ML20198S274
| ML20198S274 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 01/06/1999 |
| From: | Jacob Zimmerman NRC (Affiliation Not Assigned) |
| To: | Dennis Morey SOUTHERN NUCLEAR OPERATING CO. |
| References | |
| TAC-MA3342, NUDOCS 9901110233 | |
| Download: ML20198S274 (7) | |
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January 6, 1999 Mr. D. N. Morey Vice President - Farley Project Southern Nuclear Operating Company, Inc.
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Post Office Box 1295 1
Birmingham, Alabama 35201-1295
SUBJECT:
JOSF SH M. FARI:' Y NUCLEAR PLANT, UNIT 2 - REVIEW OF STEAM GENERATOR 90-0AV HEPORT (TAC NO MA3342)
Dear Mr. Morey:
By letter dated August 7,1998, Southern Nuclear Operating Company, Inc. (SNC), submitted its Farley Nuclear Plant, Unit 2, steam generator 90-day report, "Farley Unit 21998 Voltage-Based Repair Criteria 90 Day Report." The report summarizes the results of SNC's assessment of the eddy current inspection results with respect to the guidance established for voltage-based tube repair criteria applied to indications located at the tube support plate intersections, and attributed to outside diameter stress corrosion cracking.
The staff has reviewed your submittal and found the assessment to be acceptable. The staff's review is provided in the enclosure.
i Sincerely, ORIGINAL SIGNED BY-Jacot: 1. Zimmerman, Project Manager 9901110233 990106 Project Directorate li-2 ADOCK0500g4 Division of Reactor Projects - I/II i
PDR P '.
Office of Nuclear Reactor Regulation 4
Docket No. 50-364
Enclosure:
As stated L
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January 6, 1999 Mr. D. N. Morey Vice President - Farley Project Southern Nuclear Operating Company, Inc.
Post Office Box 1295
. Birmingham, Alabama 35201-1295
SUBJECT:
JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 2 - REVIEW OF STEAM GENERATOR 90-DAY REPORT (TAC NO. MA3342)
Dear Mr. Morey-By letter dated August 7,1998, Southern Nuclear Operating Company, Inc. (SNC), submitted its Farley Nuclear Plant, Unit 2, steam generator 90-day report, "Farley Unit 21998 Voltage-Based Repair Criteria 90 Day Report." The report summarizes the results of SNC's assessment of the cddy current inspection results with respect to the guidance established for voltage-based tube repair criteria applied to indications located at the tube support plate intersections, and attributed to outside diameter stress corrosion cracking.
1 The staff has reviewed your submittal and found the assessment to be acceptable. The staff's
- review is provided in the enclosure.
Sincerely, acob 1. Zimmerman, Project Manager Project Directorate l'-2 Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation m
l Docket No. 50-364
Enclosure:
As stated cc: w/enci: See next page i
4 4
Joseph M. Farley Nuclear Plant.
cc:
' Mr. L'. M. Stinson Rebecca V. Badham General Manager.
SAER Supervisor
. Southern Nuclear Operating Company Southern Nuclear Operating Company j
~ Post Office Box 470 P. O. Box 470
' Ashford, Alabama 36312 '
Ashford, Alabama 36312 l
l Mr. Mark Ajiuni, Licensing Manager j
Southern Nuclear Operating Company
. Post Office Box 1295 Birmingham, Alabama 35201-1295 Mr. M. Stanford Blanton.
Balch and Bingham Law Firm Post Office Box 306 1710 Sixth Avenue North Birmingham, Alabama 35201-Mr. J. D. Woodard
~ Executive Vice President l
Southern Nuclear Operating Company -
Post Office Box 1295 Birmingham, Alabama 35201 State Health Officer Alabama Department of Public Health 434 Monroe Street -
Montgomery, Alabama 36130-1701 Chairman Houston County Commission l
Post Office Box 6406
'Dothan Alabama 36302 -
Regional Administrator, Region 11 l
~ U.S. Nuclear Regulatory Commission L
Atlanta Federal Center 61 Forsyth Street, S.W., Suite 23T85 Atlanta, Georgia 30303 i
Resident Inspector L
U.S. Nuclear Regulatory Commission 7388 N. State Highway 95 '
Columbia, Alabama 36319 J
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l REVIEW OF THE FARLEY UNIT 2 STEAM GENERATOR 90-DAY REPORT l
In a letter dated August 7,1998, Southern Nuclear Operating Company, Inc. (SNC) submitted its steam generator (SG) 90-day report, "Farley Unit 21990 Veltage-Based Repair Criteria 90-day Report"(Reference 1). The staff reviewed the rabmittal using criteria from References 2 and 3 and found SNC's assessment te ' e acceptable.
