ML101480018

Request for Additional Information Regarding the 90-Day and 180-Day Steam Generator Tube Inspection Reports for Cycle 16 Refueling Outage

ML101480018
Person / Time
Site:	Sequoyah
Issue date:	06/07/2010
From:	Lingam S P Plant Licensing Branch II
To:	Krich R M Tennessee Valley Authority
Lingam, S NRR/DORL 415-1564
References
TAC ME3400, TAC ME3971
Download: ML101480018 (5)

Text

UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 June 7, 2010 Mr. Rodney 1\/1. Krich Vice President, Nuclear Licensing Tennessee Valley Authority 3R Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 SEQUOYAH NUCLEAR PLANT, UNIT 2 -REQUEST FOR ADDITIONAL INFORMATION REGARDING THE 90-DAY AND 180-DAY STEAM GENERATOR TUBE INSPECTION REPORTS FOR CYCLE 16 REFUELING OUTAGE (TAC NOS. ME3400 AND ME3971)

Dear Mr. Krich:

By letters dated February 19 and May 19, 2010, Tennessee Valley Authority submitted 90-day and 180-day steam generator tube inspection reports, respectively for the Cycle 16 refueling outage (fall 2009) in accordance with Technical Specification Section 6.9.1.16.2 for Sequoyah Nuclear Plant, Unit 2. The Nuclear Regulatory Commission staff is reviewing the submittal and has determined that additional information is required to complete its evaluation.

The enclosed request for additional information was discussed with Mr. Rodney Cook of your staff on June 2, 2010, and it was agreed that a response would be provided by July 16, 2010. If you have any questions regarding this matter, I can be reached at 301-415-1564.

Sincerely, Siva P. Lingam, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-328

Enclosure:

Request for Additional Information cc w/encl: Distribution via Listserv REOUEST FOR ADDITIONAL INFORMATION REGARDING THE CYCLE 16 90-DAYand 180-DAY STEAM GENERATOR TUBE INSPECTION REPORTS SEOUOYAH NUCLEAR PLANT, UNIT 2 DOCKET NO. 50-328 By letters dated February 19 and May 19, 2010 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML 100550767 and ML 101450411, respectively), Tennessee Valley Authority submitted the Cycle 16 refueling outage (fall 2009) 90-day and 180-day steam generator tube inspection reports, respectively, per Technical Specification Section 6.9.1.16.2 for Sequoyah Nuclear Plant (SON), Unit2. In addition to these reports, the U.S. Nuclear Regulatory Commission (NRC) staff summarized additional information concerning the 2009 SG tube inspections at SON, Unit 2 in a letter dated December 10, 2009 (ADAMS Accession No. ML093360644).

In order to complete its review of the above documents, the NRC staff needs the following additional information: Please discuss the scope and results of your secondary side steam drum inspections. Prior to shutting down for your fall 2009 steam generator tube inspections, a small primary-to-secondary leak rate existed. Please discuss whether any leakage was observed after starting up from the fall 2009 refueling outage (RFO). If so, discuss the possible source of any leakage and any implications for your inspection and repair criteria. On page E 1-10 of your February 19, 2010, letter, it was indicated that the voltage growth was determined based on the historic review of 3743 distorted support indications.

Please confirm this number since you detected 3747 indications at the supports, which included five indications that were only identified with a rotating probe. Please confirm that for every support indication identified during the fall 2009 RFO, you reviewed the historic data to determine if an indication was present; and, if an indication was present, you determined the growth rate of that indication and included it in your growth rate distribution. From Table 4-7 of your February 19, 2010, letter, the number of tubes examined on the cold-leg was calculated to be 3189. This is two tubes less than the number of tubes in service. Please discuss whether the cold-leg portion of each tube was examined with a bobbin coil probe. If not, discuss how this was accounted for in your operational assessment. In steam generators 2 and 3 (Tables 4-7 and 4-8 of your February 19, 2010, letter, respectively) the ratio of new indications in tubes tested with worn probes is higher than the ratio of new indications in tubes tested with good probes. This possibly indicates that the worn probes are missing indications (although the overall average from all four steam generators indicates that the ratios of these two quantities are comparable).

Please Enclosure

-2 discuss any corrective action taken in response to these results or discuss why no corrective action was needed. The largest indication of outside diameter stress corrosion cracking (ODSCC) at the tube support elevations grew from approximately

0.4 volts

in 2008 to 6.6 volts in 2009. The 0.4 volt indication in 2008 had been inspected with a worn probe. Please discuss any insights on the reason for such a high growth rate. For example, is the growth rate of indications in tubes inspected with a worn probe significantly higher than the growth rate of indications in tubes inspected with "good" probes (Le., probes that passed the probe wear criterion)? During the fall 2009 RFO, the voltages of two of the indications of outside diameter stress corrosion cracking at the tube support plates exceeded your projections.

