ML24291A286
| ML24291A286 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 11/01/2024 |
| From: | David Wrona Plant Licensing Branch II |
| To: | Coffey B Florida Power & Light Co |
| Jordan N | |
| References | |
| EPID L-2024-LLR-0004 | |
| Download: ML24291A286 (1) | |
Text
November 1, 2024 Robert Coffey Executive Vice President, Nuclear and Chief Nuclear Officer Florida Power & Light Company 700 Universe Blvd.
Mail Stop: EX/JB Juno Beach, FL 33408
SUBJECT:
ST. LUCIE PLANT, UNIT NO. 1 - RELIEF REQUEST: PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(z)(1), EXTENSION OF INSPECTION INTERVAL FOR REACTOR PRESSURE VESSEL WELDS FROM 10 TO 20 YEARS (EPID L-2024-LLR-0004)
Dear Robert Coffey:
By letter dated January 18, 2024, as supplemented by letter dated June 26, 2024, Florida Power and Light Company (the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for relief from certain American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI requirements for the St. Lucie Plant (St.
Lucie), Unit No. 1 facility. The licensee requested relief from the requirements of the ASME Code,Section XI, Subsection IWB-2411, Inspection Program. The proposed alternative would extend the reactor pressure vessel (RPV) volumetric examination requirements for Examination Categories B-A and B-D welds from once each 10-year inservice inspection interval to once every 20 years.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), the licensee submitted Relief Request RR#7 requesting relief from the RPV examination requirements of American Society of Mechanical Engineers (ASME),Section XI, Subsection IWB-2411, Inspection Program, for St. Lucie, Unit No. 1, on the basis that the alternative provides an acceptable level of quality and safety.
The NRC staff has reviewed the proposed alternative, and as set forth in the enclosed safety evaluation, determined that the proposed alternative provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Therefore, the NRC staff authorizes the use of proposed alternative RR#7 at St. Lucie, Unit No. 1, for the remainder of the Fifth 10-Year ISI interval and for the duration of the Sixth 10-Year ISI interval of the unit.
All other ASME BPV Code,Section XI, requirements for which relief was not specifically requested and approved remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
If you have any questions, please contact the Project Manager, Natreon Jordan, at 301-415-7410 or Natreon.Jordan@nrc.gov.
Sincerely, David Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-335
Enclosure:
Safety Evaluation cc: Listserv DAVID WRONA Digitally signed by DAVID WRONA Date: 2024.11.01 15:29:30 -04'00'
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST RR#7 EXTENSION OF INSPECTION INTERVAL FOR REACTOR PRESSURE VESSEL WELDS FROM 10 TO 20 YEARS FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT, UNIT NO. 1 DOCKET NO. 50-335
1.0 INTRODUCTION
By letter dated January 18, 2024 (Agencywide Documents and Access Management System (ADAMS) Accession No. ML24018A064), as supplement by letter dated June 26, 2024 (ML24178A269), Florida Power & Light Company (the licensee) proposed an alternative to the examination requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Subsection IWB-2411, Inspection Program, for St. Lucie Plant (St. Lucie), Unit No. 1. The proposed alternative would extend the inspection of welds and nozzle inner radius under Examination Categories B-A and B-D of the ASME Code,Section XI from once each 10-year inservice inspection (ISI) interval to once every 20 years.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1), the licensee requested to use the proposed alternative on the basis that the alternative provides an acceptable level of quality and safety.
2.0 REGULATORY EVALUATION
2.1 Regulatory Requirements The ISI interval of ASME Code Class 1, 2, and 3 components is to be performed in accordance with Section XI of the ASME Code and applicable editions and addenda as required by 10 CFR 50.55a(g), Preservice and inservice inspection requirements, except where specific relief has been granted by the Nuclear Regulatory Commission (NRC) pursuant to 10 CFR 50.55a(g)(6)(i), Impractical ISI requirements: Granting of relief.
Pursuant to 10 CFR 50.55a(g)(4), lnservice inspection standards requirement for operating plants, ASME Code Class 1, 2, and 3 components (including supports) must meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year ISI interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(a)(1)(ii), 18 months prior to the start of the 120-month interval, subject to the conditions listed in 10 CFR 50.55a(b)(2).