o 1.0 GENERAL PLANT DESCRIPTION The Joseph M. Farley Nuclear Plant, Unit 2 (Farley-2) has three Westinghouse model 51 steam generators (SGs) with 7/8-inch diameter tubes. During the last refueling outage at the end-of-cycle 12 (EOC-12), SNC implemented a voltage-based Alternate Repair Criteria (ARC) to be applied to outside diameter stress corrosion cracking (ODSCC) at the tube support plate (TSP) intersections. References 2 and 3 describe the ARC methodology in more detail.
SNC used a lower voltage repair limit of 2.0 volts end determined an upper voltage repair limit of 5.3 volts to disposition ODSCC indications at TSP intersections. SNC left in service l
indications with bobbin coil voltages less than or equal to 2.0 volts. SNC removed from service indications with bobbin coil voltages greater than 5.3 volts (although there were no such indications found during this inspection) and also removed from service indications with bobbin coil voltages between 2.0 and 5.3 volts if confirmed with a rotating pancake coil (RPC) probe.
2.0 STEAM GENERATOR TUBE EDDY CURRENT INSPECTION SCOPE AND RESULTS SNC inspected 100 percent of the Farley-2 SG tubes fulllength using a 0.720-inch diameter bobbin coil probe at allintersections at which the voltage-based ARC were applied. SNC used an RPC probe to inspect 100 percent of the indications with bobbin coil voltages greater than
'2.0 volts in all three SGs. SNC reported a total of 33 indications with bobbin coil voltages that axceeded the 2.0 volt criteria; two were confirmed with the RPC probe and subsequently plugged.
SNC reported a total of 507 ODSCC indications at TSP intersections and returned 497 indications to service at Farley-2. Of the ten ODSCC indications removed from service, eight indications were in tubes plugged for degradation mechanisms other than ODSCC at the TSPs. The remaining two indications were above the 2.0-volt limit and were confirmed with the RPC probe. Neither indication was above the upper voltage limit of 5.3 volts, in past inspections, SNC dispositioned TSP indications with near 0 percent depth (regardless of voltage) as " unusual outside diameter phase angles" (UOAs) and left such indications in service if not confirrned by RPC. In a request for additional information dated February 5,1998, the staff requested that SNC provide additional information to justify its practice of leaving UOAs in service (e.g., tube pulls, eddy current historical reviews, RPC inspection results, etc.). SNC responded in Appendix A to Reference 1. SNC performed RPC sampling of UOAs each refueling outage and found little, if any, RPC confirmation, and on this basis, left such indications in service. This practice is not consistent with staff expoctations. No tube pulls Enclosuro l
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. i have been done to confirm the existence or nonexistence of defects associated with UOAs.
For the current refueling outage, SNC discontinued its past practice of categorizing eddy current indications at the TSPs as either " potential indications" and subject to the 2.0-volt repair l
criteria or UOAs, which were not repaired. SNC now dispositions all TSP outside diameter indications in accordmice with the 2.0-volt repair criteria, which is consistent with staff expectations.
SNC identified through its RPC inspections an ODSCC flaw at TSP 2H in SG "B." The bobbin l
signal for the indication had a high residual signal and did not identify the flaw. SNC expanded f
the RPC inspection of support plate residuals to 66 more intersections in SG "B." SNC did not identify any other ODSCC indications. SNC repaired the subject indication.
The staff concludes that SNC's bobbin and RPC probe inspections were consistent with the l
guidance in References 2 and 3 and, thus, are acceptable.
l 3.0 COMPARISON BETWEEN ACTUAL AND PREDICTED EOC-12 CONDITIONAL PROBABILITY OF BURST AND TOTAL LEAK RATE UNDER POSTULATED MAIN STEAMLINE BREAK CONDITIONS l
In its previous 90-day report (Reference 4), SNC found that tubes that were deplugged at l
EOC-10 and returned to service for cycle 11 exhibited much higher growth tates than tubes that l
have always been in service. Because SNC applied a single growth rate distribution to all l
indications (i.e., SNC assumed growth rates for tubes that have always been in service would j
be applicable to deplugged tubes), SNC obtained a nonconservative prediction of the EOC-11 j
voltage distribution for the limiting Farley-2 SG "C." in its current 90-day report, SNC reported that higher growth rates associated with deplug3ad tubes was not sustained. SNC's prediction for the EOC-12 bobbin voltage distributions was found to be a conservative representation of the actual EOC-12 bobbin voltage distribution.
