RNP-RA/08-0047, Response to NRC Request for Additional Information on the Steam Generator Inservice Inspection Results

Response to NRC Request for Additional Information on the Steam Generator Inservice Inspection Results

ML081270049
Person / Time
Site:	Robinson
Issue date:	04/30/2008
From:	Castell C Progress Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RNP-RA/08-0047
Download: ML081270049 (11)

Text

10 CFR 50.55a TS 5.6.8 SProgress Energy Serial: RNP-RA/08-0047 APR 3 0I2008 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261/LICENSE NO. DPR-23 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION ON THE STEAM GENERATOR INSERVICE INSPECTION RESULTS Ladies and Gentlemen:

The steam generator inservice inspection results for Refueling Outage 24 (RO-24) at H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, were previously submitted to the NRC by letters dated August 2, 2007, and November 1, 2007. An electronic mail message received from the NRC Project Manager for HBRSEP, Unit No. 2, on March 17, 2008, requested additional information pertaining to the RO-24 results.

The response to the request for additional information is provided in the attachment to this letter.

If you have any questions regarding this matter, please contact me at (843) 857-1626.

Sincerely, Curt Castell Supervisor - Licensing/Regulatory Programs CAC Attachment c: V. M. McCree, NRC, Region II M. G. Vaaler, NRC, NRR NRC Resident Inspector Progress Energy Carolinas, Inc.

Robinson Nuclear Plant 7 3581 West Entrance Road Hartsville, SC 29550 ks

United States Nuclear Regulatory Commission Attachment to Serial: RNP-RA/08-0047 Page 1 of 10 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION By letter dated August 2, 2007 and November 1, 2007, Carolina Power & Light (the licensee), now doing business as Progress Energy Carolinas, Inc., submitted information pertaining to the 2007 steam generator (SG) tube inspections performed at the H.B.

Robinson Steam Electric Plant (HBRSEP), Unit No. 2, during Refueling Outage 24 (RO-24). An electronic mail message received from the NRC Project Manager for HBRSEP, Unit No. 2, on March 17, 2008, requested additional information pertaining to the RO-24 results. The response to the request for additional information is hereby provided.

NRC Request 1:

The licensee implied that no "active degradation" was identified in the HBRSEP SGs during the 24th refueling outage SG tube inspections; however, it was also indicated that several tubes with wear at the anti-vibration bars, at tube supports, and associated with loose parts had been identified.

HBRSEP's definition of "active degradation mechanism" is based on the industry's definition of "active degradation mechanism." The NRC staff has found that the industry's (Electric Power Research Institute's) definition of active degradation in Revision 6 to the Pressurized Water Reactor Steam Generator Examination Guidelines is misleading since tubes could have degradation that is progressing (or present on the tubes) but the degradation could be classified as "not active" (refer to Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML010320218 and ML012200349).

As a result, please confirm that other than wear at the anti-vibration bars, at tube supports, and associated with loose parts, the licensee did not find any other service-induced indications during the SG tube inspections. If other indications were found, please provide the location, orientation, and measured sizes of these indications.

HBRSEP. Unit No. 2, Response:

Three types of service induced indications have been previously identified at HBRSEP, Unit No. 2, and are considered existing. No active degradation mechanisms, as defined in Revision 6 to the Pressurized Water Reactor Steam Generator Examination Guidelines, were identified during the RO-24 SG tube inspections.

Wear can occur at anti-vibration bars (AVBs) and tube supports. Wear can also occur

United States Nuclear Regulatory Commission Attachment to Serial: RNP-RA/08-0047 Page 2 of 10 due to loose parts or outage maintenance activities. Wear due to loose parts or outage maintenance activities are volumetric indications and are not reported with an orientation as would apply to a crack-like indication.

It is judged that wear detected in outer periphery tubes is predominantly due to outage maintenance activities. The location of the wear indications is consistently about 0.5 inches above the tubesheet and affects only the outermost tubes in the bundle.

Indications in Row 1 tubes in the blowdown lane are at approximately the same elevation as the sludge lance equipment and may be the result of operation of that equipment during previous refueling outages. Some of these indications have been measured in sequential inspections and it was confirmed that no growth was indicated.