Pursuant to 10 CFR 50.55a(z), Alternatives to codes and standards requirements, alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (1) the proposed alternatives would provide an acceptable level of quality and safety, or (2) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
The regulations in 10 CFR 50.61, Fracture toughness requirements for protection against pressurized thermal shock events, require fracture toughness for the reactor vessel shell material to protect the reactor vessel from pressurized thermal shock events.
The regulations in 10 CFR 50.61a, Alternate fracture toughness requirements for protection against pressurized thermal shock events, provide alternative methodology to calculate fracture toughness to meet the fracture toughness requirements to protect the reactor vessel shell from pressurized thermal shock events.
2.2 Regulatory Guidance The NRC staff used the following guidance in the evaluation of this request:
NRC Regulatory Guide (RG) 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988 (ML003740284).
NRC RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, January 2018 (ML17317A256).
NRC NUREG-1874, Recommended Screening Limits for Pressurized Thermal Shock (PTS), March 2007 (ML070860156).
3.0 TECHNICAL EVALUATION
3.1 ASME Code Components Affected
The affected components are the St. Lucie, Unit No. 1 reactor vessel (RV) shell welds and nozzles associated with the following ASME Code,Section XI examination categories and item numbers. These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME Code,Section XI, where Examination Category B-A welds are defined as Pressure Retaining Welds in Reactor Vessel, and Examination Category B-D welds are defined as Full Penetration Welded Nozzles in Vessels.
Examination Category Item Number Description B-A B1.10 Shell Welds B-A B1.11 Circumferential Shell Welds B-A B1.12 Longitudinal Shell Welds B-A B1.20 Head Welds B-A B1.21 Circumferential Head Welds B-A B1.22 Meridional Head Welds B-A B1.30 Shell-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inside Radius Section 3.2
Applicable Code Edition and Addenda
The ASME Code,Section XI, 2007 Edition through 2008 Addenda.
3.3
Applicable Code Requirement
The ASME Code,Section XI, IWB-2411 requires volumetric examination of essentially 100 percent of RV pressure-retaining welds identified in Table IWB-2500-1 once each 10-year ISI interval. The licensee stated that the fourth 10-year ISI interval for St. Lucie, Unit No. 1, is scheduled to end on February 10, 2028. The licensee further stated that the applicable Code for the fifth 10-year ISI interval will be selected in accordance with the requirements of 10 CFR 50.55a.
3.4
Reason for Request
The licensee stated that extension of the interval between examinations of the subject welds from 10 years to up to 20 years will result in a reduction in man-rem exposure and examination costs.
3.5 Proposed Alternative and Basis for Use The licensee proposed not to perform the ASME Code-required volumetric examination of the RV full penetration pressure-retaining Examination Category B-A and B-D welds, currently scheduled for 2025, during the fifth ISI interval. The licensee stated that it will perform the fifth ASME Code-required volumetric examination of the subject welds during the sixth ISI interval in 2037. The licensee further stated that the proposed inspection date is consistent with the latest revised implementation plan from PWR Owners Group (PWROG), Letter No. OG-10-238, Revision to the Revised Plan for Plant Specific Implementation of Extended Inservice Inspection Interval per WCAP-16168-NP, Revision 1, Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval, PA-MSC-0120, July 12, 2010 (ML11153A033).
The licensee proposed this alternative on the basis that the current 10-year examination interval can be revised to 20 years with negligible change in risk by satisfying the risk criteria specified in RG 1.174.
The licensee based its analysis on the topical report WCAP-16168-NP-A, Revision 3, Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval (ML11306A084).
This study focuses on risk assessments of materials within the beltline and extended beltline regions of the RV wall. The licensee compared the results from the plant-specific calculations for St. Lucie, Unit No. 1, to those obtained from the Combustion Engineering (CE) pilot plant evaluated in WCAP-16168-NP-A, Revision 3. Appendix A of the WCAP identifies the parameters to be compared. The licensee demonstrated that the parameters for St. Lucie, Unit No. 1, are bounded by the results of the CE pilot plant. As such, the licensee stated that St.
