ML14174B123

Review of 18th Steam Generator Tube Inspection Report

ML14174B123
Person / Time
Site:	Wolf Creek
Issue date:	06/25/2014
From:	Lyon C Plant Licensing Branch IV
To:	Heflin A Wolf Creek
Lyon C
References
TAC MF2837
Download: ML14174B123 (4)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 25, 2014 Mr. Adam C. Heflin President, Chief Executive Officer, and Chief Nuclear Officer Wolf Creek Nuclear Operating Corporation Post Office Box 411 Burlington, KS 66839

SUBJECT:

WOLF CREEK GENERATING STATION- REVIEW OF THE 2013 STEAM GENERATOR TUBE INSPECTION REPORT (TAC NO. MF2837)

Dear Mr. Heflin:

By letter to the U.S. Nuclear Regulatory Commission (NRC) dated September 30, 2013, as supplemented by letter dated February 6, 2014, Wolf Creek Nuclear Operating Corporation (the licensee) submitted its inspection results for the 18th steam generator tube inservice inspection at Wolf Creek Generating Station, which was conducted during refueling outage 19.

Conference calls between the licensee and NRC staff during your inspection were documented in the NRC letter dated April 1, 2013. Your report was submitted in accordance with Technical Specification (TS) 5.6.1 0, "Steam Generator Tube Inspection Report."

Based on its review, the NRC staff concludes that the licensee has provided the information required by TS 5.6.1 0. A summary of the NRC staff's review is enclosed. If you have any questions, please contact me at 301-415-2296 or via e-mail at Fred. Lyon@nrc.gov.

Sincerely, Carl F. Lyon, Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-482

Enclosure:

As stated cc w/encl: Distribution via Listserv

SUMMARY

OF THE REVIEW OF THE 2013 REFUELING OUTAGE 19 STEAM GENERATOR TUBE INSERVICE INSPECTION REPORT WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50-482 By letters dated September 30, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13277A558), and February 6, 2014 (ADAMS Accession No. ML14049A286), Wolf Creek Nuclear Operating Corporation (WCNOC, the licensee),

submitted information summarizing the results of the 2013 steam generator (SG) tube inspections performed at Wolf Creek Generating Station (WCGS) during refueling outage 19 (RF019). Conference calls regarding the SG inspections were documented by the U.S. Nuclear Regulatory Commission (NRC) staff in its letter dated April 1, 2013 (ADAMS Accession No. ML13077A073).

WCGS has four Westinghouse Model F SGs. Each SG contains 5,626 thermally-treated Alloy 600 tubes. Each tube has a nominal outside diameter of 0.688 inches and a nominal wall thickness of 0.040 inches. The tubes are supported by stainless steel tube supports with quatrefoil-shaped holes and V-shaped chrome-plated Alloy 600 anti-vibration bars.

The licensee provided the scope, extent, methods, and results of its SG tube inspections in the documents referenced above. In addition, the licensee described corrective actions, such as tube plugging, taken in response to the inspection findings.

Based on the NRC staff's review of the information provided by the licensee, the staff has the following observations/comments:

• A flaw in the divider plate-to-channel head weld was noted in the stainless steel cladding in the hot-leg channel head in SG A. The flaw was approximately 0.1 inches deep and approximately 2 inches long. The licensee performed an evaluation that determined it was acceptable to operate for the next operating cycle. The licensee plans to perform a detailed analysis of the flaw prior to refueling outage 20 (RF020), and ultrasonic characterization of the flaw is planned for RF020, to determine if the flaw has grown.

• A circumferential primary water stress-corrosion cracking (PWSCC) indication was found in one tube in SG B during RF019. The indication was located approximately 6 inches below the top of the tubesheet, at a previously identified bulge. The tube was removed from service by plugging. In SG B, 100 percent of the hot-leg tubes with bulge (BLG) or overexpansion (OXP) signals were inspected from three inches above the top of the tubesheet to 15.21 inches below the top of the tubesheet (+3/-15.21 inches) with a rotating probe equipped with a +Point' coil. The licensee confirmed that at least 20 percent of the tubes with BLG and OXP signals in the remaining three SGs were also inspected from Enclosure

+3/-15.21 inches with a rotating probe equipped with a +Point' coil. Although no additional PWSCC indications were detected in the remaining three SGs, the NRC staff notes that a sampling approach may have limitations when only a few indications may be present.

• Based on operating experience from another plant, the licensee repeated the screening for tubes with high residual stresses caused by non-optimal processing during tube manufacturing (such as those found at Seabrook Nuclear Plant).

The repeat screening was performed using a manual analysis technique and five tubes were identified that exhibited similarities to the tubes found at Seabrook, suggesting an unusual manufacturing process. All five of the tubes were plugged during RF019.

• Visual and ultrasonic inspections of the upper steam drum components were performed in SGs A and D. Minor degradation is being monitored at a few J-nozzle-to-feedring interface locations; however, all results were within acceptance criteria. No other degradation was observed during these inspections.

• A study of the as-built anti-vibration bar positions has been completed. A plant-specific U-bend fatigue analysis is in progress and is expected to be complete by the summer of 2014.

Based on a review of the information provided, the NRC staff concludes that the licensee provided the information required by its technical specifications. In addition, the staff concludes that there are no technical issues that warrant follow-up actions 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.

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