ML051810711

Ltr., SG ISI Reports Summary for Cycle 12 (Tac No. MC4545)

ML051810711
Person / Time
Site:	Watts Bar
Issue date:	07/08/2005
From:	Pickett D NRC/NRR/DLPM/LPD2
To:	Singer K Tennessee Valley Authority
Pickett D, NRR/DLPM, 415-1364
References
GL-95-005, TAC MC4545
Download: ML051810711 (6)

Text

July 8, 2005 Mr. Karl W. Singer Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, Tennessee 37402-2801

SUBJECT:

SEQUOYAH NUCLEAR PLANT, UNIT 2 REVIEW

SUMMARY

OF THE STEAM GENERATOR TUBE INSERVICE INSPECTION REPORTS FOR THE END-OF CYCLE 12 REFUELING OUTAGE IN 2003 (TAC NO. MC4545)

Dear Mr. Singer:

In a letter dated December 12, 2003, Tennessee Valley Authority (TVA) submitted the 15-day steam generator (SG) plugging report for Sequoyah Nuclear Plant, Unit 2, in accordance with Technical Specification 4.4.5.5.a. By letter dated March 9, 2004, TVA submitted the 90-day SG voltage-based alternate repair criteria report in accordance with Sequoyah Unit 2 License Condition 2.C.(8)(b) and U.S. Nuclear Regulatory Commission (NRC) Generic Letter 95-05, , Section 6.b. By letter dated September 20, 2004, TVA submitted the 12-month SG tube inspection report in accordance with TS 4.4.5.5.b.

On November 24, November 26, and December 1, 2003, NRC staff participated in conference calls with TVA to discuss the ongoing SG inspection activities for Sequoyah Unit 2 during their end-of-cycle 12 refueling outage. These conference calls were summarized in an NRC letter dated April 20, 2004. Finally, TVA submitted additional information concerning their 2004 outage by letter dated April 22, 2005. Enclosed is a brief summary of the review of the SG inservice inspection reports prepared by the NRC staff.

Sincerely,

/RA/

Douglas V. Pickett, Senior Project Manager, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-390

Enclosure:

As stated cc w/enclosure: See next page

ML051810711 NRR-106 OFFICE LPD2-2 PM:LPD2-2 LA:LPD2-2 SC:EMCB SC:LPD2-2 BMozafari for NAME MVaaler DPickett BClayton LLund MMarshall DATE 06/ 29/ 05 07/ 07/ 05 07/ 07/ 05 05 / 31 /05 07/ 08/ 05

Mr. Karl W. Singer Tennessee Valley Authority SEQUOYAH NUCLEAR PLANT cc:

Mr. Ashok S. Bhatnagar, Senior Vice President Mr. Paul L. Pace, Manager Nuclear Operations Licensing and Industry Affairs Tennessee Valley Authority ATTN: James D. Smith 6A Lookout Place Sequoyah Nuclear Plant 1101 Market Street Tennessee Valley Authority Chattanooga, TN 37402-2801 P.O. Box 2000 Soddy Daisy, TN 37384-2000 Mr. Larry S. Bryant, General Manager Nuclear Engineering Mr. David A. Kulisek, Plant Manager Tennessee Valley Authority Sequoyah Nuclear Plant 6A Lookout Place Tennessee Valley Authority 1101 Market Street P.O. Box 2000 Chattanooga, TN 37402-2801 Soddy Daisy, TN 37384-2000 Mr. Randy Douet Senior Resident Inspector Site Vice President Sequoyah Nuclear Plant Sequoyah Nuclear Plant U.S. Nuclear Regulatory Commission Tennessee Valley Authority 2600 Igou Ferry Road P.O. Box 2000 Soddy Daisy, TN 37379 Soddy Daisy, TN 37384-2000 Mr. Lawrence E. Nanney, Director General Counsel Division of Radiological Health Tennessee Valley Authority Dept. of Environment & Conservation ET 11A Third Floor, L and C Annex 400 West Summit Hill Drive 401 Church Street Knoxville, TN 37902 Nashville, TN 37243-1532 Mr. John C. Fornicola, Manager County Mayor Nuclear Assurance and Licensing Hamilton County Courthouse Tennessee Valley Authority Chattanooga, TN 37402-2801 6A Lookout Place 1101 Market Street Ms. Ann P. Harris Chattanooga, TN 37402-2801 341 Swing Loop Road Rockwood, Tennessee 37854 Mr. Fredrick C. Mashburn Senior Program Manager Nuclear Licensing Tennessee Valley Authority 4X Blue Ridge 1101 Market Street Chattanooga, TN 37402-2801

