ML20217Q419
| ML20217Q419 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 08/29/1997 |
| From: | Dennis Morey SOUTHERN NUCLEAR OPERATING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20217Q420 | List: |
| References | |
| GL-95-05, GL-95-5, NUDOCS 9709030073 | |
| Download: ML20217Q419 (2) | |
Text
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Dave Morey S:uthern Nucleat 4
%ce Pre @nt Operating Company Failey Project -
P.O. Box 1295 j
Birmingham./Jabama 35201 Tel 205.992 5131 August 29, 1997
[
k SOUTHERN h COMPANY Energy to Sme YourWorld' Docket No.:
50-348 U. S. Nuclear Regulatory Commission ATIN.: Document Control Desk Washington, DC 20555 Joseph M. Farley Nuclear Plant - Unit i Sicas Ocncrator Tube Voltage-Based Mtnuate Repair Criteria _ Data Ragt Ladies and Gentlemen:
Generic Letter 95-05, " Voltage-Based Repair Criteria for the Repair of Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking," requires the submittal of an associated report within 90 days of the restart following a steam generator inspection. The required report for the Spring 1997 Farley Unit I inspection is attached.
As reported earlier, an approximately 14 volt bobbin indication wm found at the first tube support plate of steam generator C. Although outside the nomial region where tube pulls are attempted, the tube was successfully pulled and has been metallurgically examined as discussed in Section 3,0 of the attached report. Burst pressure of the tube exceeded the 1.4 X steam line break differential pressure thereby meeting the requirements of Generic L etter 95-05. Currently, the rapid growth in bobbin voltage is possibly attributed to two causes:
- 1. Stresses imparted by pressure pulse cleaning changed the mechanical ch 'Jacteristics of an existing flaw resulting in increasing the magnitude of *he bobbin voltage, reducing the burst pressure, and/or increasing the associated leak rate; or
- 2. A lead related corrosion mechanism causing rapid corrosion, o
The calculations contained in the attached report assume that the 14 volt indicr. tion was the result of outside dianuter stress wrmsion cracking; although evidence exists that the magnitude of the bobbin voltage may have been afTected by pressure pulse cleaning. Because of the currently planned repl cement ot'the Farley steam generators, pressure pulse cleaning is not currently scheduled for the next Unit I outage.
Additionally as reported earlier, an inconsistency was discovered in the methodologies used to develop the steam line break primany-to-secondary leakage limit and to project leakage in the event of a steam line break. Temperature corrections were not applied consistently for the two methodologies. Based on the dose equivalent iodine limit approved dunng the last outage, the room temperature leakage limit for operational assessment is 13.7 gpm (19gpm at operating temperature).
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0 Leakage projections are calculated using the standard approved methodology of WCAP 14277, Revision 1. The results of an alternative method submitted as part of EPRI Report NP 7480-L,
" Steam Generator Outside Diameter Stress Cracking at Tube Support Plates Database for Alternate Repair Criteria," October 1996, is also provided for information. While the WCAP-14277 methodology results in a 15.7 gpm projected endef-cycle operational assessment leak rate, the alternate method results in an 11.4 gpm projected leak rate. It should be noted that use of the database proposed by NEl/EPRI in September 1996 results in a projected leak rate of 3.8 gpm. As a result of the projected steamline break leakage exceeding the 13.7gpm limit at approximately 305 EFPD of operation, Southern Nuclear will work with the NRC staff to address any concerns.
Po sible approaches include approval of the NEl/EPRI proposed database, approval of altemative calculation methodologies, or tecimical specifications amendments changes for dose equivalent i-dine.
As stated in the attached report, calculations are performed using the NRC-approved burst and leakage databases. liowever, the inclusion of the Unit i pulled tube data point will have a significant impact on these calculations. Calculations incorporating both the last Farley Unit I and Unit 2 pulled tubes are currently in progress. The results of these calculations will be provided to the NRC Staff on completion (no more than 60 days of the date of this letter).
If there are any questions, please advise.
Respectfully submitted, SOUTIIERN NUCLEAR OPERATING COMPANY g7}
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Dave Morey REM maf 90DAYNRC. DOC Attachment cc:
Mr. L. A. Reyes, Region 11 Administrator Mr. J.1. Zimmerman, NRR Project Manager Mr. T. M. Ross, Plant Sr. Resident hispector Dr. D. E. Williamson, State Department of Public Health 3