Mechanical wear indications due to loose parts and maintenance activities cannot be readily differentiated. The data associated with these indications were provided in the letter dated November 1, 2007. The following tables summarize the mechanical wear indications in the SGs (these tables are the same tables provided in the November 1, 2007, letter with the AVB and tube support wear indications deleted):

Steam Generator "A" Row Column  % Depth Location Inches 1 1 13 CTS 13.72 7 1 18 CTS 0.68 7 1 14 CTS 0.69 1 2 13 CTS 13.74 11 2 17 CTS 0.54 1 3 17 CTS 15.66 13 3 16 CTS 0.45 1 4 18 CTS 15.78 16 4 19 CTS 0.41 16 4 13 CTS 0.84 1 5 16 CTS 15.76 1 6 17 CTS 15.57 23- 7 17 CTS 0.5 26 9 16 CTS 0.5 23 14 23 HTS 0.05 33 15 16 CTS 0.48 37 20 15 HTS 0.63 40 25 18 CTS 0.47 40 25 14 CTS 0.68

United States Nuclear Regulatory Commission Attachment to Serial: RNP-RA/08-0047 Page 3 of 10 Steam Generator "A" Row Column  % Depth Location Inches 42 30 13 HTS 0.63 42 30 14 HTS 0.95 42 30 16 CTS 0.47 43 33 14 CTS 0.56 43 33 13 CTS 0.57 44 36 18 CTS 0.64 44 36 15 CTS 1.69 43 37 15 CTS 0.55 43 37 16 CTS 1.75 33 41 22 HTS 0.1 34 41 23 HTS 0.09 45 41 17 CTS 0.61 45 47 15 CTS 2.72 45 47 18 CTS 6.77 45 52 18 CTS 0.57 45 52 19 CTS 0.63 45 52 15 CTS 3.78 41 53 27 HTS 0.07 44 57 16 CTS 0.6 44 57 20 CTS 0.61 44 57 13 CTS 1 44 57 16 CTS 1.97 44 57 13 HTS 0.67 44 57 15 HTS 1.11 43 60 19 CTS 0.61 42 63 15 CTS 0.62 42 63 18 CTS 0.7 42 63 12 HTS 0.64 40 68 17 HTS 0.65 40 68 17 HTS 0.73 36 74 13 HTS 0.62 31 80 16 CTS 0.59 26 84 15 CTS 0.61 23 86 17 CTS 0.61

United States Nuclear Regulatory Commission Attachment to Serial: RNP-RA/08-0047 Page 4 of 10 Steam Generator "A" Row Column  % Depth Location Inches 23 86 13 CTS 3.16 19 87 16 CTS 7.69 1 89 14 HTS 15.54 11 91 18 CTS 0.56 11 91 17 CTS 0.67 7 92 15 CTS 0.61 Steam Generator "B" Row Column  % Depth Location Inches 1 1 13 CTS 9.13 1 1 17 CTS 12.1 1 1 21 CTS 12.63 1 1 14 CTS 13.1 1 1 17 CTS 15.15 2 1 15 CTS 5.49 2 1 17 CTS 10.17 2 1 17 CTS 11.64 6 1 15 CTS 0.48 7 1 13 CTS 0.64 7 1 13 CTS 1.12 7 1 16 CTS 12.4 11 2 17 CTS 0.56 16 4 17 CTS 0.5 23 7 14 HTS 0.68 26 9 19 HTS 0.62 29 14 31 FBH 0.46 30 14 15 FBH 0.42 30 14 22 FBH 0.44 29 15 36 FBH 0.43 7 16 18 CTS 16.11 33 17 20 FBH 0.72 37 20 17 HTS 0.6 37 20 15 CTS 0.51

