ML20080T689

Proposed Tech Specs,Modifying SG Tube Plugging Criteria in TS 3.4.5,SGs & Allowable Leakage for Unit 1 in TS 3.4.6.2, Operational Leakage & Associated Bases

ML20080T689
Person / Time
Site:	South Texas
Issue date:	03/01/1995
From:	HOUSTON LIGHTING & POWER CO.
To:
Shared Package
ML20080T682	List: ML20080T689 ST-HL-AE-5027, Application for Amend to License NPF-76,modifying SG Tube Plugging Criteria in TS 3.4.5,SGs & Allowable Leakage for Unit 1 in TS 3.4.6.2,operational Leakage & Associated Bases
References
NUDOCS 9503140069
Download: ML20080T689 (9)

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i ATTACHMENT 3 PROPOSED TECHNICAL SPECIFICATION CHANGES  !

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' Attachment 3. l Page 1 of 8 l

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REACTOR COOLANT SYSTEM ,

STEAM GENERATORS '  !

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SURVEILLANCE REQUIREMENTS (Continued) : -i 3)' l A tube inspection (pursuan't to Specification 4.4.5.4a.8) shall be performed. l on each selected tube. . If any. selected tube does not permit the passage of - i 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.

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4) - Tubes ~left iriMrvice'asTisultlof ^applic~ation'of th(tubs su~pport) late

' lugging criteria shall be inspected. by,bobbinicoil probe during all;futup p l refueling outages. j

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

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-1) The tubes selected for these samples include the tubes from those areas of. j the tube sheet array where tubes with imperfections were previously found, .  !

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

l d: Forl Unit 1RCpcle'6,'implementationTof'theLtube" support ~ plate laltemate; plugging criteria limit requires a"100 percent bobbin ~ coil inspection for all hot leg tube  !

support plate intersections and all cold leg' intersections down to'the lowesticold 16g  :!

tube. support plate with outer' diameter stress' corrosion; cracking (ODSCC)

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IndicationsEThe determination'of the tubeLsupport plate intersections havi6g ODSCC indications shall'be based on the performanna of.at.least;a;20% randoni -!

' sampling Lof.. tubes ~ inspected over their full, length; j 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 q none of the inspected tubes are defective. -

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

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

'l Note: In all inspections, previously degraded tubes must exhibit significant (greater  ;

than 10%) further wall penetrations to be included in the above percentage  !

calculations.

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< Attachment 3 Page 2 of 8 L

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 -

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;

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

4)  % Degradation 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) Plugging 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 (li Cycle'6, this ' definition does "nof apply .to the fesion of the' tube subject to the tube suppoit plate l alternate plugging criteria limit; i.e.~ '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 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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Attachment 3 ,

Page 3 of 8- L REACTOR ~ COOLANT SYSTEM  !

i STEAM GENERATORS  :

SURVEILLANCE REQUIREMENTS '(Continued) y 4.4.5.4. Acccotance Criteria  !

10)- For.Unii'1[ Cycle 6ftlETube Sannart Plate %hehate PinWriairriearis j Mg3h is' used for the. disposition of a steam generator tube for continued  !

service that is' experiencing outer diameter ' stress 'conosion cracking confined .

within the thicknessLof the tube suppoit plates.fAtit6e support plate .

intenections, the repair limit is based on;maintainingisteam generator tube  :

serviceability;as described.belowi  ;

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a) D6 gradation' attributed to'outside~diametef stress conosion'dracking

'within the bounds of the tube support plate with bobbin voltagejless  ;

than .or.ejjuallto;1.0 yoit _will be allowed to . remain in'servicei j b) t Degradation attributed to'6utside* diameter"s'iess" corrosion ~ cracking within the bounds of the tube support plate with bobbin.. voltage,  ;

greater than L1.0. volt lwill be plugged except as'.noted'in 4.4.5.4.a.10.c.

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k c) Indications of potential degradation' attributed to'outside^diameteF t stress conosion cracking within the bounds'of the tube ~ support plate j

. with' a bobbiti voltage greater than 1.0 volt. but lessLthan or equal to 2.3 volts may semain in service if a rotating naac=ka coil inspection does not detect degradation. LIndications of outside diameter stress ,

corrosion cracking degradation with bobbin ~ voltagefgreater. than 2.3  ;

. v' olts will be plugged.

