ML20117F305

Amend 82 to License NPF-76 Revising TS Re SG Tube Plugging Criteria & Associated Bases to Allow Implementation of Alternate SG Tube Plugging Criteria for tube-to-tubesheet Joints for Unit 1

ML20117F305
Person / Time
Site:	South Texas
Issue date:	05/14/1996
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20117F306	List: ML20117F305 ML20117F311
References
NUDOCS 9605170327
Download: ML20117F305 (10)

Text

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UNITED STATES s

j NUCLEAR REGULATORY COMMISSION 2

WASHINGTON, D.C. 38864001

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HOUSTON LIGHTING & POWER CONPANY CITY PUBLIC SERVICE BOARD OF SAN ANTONIO CENTRAL POWER AND LIGHT COMPANY CITY OF AUSTIN. TEXAS DOCKET NO. 50-498 SOUTH TEXAS PROJECT. UNIT 1 AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No. 82 License No. NPF-76 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Houston Lighting & Power Company *

(HL&P) acting on behalf of itself and for the City Public Service Board of San Antonio (CPS), Central Power and Light Company (CPL),

and City of Austin, Texas (C0A) (the licensees), dated January 22, 1996, as supplemented by letter dated April 18, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance:

(1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

• Houston Lighting & Power Company is authorized to act for the City Public Service Board of San Antonio, Central Power and Light Company and City of Austin, Texas and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.

j 9605170327 960514 R

PDR ADOCK 05000498 K

P pop y

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Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license a;nendment l

and Paragraph 2.C.(2) of Facility Operating License No. NPF-76 is hereby amended to read as follows:

2.

Technical Specifications l

The Technical Specifications contained in Appendix A, as revised through Amendment No. 82, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specificatiens and the Environmental Protection Plan.

3.

The license amendment is effective as of its date of issuance to be

' implemented within 10 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

.QYi&

LGU Thomas W. Alexion, Project Manager Project Directorate IV-1 Division of Reactor Projects III/IV l

Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

May 14, 1996 l

i

r ATTACHMENT TO LICENSE AMENDMENT NO. 82 FACILITY OPERATING LICENSE NO. NPF-76 DOCKET NO. 50-498 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain marginal lines indicating the areas of change. The corresponding overleaf pages are also provided to maintain document completeness.

REPOVE INSERT-3/4 4-13 3/4 4-13 3/4 4-16 3/4 4-16 l

B 3/4 4-2a B 3/4 4-2a l

B 3/4 4-3 B 3/4 4-3 l

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REACTOR COOLANT SYTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) 3)

A tube inspection (pursuant to S>ecification 4.4.5.4a.8) shall be performed on each selected tu >e.

If any selected tok 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.

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, any tube allowed to remain in service per Acceptance Criterion 10 (of Technical Specification 4.4.5.4) shall be inspected via the rotatina pancake coil (RPC) addy current method over the F*

distance. Such tubes are exempt from eddy current inspection over the portion of the tube below the F* distance which is not structurally relevant.

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

Cateaory Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and one 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 Nore 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.

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

REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.3 Inspection Frequencies - 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 af ter the previous inspection.

If two consecutive inspections, not including the preser-vice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months; b.

If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months.

The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3a. ; the interval may then be extended to a maximum of once per 40 months; and c.

Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown s

equent tc any of the following conditions:

1)

Primary-to-secondary tube leaks (not includire leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2, or 2)

A seismic occurrence greater than the Operating Basis Earthquake, or 3)

A loss-of-coolant accident requiring actuation of the Engineered Safety Features, or 4)

A main steam line or feedwater line break.

't SOUTH TEXAS - UNITS 1 & 2 3/4 4-14

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REACTOR COOLANT SYSTEM i

STEAM GENERATORS l

4 1

2 SURVEILLANCE REQUIREMENTS (Continued) i 4.4.5.4 Acceptance Criteria j

a.

As used in this specification:

i-1 1)

Imperfection means an exception to the dimensions, finish, or l

contour of a tube from that required by fabrication drawings or 4

specifications.

