ML091130149
| ML091130149 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 05/08/2009 |
| From: | Nadiyah Morgan Plant Licensing Branch 1 |
| To: | Sena P FirstEnergy Nuclear Operating Co |
| morgan n | |
| References | |
| TAC MD9559, TAC ME0097 | |
| Download: ML091130149 (4) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 8, 2009 Mr. Peter P. Sena III Site Vice President FirstEnergy Nuclear Operating Company Beaver Valley Power Station Mail Stop A-BV-DEB1 P.O. Box 4 Route 168 Shippingport, PA 15077
SUBJECT:
BEAVER VALLEY POWER STATION, UNIT NO.2 - STAFF EVALUATION RE:
2008 STEAM GENERATOR TUBE INSPECTION REPORTS (TAC NOS.
MD9559 AND ME0097)
Dear Mr. Sena:
By letter dated August 7,2008, as supplemented by letters dated October 28,2008, and February 20,2009, FirstEnergy Nuclear Operating Company, (licensee), submitted information summarizing the results of the 2008 steam generator (SG) tube inspections at Beaver Valley Power Station, Unit NO.2 (BVPS-2). In addition to these reports, the Nuclear Regulatory Commission (NRC) staff summarized additional information concerning the 2008 SG tube inspections at BVPS-2 in a letter dated May 27, 2008.
The NRC staff has completed its review of these reports and concludes that the licensee provided the information required by their Technical Specifications and that no additional follow up is required at this time. The NRC staff's review of the reports is enclosed.
Please contact me at (301) 415-1016, if you have any questions regarding this issue.
Sincerely, adiyah S. Morgan, Project Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Plant Licensing Branch 1-1 Docket No. 50-412
Enclosure:
As stated cc w/encl: Distribution via Listserv
STAFF EVAULATION REGARDING THE 2008 STEAM GENERATOR TUBE INSPECTION REPORTS BEAVER VALLEY POWER STATION, UNIT NO.2 DOCKET NO. 50-412 By letters dated August 7,2008 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML082240290), as supplemented by letters dated October 28, 2008 (ADAMS Accession No. ML083050508). and February 20, 2009 (ADAMS Accession No. ML090550714), FirstEnergy Nuclear Operating Company, (licensee), submitted information summarizing the results of the 2008 steam generator (SG) tube inspections at Beaver Valley Power Station, Unit NO.2 (BVPS-2). In addition to these reports, the Nuclear Regulatory Commission (NRC) staff summarized additional information concerning the 2008 SG tube inspections at BVPS-2 in a letter dated May 27,2008 (ADAMS Accession No. ML081410092).
The SGs at BVPS-2 are Westinghouse model 51 SGs. Each SG contains 3,388 mill annealed Alloy 600 tubes. Each tube has a nominal outside diameter (00) of 0.875 inch and a nominal wall thickness of 0.050 inch. The tubes are supported by a number of carbon steel tube support plates and Alloy 600 anti-vibration bars. The tubes were roll expanded into the tubesheet at both ends for the full length of the tubesheet. The entire length of tube within the tubesheet was shot-peened on both the hot and cold-leg side of the SG prior to operation. In addition, the U-bend region of the small radius tubes were in-situ stress relieved prior to operation. There are no sleeves installed in the BVPS-2 SGs as of spring 2008 refueling outage.
In addition to the depth-based tube repair criteria, the licensee is also authorized to apply the voltage-based tube repair criteria for predominantly axially oriented 00 stress-corrosion cracking at the tube support plate elevations. Although authorized to implement the voltage-based repair criteria, the licensee has not found it necessary to implement these criteria since few indications subject to this repair criteria have been identified.
The licensee provided the scope, extent, methods, and results of their SG tube inspections in the documents referenced above. In addition, the licensee described corrective actions (e.g.,
tube plugging) taken in response to the inspection findings.
Based on its review of the reports submitted, the NRC staff has the following observations and comments:
- An in-situ pressure test was performed on one tube with a circumferential indication associated with a free span ding. The tube passed the in-situ pressure test with no leakage at any of the test pressures.
