ML20065C660

Proposed Tech Specs Re Steam Generator Tube Support Plate Interim Repair Criteria

ML20065C660
Person / Time
Site:	Farley
Issue date:	04/01/1994
From:	SOUTHERN NUCLEAR OPERATING CO.
To:
Shared Package
ML20065C658	List: ML20065C655 ML20065C660
References
NUDOCS 9404050271
Download: ML20065C660 (1)

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. REACTOR COOLANT SYSTEM BASES ,

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3/4.4.6 STEAM GENERATORS, The Surveillance Requirements for inspection of the steam generator tubes

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ene9te that the structural integrity of this portion of the RCS will be ma2ntained. 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 ~,ans of characterizing the nature and cause of any tube degradation s. 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 result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likelv 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' primary coolant system and the secondary coolant system (primary-to-secondary leakage = 140 gallons per day per steam generator) .

- 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. Operational leakage of this magnitude can be readily detected by existing Farley Unit 1.

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

For the Thirteenth Operating Cycle only,-the. repair limit for tubes with flaw indications contained within the bounds of a tube support plate has been "

provided to the NRC in Southern Nuclear Operating Company letters dated December 09, 1993 and February 23, 1994. The repair limit is based on the analysis contained in WCAP-12871, Revision 2, "J. M.'Farley Units 1 and 2 SG Tube Plugging Criteria for ODSCC at Tube Support Plates," and documentation contained in EPRI Report TR-100407, Revision.1, "PWR Steam Generator Tube ,

Repa!r Limit: Trehnical Support Document f or Cut side Diameter SFicas Corrosion Cracking at- Tube Support Plates." The application of this criteria is based on limiting primary-to-secondary leakage during a steam line break to ensure the applicable Part 100 limits are not exceeded.

4 Wastage-type defects are unlikely with proper chemistry treatment of the l secondary coolant. However,:even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations. lj

-Plugging or repair will be required for all tubes with imperfections exceeding 40% of the tube nominal wall thickness. If a sleeved tube is found to have: H through wall penetration of greater than or equal to 31%'for the mechanical '

sleeve and 37V for the laser welded sleeve of sleeve nominal wall thickness in the sleeve, it must be plugged. The 31% and 37% limits are derived from R.G.

1.121' calculations with 20% added for conservatism. The portion'of the tube and the sleeve for which indications of wall degradation must be evaluated'can-be' summarized as follows: l FARLEY-UNIT 1 B3/4 4-3 AMENDMENT NO.

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