ML101620623

Summary of Conference Telephone Call Regarding the 2010 Steam Generator Tube Inspections

ML101620623
Person / Time
Site:	Prairie Island
Issue date:	06/18/2010
From:	Thomas Wengert Plant Licensing Branch III
To:	Schimmel M Northern States Power Co
Wengert, Thomas J, NRR/DORL, 415-4037
References
TAC ME3558
Download: ML101620623 (4)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 18, 2010 Mr. Mark A. Schimmel Site Vice President Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota 1717 Wakonade Drive East Welch, MN 55089-9642 SUB"IECT: PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 2 -

SUMMARY

OF CONFERENCE TELEPHONE CALL REGARDING THE 2010 STEAM GENERATOR TUBE INSPECTIONS (TAC NO. ME3558)

Dear Mr. Schimmel:

On April 28, 2010, the U.S. Nuclear Regulatory Commission (NRC) staff participated in a conference call with representatives of Northern States Power Company - Minnesota (NSPM) regarding the ongoing steam generator (SG) inspection activities at the Prairie Island Nuclear Generating Plant, Unit 2, conducted during their 2010 outage. The NRC follows the results of the industry's SG inspections in order to maintain an awareness of the condition of the SGs and the types of tube degradation mechanisms that are active.

The enclosed documentation of the phone call is provided to NSPM for information. The slides provided by NSPM in support of this discussion (Agencywide Documents Access and Management System (ADAMS) Accession No. ML101580107) are available to the public.

Based on the information provided during the conference call, the NRC staff did not identify any issues that warranted additional follow-up action at this time. If there are any questions, please contact me at 301-415-4037.

Sincerely, Thomas J. Wengert, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-306

Enclosure:

Conference Call Summary cc w/encl: Distribution via ListServ

SUMMARY

OF CONFERENCE CALL WITH NORTHERN STATES POWER COMPANY - MINNESOTA REGARDING THE SPRING 2010 STEAM GENERATOR TUBE INSPECTION RESULTS PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 2 DOCKET NO. 50-306 On April 28,2010, the U.S. Nuclear Regulatory Commission (NRC) staff participated in a conference call with representatives of Northern States Power Company, Minnesota (the licensee) regarding the ongoing steam generator (SG) inspection activities at Prairie Island Nuclear Generating Plant (PINGP), Unit 2. Prior to the conference call, the licensee provided several slides to facilitate the discussion. These slides are located in the NRC's Agencywide Documents Access and Management System under Accession No. ML101580107.

PINGP Unit 2 has two Westinghouse model 51 SGs. Each SG contains 3388 mill-annealed Alloy 600 tubes. Each tube has a nominal outside diameter of 0.875 inches and a nominal wall thickness of 0.050 inches. The tubes were roll-expanded into the tubesheet at both ends for approximately 2.75 inches (Le., they are expanded for only a fraction of the tubesheet thickness and are considered partial depth hard-rolled tubes). The tubes are supported by a number of carbon steel tube support plates. The original anti-vibration bars were removed and replaced.

The row one and two tubes were subjected to an in-situ thermal stress relief in May 2000. Many tubes have been roll-expanded into the tubesheet above the original factory roll expansion to permit defects below these re-rolled locations to remain in service.

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 outside diameter stress corrosion cracking (ODSCC) 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, if any, indications subject to this repair criteria have been identified at Unit 2. In addition, the licensee is authorized to leave flaws within the tubesheet region in service provided they satisfy the F*/EF* repair criterion. The major cause of degradation within the tubesheet region is primary water stress corrosion cracking at the roll transition zones. Secondary side intergranular attack and ODSCC have also been observed at this location.

The licensee plans to replace their SGs during their 2013 refueling outage.

Additional clarifying information or information not included in the document provided by the licensee is summarized below:

• The primary-to-secondary leakage trend for the recently completed cycle was similar to the leakage trend from previous cycles. The licensee noted that the upward trend in SG tube leakage during the recently completed cycle was a result of dilution of the tritium in the reactor coolant.

• The scope of the inspections was similar to those in prior outages.

Enclosure

-2

• On Slide 18 of the licensee's discussion slides, the licensee clarified that they planned to plug 12 tubes in SG 21 and 9 tubes in SG 22. The licensee stated that these tubes were being plugged as a result of tube degradation, and were not being plugged due to flaws in (already existing) plugs.

• The height of the sludge pile is determined every few years. Rotating probe inspections extend approximately 6-inches above the top of the tubesheet in the box area (area of highest sludge deposition) and approximately 3-inches outside the box area. This extent of inspection is intended to bound the sludge pile.

• Sludge lancing was only performed in SG 22 during the 2010 outage.

• At the time of the call, no degradation of the secondary side internals had been discovered.

• The feedwater heater contains holes that permit the feedwater to enter the SG.

• All indications detected to date were small (i.e., no need to profile any of the indications).

• The F* criterion only applies to the hot-leg side of the SG. If cracking is detected on the cold-leg side, the tube would be plugged and the inspection would be expanded, as appropriate.

• At the time of the call, no inspections of the low row U-bends had been performed.

• The noise levels in the eddy current data were qualitatively assessed as being somewhat lower in the 2010 outage, as compared to noise levels observed in previous inspections. This was attributed to using inspection hardware that did not have any slip rings or extension cables.

• On Slide 14 of the licensee's discussion slides, when two numbers are provided in the voltage and depth/length columns, the first number in the voltage column corresponds to the first number in the depth or length column. The indication with the largest voltage and largest depth/length are provided for select indications.

SUbsequent to the call, the licensee provided the following information:

• The 22 SG girth weld inspection was a visual examination.

• It was determined that an indication on the tube located at row 3, column 38 in SG 21 had a free span single axial indication. This indication required in-situ pressure testing.

~ The tube located at row 3, column 38 in SG 21 passed the in-situ pressure test

• The tube located at row 33, column 27 in SG 22 had a possible loose part indication and showed tube damage (mechanical wear).

The NRC staff did not identify any issues that required follow-up action at this time; however, the staff asked to be notified in the event that any unusual conditions were detected during the remainder of the outage.

ML101620623 Slid I es Accession

. No. ML101580107 *B~y memo d ate d 5/14/10 NRR 106 OFFICE LPL3-1/PM LPL3-1/LA DCI/CSGB/BC LPL3-1/BC NAME TWengert BTuily RTaylor* RPascarelli DATE 06/18/10 06/18/10 05/14/10 06/18/10