SUMMARY

OF STAFF REVIEW TENNESSEE VALLEY AUTHORITY STEAM GENERATOR TUBE INSPECTION REPORTS FOR THE END-OF-CYCLE 12 REFUELING OUTAGE SEQUOYAH NUCLEAR PLANT, UNIT 2 TAC NO. MC4545 By letter dated December 12, 2003 (ML033510702), Tennessee Valley Authority (TVA, the licensee) submitted the 15-day steam generator (SG) plugging report in accordance with Technical Specification (TS) 4.4.5.5.a. By letter dated March 9, 2004 (ML040710360), TVA submitted the 90-day SG voltage-based alternate repair criteria report in accordance with Sequoyah Nuclear Plant, Unit 2 License Condition 2.C.(8)(b) and U.S. Nuclear Regulatory Commission (NRC) Generic Letter (GL) 95-05, Attachment 1, Section 6.b. By letter dated September 20, 2004 (ML042720448), TVA submitted the 12-month SG tube inspection report in accordance with TS 4.4.5.5.b. On November 24, 2003, November 26, 2003, and December 1, 2003, NRC staff participated in conference calls with TVA to discuss the ongoing SG inspection activities for Sequoyah Unit 2. These conference calls were summarized in NRC letter dated April 20, 2004 (ML040680349, ML040680360, and ML040680369). TVA submitted additional information concerning their 2004 outage by letter dated April 22, 2005 (ML051120459).

Sequoyah Unit 2 has Westinghouse Model 51 SGs, which are designated SG1, SG2, SG3, and SG4. All four SGs were inspected during the November 2003 refueling outage. The Westinghouse Model 51 SG consists of approximately 3300 tubes which have an outside diameter of 7/8-inch and a wall thickness of 0.050-inch. The tubes have been explosively expanded into the tubesheet and are supported by several 3/4-inch carbon steel tube support plates which contain drilled holes through which the tubes pass. The Sequoyah Unit 2 SGs began operation in 1981 and have mill-annealed Inconel 600 tubing.

The licensee provided the scope, extent, methods, and results of Sequoyah Unit 2 SG tube inspections in the documents referenced above. In addition, the licensee described corrective actions (i.e., tube plugging or repair) taken in response to the inspection findings.

The staff has the following notes/observations as a result of reviewing the aforementioned submittals.

1. In their submittal, TVA indicated that they are currently evaluating a license amendment request to utilize the probability of prior cycle detection methodology Enclosure

in order to address the under prediction in the number of indications observed in the 2003 outage. The NRC staff notes that currently the safety implications of the under prediction in the number of indications are not significant. However, these under predictions may become significant as more and larger indications are left in service.

1. In their submittal, the licensee indicated that they implement the probe wear criterion outlined in GL 95-05 and letters between the NRC and the Nuclear Energy Institute in 1996. They also indicated that they would include a discussion of probe wear in future reports. The reporting criteria outlined in the 1996 letters include the following;

1. The 90-day report should include a comparison between the actual and projected end-of-cycle voltage distributions and a root cause should be evaluated and reported to the NRC if any significant differences exist (i.e.,

number of indications, distribution of indications, size of largest indication, etc.).

Probe wear should be considered in this root cause and if determined to be a contributing factor for the difference, then corrective actions should be taken to prevent recurrence.

1. Tubes identified since the last successful probe wear check with indications above 75-percent of the repair limit should be reinspected full-length with a new probe and all of the corresponding data should be evaluated (not just for the indications above 75-percent of the repair limit). An assessment of the significance of large indications that were detected with the good probe but missed by the worn probe should be performed during the outage and reported.

1. A significance assessment of large indications and/or a nonproportionate number of new indications identified in tubes that were inspected with a worn probe during the previous outage should be performed to determine the need for the probe wear criteria to be more restrictive.

1. Data should be continuously evaluated during the outage to ensure the adequacy of the 75-percent criteria and the analyses should be reported.

1. In their submittal the licensee indicated that (a) TVA performs SG growth rate predictions in accordance with Westinghouse WCAP-14277, Revision 1 methodology, (b) TVA believes probe wear is not a major contributor to the probability of burst (POB) and accident induced leakage under predictions, (c) TVA does not believe that mixed residuals are a major contributor to the POB and accident induced leakage under predictions due to the mixed residual indications being investigated each outage, (d) Sequoyah does not place axial outside diameter stress corrosion cracking indications at tube support plates back into service by de-plugging tubes, and (e) cycle to cycle growth rate increases are not considered the most significant cause of POB and accident induced leakage under predictions. For many of these issues, the licensee did not provide a technical analysis to support their reason that these issues did not adversely affect their projections. The staff notes that several plants that

implement (or have implemented) a similar repair criteria to what is implemented at Sequoyah Unit 2 have observed these effects. However, given the margin to the POB and accident induced leakage limits, the staff finds this acceptable for the 2003 outage.

Based on a review of the information provided, the staff concludes that the licensee provided the information required by their TSs. In addition, the staff concludes that there are no technical issues that warrent follow-up action at this time since the inspections appear to be consistent with the objective of detecting potential tube degradation and the inspection results appear to be consistent with industry operating experience at similarly designed and operated units.