ML073310253

Request for Additional Information Regarding Generic Letter Nos. 95-05, 95-05, 97-05, 97-06, 2004-01, and 2006-01 Pertaining to Stream Generator Tube Integrity

ML073310253
Person / Time
Site:	Watts Bar
Issue date:	11/27/2007
From:	Chernoff M Plant Licensing Branch III-2
To:	Bhatnagar A Tennessee Valley Authority
Chernoff M, NRR/DORL, 415-4041
References
GL-04-001, GL-06-001, GL-95-003, GL-95-005, GL-97-005, GL-97-006, TAC MD6715, TAC MD6716, TAC MD6720, TAC MD6721
Download: ML073310253 (7)

Text

November 27, 2007 Mr. Ashok S. Bhatnagar Senior Vice President Nuclear Generation Development and Construction 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801

SUBJECT:

WATTS BAR NUCLEAR PLANT, UNIT 2 - REQUEST FOR ADDITIONAL INFORMATION REGARDING GENERIC LETTER NOS. 95-03, 95-05, 97-05, 97-06, 2004-01, AND 2006-01 PERTAINING TO STEAM GENERATOR TUBE INTEGRITY (TAC NOS. MD6715, MD6716, MD6720, MD6721, MD6725, AND MD6727)

Dear Mr. Bhatnagar:

By letter dated September 7, 2007 (ADAMS Accession No. ML072570676), the Tennessee Valley Authority (TVA) submitted responses to several bulletins and generic letters (GLs). The Nuclear Regulatory Commission staff has reviewed TVAs responses to GL 95-03, Circumferential Cracking of Steam Generator Tubes, GL 95-05, Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking, GL 97-05, Steam Generator Tube Inspection Techniques, GL 97-06, Degradation of Steam Generator Internals, GL 2004-01, Requirements for Steam Generator Tube Inspections, and GL 2006-01, Steam Generator Tube Integrity and Associated Technical Specifications, and determined that additional information is required in order to complete the evaluation. The specific information is delineated in the enclosure to this letter.

This request for additional information was discussed with Mr. Gordon Arent and others of your staff on November 26, 2007. It was agreed that a response would be provided by January 11, 2008.

If you have questions regarding these information requests, please contact me at 301-415-4041.

Sincerely,

/RA/

Margaret H. Chernoff, Senior Project Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-391

Enclosure:

Request for Additional Information cc w/encl: See next page

ML072570676), the Tennessee Valley Authority (TVA) submitted responses to several bulletins and generic letters (GLs). The Nuclear Regulatory Commission staff has reviewed TVAs responses to GL 95-03, Circumferential Cracking of Steam Generator Tubes, GL 95-05, Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking, GL 97-05, Steam Generator Tube Inspection Techniques, GL 97-06, Degradation of Steam Generator Internals, GL 2004-01, Requirements for Steam Generator Tube Inspections, and GL 2006-01, Steam Generator Tube Integrity and Associated Technical Specifications, and determined that additional information is required in order to complete the evaluation. The specific information is delineated in the enclosure to this letter.

This request for additional information was discussed with Mr. Gordon Arent and others of your staff on November 26, 2007. It was agreed that a response would be provided by January 11, 2008.

If you have questions regarding these information requests, please contact me at 301-415-4041.

Sincerely,

/RA/

Margaret H. Chernoff, Senior Project Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-391

Enclosure:

Request for Additional Information cc w/encl: See next page Distribution: Rids Ogc Rp RidsAcrsAcnwMailCenter NON-PUBLIC RidsRgn2MailCenter RidsNrrDciCsgb LRaghavan RidsNrrLABClayton RidsNrrPMJWilliams RidsNrrPMMChernoff WBN-2 Reading File Accession Number: ML073310253 OFFICE NRR NRR DCI/CSGB NRR NAME MChernoff BClayton AHiser memo LRaghavan DATE 11/ 27 /07 11/ 27 /07 10/ 26 /07 11/ 27 /07 Ashok S. Bhatnagar Tennessee Valley Authority WATTS BAR NUCLEAR PLANT cc:

Mr. Gordon P. Arent Mr. John C. Fornicola, General Manager New Generation Licensing Manager Nuclear Assurance Tennessee Valley Authority Tennessee Valley Authority 5A Lookout Place 3R Lookout Place 1101 Market Street 1101 Market Street Chattanooga, TN 37402-2801 Chattanooga, TN 37402-2801 Mr. Masoud Bajestani, Vice President General Counsel Watts Bar Unit 2 Tennessee Valley Authority Watts Bar Nuclear Plant 6A West Tower Tennessee Valley Authority 400 West Summit Hill Drive P.O. Box 2000, EQB 1B Knoxville, TN 37902 Spring City, TN 37381 Mr. Michael J. Lorek, Plant Manager Senior Resident Inspector Watts Bar Nuclear Plant Watts Bar Nuclear Plant Tennessee Valley Authority U.S. Nuclear Regulatory Commission P.O. Box 2000 1260 Nuclear Plant Road Spring City, TN 37381 Spring City, TN 37381 Mr. Lawrence E. Nanney, Director Mr. Michael K. Brandon, Manager Tennessee Dept. of Environmental Health &

Licensing and Industry Affairs Conservation Watts Bar Nuclear Plant Division of Radiological Health Tennessee Valley Authority 3rd Floor, L&C Annex P.O. Box 2000 401 Church Street Spring City, TN 37381 Nashville, TN 37243-1532 Mr. William R. Campbell Mr. Larry E. Nicholson, General Manager Chief Nuclear Officer and Performance Improvement Executive Vice President Tennessee Valley Authority Tennessee Valley Authority 4X Blue Ridge 3R Lookout Place 1101 Market Street 1101 Market Street Chattanooga, TN 37402-2801 Chattanooga, TN 37402-2801 Mr. Michael D. Skaggs County Executive Site Vice President 375 Church Street Watts Bar Nuclear Plant Suite 215 Tennessee Valley Authority Dayton, TN 37321 P.O. Box 2000 Spring City, TN 37381 County Mayor P. O. Box 156 Decatur, TN 37322

REQUEST FOR ADDITIONAL INFORMATION WATTS BAR NUCLEAR PLANT, UNIT 2 RESPONSES TO VARIOUS GENERIC LETTERS PERTAINING TO STEAM GENERATOR TUBE INTEGRITY TAC NOS. MD6715, MD6716, MD6720, MD6721, MD6725, MD6727 DOCKET NO. 50-391 By letter dated September 7, 2007 (ADAMS Accession No. ML072570676), the Tennessee Valley Authority (TVA) submitted responses to several bulletins and generic letters (GLs). The Nuclear Regulatory Commission (NRC) staff has reviewed TVAs responses to GL 95-03, Circumferential Cracking of Steam Generator Tubes, GL 95-05, Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking, GL 97-05, Steam Generator Tube Inspection Techniques, GL 97-06, Degradation of Steam Generator Internals, GL 2004-01, Requirements for Steam Generator Tube Inspections, and GL 2006-01, Steam Generator Tube Integrity and Associated Technical Specifications. In order for the staff to complete its review of these GL responses, the following information is needed.

GL 95-03: Circumferential Cracking of Steam Generator Tubes

1. With respect to the Westinghouse Model D3 steam generators that are to be used in Watts Bar Unit 2, please address the following:

a. Discuss whether all tubes were subject to rotopeening at the hot-leg expansion transition and shotpeening at the cold-leg expansion transition.

b. Discuss whether the U-bend region of the low row tubes (e.g., rows 1 and 2) were heat treated.

c. Discuss the extent to which any tubes were expanded into the tube support plates. In addition, confirm that the design modifications discussed in NUREG-0966, Safety Evaluation Report related to the D2/D3 Steam Generator Design Modification, were implemented at Watts Bar Unit 2.

2. It was indicated that 100 percent of the tubes would be inspected prior to fuel load.

Please discuss the specific probe types that will be used during this inspection (e.g., will the full length of 100 percent of the tubes be examined with a bobbin coil? Will 100 percent of the tubes be inspected at both the hot-leg and cold-leg expansion transition with a rotating probe equipped with a +PointTM coil?).

