ML063040307

Response to Request for Additional Information Regarding Report of Inservice Inspection of Steam Generator Tubes, Cycle 14

ML063040307
Person / Time
Site:	San Onofre
Issue date:	10/26/2006
From:	Katz B Southern California Edison Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC MC8395
Download: ML063040307 (6)

Text

SOUTHERN CALIFORNIA Brian Katz E DISON aVice President An EDISON INTERNATIONAL Company October 26, 2006 U. S. Nuclear Regulatory Commission Document Control Desk Washington, D. C. 20555

Subject:

San Onofre Nuclear Generating Station, Unit 2 Docket No. 50-361 Response to Request for Additional Information Regarding Report of Inservice Inspection of Steam Generator Tubes, Cycle 14

References:

1. Letter from B. Katz (SCE) to Document Control Desk (NRC) dated February 7, 2006,

Subject:

Special Report: Inservice Inspection of Steam Generator Tubes, Cycle 14

2. Letter from N. Kalyanam (NRC) to Richard M. Rosenblum (SCE), dated September 28, 2006;

Subject:

San Onofre Nuclear Generating Station, Unit 2 (SONGS 2) - Request for Additional Information (RAI) Regarding the 2006 Steam Generator Tube Inspections (TAC NO. MC8395)

Dear Sir or Madam:

By Reference 1, Southern California Edison (SCE) submitted the reports required by Technical Specification 5.7.2.c of the inservice inspection of steam generator tubes at San Onofre Nuclear Generating Station Unit 2. Subsequently, by Reference 2, the NRC staff requested certain additional clarifying information. The requested information is provided in the enclosure, with the exception of Request Number 1, regarding sleeves. SCE is working with the sleeve supplier, Westinghouse, to assemble this information. SCE submittal of a response for Request Number 1 is targeted for late December 2006.

If you have any questions or would like additional information concerning this subject, please contact Ms. Lynn Pressey at (949) 368-6351.

Sincerely, Enclosure cc: B. S. Mallett, NRC Regional Administrator, Region IV N. Kalyanam, NRC Project Manager, San Onofre Units 2 and 3 C. C. Osterholtz, NRC Senior Resident Inspector, San Onofre Units 2 and 3 P.O. Box 128 San Clemente, CA 92674-0128 949-368-9275 Fax 949-368-9881 A~J

REQUEST FOR ADDITIONAL INFORMATION SAN ONOFRE NUCLEAR GENERATING STATION. UNIT 2 2006 STEAM GENERATOR TUBE INSPECTIONS

1. During a conference call in which the 2006 steam generator tube inspections were discussed (refer U.S. Nuclear Regulatory Commission (NRC) letter dated April 3, 2006 - Agencywide Documents Access and Management System Accession No. ML060950305), there was some discussion concerning nickel migration associated with sleeved tubes. Please provide additional information concerning the nature of this phenomenon and the extent to which it could affect the detectability of flaws in the pressure boundary portion of the tube and the sleeve.

Southern California Edison (SCE) Response SCE is working with the sleeve supplier, Westinghouse, to assemble this information. SCE submittal of a response for Request Number 1 is targeted for late December 2006.

2. Please confirm that no cracking was identified in dents and dings during your 2006 inspections. If cracking was identified, please provide the size of the dent/ding, the location of the flaw, and its estimated size. The staff recognizes that certain dentsldings were not inspected with a rotating probe because of interferences associated with sleeves installed in the tube (and, as a result, a flaw could not positively be identified at the dents/dings in these tubes).

SCE Response Axially oriented, inside-diameter-initiated cracking was identified at dents. No cracking was identified at dings. The following Tables provide the requested information.

