LR-N25-0084, Steam Generator Tube Inspection Report - Twenty-Seventh Refueling Outage
| ML25279A204 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 10/06/2025 |
| From: | Jurek S Public Service Enterprise Group |
| To: | Document Control Desk |
| References | |
| LR-N25-0084 | |
| Download: ML25279A204 (1) | |
Text
Shane Jurek Regulatory Programs Manager - Licensing - PSEG Nuclear PO Box 236 Hancocks Bridge, New Jersey 08038-0221 Shane.Jurek@pseg.com LR-N25-0084 Technical Specification 6.9.1.10 October 6, 2025 US Nuclear Regulatory Commission ATTN : Document Control Desk Washington, DC 20555-0001 Salem Generating Station, Unit 2 Renewed Facility Operating License No. DPR-75 NRC Docket No. 50-311 Subject :
Steam Generator Tube Inspection Report - Twenty-Seventh Refueling Outage
References:
- 1. PSEG letter to NRC, Steam Generator Tube Inspection Report - Twenty-seventh Refueling Outage dated May 15, 2025 (ADAMS Accession No. ML25135A078).
- 2. NRC email to PSEG, NRC Request for Additional Information for Salem 2R27 SGTIR (L-2025-LRO-0032) dated September 12, 2025 (ADAMS Accession No. ML25255A067).
PSEG Nuclear LLC (PSEG) submitted, in Reference 1, an updated steam generator (SG) tube inspection report (SGTIR) per technical specification 6.9.1.10. The NRC notified PSEG in Reference 2 of the need for additional information to complete its review. The attachment to this letter provides the PSEG response to the additional information requests.
There are no new or revised regulatory commitments contained in this letter.
If there are any questions or if additional information is needed, please contact Mr. Harry Balian at Harry.Balian@pseg.com.
Very truly yours, Shane Jurek Regulatory Programs Manager - Licensing PSEG Nuclear LLC o PSEG I NUCLEAR
October 6, 2025 Page 2 LR-N25-0084
Attachment:
Response to Requests for Additional Information cc:
USNRC Regional Administrator Region 1 USNRC NRR Project Manager - Salem USNRC Senior Resident Inspector - Salem NJ Department of Environmental Protection, Bureau of Nuclear Engineering NJ Occupational Safety and Health Bureau of Boiler and Pressure Vessel Compliance
LR-N25-0084 Attachment Page 1 of 2 ATTACHMENT Response to Requests for Additional Information PSEG Nuclear LLC (PSEG) submitted an updated steam generator (SG) tube inspection report (SGTIR) per technical specification 6.9.1.10 to the United States Nuclear Regulatory Commission (NRC) under PSEG letter LR-N25-0048, Steam Generator Tube Inspection Report
- Twenty-seventh Refueling Outage on May 15, 2025 (ADAMS Accession No. ML25135A078).
The NRC notified PSEG of the need for additional information to complete its review by electronic mail, NRC Request for Additional Information for Salem 2R27 SGTIR (L-2025-LRO-0032) dated September 12, 2025 (ADAMS Accession No. ML25255A067). This attachment provides the PSEG response to the additional information requests.
Question 1 Table 3 of attachment 1 to LR-N25-0048 provides the Salem, Unit 2 condition monitoring summary. For tube wear at the tube support plates (TSPs), the 2R22 to 2R27 operational assessment projected depth was 51.1 percent through-wall. The staff notes the maximum depth of tube wear at TSPs measured during 2R27 was 13 percent through-wall. The condition monitoring limit for TSP wear in Table 3 is shown as 47.5 percent through-wall, which is less than the conservative 2R22 operational assessment projected depth. The updated steam generator inspection report from 2R22 (ADAMS Accession No. ML21305A003) stated that a conservative structural depth for wear at TSPs was 52 percent through-wall. Discuss if the 47.5 percent through-wall condition monitoring limit shown in Table 3 is correct.
PSEG Response The 47.5 percent through-wall (%TW) condition monitoring limit shown in Table 3, provided in LR-N25-0048, is correct. Values in both Table 3 and Table 4 of LR-N25-0048, and the conservative structural depth (i.e. - OA Structural Limit) for wear at TSPs of 52%TW as documented in response to question d of the 2R22 report (ADAMS Accession No. ML21305A003), are also correct. A key difference in the condition monitoring (CM) limit values as compared to the operational assessment (OA) structural limit values is a CM limit includes applicable eddy current test (ECT) technique regression and uncertainty while the OA structural limit does not include ECT uncertainty. Another difference is the normal operating pressure differential (NOPD) applied in the calculation of the CM Limit or OA Structural Limit can vary with operating interval and conservatism. Note 1 in Table 3 and Note 1 in Table 4 of LR-N25-0048 also provide similar clarification as provided above. The calculations of CM limits and OA structural limits are consistent with guidance provided in the Electric Power Research Institute, Inc. (EPRI) topical report 3002020909, Steam Generator Management Program: Steam Generator Integrity Assessment Guidelines - Rev 5. As the staff also noted, the 2R27 maximum depth of TSP wear identified in outage 2R27 is documented as 13%TW. This is less than the CM Limit of 47.5%TW as shown in Table 3 of LR-N25-0048 and therefore the tube integrity performance criteria in TS 6.8.4.i.b are satisfied. This also demonstrates the analysis of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment), as summarized in response to question d within the 2R22 report, was very conservative.
LR-N25-0084 Attachment Page 2 of 2 Question 2
, section f of LR-N25-0048 provided the results of any steam generator secondary side inspections. Secondary side visual inspections included all steam drum components in all four steam generators. Discuss if any degradation was detected in the inspection of the feed ring, J-nozzles, moisture separator equipment, flow restrictor outlet nozzles, and other internal supports.
PSEG Response No degradation was detected in the inspection of the feed rings, J-nozzles, moisture separator equipment, flow restrictor outlet nozzles, or other internal supports.
Question 3
, section f of LR-N25-0048, provides a discussion of the secondary side foreign object search and retrieval (FOSAR) and indicates that no eddy current loose parts were reported in 2R27 and that no loose part wear was identified on any tube. In addition, all foreign material remaining in the steam generators has been evaluated and determined to be acceptable for the next five cycles. This section also states, Appendix F contains a summary of the SSI [secondary-side inspection] findings (FOSAR). There is no Appendix F in the submittal.
Therefore, the NRC requests the licensee to indicate if this sentence was inadvertently included; otherwise, the NRC staff requests the licensee to submit Appendix F.
PSEG Response This sentence was inadvertently included.
Question 4 Confirm whether the title of the graphic in Attachment 3 of LR-N25-0048 contains a typographical error and should state Model 61/19T instead of Model F.
PSEG Response The title of the graphic in Attachment 3 of LR-N25-0048 contains a typographical error and should state Model 61/19T instead of Model F. Also note page 1 of PSEG letter LR-N25-0048 refers to Appendix B and this was intended to state See Attachment 3.