ML25337A339
| ML25337A339 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 12/15/2025 |
| From: | V Sreenivas Plant Licensing Branch 1 |
| To: | Coffey R Florida Power & Light Co |
| Lantigua, R, NRR/DORL/LPLI, | |
| References | |
| EPID L-2025-LRO-0027 | |
| Download: ML25337A339 (0) | |
Text
December 15, 2025 Mr. Robert Coffey Executive Vice President, Nuclear Division and Chief Nuclear Officer Florida Power & Light Company Mail Stop: EX/JB 700 Universe Blvd.
Juno Beach, FL 33408
SUBJECT:
SEABROOK STATION, UNIT 1 - REVIEW OF THE FALL 2024 STEAM GENERATOR TUBE INSPECTION REPORTS (EPID L-2025-LRO-0027)
Dear Mr. Coffey:
By letters dated May 5, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25125A317), and June 4, 2025 (ML25155A139), NextEra Energy Seabrook, LLC (the licensee) submitted information summarizing the results of the fall 2024 steam generator tube inspections performed at Seabrook Station, Unit 1, during refueling outage 23.
The U.S. Nuclear Regulatory Commission (NRC) staff has completed its review of the submittal and concludes that the licensee provided the information as required. No additional follow-up is required at this time. The results of the NRC staff's review are enclosed.
If you have any questions, please contact me at 301-415-2597 or via email at v.sreenivas@nrc.gov.
Sincerely,
/RA/
V. Sreenivas, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-443
Enclosure:
Review of the Steam Generator Tube Inspection Report cc: Listserv
Enclosure SEABROOK STATION, UNIT 1 REVIEW OF THE FALL 2024 STEAM GENERATOR TUBE INSPECTION REPORT DOCKET NO. 50-443 By letters dated May 5, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25125A317), and June 4, 2025 (ML25155A139), NextEra Energy Seabrook, LLC (the licensee) submitted information summarizing the results of the fall 2024 steam generator (SG) tube inspections performed at Seabrook Station during refueling outage (RFO) 23.
Seabrook Station, Unit 1 (Seabrook), has four Westinghouse Model F SGs, each of which contains 5,626 thermally treated Alloy 600 (Alloy 600TT) 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 ends were hydraulically expanded over the full depth of the tubesheet.
Eight horizonal stainless steel structures support the vertical section of the tubes. The lowest support is a flow distribution baffle with drilled holes. The other seven are tube support plates (TSPs) with broached quatrefoil holes. Chrome-plated Alloy 600 anti-vibration bars support the U-bend section of the tubes. The first 10 rows of tubes received a thermal stress-relief treatment to improve the corrosion resistance of the bend region.
The licensee provided the scope, extent, methods, and results of their SG tube inspections. In addition, the licensee 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:
For the Seabrook SGs, tube sub-populations with increased degradation susceptibility include tubes with ding/dent indications, and tubes with potentially high residual stress.
There are 67 tubes in service with potentially high residual stress (15 in SG A, 26 in SG B, 18 in SG C and 8 in SG D).
Operating leakage in SG B varied between 0.21 gallons per day (gpd) and 1.1 gpd during the operating cycles preceding the RFO 23 inspection. The leakage has been detected for several cycles, fluctuating between 0 and 1.2 gpd. No operating leakage has been detected in SG A, SG C, or SG D.
An upper-bundle flush, sludge lancing, and foreign object search and removal were performed in all four SGs. Approximately 166 pounds of sludge were removed from the four SGs after the upper-bundle flush (UBF) and sludge lancing. There was some hard sludge/scale during the top of tubesheet in-bundle exams of the sludge pile region.
