ML20101Q318

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Proposed Tech Specs 3.4.5 & 3.4.6.2,incorporating voltage- Based Repair Criteria
ML20101Q318
Person / Time
Site: South Texas STP Nuclear Operating Company icon.png
Issue date: 04/04/1996
From:
HOUSTON LIGHTING & POWER CO.
To:
Shared Package
ML20101Q317 List:
References
NUDOCS 9604110123
Download: ML20101Q318 (10)
v • d • e


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ATTACHMENT 3 PROPOSED TECHNICAL SPECIFICATION l

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9604110123 960404 l

PDR ADOCK 05000498 p PDR TSC-%\5332.w

ST-HL-AE-5332 l' Attachment 3 Page 1 of 9 i REACTOR COOLANT SYSTEM l STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) l l

3) A tube inspection (pursuant to Specification 4.4.5.4a.8) shall be performed I on each selected tube. If any selected tube does not permit the passage of
the eddy currer.t probe for a tube inspection, this shall be recorded and an l adjacent tube shall be selected and subjected to a tube inspection.

i 4) Indications left in service as a result of application of the tube support plate l

voltage-based repair criteria shall be inspected by bobbin coil probe during all future refueling outages.

c. The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:
1) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found, and
2) The inspections include those portions of the tubes where imperfections were previously found.
d. 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 tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of the lowest cold-leg tube support plate intersecticns 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:

i Category Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

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

l C-3 More than 10% of the total tubes inspected are degraded tubes or I

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 SOUTH TEXAS - UNITS 1 & 2 3/4 4-13 TsC-960332.w

ST-HL-AE-5332 l Attachment 3 Page 2 of 9 REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria

a. As used in this specification:
1) Imperfection 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 l detectable, may be considered as imperfections;
2) Degradation means a service-induced cracking, wastage, wear, or general ,

corrosion occurring on either inside or outside of a tube; I

3) Degraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation;
4)  % Decradation 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) Pluccing Limit means the imperfection depth at c: beyond which the tube shall be removed from service and is equal to 40% of the nominal tube wall thickness. For Unit 1, this definition does not apply to tube support plate intersections for which the voltage-based repair criteria are being applied.

Refer to 4.4.5.4.a.10 for the repair limit applicable m these intersections.

7) 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;
8) Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg; 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.

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ST-HL-AE-5332 Attachment 3 Page 3 of 9 REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria

10) For Unit.1, Tube Support Plate Plunging 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 confm' ed within the thickness of the tube support plates. At tube support plate intersections, the plugging (repair) limit is based on maintaining steam generator tube serviceability as described below:

a) Steam generator tubes, whow 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 in 4.4.5.4.a.10.c below.

c) Ste'am generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair limit (Note 1) but less than or equal to the upper repair voltage limit (Note 2), may remain in service if a rotating pancake coil inspection does not detect degradation. Steam generator trbes, with indications of outside diameter stress corrosion cracking degradation with bobbin voltage greater than the upper voltage repair limit (Note 2) will be_ plugged or repaired, d) Certain intersections as identified in Framatome Technologies, Inc.

Topical Report BAW-10204P, " South Texas Project Tube Repair Criteria For ODSCC At Tube Support Plates" will be excluded from application of the voltage-based repair criter:a as it is determined that these intersections may collapse or deform following a postulated LOCA + SSE event.

e) If an unscheduled mid-cycle inspection is performed, the mid-cycle repair limits apply instead of the limits identified in 4.4.5.4.10.a, 4.4.5.4.10.b, and 4.4.5.4.10.c. The mid-cycle repair limits will be determined from the equations for mid-cycle repair limits of NRC Generic Letter 95-05, Attachment 2, page 3 of 7. Implementation of these mid-cycle repair limits should follow the same approach as in TS 4.4.5.4.10.a, 4.4.5.4.10.b, and 4.4. 5.4.19.c.

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ST-IIL-AE-5332 Attachment 3 Page 4 of 9 REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued)

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.

Note 2: The upper voltage repair limit is calculated according to the methodology in ,

Generic Letter 95-05 as supplemented. Vuni may differ at the _TSPs and flow distribution baffle.

b. The steam generator shall be determined OPERABLE after completing the i 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

a. Within 15 days following the completion of each inservice inspection of steam j generator tubes, the number of tubes plugged in each steam generator shall be l reported to the Commission in a Special Report pursuant to Specification 6.9.2; j 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

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within 12 months following the completion of the inspection. This Special Report i shall include: I

1) Number and extent of tubes inspected, j
2) Location and percent of wall-thickness penetration for each indication of an imperfection, and
3) Identification of tubes plugged.