The methodology for projecting the end-of-cycle voltage distribution for such indications was intended to be conservative in terms of projecting the number and severity of the flaws (and therefore conservative in estimating the accident induced leakage and burst probability).

This under prediction in the severity of the indications led to under predicting the burst probability in steam generator

4. Although no performance criteria were exceeded, the results appear to question the conservatisms of the methodology and may become a safety concern if your projections become closer to the performance criteria.

Given these results, please discuss whether any changes to your assessment methodology are needed to ensure your projections will be conservative. On page E2-2 of your February 19, 2010, letter, you reported the operational assessment leakage for "GL [Generic Letter] 95-05" flaws as 1.760 gallons per minute. This value does not match the most limiting value reported on page E 1-82 of that letter. Please clarify. Please clarify the first sentence of the second paragraph on page E2-2 of your February 19, 2010, letter. In particular, confirm that you assessed the leakage contribution from all primary water and outside diameter stress corrosion cracking indications at or below the top of the tubesheet in your condition monitoring and operational assessment.

10. Please discuss whether any of the tubes had the bottom of the WEXTEX transition located more than 2.88-inches below the top of the tubesheet.

If so, discuss how many tubes had this condition.

11. Please clarify the indication and location columns in Table 2 on page E2-5 of your February 19, 2010, letter. In particular address why there are two indication columns, two "location 1" columns, and two "location 2" columns. 12. Several ODSCC indications were reported in the tubesheet region. Please discuss whether these indications were below the bottom of the expansion transition.

If so, discuss how a corrosive environment was achieved below the bottom of the WEXTEX transition (e.g., did the tube lose contact with the tubesheet).

If the tube is not in contact with the tubesheet, discuss any implications to W*.

-13. Two indications were attributed to wear from a loose part in steam generator

4. A possible loose part signal was not evident in the eddy current data. Please discuss whether a visual inspection was performed to confirm the absence of a loose part at these locations.

Since a loose part may not be conductive or may be a small distance away from the tube and therefore not detected during the eddy current examination, discuss whether an assessment was performed for the continued wear of these tubes. If not, discuss why not. 14. On page 6 of 101 in the May 19, 2010, letter, it was indicated that twelve indications of axial OOSCC was detected in the free span region in three tubes. This appears to contradict the information on page 4 of 101 where four tubes are identified as being plugged for this degradation mechanism.

Is this difference a result of counting the one tube that was plugged for axial OOSCC in the sludge pile region as an axial OOSCC indication in the free span? If not, please explain the difference.

15. On page 8 of 101 in the May 19, 2010, letter, it was indicated that the U2C15 operational assessment predicted the limiting accident leakage to be 1.34 gallons per minute and the limiting burst pressure as 4.55E-4. These values do not appear to match those reported on page 5-6 of your February 19, 2010, letter. Please clarify. 16. Please clarify the nature of the geometry affect at the first tube support plate intersection in the tube that was preventively plugged.

Mr. Rodney M. Krich Vice President, Nuclear Licensing Tennessee Valley Authority 3R Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 SEQUOYAH NUCLEAR PLANT, UNIT 2 -REQUEST FOR ADDITIONAL INFORMATION REGARDING THE 90-DAY AND 180-DAY STEAM GENERATOR TUBE INSPECTION REPORTS FOR CYCLE 16 REFUELING OUTAGE (TAC NOS. ME3400 AND ME3971)

Dear Mr. Krich:

By letters dated February 19 and May 19, 2010, Tennessee Valley Authority submitted 90-day and 180-day steam generator tube inspection reports, respectively for the Cycle 16 refueling outage (fall 2009) in accordance with Technical Specification Section 6.9.1.16.2 for Sequoyah Nuclear Plant, Unit 2. The Nuclear Regulatory Commission staff is reviewing the submittal and has determined that additional information is required to complete its evaluation.

The enclosed request for additional information was discussed with Mr. Rodney Cook of your staff on June 2, 2010, and it was agreed that a response would be provided by July 16, 2010. If you have any questions regarding this matter, I can be reached at 301-415-1564.

Sincerely, IRA! Siva P. Lingam, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-328

Enclosure:

Request for Additional Information cc w/encl: Distribution via Listserv DISTRIBUTION:

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