti6ns' thst fall'withis the tube supp6ft plats plastic

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" deformation' exclusion zones will beLexclsded from application of the

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voltag6 based plugging criteria) c)j flf An unsch6daled mid cycliinspection is performedfthiinid cpcle?

repair limits apply instead ;of the limits identified-inL4;4.5.4.10.a; 1

4.4.5.4.10.b, and 4.4.5.4.10.cQ:The mid-cycle repair limits;,will be determined. from the equations for:mid-cy'cle repair 1imitslof NRC Generic Letter'95-05, Attachmenf 2{

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ST-HL-AE-5269 Page 4 of 8 REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) b.

The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2.

4.4.5.5 Reports Within 15 days following the completion of each inservice inspection of steam a.

generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2; 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 3)

Identification of tubes plugged.

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 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'-basid ~ repair criteria to tube l support' plate intersections,. reports to the.. Staff shall be made;as follows:

1);

Notify the Staff prior to returning thel steam l generators'to sefvicEshould'ahy of the following conditions arise

a)

If estimated leakage based'on the actdal measured:end-of-cycle voltage distribhtionLwould have exceeded the leak limit (for'

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postulated main steam line; break utilizing licensing basis assumptions) during the previous: Cycle.

SOUTil TEXAS - UNITS 1 & 2 3/4 4-16 TSC 963269 =

ST-HL-AE-5269 Page 5 of 8 REACTOR ^COOEANT SYSTEM.

STEAM 2 GENERATORS SURVEILLANCE LREQUIREMENTSll(Continued).

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? If cifcsinferential crack ~ liks indications are'dstected at the thbs

' support plate'; intersections. ~

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11findisations'sisidentifisd(that stend be 66d.;ths::c6nfiss[of ths tube support l plate, d)"?

TIfindications'aFe identified ars'at the tsbe"shpp6rt'plsts'eleVsti6ns

~that arelstiributableLtolprimaryjyvaterl. stress'sorrosionlWackingf

If theTealculstsd's6nditional b. fstTprobability'sk6sedil[i?l0'9 a

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" notify the NRC and provide an sssessment 'of the~ safety)signifiesnee of thel occurrence) ~ ~

2p" ^!ThB final results bf thefinspe' tion and tha isbiintsgrifylevaluitionishall be c

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' reported t6 the Staff p~ursuant Ltd Sp"ecification 6.9.2.withis90 days following restartJMods11).

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SOUTH -; TEXAS t^ U, NIT.2: 1 & -2,,i ' ',, l 3/4. '4.. :16s

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I ST-HL-AE-5269 Page 6 of 8 REACTOR COOLANT SYSTEM

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OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

3 a.

No PRESSURE BOUNDARY LEAKAGE, b.

I gpm UNIDENTIFIED LEAKAGE, I

Foi).UnitiR60075allohspsFl day 16tsi ii ssts6t6:ssb6ndary?!caksie th~r6sglir llideais e

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gena @tgpasd!1pagallonslperfdayy6ugl[thr@ough'lill steam generato an 241 gpm total reactor-to-secondary liakage hallons per day through any one steam generator, j

d.

10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, and 3

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 i 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.*

APPLICABILITY: MODES 1,2,3, and 4.

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2 ACTION:

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2 a.

With any PRESSURE BOUNDARY LEAKAGE, be in at least IIOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.

With any Reactor Coolant System leakage greater than any one of the above limits, l

excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 4

hours or be in at least IIOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SilUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c.

With any Reactor Coolant System Pressure Isola' ion Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two closed manual or deactivated automatic valves, or be in at least IIOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SilUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

  • Test pressures less than 2235 psig but greater than 150 psig are allowed. Observed leakage shall be adjusted for the actual test pressure up to 2235 psig assuming the leakage to be directly proportional to pressure differential to the one-half power.

i-SOUTH TEXAS - UNITS 1 & 2 3/4 4-20 nc ss2..

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ST-IIL-AE-5269 Page 7 of 8 REACTOR COOI. ANT SYSTEM BASES STEAM GENERATORS (Continued)

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 neg!!gible minimizs corrosion of the steam generator tubes. Localized corrosion may likely result in stress coirosio~n cracking. The extent of cracking during plant operation would be limited by the 3A6.2M limitation of steam generator tube leakage between the Reactor Coolant System and the Secoridary Coolant System (primary te =cendary !:ahage - 500 gallens per day per steam generater). 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 of-500 is 16(f sisali 6150 gallons per day per steam generator can readily be detected by radiation mcnif5:f6 generator b!c vdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.

Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. Ilowever, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations. LExcept as discussed below, plugging will be

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required for all tubes with imperfections exceidins the plugging limifor 40% of the tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.

For Unit 1;; tubes experiencing outer;dianieterl stress"c6rrosion cracking ~at ths; tube support plates (TSPs),1where such cracking is confined to the thickness of the TSPs, will be dispositioned inLaccordance with Specification 4.4.5.4.a.114 Testing bf tubes with ODSCC has. demonstrated.a high margin to faihire and evaluations have shown that existing' tube plugging criteria would cause unnecessary and inappropriate t_ube plugging.? Unnecessarily plugged. tubes can reduce steam generator heat removal capacity in both; accident conditions _and normal operations.

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.

SOUTil TEXAS - UNITS 1 & 2 B 3/4 4-3 T5CM32new

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ST-HL-AE-5269 l

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REACTOR COOLANT SYSTEM j

BASES 3/4,4.6 REACTOR COOLANT SYSTEM LEAKAGE j

3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS Leakage Detection Systems required by this specification are provided to monitor i

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 j

Boundary Leakage Detection Systems," May 1973.

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3/4.4.6.2 OPERATIONAL LEAKAGE 1

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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 3

PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD S11UTDOWN.

j 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 1 l

gpm. This threshold value is sufficiently low to ensure early detection of additional leakage.

t ForLUnit 1Tmaintaining;an operating leakage ~ limit of 15.0'gpd psr steamienerat6r;(600 1

hpd total)?will ininimize the potential for a large leakage event during imain stedm'linelbrealc9 3

i Based on the hon.<iestructive examination uncertainties, bobbi6 coil { voltage'distributi6ntandWack growth rate from the previoufi inspecti6n~Tthe sxpected leakirate following:a" steam;line supturiis~

i limited to below the applicable l dose limits in the faulted loopN Lsakage in thelintast loops ;will. be limited to the operating 11imit of 150 gpd;JIf ths projected end-of-cycle ~ distribution ~ of stack indications results in primary-to-secondary;1eakage' greater than the applicable dose limits in the faulted loop during a postulated steam'line break eventfadditionalttubes must be removed frbm service in order;to reduce thel postulated steam' line breakLleakage to below;the. applicable l dose limits;

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i F6f Unit 2lthe total steam generator tube leakage limit of I gpm for all steam generators not isolated 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 values in the event of either a steam generator tube rupture or steam line break. The 1 gpm limit is consistent with the assumptions used in the analysis of these accidents. The 500 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.

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SOUTH TEXAS QUNffSll'&. 21, j B"3/(413a 150-Ge\\S260 w

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ATTACHMENT 5

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l SOUTH TEXAS PROJECT TUBE REPAIR CRITERIA 1

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i TUBE SUPPORT PLATES

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i BAW-10204, REVISION 2 i

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