Steam Generator E-088 Tube Tube Flaw Plus Depth Length Dent Row Column Location Point (% (inches) Size Probe through (volts)

Volts wall) 17 55 06H+0.3 0.99 22 0.29 3.18 17 119 06H+0.48 0.88 60 0.3 21.46 31 113 06H+0.4 1.16 57 0.29 7.82 130 124 06H+0.11 1.44 21 0.13 24.74 1

Steam Generator E-089 Tube Tube Flaw Plus Depth Length Dent Row Column Location Point (% (inches) Size Probe through (volts)

Volts wall) 87 25 04H+0.3 1.6 53 0.22 2.24 87 25 04H+0.68 1.36 44 0.17 2.24 91 35 05H+0.18 1.64 49 0.29 <2 (Note 1)

Note 1: The bobbin probe is qualified for detection of axially oriented, inside-diameter-initiated cracking at locations where denting less than 2 volts may exist. This axially oriented, inside-diameter-initiated crack was detected by the bobbin probe, requiring the confirmation rotating plus point probe examination. The dent value (< 2 volts) was obtained because of the bobbin probe detection of the axially oriented inside-diameter-initiated cracking, not during the bobbin probe examination where dents greater than or equal to 2 volts are measured. The other cracks shown in the Tables were at locations with dents greater than 2 volts; thus the qualified detection tool for them was the rotating plus point probe.

Everything except the size (voltage) of the associated dents (the far-right-column in each Table) was taken from Reference 1.

This is a guide to the information in Reference 1:

a. Table 3, Row 1 shows that there are 6 applicable tubes.

b. In Tables 6 and 7, for Steam Generators (SGs) E-088 and E-089, respectively, the 6 applicable tube identification numbers ("row" and "column") can be found by searching for the Table 3 cross-reference ("ID Axial @ Support") in the "Reason for Plugging Tube (per Table 3)" Column.

c. Tables 8 and 9, for SGs E-088 and E-089, respectively, provide the tube support location and measured sizes of the axially oriented indications.

3. Please discuss the results of your inspections associated with the scallop bars.

SCE Response The inspection probe and location, shown in Tables 1A and 1B of Reference 1, was "Plus Point Probe examinations of hot leg scallop bar supports with the above adjacent hot leg square bend and the below one support elevation of tubing freespan." This inspection was done in 81 tubes (a 22% sample) in SG E-088. This inspection was done in 74 tubes (a 20%

sample) in SG E-089. The results of these inspections were that no crack indications were detected.

2

4. Regarding the eight tubes that were not inspected with a rotating probe due to interference associated with the sleeves installed in the tubes, please provide the following information: the voltage for these indications, a summary of whether the dent has changed in size/characteristic since the prior inspection, and the date of the last rotating probe examination. These eight tubes were not in-situ pressure tested.

SCE Response SG Tube Tube Dent in Last Change in Notes Row Col- 2006 Rotating Size or umn (volts) Plus Point Character-Exam (year) istic E-088 30 118 2.60 2004 no E-088 53 79 3.73 2004 no E-088 56 92 2.73 2004 no E-088 64 52 2.60 2002 no Note 1 E-088 80 104 3.98 2004 no E-088 81 97 2.78 2004 no E-088 83 111 2.37 2002 no Note 1 E-089 71 63 2.23 2004 no Note 1: In 2004 the Dent signal was below the 2 volt threshold for reporting and follow-up rotating probe exam.

5. Table 3 of the February 7, 2006, letter implies that 145 tubes were plugged for active degradation in steam generator E-088 and 76 tubes were plugged for active degradation in steam generator E-089. Tables 8 and 9, however, only provide the sizes of approximately 40 indications in steam generator E-088 and 15 indications in steam generator E-089. Please provide the sizes (e.g., percent wall thickness) of all detected indications.

SCE Response Table 3 of Reference 1 contains a category of "Tubes with indications of wear at tube support locations" which is 22 tubes in SG E-088 and 9 tubes in SG E-089. The size of indications in these tubes is provided in Reference 1 (Appendix 3 for SG E-088 and Appendix 4 for SG E-089).

Appendices 3 and 4 of Reference 1 also provide the size of indications of wear at tube support locations in tubes that were not repaired, and thus are not in Table 3 of Reference 1.

For indications, other than wear at tube support locations, size information is available in Reference 1 in two places for SG E-088 and two corresponding places for SG E-089.