Lucie, Unit No. 1, is qualified for an ISI interval extension for the inspection of the subject welds.
Table 1 of the licensees Relief Request #7 (RR#7) lists the critical parameters investigated in the WCAP and compares the results of the Westinghouse pilot plant to those of St. Lucie, Unit No. 1. Tables 2 and 3 of RR#7 provide additional information that was requested by the NRC during the review of the WCAP and included in Appendix A of WCAP-16168-NP-A, Revision 3.
Table 2 of RR#7 provides a summary of the latest reactor vessel inspection for St. Lucie, Unit No. 1, and an evaluation of the recorded indications. The licensee stated that this information confirms that it has performed satisfactory examinations on the subject welds.
3.6 Duration of Proposed Alternative The licensee requested relief to be applicable to the St. Lucie, Unit No. 1, ISI program for the fifth and sixth 10-year inspection intervals.
3.7
NRC Staff Evaluation
3.7.1 Background The NRC staff reviewed Relief Request RR#7 to determine its consistency with topical report WCAP-16168-NP-A, Revision 3. Henceforth, WCAP-16168-NP-A, Rev. 3 will be referred to as WCAP-A. The topical report provides a basis for the acceptability of the proposed inspection intervals for Category B-A and B-D components at U.S. pressurized-water reactors (PWRs) designed by Westinghouse, Combustion Engineering, and Babcock and Wilcox (B&W) through the use of risk-informed analyses and probabilistic fracture mechanics for a pilot plant of each design. WCAP-A also contains the NRC staffs safety evaluation (SE) approving its generic use. The NRCs SE for WCAP-A finds the proposal acceptable for use based on consistency with the principles contained in RG 1.174. However, the NRC staffs SE imposes a condition that requires licensees to provide plant-specific information in six areas to demonstrate the applicability of WCAP-A to the licensees plant. The plant-specific information required by the condition is:
(1) Licensees must provide the 95th percentile total through-wall cracking frequency (TWCFTOTAL) and its supporting material properties at the end of the proposed 20-year ISI interval. The 95th percentile TWCFTOTAL must be calculated using the methodology in NUREG-1874, which is frequently referred as the NRC PTS Risk Study. The RTMAX-X and the shift in the Charpy transition temperature produced by irradiation defined at the 30 foot to pound (ft-lb) energy level, T30, must be calculated using the latest revision of RG 1.99 or other NRC-approved methodology.
(2) Licensees must report whether the frequency of the limiting design basis transients during prior plant operation are less than the frequency of the design basis transients identified in the PWROG fatigue analysis as significant contributors to fatigue crack growth.
(3) Licensees must report the results of prior ISI of RV welds and the proposed schedule for the next 20-year ISI interval. Each licensee shall identify the years in which future inspections will be performed, and the dates provided must be within plus or minus one refueling cycle of the dates identified in the implementation plan provided to the NRC in PWROG Letter No. OG-10-238 (ML11153A033).
(4) Licensees with B&W plants must (a) verify that the fatigue crack growth of 12 heat-up/cool-down transients per year that was used in the PWROG fatigue analysis bound the fatigue crack growth for all of its design basis transients, and (b) identify the design bases transients that contribute to significant fatigue crack growth.
(5) Licensees with RVs having forgings that are susceptible to underclad cracking and with RTMAX-FO values exceeding 240 °F must submit a plant-specific evaluation because the analyses performed in the WCAP-A are not applicable.
(6) Licensees seeking second or additional interval extensions shall provide the information and analyses requested in Section (e) of 10 CFR 50.61a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events.
The licensees proposed extended ISI interval is based on a risk-informed RV fracture mechanics analysis that was performed in accordance with the NRC staff-approved, risk-informed flaw analysis methods in WCAP-A. The methodology in WCAP-A was developed by the PWROG to satisfy through-wall cracking frequency (TWCF) criteria, specifically the total 95th percentile TWCF, TWCF95-TOTAL, for PWRs established in NRC NUREG-1874 and the delta large early release frequency (LERF) criteria specified in RG 1.174.