Enclosure

3. Please discuss the extent to which circumferential cracking could occur in larger radius U-bends (i.e., larger than those in row 2). Please refer to NRC Information Notice 2003-13, Steam Generator Tube Degradation at Diablo Canyon. If circumferential cracking can occur at these locations, discuss your plans to inspect these locations with a probe capable of detecting this form of degradation.

4. Please discuss the extent to which circumferential cracking could occur in dings. If circumferential cracking can occur at these locations, discuss your plans to inspect these locations with a probe capable of detecting this form of degradation.

5. With respect to your inspection program for dents, you indicated that a typical inspection plan would be to inspect a 20 percent sample of the dents greater than or equal to 5 volts (as determined with a bobbin coil) with a technique qualified for crack detection at dents. Please provide the basis for limiting this sample to the first two support plates on the hot-leg side of the steam generator given that these locations are susceptible to cracking and the bobbin coil is not qualified for inspecting dents greater than 5 volts (i.e., shouldnt the 20 percent sample be for all dents greater than 5 volts?). In addition, discuss the extent to which circumferential cracking could occur in dents less than 5 volts. If circumferential cracking can occur at these locations, discuss your plans to inspect these locations with a probe capable of detecting this form of degradation. If your scope is limited to just the lower hot-leg tube support plates, discuss the basis for this approach (given more recent operating experience that indicates cracking can occur in dings on the cold leg before it is observed on the hot-leg since the potential for cracking relies on not only the temperature, but also the stresses and the material, which can vary).

6. If any tubes were expanded into the tube support plates (refer to previous question),

please discuss the extent to which circumferential cracking could occur at these locations. If circumferential cracking can occur at these locations, discuss your plans to inspect these locations with a probe capable of detecting this form of degradation.

7. Please clarify the statement that the minimum inspection scope at the first refueling outage will be based on the current version of the Technical Specifications and Electric Power Research Institute (EPRI) guidelines. In particular, clarify that the current version of the technical specifications that you are referring to are essentially identical to those contained in Technical Specification Task Force Traveler 449 (TSTF-449),

Revision 4, Steam Generator Tube Integrity.

8. Clarify the statement that if cracking is detected, the examination expansion requirements of the technical specifications will be fulfilled. The staff notes that there are no sample expansion requirements in the technical specifications that are modeled after TSTF-449 (which the staff is under the impression you will adopt prior to commencing commercial operation per your response to GL 2006-01, Steam Generator Tube Integrity and Associated Technical Specifications). Rather, technical specifications modeled after TSTF-449 require inspections to be performed to ensure tube integrity.

9. During the Maine Yankee outage in July/August 1994, several weaknesses were identified in their eddy current program as detailed in NRC Information Notice 94-88, Inservice Inspection Deficiencies Result in Severely Degraded Steam Generator Tubes. In Information Notice 94-88, the NRC staff observed that several circumferential indications could be traced back to earlier inspections when the data was reanalyzed using terrain plots. These terrain plots had not been generated as part of the original field analysis for these tubes. For the rotating probe (or equivalent) examinations to be performed at your plant, discuss the extent to which terrain plots will be used to analyze the data at locations susceptible to circumferential cracking.

GL 97-05, Steam Generator Tube Inspection Techniques Regarding the types of indications that you will size and leave in service, please confirm that the only indications that you size for purposes of leaving in service are wear attributed either to loose parts or to support structures (i.e., antivibration bars or tube support plates). Confirm that these wear indications are sized using methods that were evaluated using the EPRI Pressurized Water Reactor Steam Generator Examination Guidelines: Revision 6.

GL 97-06, Degradation of Steam Generator Internals In your March 30, 1998, response to GL 97-06, you indicated that one end of the blowdown pipe in two of the four steam generators in Watts Bar Unit 1 had severed after operating one cycle.

Please discuss whether a similar situation could occur at Watts Bar Unit 2. If so, discuss any corrective action that you have planned and/or any inspection program for assessing this potential degradation mechanism. If no corrective action is planned prior to commencing commercial operation, please describe the basis for concluding that operating with this condition is acceptable.