Eddy-current testing indications below 20% of i

the nominal tube wall thickness, if detectable, may be l

considered as imperfections;

' 21 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 i

than or equal to 20% of the nominal wall thickness caused by degradation; 4)

% Degradation means the percentage of the tube wall thickness j

affected or removed by degradation; 4

j 5)

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

6)

Plugging Limit means the imperfection depth at or beyond which 1,

the tube shall be removed from service and is equal to 40% of the nominal tube wall thickness; j

7)

Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integ-j rity 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 I

each tube in each steam generator performed by addy 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 - UNITS 1 & 2 3/4 4-15

REACTOR COOLANT SYSTEM

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STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued)

(10) F* criteria IFor Unit 1 on1vl Tube degradation below a s ucified distance from the hard roll contact point at or near tte top-of-tubesheet (the F* distance) can be excluded from consideration to the acceptance criteria stated in this section (i.e., plugging of such tubes is not required). The methodology for determination.for 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 0.

b.

The steam generator shall be determined GPERABLE 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 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 Specirl 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 i

investigations conducted to determine cause of the tube c3 gradation and corrective measures taken to prevent recurrence.

4 SOUTH TEXAS - UNITS 1 & 2 3/4 4-16 Unit 1 - Amendment No. 82

RfACIQF_IMI.aHI 1YSlIM RASES 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) isolate the PORV with excessive seat leakage (Item 8).

D.

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

3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure th~at the structural integrity of this portion of the RCS will be maintained." The program for inservice inspection of steam generator tubes is bastrd 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 er 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 l

coolant will be maintaineo within those chemistry limits found to result in i

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 limitation of steam generator tube

-leakage between the Reactor Coolant System and the Secondary Coolant System (primary-to-secondary leakage = 500 gallons per day per steam generator).

Cracks naving a primary-to-secondary leakage less than this limit during operation will have an adequata 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 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown.

Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.

SOUTH TEXAS - UNITS 1 & 2 8 3/4 4-2a Unit 1 - Amendment No. 66,82 Unit 2 - Amendment No. 44

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REACTOR COOLANT SYSTEM RASES i

STEAM GENERATORS (Continued) l Wastage-type defects are unlikely with pro >er chemistry treatment of the secondary coolant. However, even if a defect s sould develop in service, it will be found during scheduled inservice steam generator tube examinations.

Plugging will be required for all tubes with imperfections exceeding the j

plugging limit of 4M of the tube nominal wall thickness. Steam generator j

tube inspections of operating plants have demonstrated the capability to i

reliably detect degradation that has penetrated 2M of the original tube wall

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thickness.

i Exclusion of certain areas of Unit I tubes from consideration has been i

analyzed using an F* criteria.

The criteria allows service induced i

degradation deep within the tubesheet to remain in service. The analysis methodology determines the length of sound fully rolled expanded tubing i

required in the uppermost area within the tubesheet to preserve needed j

structural margins for all service conditions. The rema,inder of the tube,

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below the F* distance, is considered not structurally relevant and is excluded i

from consideration to the customary plugging criteria of 40% throughwall.

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 secondary leakage under accident conditions. Primary to secondary leakage j

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.

i Whenever the results of any steam generator tubing inservice inspection i

fall into Category C-3, these results will be promptly reported to the l

Commission in a Special Report pursuant to Specification 6.9.2 within 30 days i

and prior to resumption of plant operation. Such cases will be considered by j

the Commission on a case-by-case basis and may result in a requirement for s

analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE

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3/4.4.6.1 LEAKAGE DETECTION SYSTEMS l

The RCS Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure

>oundary. These Detection Systems are consistent with the recommendations of i

Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection i

Systems," May 1973.

j 3/4.4.6.2 OPERATIONAL LEAKAGE j

PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Therefore, f

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

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REACTOR C0OLANT* SYSTEM i

RASES l

OPERATIONAL LEAKAGE (Continued) i the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly j

placed in COLD SHUTDOWN.

l Industry experience has shown that while a limited amount of leakage is i

expected from the RCS, the unidentified portion of this leakage can be reduced l

to a threshold value of less than 1 ppa. This threshold value is sufficiently low to ensure early detection of add'tional leakage.

l The total steam generator tube leakage limit of I gpa for all steam

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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 guidelipe 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 i

event of a main steam line rupture or under u'CA conditions.

t The 10 gpm IDENTIFIED LEAKAGE limitation providet allowance for a limited amount of 1c.1kage from known sources whose presence will not interfert with the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems.

Tne 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 assur.ince of valve integrity thereby reducing the protiability of gross valve failur and consequent intersystem LOCA.

Leakage from the RCS pressure isolation vake 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 chmistry ensure that corrosion of the Reactor Coolant System is minimized and rMuces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining SOUTH TEXAS - LNITS 1 & 2 B 3/4 4-4 m