Visual inspections were performed in all four SGs, including the steam dryers, primary moisture separators, swirl vanes, feedring, and J-nozzles. Pre-and post-bundle flush visual exams of peripheral tubes, anti-vibration bars, and sections of the uppermost TSPs were also completed. Some moderate buildup of debris and tube scale was noted on some tubes. The broached openings of the TSPs were relatively clear and unobstructed after the UBF. In SG B, a discolored band on the steam drum wall, just above the feedring, was observed. The discoloration was reported during RFO 21, and no changes were noted between the RFO 21 and RFO 23 inspections. The discoloration appears to have been caused by the spalling of a thin magnetite layer from the wall; however, the base material appears to be intact. A minimal amount of erosion was observed inside some of the J-tubes (in all four SGs) at the interface with the feedring; the condition was typical as observed in prior outages. No loose parts, foreign objects, broken welds, or any other abnormal conditions were found during the upper steam drum exams.
Nine outside diameter stress corrosion cracking (ODSCC) indications were reported during the RFO 23 SG inspection.
o None were in tubes with potentially high residual stress.
o Two of the crack-like indications were single axial indications (SAls) located in the sludge-pile at the hot-leg (HL) expansion transition region.
o Six of the remaining seven flaws were located at the uppermost TSP in SG B and were classified as SAls. Most indications were coincident with dents at the upper or lower edge of the TSP, and all except one was in the HL.
o One crack-like flaw indication in row 4 column 119 (R4C119) of SG B was detected during the full-length array exam and was influenced by an overlapping dent/ding signal that was traceable to RFO 03. Because of the interaction between the dent/ding and the new crack-like indication, the eddy current (EC) signal was complex, making it difficult to characterize the flaw.
The indication showed attributes of both volumetric wear and circumferential cracking. Diagnostic exams were performed with the +Point' as well as the Ghent probe. The most conservative approach was taken to classify the indication as a single circumferential indication (SCI) at a dent/ding. The tube was plugged and removed from service. There is no known history of an SCI at a dent/ding in Alloy 600TT SG tubing; therefore, this was treated as a potentially new degradation mechanism. Since this was the first reported occurrence of this mechanism in Alloy 600TT SG tubing, existing data and qualified techniques did not exist for developing the inputs for probabilistic models for initiation, growth, crack length and depth characteristics and probability of detection. Therefore, a deterministic approach was used to evaluate the forward-looking tube integrity assessment for this degradation mechanism in the tubes remaining in service.
Two tubes with indications required in situ pressure testing (ISPT) to demonstrate meeting condition monitoring (CM) limits, because tube integrity could not be demonstrated analytically.
o The tube in R4C119 of SG B required ISPT because there was no qualified EC technique for SCI at the Dent/Ding. Since this is the first indication of this type in A600TT and there are many instances of this degradation in mill-annealed Alloy 600, a CM limit curve for structural integrity performance criteria (SIPC) was plotted for ETSS 21410.1 using non-destructive evaluation (NDE) percent degraded area (PDA) as the structural variable.
With the NDE PDA of the flaw plotted on the burst pressure line, the indication fell well below the limiting PDA value, thereby analytically meeting the CM requirements for SIPC. However, since no industry/EPRI guidance existed for accident-induced leakage performance criteria (AILPC) screening of circumferential ODSCC at a dent/ding, the flaw was selected for ISPT to demonstrate that it met the AILPC. During ISPT, the tube did not leak or burst when tested at hold points to simulate normal operating pressure differential (NOPD), main steam line break (MSLB), and 3xNOPD conditions, demonstrating AILPC and SIPC.
o A tube in R28C57 of SG C with a 2.52 volt indication also required ISPT. This was because the indication exceeded the 2.4 Vcrit criterion in the EPRI ISPT guidelines for AILPC and because the structural equivalent depth of 81.1 percent through-wall (TW) exceeded the CM limit of 77.4 percent TW.
During ISPT, the maximum leak rate was 68.5 gpd at MSLB conditions, which was less than the 500 gpd AILPC. The tube did not burst and therefore met AILPC and SIPC.
Based on a review of the information provided, the NRC staff conclude 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 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.
ML25337A339 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DNRL/NCSG/BC NAME VSreenivas KEntz SBloom DATE 12/01/2025 12/4/2025 11/18/2025 OFFICE NRR/DORL/LPL1/BC (A)
NRR/DORL/LPL1/PM NAME UShoop VSreenivas DATE 12/12/2025 12/15/2025