The overall shape of the actual distribution of vol. ages was similar to the predicted distribution of voltages. SNC overpredicted the total number of indications, the number of indications greater than 1.0 volt, and the peak voltage indication for all three SGs. SNC underpredicted the number of indications less than 1.0 volt. Licensees implementing voltage-based repair criteria typically underpredict the low voltage indications. The underprediction is not significant because of the negligible contribution to tube burst probability and accident leak rate from indications less than 1.0 volt.
SNC calculated the conditional probability of burst and the total leak rate under postulated main steamline break (MSLB) conditions using the actual EOC-12 bobbin voltage distribution and then compared these values to those predicted for EOC-12. The limiting conditional tube burst l
probability for one tube was 2.3 x 10" based on the actual EOC-12 voltage distribution l
compared with 2.0 x 10 8 based on the predicted EOC-12 voltage distribution. The limiting MSLB !eak rate was 2.4 gallons per minute (gpm) based on the actual EOC-12 voltage distribution compared with 9.3 gpm based on the predicted EOC-12 voltage distribution. The j
actual values in both cases were below the technical specification reporting threshold values of 1 x 10-2 for the conditional probability of burst and 23.8 gpm for the accident leak rate. All leak rates are ai reference to room temperature conditions.
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3-4.0 TUBE INTEGRITY EVALUATIONS FOR EOC-13 For the EOC-13 predictions of the voltage distribution of indications, SNC used the most limiting growth rates observed during the last two inspection cycles. More specifically, SNC used the cycle 11 SG "C" growth distribution to predict for SG "C" and used the cycle 11 composite SG growth rates to predict for SGs "A" and "B."
The conditional probability of burst refers to the probability that the burst pressures associated with one or more indications in the faulted SG will be less than the maximum pressure differential associated with a postulated MSLB assumed to occur at EOC. For ODSCC at the TSP intersectiors, the staff considers an acceptable level of structural margin consistent with the applicable General Design Criteria (GDC) of 10 CFR Part 50, Appendix A to be met with a conditional burst probability of less than 1 x 102 SNC performed this assessment using a methodology previously approved by the NRC staff in Reference 2. SNC reported a limiting d
burst probability of 6.5 x 10, which is below the threshold value of 1 x 102 and is, therefore, acceptable.
The predicted MSLB leak rate is calculated to ensure leakage from indications under worst case MSLB conditions will not result in offsite and control room dose releases that exceed the guidelines of 10 CFR Part 100, GDC 19, and the Farley-2 site-specific limit. SNC performed this assessment using a methodology previously approved by the NRC staff in Reference 2.
SNC reported a limiting MSLB leak rate of 1.4 gpm, which is below the site's maximum allowable leak rate of 23.8 gpm (room temperature conditions), and is, therefore, acceptable.
(The staff notes that the value of 1.4 gpm is actually less than the limiting EOC-12 accident leak rate because the former value was obtained using a leak rate versus bobbin coil voltage correlation whereas the latter value was obtained with leak rates independent of voltage.)
5.0 TUBE PULL RESULTS Reference 3 requires periodic tube pulls to confirm the ODSCC morphology and to obtain additional data for inclusion in the correlations relating bobbin coil voltage amplitude to tube burst pressure, probability of leakage, and leak rate. SNC did not pull tubes at this outage.
SNC removed one tube from SG "C" during the previous EOC-11 refueling outage as discussed in Reference 4. SNC is following the tube pull scheduling guidance in References 2 and 3; therefore, the staff finds this acceptable, i
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o i References
- 1. "Farley Unit 21998 Voltage Based Repair Criteria 90 Day Report," Westinghouse Electric Corporation, SG 98-07-012, July 1998.
- 2. Letter from J. l. Zimmerman (NRC) to D. N. Morey (SNC), " Issuance of Amendment -
Joseph M. Farley Nuclear Plant, Unit 2 (TAC NO. M95146)," dated October 11,1996.
? Generic Letter 95-05, " Voltage-Based Repair Criteria for Westinghouse Steam Generator ubes Affected by Outside Diameter Stress Corrosion Cracking," August 3,1995.
- 4. "Farley Unit 21996 Alternate Repair Criteria 90 Day Report," Westinghouse Electric Corporation,' SG-97-03-001, March 1997.
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