In Section 3.4 of the NRC staffs SE dated July 26, 2011, for WCAP-A (hereafter WCAP-A SE), the NRC staff specified plant-specific information that licensees must submit for alternative requests that are based on the methodology of WCAP-A. Tables A-1, A-2, and A-3 in Appendix A of WCAP-A show the format used for providing the plant-specific information. The licensee provided the plant-specific information in Tables 1, 2, and 3 of Section 5 of RR#7. The NRC staff evaluated the plant-specific information in the following subsections.
3.7.2 TWCF Evaluation Section 3.4 of the WCAP-A SE establishes the NRC staffs position that the maximum adjusted reference temperatures and 30 ft-lb shifts in adjusted referenced temperature values (i.e.,
RTMAX-X and T30 values), as defined in 10 CFR 50.61(a) may be calculated using the methods documented in RG 1.99, Revision 2, or in an alternate NRC-approved methodology using these types of parameters. The WCAP-A SE states that licensees submittals should include the material property and fluence information related to these parameters and that Table A-3 of WCAP-A identifies the information needed to be submitted.
Table 3 of RR#7 includes the material property and neutron fluence data, T30 values, and RTMAX-XX values for the RV base metal and weld components of St. Lucie, Unit No. 1, at 72 effective full-power years. The licensee stated that material properties and fluence inputs are based on WCAP-18609-NP, St. Lucie Units 1 & 2 Subsequent License Renewal: Time-Limited Aging Analyses on Reactor Vessel Integrity, July 2021.
The NRC staff verified the licensee calculated TWCF95-TOTAL value of 4.45x10-8 events per year for the RV, as shown in Table 3 of RR#7 were calculated in accordance with NUREG-1874, which is consistent with the condition for WCAP-A. The NRC staff verified that the TWCF95-TOTAL for the RPV is well within by the maximum (upper-bound) TWCF95-TOTAL value of 3.16x10-7 through-wall cracking events per reactor year that is established for CE-designed PWRs in WCAP-A; therefore, the licensees TWCF95-TOTAL for St. Lucie, Unit No. 1, was determined to be acceptable.
The NRC staff noted that the methodology in WCAP-A conservatively sets the TWCF95-TOTAL equal to the LERF value for the RV that may result from initiation of the postulated, limiting PTS event at a plant. Thus, based on the NRC staffs verifications discussed above, the NRC staff determined that the TWCF95-TOTAL value for the St. Lucie, Unit No. 1, RV meets the limit of 1x10-7 early release events per reactor year that is established for LERF values in RG 1.174, Revision 3.
3.7.3 Identification of Limiting Design Basis Transients Design Basis Transients Regarding the PTS transients, the licensee identified in Table 1 of RR#7 that the transients are defined in NRC letter report, Generalization of Plant-Specific Pressurized Thermal Shock (PTS)
Risk Results to Additional Plants (ML042880482), hereafter PTS Generalization Study, and that those transients serve as the limiting design basis transients for the subject welds. The NRC staff verified that for CE-designed PWRs, such as St. Lucie, Unit No. 1, the PWROGs methodology in WCAP-A uses the PTS transients that were defined in NUREG-1874 and clarified in the PTS Generalization Study as the limiting PTS transients for the PWROGs risk-informed flaw analysis that was included in WCAP-A. Therefore, the NRC staff finds the licensees transient basis to be acceptable based on the information in NUREG-1874 and the PTS Generalization Study. The NRC staff concluded in the WCAP-A SE that the PTS transient characteristics for a given nuclear steam supply system (NSSS) design of PWR are generically applicable for all PWRs designed by the same reactor NSSS vendor (i.e., CE for St. Lucie, Unit No. 1).
Regarding the cladding layers, the licensee reported in Table 1 of RR#7 that the cladding for the RV at St. Lucie, Unit No. 1, was deposited using a single layer. Thus, for the proposed alternative, the NRC staff concludes that the licensee did not need to evaluate the impacts that multiple pass layers would have on the design of the RV cladding because the cladding layer at St. Lucie, Unit No. 1, was deposited as a single layer, and is consistent with and bounded by the NRC staffs evaluation of the cladding layer in the WCAP-A SE.