- The circumferential indication that was in-situ pressure tested was associated with a ding. A second ding was located several inches away and on the opposite side of the tube. There were no crack-like indications detected at the second ding. Given the orientation of these dings. it is similar to the "ding pairs" that have been observed at other plants and which have Enclosure
-2 exhibited cracking (although the axial separation of the dings in this tube was greater than what would typically be observed in a ding pair). Ding pairs are predominantly located in the upper region of the SG. As a result of finding the circumferential indication in a ding that is similar to a "ding pair," the licensee concentrated their supplemental examinations in the upper region of the SG even though stress-corrosion cracking is also dependent on temperature (which is higher in the lower regions of the hot-leg side of the SG). No additional indications associated with dings were detected.
- A small scale chemical cleaning was performed during the outage using Westinghouse's Advanced Scale Conditioning Agent. The cleaning was performed at the top of the tubesheet to remove soft sludge.
- The inspection plan includes criteria (voltage, phase angle) for when a support plate mix residual signal should be inspected with a rotating probe. In addition, criteria exist (Le., change in voltage or phase rotation) for when a historic indication should be inspected with a rotating probe. The NRC staff did not review the basis for these criteria; however, the criteria that are implemented are similar to what has been used in other inspections.
Indications of possible tube support plate indications were re-inspected with a +Point probe.
If the +Point probe indicated a tube support plate ligament breach, it was evaluated for acceptability. All ligament breaches were found to be acceptable by the licensee. The licensee has not implemented the voltage-based alternate repair criteria (which relies, in part, on the presence of the support plate to prevent burst during normal operation).
- An upper steam drum inspection was performed in all three SGs. Minor erosion/corrosion of carbon steel materials was reported. For example, minor erosion/corrosion at the feedwater header to J-nozzle (which are made from Alloy 600) interface was noted in SG A. In addition, a small through-wall penetration of the feedwater header was noted near the J-nozzle 33 location in SG A. This specific area was considered to have the highest velocities within the feedwater header as a result of a factory modification associated with removing and relocating the J-nozzle to accommodate the position of a seismic restraint.
The through-wall penetration is associated/near the plug installed in the location where the J-nozzle was originally located.
- Noise levels were not monitored in the U-bend region of the row 2 tubes since the noise levels in row 2 are typically less than that in row 1. In addition, no tubes in row 1 had noise levels that would have required inspection with a high-frequency +Point probe. The licensee indicated that primary water stress-corrosion cracking is not expected in the small radius U-bends due to the heat treatment (stress relief) that the U-bends received prior to plant operation.
Based on a review of the information provided, the NRC staff concludes that the licensee provided the information required by their Technical Specifications. In addition, the NRC staff concludes that there are no technical issues that warrant 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.
May 8,2009 Mr. Peter P. Sena III Site Vice President FirstEnergy Nuclear Operating Company Beaver Valley Power Station Mal Stop A-BV-DEB1 P.O. Box 4 Route 168 Shippingport, PA 15077
SUBJECT:
BEAVER VALLEY POWER STATION, UNIT NO.2 - STAFF EVALUATION RE:
2008 STEAM GENERATOR TUBE INSPECTION REPORTS (TAC NOS.
MD9559 AND ME0097)
Dear Mr. Sena:
By letter dated August 7, 2008, as supplemented by letters dated October 28, 2008, and February 20,2009, FirstEnergy Nuclear Operating Company, (licensee), submitted information summarizing the results of the 2008 steam generator (SG) tube inspections at Beaver Valley Power Station, Unit NO.2 (BVPS-2). In addition to these reports, the Nuclear Regulatory Commission (NRC) staff summarized additional information concerning the 2008 SG tube inspections at BVPS-2 in a letter dated May 27,2008.
The NRC staff has completed its review of these reports and concludes that the licensee provided the information required by their Technical Specifications and that no additional follow up is required at this time. The NRC staff's review of the reports is enclosed.
Please contact me at (301) 415-1016, if you have any questions regarding this issue.
Sincerely, IRA!
Nadiyah S. Morgan, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-412
Enclosure:
As stated cc w/encl: Distribution via Listserv DISTRIBUTION:
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