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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 i a description of investigations conducted to determine cause of the tube degradation  ;

and corrective measures taken to prevent recurrence. 1

d. For Unit 1,' implementation of the voltage-based repair criteria to tube support plate  !

intersections, notify the Staff prior to retuming 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 not practical, ,

i using the actual measured end-of-cycle) voltage distribution exceeds the leak limit (determined from the licensing basis ' dose calculation for the postulated main stream line break) for the next operating cycle.

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ST-HL-AE-5332 Attachment 3 Page 5 of 9 REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued)

2. If circumferential crack-like' indications'are detected at the tube support plate intersections.
3. If indications are identified that extend beyond the confines of the tube support plate.
4. If indications are identified at the tube support plate elevations that'are attributable.to primary water stress corrosion cracking.
5. If the calculated conditional burst probability based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds 1 x 10-2, notify the NRC and provide an assessment of the' safety significance of the occurrence.

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ST-HL-AE-5332 Attachment 3 Page 6 of 9 REACTOR COOLANT SYSTEM i OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE, l
b. I gpm UNIDENTIFIED LEAKAGE,
c. For Unit 1,'.150 gallor.s per day of primary-to-secondary leakage through any one l

steam generatoi, and for _ Unit 2,-l gpm total reactor-to-secondary leakage through all steam generators and 500 gallons per day through any one steam generator,

d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, and ,
e. 0.5 gpm leakage per nominal inch cf valve size up to a maximum of 5 gpm at a Reactor Coolant System pressure of 2235120 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.*

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

1 ACTION l a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY l 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, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT 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 SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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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 l

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 HOT 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 SHUTDOWN 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 i 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.

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ST-HL-AE-5332 Attachment 3 Page 7 of 9 REACTOR COOLANT 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 ::: !: in neg!!gib!: ininimize corrosion of the steam generator tubes. 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 (prima j :c ;::endarf :hage 1 = 500 g;!!cn; per day per :::= g:ncreter-). 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 ef-500 as low as 150.

gallons per day per steam generator can readily be detected by radiatica monitors of ;;;=

gen =:c b!cwdc=. 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. However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube e-xaminations. :Except as. discussed below, plugging will be required for all tubes with imperfections exceeding the plugging limit of 40% of the tube nominal wall ihickness. 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.

Fo'r 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

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diameter stress corrosion cracking (ODSCC) located at the tube-to-tube support plate intersections. I i The voltage-based _ repair limits are not applicable to other forms of SGdsoe degradation nor are l 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 I for additional description of 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 subsequent derivation of the voltage repair  !

! limit from the structural limit ('which is then implemented ; by this surveillance).

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ST-HL-AE-5332 Attachment 3 Page 8 of 9 1 REACTOR COOLANT SYSTEM BASES STEAM GENERATORS (Continued)

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 95/952 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; Vuat, is determined from the structural . voltage limit by applying the following equation:

Vunt = Vst Von - V uos where Van represents the allowance for flaw grow h between inspections and.Vuos' represents the allowance for potential sources of error in the measurement of the bobbin coil voltage. ,Further discussion of the assumptions necessary to determine the voltage repair limit are discussed in GL 95-05.

The 'mid-cycle' equation in' SR"4.4.5.4.10.e should only be used during~unplanneo inspections in which. eddy current data is acquired for indications at the tube support plates.

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SR 4.4.5.5 implements several reporting requirements recommended by GL 95-05 for situations which the NRC wants to be notified prior to returning.the SGs to service. For the purpose of this reporting requirement, leakage and conditional burst probability can be calculated based on the as-found voltage distribution rather thvi the projected end-of-cycle voltage .

distribution (refer to GL 05-05 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 thit if leakage and conditional burst probability were 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.

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ST-HL-AE-5332 Attachment 3 Page 9 of 9 REACTOR COOLANT SYSTEM BASES 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 -

I 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 1

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

For Unit 1; the le~akage' limits incorporated into_ SR'4.4.6~are mom restrictive than the standard operating leakage limits and am intended to provide ~ an additional margin' to accommodate a crack which might grow at a greater than e.xpected rate or unexpectedly ~ extend outside the thickness of the tube support plate. : Hence,' the reduced leakage limit, when combined

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j with an effective leak rate monitonng program, provides additional assurance that should a l significant leak be experienced in' service,'it will be detected, and the plant shut down.in a timely manner.

Foi Unit .2,t.the 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 I gpm limit is consistent with the assumptions l 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 ;- UNITS 1 & 2 B 3/4 4-3a TsC-%\5332.w

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