Frequency and Severity of Design Basis Transients Table 1 of RR#7 indicates that the plant-specific basis for frequency and severity of design basis transients is bounded by the 13 cycles/reactor year of plant heatup and cooldown transients of the associated pilot plant assessed in WCAP-A. To demonstrate the bounding of the cycles at St. Lucie, Unit No. 1, in its supplement dated June 26, 2024, the licensee cited similar precedence described in Section 3.7.4 of the SE for St. Lucie, Unit No. 2 (ML21236A131). Chapter 3 of the St. Lucie, Unit No. 1, Updated Final Safety Analysis Report (UFSAR) establishes an upper bound limit of 500 heatup and cooldown cycles for the reactor coolant system (RCS) over the design life of the plant. This corresponds to an average of 8.33 for a cumulative 60-year licensing term and that this yearly frequency value for St. Lucie, Unit No. 1, is well within the maximum frequency limit of 13 cycles per reactor year assessed in WCAP-A.
The NRC staff confirmed in the WCAP-A SE that the PWROG established that the RCS heatup and cooldown transients are the limiting design transients for RPV fatigue flaw growth in CE-designed PWRs (e.g., St. Lucie, Unit No. 1). The NRC staff also confirmed that the PWROG established 13 cycles/reactor year as the maximum bounding number of heatup and cooldowns that could occur per reactor year for CE-designed PWRs. The NRC staff confirmed that the design basis transients used as input to the licensees equivalent fatigue crack growth analysis for St. Lucie, Unit No. 1, is consistent with design transients specified in the St. Lucie, Unit No. 1 UFSAR. The NRC staff noted that the St. Lucie, Unit No. 1, equivalent fatigue crack growth value for 13 heatup and cooldown cycles per year was less than the corresponding value for the CE pilot plant analyzed in WCAP-A and, therefore, is acceptable.
Based on the discussion above, the NRC staff finds that the licensees equivalent fatigue crack growth analysis for St. Lucie, Unit No. 1, is acceptable because it is sufficiently bounded by that analyzed and established for the pilot plant in WCAP-A.
3.7.4 Inservice Inspection Results and Proposed Future Inspections Inservice Inspection Results Section 3.4 of the WCAP-A SE establishes the NRC staffs position that licensees submitting risk-informed ISI extensions for their RV welds should report the results of prior ISI of the applicable RV weld locations.
Table 2 of RR#7 identifies the results of previously performed inspections of the subject welds.
The licensee stated that it has performed a preservice inspection, and four 10-year ISIs have been performed in 1983, 1996, 2008, and 2018.
The licensee stated that the latest RV ISI for St. Lucie, Unit No. 1, was conducted in accordance with the requirements of Appendix VIII of the ASME Code,Section XI, 1995 Edition with Editions and Addenda through 2000, as modified by the Performance Demonstration Initiative program and 10 CFR 50.55a(b)(2)(xiv, xv, and xvi). The licensee evaluated recordable indications to the acceptance standards of the ASME Code,Section XI, 2001 Edition with Addenda 2003 Addenda. The licensee stated that future ISIs will be performed to the ASME Section XI, Appendix VIII requirements.
As shown in Table 2 of RR#7, the licensee stated that it identified 41 total indications in the RV beltline and extended beltline regions during the most recently completed ISI. These subsurface indications are located in the upper-to-intermediate shell circumferential weld seam (Item 10 in Table 3), the intermediate-to-lower shell circumferential weld seam (Item 11 in Table 3), and the longitudinal weld seams in the upper shell, intermediate shell, and lower shell (Items 12, 17, 18, and 19 in Table 3). The licensee further stated that all 41 indications are acceptable per Table IWB-3510-1 of Section XI of the ASME Code. Of the 41 indications, there are 9 indications within the inner 1/10th or inner 1-inch of the reactor vessel wall thickness. Four flaws were located in the weld material, and five flaws were located in the plate material.
