ML23066A290

Review of the Refueling Outage 2R22 Steam Generator Tube Inspection Report

ML23066A290
Person / Time
Site:	Vogtle
Issue date:	03/16/2023
From:	John Lamb NRC/NRR/DORL/LPL2-1
To:	Brown R Southern Nuclear Operating Co
References
EPID L-2022-LRO-0120
Download: ML23066A290 (1)

Text

March 16, 2023

Mr. R. Keith Brown Regulatory Affairs Director Southern Nuclear Operating Company, Inc.

P. O. Box 1295, Bin 038 Birmingham, AL 35201-1295

SUBJECT:

VOGTLE ELECTRIC GENERATING PLANT, UNIT 2 - REVIEW OF THE REFUELING OUTAGE 2R22 STEA M GENERATOR TUBE INSPECTION REPORT (EPID: L-2022-LRO-0120)

Dear Mr. Brown:

By letter dated September 30, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22273A160), as supplemented by letter dated January 27, 2023 (ML23027A093), Southern Nuclear Operating Co mpany (SNC, the licensee) submitted information summarizing the results of the refueling outage (RFO) 2R22 steam generator (SG) inspections performed at Vogtle Electric Generating Plant (Vogtle), Unit 2. The inspections were performed during RFO 2R22. SNC provided additional information concerning the inspections in a letter dated January 27, 2023. The SG tube inspection report was submitted in accordance with Technical Specification (TS) 5.6.10, Steam Generator Tube Inspection Report.

The U.S. Nuclear Regulatory Commission (NRC) staff has completed its review of the information provided by SNC and concludes that the licensee provided the information required by Vogtle, Unit 2, TS and no follow-up is required at this time. The NRC staffs review of the report is enclosed.

If you have any questions, please contact me at (301) 415-3100 or via email at John.Lamb@nrc.gov.

Sincerely,

/RA/

John G. Lamb, Senior Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Docket No. 50-425

Enclosure:

Review of the Refueling Outage 22 Steam Generator Tube Inspection Report

cc: Listserv REVIEW OF THE REFUELING OUTAGE 22 STEAM GENERATOR

TUBE INSPECTION REPORT

SOUTHERN NUCLEAR OPERATING COMPANY

VOGTLE ELECTRIC GENERATING PLANT, UNIT 2

DOCKET NO. 50-425

By letter dated September 30, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22273A160), as supplemented by letter dated January 27, 2023 (ML23027A093), Southern Nuclear Operating Co mpany (SNC, the licensee) submitted information summarizing the results of the refueling outage (RFO) 2R22 steam generator (SG) inspections performed at Vogtle Electric Generating Plant (Vogtle), Unit 2. The inspections were performed during RFO 2R22. SNC provided additional information concerning the inspections in a letter dated January 27, 2023. The SG tube inspection report was submitted in accordance with Technical Specification (TS) 5.6.10, Steam Generator Tube Inspection Report.

Vogtle, Unit 2, has four Westinghouse Model F SGs, each of which contains 5,626 U-bend 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. During SG fabrication, the tubes were hydraulically expanded, at both ends, over the full depth of the tubesheet. Type 405 stainless steel support plates, which have broached quatrefoil holes, support the vertical section of the tubes, and anti-vibration bars support the U-bend section of the tubes.

The licensee provided the scope, extent, methods, and results of the SG tube inspections in the letters referenced above. In addition, SNC described corrective actions (e.g., tube plugging), if any were taken in response to the inspection findings.

Based on the review of the information provided, the U.S. Nuclear Regulatory Commission (NRC) staff has the following observations:

Circumferential outside diameter stress corrosion cracking (ODSCC) indications were reported in two tubes at the top of the tubesheet (TTS) hot leg expansion transition. The tubes were in SG 1 and SG 4. For the indication in SG 1, the estimated maximum through-wall (TW) depth and percent degraded area (PDA) from rotating probe

(+Point') sizing were 35-percent and 9.4-percent, respectively. For the indication in SG 4, the estimated maximum depth and PDA were 22-percent TW and 4.5-percent, respectively. Both tubes were stabilized and plugged. The indications were not detected by the array probe.

An axial ODSCC indication was reported in one tube at a dent at the bottom of the seventh (uppermost) tube support plate (TSP) on the hot leg. The bobbin probe voltage of the dent was 2.81 volt, and the bobbin coil probe signal resulted in additional examination with the +Point probe. The estimated length and maximum TW depth were 0.11 inch and 58-percent, respectively, based on +Point' sizing. The tube was plugged.

This indication was not detected by the array probe. Axial ODSCC associated with denting at the seventh TSP on the hot leg has been observed previously in a Model F SG at another unit.

Enclosure

No tubes exhibited degradation exceeding the condition monitoring limits; therefore, no in-situ pressure tests were required.

Vogtle, Unit 2, reported one prior indication of stress corrosion cracking (SCC)

(circumferential ODSCC) at the TTS on the hot leg in SG 2 during RFO 16 in 2013.

Based on a review of the information provided, the NRC staff concludes that SNC provided the information required by the Vogtle, Unit 2, TSs. In addition, the NRC staff concludes that there are no technical issues that warrant additional 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.

ML23066A290 OFFICE DORL/LPL2-1/PM DORL/LPL2-1/LA DMLR/MCCB/BC DORL/LPL2-1/BC DORL/LPL2-1/PM NAME JLamb KGoldstein SBloom MMarkley (EMiller for) JLamb DATE 03/06/2023 03/09/2023 03/06/2023 03/15/2023 03/16/2023