3

• Size information (the plus point probe voltage) for all these indications is in Appendix 3 for SG E-088 and Appendix 4 for SG E-089.

" Additional size information for appropriate indications is in Table 8 for SG E-088 and Table 9 for SG E-089.

o The industry guidance for determining the indications for which Tables 8 and 9 sizing (beyond voltage measurement) is applicable to assess the capability of the SG tubing to meet structural and accident induced leakage performance criteria, is in the Electric Power Research Institute's "Steam Generator In Situ Pressure Test Guidelines", Revision 2, Report 10079004 dated August 2003.

The applicable guidance in this guideline is the first sentence in Section 4.4 and the third sentence in Section 4.5.B.

6. Please discuss how the operating experience at Waterford Steam Electric Station, Unit 3 (refer to NRC Information Notice 2005-29, "Steam Generator Tube and Support Configuration") was factored into your 2006 steam generator tube inspections.

SCE Response SCE documented evaluation of this industry experience within SCE's Action Request process, and the Steam Generator Degradation Assessment.

Tube support locations receive 100% inspection for wear of the tubing. This is shown in detail in Tables 1A and 1 B of Reference 1.

SCE obtained the applicable eddy current raw data from Waterford 3. SCE had the benefit of analyst experience with this particular experience because San Onofre and Waterford use the same supplier of secondary party data analysis services. The secondary data analysis supplier used the Waterford eddy current raw data to produce instructional example graphics, with emphasis on the tube support signals and locations. Key points were annotated on the graphics to maximize their instructional value.

The San Onofre Units 2 and 3 Steam Generator data analysis guideline was updated with a summary of the Waterford 3 experience and the instructional eddy current signal graphics discussed in the previous paragraph. This provides San Onofre data analysts with a tool to recognize and evaluate this condition.

With extensive inspection, and this tool, no such conditions were found at San Onofre Unit 2 in January 2006.

7. In the NRC's summary of a conference call held to discuss the results of your 2006 steam generator tube inspections, there was reference to a discussion concerning recent findings at Catawba Nuclear Station, Unit 2 (Catawba 2). The 4

reference was made concerning inspections adjacent to tie rods and tubes surrounding all new and previously identified loose parts. Please confirm that Catawba 2 is the correct reference. If it is, please discuss the specific operating experience that is relevant to San Onofre Unit 2. If Catawba 2 is not the correct reference, please provide the correct reference, if any.

SCE Response The Catawba 2 reference was not made concerning inspections adjacent to tie rods and tubes surrounding all new and previously identified loose parts. These inspections are related to previous San Onofre experience and other industry experience.

Catawba 2 is the correct reference for conference call discussion regarding inspection in tubing expanded within the tubesheet.

The relevant operating experience is in NRC Information Notice 2005-09 dated April 7, 2005;

Subject:

Indications in Thermally Treated Alloy 600 Steam Generator Tubes and Tube-to-Tubesheet Welds. Catawba 2 found Primary Water Stress Corrosion Cracking (PWSCC) indications in Inconel 600 thermally treated tubing expanded within the cold leg tubesheet.

Also, the indications were at locations other than the expansion transition (at the top of the tubesheet).

Prior to the San Onofre Unit 2 inspection in January 2006, this experience was incorporated into the Degradation Assessment. In the San Onofre Degradation Assessment, this experience was classified as a potential degradation mechanism. San Onofre Unit 2 potential was based on the key factors of having tubing material more susceptible to PWSCC (Inconel 600 mill annealed) and similar steam generator operation time. The applicable inspection update was to extend the existing rotating plus point probe inspection at the cold leg top-of-tubesheet to the applicable region within the tubesheet. As shown in Tables 1A and 1 B of Reference 1, this was approximately 30% of the tubes, from 2 inches above to 13 inches below the cold leg top-of-tubesheet. The 13 inch lower inspection bound is based on a methodology commonly called C-star, which determines the region of explosively expanded tubing within the tubesheet where inspection is applicable for tubing structural integrity and accident induced leakage considerations.

5