According to the licensee, the fourth 10-year inspection was the first ISI examination that detected the nine indications described above. There is no site-specific flaw growth data since these indications were evaluated as acceptable per the ASME Section XI Table IWB-3510-1 and are indicative of fabrication flaws typical of small slag inclusions.
A disposition of the nine flaws against the limits of the Alternate PTS Rule, 10 CFR 50.61a, is shown in Table 2 of RR#7 and the nine flaws were found to be acceptable per the requirements of the Alternate PTS Rule.
The NRC staff finds the licensees evaluation of its ISI results for the subject RV welds acceptable because the nine subsurface indications that were detected were acceptable per Section XI of the ASME Code and they were deemed insignificant to warrant additional analysis as specified in 10 CFR 50.61a.
Proposed Future Inspections Section 3.4 of the WCAP-A SE states that licensees should identify an ISI schedule for RV weld examinations that will be performed during the proposed 20-year ISI interval. The WCAP-A SE also established the NRC staffs position that the dates for the weld inspections must be within one refueling outage (i.e., plus or minus one refueling outage) of the revised dates identified for inspection in the implementation plan in PWROG Letter No. OG-10-238.
The licensee proposed not to perform required ASME Code volumetric examinations of the subject RV welds for the fifth 10-year ISI interval in 2025. The licensee proposed instead to perform these volumetric examinations in 2037. The licensee stated that this proposed inspection date is based on the implementation plan recommended in PWROG Letter No. OG-10-238 for St. Lucie, Unit No. 1.
The NRC staff finds the licensees proposed scope and schedule for inspecting the subject RV welds acceptable because the licensee will follow the implementation plan recommended for St. Lucie, Unit No. 1, in PWROG Letter No. OG-10-238, consistent with the WCAP-A SE.
3.7.5 Fatigue Crack Growth in B&W plants and Underclad Cracking of RV Forgings Condition Item 4 of the WCAP-A SE is related to fatigue crack growth in B&W plants. St. Lucie, Unit No. 1, is a CE-designed plant. Therefore, Condition Item 4 is not applicable to St. Lucie, Unit No. 1.
Condition Item 5 of the WCAP-A SE is related to underclad cracking of RV forgings. The St.
Lucie, Unit No. 1, RV was fabricated from rolled plates, not made of forgings, as shown in Table 3 of RR#7. Therefore, the issue of underclad cracking of RV forgings is not applicable to St. Lucie, Unit No. 1.
3.7.6 Submittal of Information Required by Section (e) in 10 CFR 50.61a Condition Item 6 of the WCAP-A SE is for licensees seeking second or additional interval extensions. In the supplement dated June 26, 2024, the licensee confirmed that the subject relief request is the first interval extension request for St. Lucie, Unit No. 1, that is based on the WCAP-A methodology. Therefore, the NRC staff determined that the licensee does not need to include the information required by Section (e) of 10 CFR 50.61a for the subject relief request.
3.7.7 NRC Staff Evaluation Summary The NRC staff determined that the licensee has satisfied all plant-specific information items specified in the WCAP-A SE. Therefore, the NRC staff finds that the licensees proposed risk-informed alternative provides an acceptable level of quality and safety in lieu of complying with the ASME Code,Section XI requirements.
4.0 CONCLUSION
The NRC staff has determined that the proposed alternative in the licensees request referenced above would provide an acceptable level of quality and safety. The NRC staff concludes that the licensee has adequately addressed the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Therefore, the NRC staff authorizes the use of proposed alternative RR#7 at St. Lucie, Unit No. 1, for the remainder of the Fifth 10-Year ISI interval and for the duration of the Sixth 10-Year ISI interval of the unit.
All other ASME BPV Code,Section XI, requirements for which an alternative was not specifically requested and authorized remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
Principal Contributors: C. Moyer, NRR J. Tsao, NRR Date: November 1, 2024
- by email OFFICE NRR/DORL/LPL2-2/PM NRR/DORL/LPL2-2/LA* NRR/DNRL/NVIB/BC*
NRR/DORL/LPL2-2/BC*
NAME NJordan ABaxter ABuford (On Yee for)
DWrona DATE 10/17/2024 10/22/2024 10/11/2024 11/1/2024