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Tube' intersections that fall ~within' thitube'siipportilate plastic i

deformation exclusion zones.will be excluded from. application of ths voltage-based plugging criteria:

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J SO,UTH, TEX. A. S_TU. NIT.1 & 2 3/4. _4.15a.

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Attachment 3 Page 4 of 8  !

REACTOR COOLANT SYSTEM STEAM GENERATORS i

' 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 l containing through-wall cracks) required by Table 4.4-2. i i'

4.4.5.5 Reports

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

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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 i and corrective measures taken to prevent recurrence.

d. For Unit 1,- Cycle 6, implementation ~'of the voltage-based repair criteria ~to' tube support plate intersections, reports to the Staff shall be made as' follows;

1) Notify the Staff prior to returning the" steam generators to service;should any ,

of the following conditions arise: ;

a) If estimated leakage based on~ the actual ~ measured end-of-cycle

• voltage distribution would have exceeded the~ leak limit (for postulated main ' steam line break utilizing licensin'g basis assumptions) during .the previous Cycle.-

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Attachment 3 Page 5 of 8 l

. REACTOR ~ COOLANT SYSTEM

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SURVEILLANCE REQUIREMENTS (Continued) ,

i b) If circumferential' crack-like indications are detected lat theLtub'e support plate intersections. t c) If indications ~are identified that Extend beyond thi^ confines of'the i tube support plate. {

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d) If the ' calculated' conditional burst probabilitj' exceeds 1 x 10-2, notify the NRC and provide an assessment of the safetyy significance of _the l occurrence. .

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2) The final results of the inspection and the tube integrity evaluation shall be reponed to the Staff pursuant to Specification .6.9.2 within 90. days following restart of Cycle 6_(Mode 1).

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ST-HL-AE-5027 Attachment 3 Page 6 of 8 REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE

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

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a. No PRESSURE BOUNDARY LEAKAGE,

b. I gpm UNIDENTIFIED LEAKAGE,

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c. For Unit 1,'600 gallons per day total reactor-to-secondary leakage through all~ steam generators and 150 gallons per day through any one steam generator, and for Unit 2,1 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 of 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.

ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT 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, 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 />.

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 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 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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REACTOR COOLANT SYSTEM l

BASES l

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 negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during 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 Seconday Coolant System (pri=ry :c seccndary

!eakage = 500 ga!!cn> per day per s:ca:n genera:cr). 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 :neniter; cf s:ca:n generater b!cwdern 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 chemisty 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 will be required for all tubes with imperfections exceeding the plugging limit of 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, Cycle 6, tubes experiencing outer diameter stress corrosion cracking at the tube support plates (TSPs) where such cracking is confined to the thickness of the TSPs will be ,

dispositioned in accordance with Specification 4.4.5.4.a.10. Testing of tubes with ODSCC has demonstrated a high margin to failure and evaluations have shown that existing tube plugging criteria would cause unnecessan and inappropriate tube 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 l C-3, these results will be promptly reported to the Commission in a Special Repon 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.I 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.

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ST-HL-AE-5027 Attachment 3 Page 8 of 8 REACTOR COOLANT SYSTEM BASES 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 value is sufficiently low to ensure early detection of additional leakage.

Maintaining an' operating leakage limit 'of 150 gpd per steam generator (600 gpd total)~ for Unit I will minimize the potential for a large leakage event during a main steam line break.

Based on the non-destructive examination uncertainties, bobbin coil voltage distribution,'and crack growth rate from the previous inspection, the expected leak rate following a steam line rupture is limited to below the applicable dose limits in the faulted loop. Leakage in.the intact loops.will be limited to the operating limit of 150 gpd. If the projected end-of-cycle distribution of crack indications results in primary-to-secondary leakage greater than the applicable dose limits in the faulted loop during a postulated steam line break event, additional tubes must be removed from service in order to reduce the postulated steam line bicak leakage.to below the applicable. dose limits.

For Unit 2, 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 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.

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 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.

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