United States Nuclear Regulatory Commission Attachment to Serial: RNP-RA/08-0047 Page 5 of 10 Steam Generator "B" Row Column  % Depth Location Inches 37 20 15 CTS 0.53 40 25 22 CTS 0.51 40 25 20 CTS 0.53 40 26 16 CTS 0.51 4 32 20 CTS 3.11 4 32 18 CTS 6.12 2 41 18 CTS 1.19 2 42 27 CTS 1.59 3 44 32 HTS 0.24 3 44 23 HTS 0.41 34 44 26 HTS 0.07 3 45 19 CTS 1.42 4 47 16 CTS 1.42 4 47 16 CTS 1.44 4 48 17 CTS 1.59 4 48 28 CTS 3.27 5 48 28 CTS 1.67 4 49 20 CTS 1.5 5 49 27 CTS 1.69 4 50 16 CTS 1.76 11 53 26 CTS 1.27 43 60 18 CTS 0.58 40 67 13 CTS 0.52 40 68 18 CTS 0.51 34 76 21 CTS 1.83 33 78 16 CTS 0.63 33 78 18 CTS 0.64 26 84 17 CTS 0.67 23 86 19 CTS 0.65 16 89 21 CTS 0.6 1 92 12 HTS 2.02 1 92 16 HTS 2.88 1 92 15 HTS 13.91 2 92 14 CTS 11

United States Nuclear Regulatory Commission Attachment to Serial: RNP-RA/08-0047 Page 6 of 10 Steam Generator "C" Row Column  % Depth Location Inches 7 1 15 CTS 0.6 44 38 17 HTS 0.55 34 50 23 HTS 0.02

34. 51 20 HTS 0.05 35 51 18 HTS 0.08 44 57 17 HTS 0.56 44 57 16 HTS 1.21 44 57 18 CTS 0.65 35 75 17 CTS 1.49 33 78 14 HTS 0.66 2 91 38 CTS 0.05 3 91 25 HTS 0.39 Abbreviations:

CTS Cold leg top-of-tubesheet HTS Hot leg top-of-tubesheet FBH Flow distribution baffle on hot leg side 02A Anti-vibration bar 2 03A Anti-vibration bar 3 04A Anti-vibration bar 4 02C Cold leg second support plate 02H Hot leg second support plate 03C Cold leg third support plate 03H Hot leg third support plate 04C Cold leg fourth support plate 04H Hot leg fourth support plate 05C Cold leg fifth support plate 05H Hot leg fifth support plate 06C Cold leg sixth support plate 06H Hot leg sixth support plate NRC Request 2:

For each refueling outage and SG tube inspection outage since the replacement of the HBRSEP SGs, please provide the cumulative effective full power months (EFPM).

United States Nuclear Regulatory Commission Attachment to Serial: RNP-RA/08-0047 Page 7 of 10 HBRSEP, Unit No. 2, Response:

The following data is provided to summarize the cumulative EFPM for the HBRSEP, Unit No. 2, SGs base on the cycle effective full power years (EFPY):

Service Time and Corresponding ISI Intervals Cycle (C) / SG EOC Refueling End-of-Cycle speon Outage (EOC) Cumulative Cumulative Inspection (RO) EFPY EFPY EFPM Outage RO-9 SGs Replaced C-10 9 RO-10 1st SG ISI 0.9 11.0 Y C-11 9.9 0.9 11.0 Y C-12 10.9 1.9 22.8 Y C-13 11.9 2.9 34.6 Y C-14 12.9 3.9 46.6 Y C-15 14.0 5.0 59.4 Y C-16 15.0 6.0 72.4 Y C-17 16.2 7.2 86.4 Y C-18 17.6 8.6 102.6 Y C-19 19.0 10.0 119.5 Y C-20 20.4 11.4 136.6 Y C-21 21.8 12.8 153.6 Y C-22 23.2 14.2 170.4 Y C-23 24.6 15.6 187.2 N C-24 26.0 17.0 204.4 Y NRC Request 3:

Regarding the scope of the SG tube inspections, discuss whether any plug or secondary side inspections (including foreign object search and retrieval inspections) were performed. If so, please discuss the results. If any degraded conditions (e.g., erosion of J-tubes) were identified, please discuss the actions taken to ensure acceptable SG performance until the next scheduled inspection.

HBRSEP, Unit No. 2, Response:

Prior to eddy current testing during RO-24, a video inspection was performed of the secondary-side tube sheet region of each steam generator. Each previously installed plug location was inspected for correct location, excessive boron build-up, and wetness. No discrepancies were identified and the inspected plugs were present and acceptable.

