L-2025-148, Deviation from EPRI MRP-227
| ML25234A065 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 08/20/2025 |
| From: | Mack K Florida Power & Light Co |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| L-2025-148 | |
| Download: ML25234A065 (1) | |
Text
L-2025-148 Enclosure Page 1 of 2 PSL Unit 1 - Deviation from MRP-227 Reason for Deviation:
EPRI MRP-227 (Reference 1) includes the following NEI 03-08 Needed requirement:
Needed: Examination results that do not meet the examination acceptance criteria defined in Section 5 of these guidelines shall be recorded and entered in the owners plant corrective action program and dispositioned. Engineering evaluations used to disposition an examination result that does not meet the examination acceptance criteria in Section 5, shall be conducted in accordance with NRC approved evaluation methods (i.e., ASME Code Section XI, PWR Owners Group topical report WCAP-17096-NP-A or equivalent method).
PSL Unit 1 is deviating from this Needed requirement for its evaluation of flaws identified on the reactor vessel core barrel during the 1R28 Spring 2018 refueling outage. Specifically, PSL is deviating from the Needed requirement to disposition the flaws in accordance with NRC approved evaluation methods such as ASME Code Sections XI and PWROG topical report WCAP-17096 (Reference 2).
Note - The need for this deviation was identified following review of EPRI letter MRP 2025-003 (Reference 3), which provided perspective and clarification on the subject Needed requirement. This review is documented by OE AR 2509729.
Expected Duration for which Justification/Deviation applies:
This deviation will apply until a WCAP-17096 compliant evaluation is completed. This evaluation is expected to be completed prior to the U1R33 Fall 2025 refueling outage. This outage also includes re-inspection of the core barrel flaws.
Technical Justification:
The core barrel flaws were initially evaluated in accordance with WCAP-17096 via Westinghouse dispositions performed in response to discovery, and following an 18-month re-inspection (References 4 and 5). These evaluations concluded the indications were acceptable for long-term operation and recommended a 10-year re-inspection frequency.
In November 2023, EPRI issued letter MRP 2023-011 (Reference 6), which notified the industry of the conclusion of an expert panel assessment of fracture toughness values for irradiated stainless steel PWR internals components. This assessment resulted in a significant reduction in the fracture toughness values that could be utilized for WCAP-17096 compliant flaw evaluations and crack growth projections. This conclusion partially invalidated the evaluations that supported the PSL U1 core barrel flaw disposition.
PSL pro-actively contracted Westinghouse to evaluate the potential impact of the fracture toughness reduction prior to the final conclusion included in Reference 6. In response, Westinghouse provided a Justification for Continued Operation (JCO) that concluded the reduction in fracture toughness did not impact the ability of the core barrel to perform its safety function. This JCO (Reference 7) has been updated for each operating cycle following notification of the potential issue.
The JCO included the following technical justifications:
L-2025-148 Enclosure Page 2 of 2
- 1. The reduction in fracture toughness results in postulated flaw sizes that cannot analytically support faulted events, specifically loss-of-coolant (LOCA) loads. The acceptability under normal/upset loadings and/or flow induced vibration (FIV) conditions are not impacted.
- 2. The impact associated with LOCA loads is mitigated by automatic insertion of the core control elememts.
- 3. The 18-month ultrasonic re-inspection identified no flaw growth during the operating cycle.
- 4. The JCO was supplemented with Neutron Noise Monitoring. This was a conservative, one-time activity meant to verify the integrity of the core barrel.
In addition to the JCO, PSL completed a leak-before-break analysis that results in reduced postulated LOCA loads. These LOCA loads are being utilized to support an update to the flaw evaluation and restore compliance to WCAP-17096. This update will be completed prior to the fall 2025 refueling outage.
Conclusion/Findings:
Deviation from the NEI 03-08 Needed requirement is technically justified by the referenced Westinghouse JCO, which describes the lack of safety and/or commercial concern during normal operating and transient loads. The impact to LOCA load conditions is mitigated by the automatic shutdown capability, as well as the pending flaw calculation update that will restore compliance with an NRC approved methodology.
References:
- 1. EPRI MRP-227, Revision 1-A, Pressurizer Water Reactor Internals Inspection and Evaluation Guidelines, December 2019.
- 2. Westinghouse Report WCAP-17096-NP-A, Rev. 03, Reactor Internals Acceptance Criteria Methodology and Data Requirements, August 2023.
- 3. EPRI Letter MRP 2025-003, MRP-227 - Use of NRC Approved Methodologues for PWR Internals Evaluations, March 5, 2025.
- 4. Westinghouse Letter LTR-AMLR-18-22, Rev.0, Disposition of Indications Observed in the Core Support Barrel at St. Lucie Unit 1, March 30 2018.
- 5. Westinghouse Report WCAP-18452, Rev. 01, St. Lucie Unit 1 Core Support Barrel and Core Shroud Flaw Analysis, March 2020.
- 6. Westinghsoue Letter MRP-2023-011, Summary of EPRI Expert Panel Assessmenr of Irradiated Stainless Steel Fracture Toughness in PWR Internals Components, November 6, 2023.
- 7. Westinghouse Letter LTR-AMLR-21-19, Rev.02, Justification for Continued Operation for St.
Lucie Unit 1 Fall 2025 Refueling Outage, February 29, 2024.