United States Nuclear Regulatory Commission Attachment to Serial: RNP-RA/08-0047 Page 8 of l0 A total of six tubes were plugged during the RO-24 Steam Generator inspection. There were five tubes plugged in Steam Generator "B." In Steam Generator "C" one tube was plugged. The wear depth for the tubes that were plugged ranged from 20% to 38%

through-wall.

A top-of-tube-sheet in-bundle foreign object search and retrieval (FOSAR) inspection was performed. Foreign material identified was classified in three categories:

• Category 1 objects may cause tube wear to exceed 40% through-wall in less than two operational cycles.

" Category 2 objects are not expected to cause tube wear to exceed 40% through-wall over two operational cycles.

• Category 3 objects are not expected to cause significant tube wear.

One Category 1 item could not be retrieved from Steam Generator "A." The object was not detectable in eddy-current test (ECT) signals, indicating it is not metallic and it partially broke-up during retrieval efforts. An engineering evaluation determined this object could remain in the steam generator without challenging the condition monitoring limit over the next two cycles of operation. One Category 2 item in Steam Generator "B" was not retrieved due to its location. No foreign material items were left vhere they would be expected to damage steam generator tubes over the next two operating cycles.

The steam drum areas of Steam Generator "B" were inspected during RO-24. No loose parts/foreign objects were identified. No significant degradation was found.

The feedring was visually inspected and ultrasonic measurements were performed in specific locations on the feedwater ring. Thickness measurements were obtained at 16 accessible locations around the feedring. The thickness measurements were within acceptable limits:

Six (6) J- Nozzles in Steam Generator "B" were inspected by means of a video probe. No anomalous conditions were found.

NRC Request 4:

The licensee indicated that no corrosion-related degradation has been observed in the HBRSEP SGs. Please provide the HBRSEP hot-leg operating temperature.

HBRSEP. Unit No. 2, Response:

The HBRSEP, Unit No. 2, hot leg reactor coolant temperature is approximately 604'F.

United States Nuclear Regulatory Commission Attachment to Serial: RNP-RA/08-0047 Page 9 of 10 NRC Request 5:

Please confirm that 100% of all row 1 and 2 U-bend region tubes were inspected during the 90 EFPM interval.

HBRSEP. Unit No. 2. Response:

Tube inspections in RO-22 and RO-24, which were within the 90 EFPM interval, included 100% inspection of the Row 1 and Row 2 U-bends region tubes.

NRC Request 6:

With respect to the tubes with potentially non-optimal tube processing, please discuss whether any rotating probe examinations were performed on these tubes. Please include the number of tubes with non-optimal tube processing and the number of rotating probe exams at each location (e.g., dents, expansion transition, U-bend (if a low row tube)),.

HBRSEP. Unit No. 2. Response:

Unexpanded and Partially Expanded Tubes:

One tube in SG "A," Rowl, Column 47 was not full depth expanded in the cold leg tubesheet. This tube was inspected from the tube end to the first support with the Plus Point probe. No degradation was found.

One tube in SG "B," Row 25, Column 10 is partially expanded in the cold leg tubesheet. This tube was inspected from the tube end to the top of tubesheet with the Plus Point probe. No degradation was found.

Bulges, Overexpansions and Dents:

Bobbin data from 1999, 2001, 2002, and 2004 outages was reviewed to identify overexpansions greater than 1.5 mils. Bulges and dents were identified if they had greater than 18 volts from the 400 kilohertz differential channel. Tubes in the planned hot leg top of tubesheet inspection that contained bulges, over-expansions, and dents that met this criterion were inspected with the Plus Point probe from 4 inches above the top of tubesheet to 17 inches below the top of tubesheet regardless of the location of the anomaly.

The following table summarizes the number of tubes on the hot leg side inspected from 4 inches above the top of tubesheet to 17 inches below the top of tube sheet that met the criteria for bulges, overexpansions, or dents.

United States Nuclear Regulatory Commission Attachment to Serial: RNP-RA/08-0047 Page 10 of 10 Steam Generator A Steam Generator B Steam Generator C 367 315 228 No degradation was reported as a result of the rotating coil examination.

Low Row U-Bends:

A total of 277 low row U-bend tubes were tested during RO-24. Ninety-three (93) of these were in Steam Generator "A" and 92 each in Steam Generator "B" and Steam Generator "C." No degradation in the U-bend area was identified.