ML24071A091
| ML24071A091 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 04/22/2024 |
| From: | Jeffrey Whited Plant Licensing Branch III |
| To: | Coffey B Point Beach |
| Wall, S P | |
| References | |
| EPID L-2023-LLR-0032 | |
| Download: ML24071A091 (11) | |
Text
April 22, 2024
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:
POINT BEACH NUCLEAR PLANT, UNIT 2 - ISSUANCE OF RELIEF REQUEST I6-RR EXTENSION OF THE UNIT 2 STEAM GENERATOR PRIMARY NOZZLE DISSIMILAR METAL WELDS SIXTH 10-YEAR INSERVICE INSPECTION PROGRA M INTERVAL (EPID L-2023-LLR-0032)
Dear Robert Coffey:
By letter dated June 27, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23178A142), NextEra Energy Point Beach, LLC (NextEra, the licensee) submitted relief requests (RRs) Nos. I6-RR-01, I6-RR-02, and I6-RR-03, to the U.S. Nuclear Regulatory Commission (NRC) for the use of alternatives to the requirements in American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),
section XI, Rules for lnservice Inspection of Nuclear Power Plant Components, at Point Beach Nuclear Plant (Point Beach), Units 1 and 2, associated with the sixth 10-year inservice inspection (ISI) interval.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1),
Acceptable level of quality and safety, the licensee requested to use the proposed alternative in RR I6-RR-03 on the basis that the proposed alternative will provide an acceptable level of quality and safety.
The NRC staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, 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 use of the licensees proposed alternative I6-RR-03 for up to and including the fall 2030 RFO, but not to exceed 9 calendar years from the prior volumetric and surface examination.
All other ASME Code, section XI, requirements for which relief was not specifically requested and approved in the subject requests for relief remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.
The RRs identified as I6-RR-01 and I6-RR-02 will be handled under separate NRC correspondence.
R. Coffey
If you have any questions, please contact the Project Manager, at 301-415-2855 or email at Scott.Wall@nrc.gov.
Sincerely,
Jeffrey A. Whited, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Docket Nos. 50-266 and 50-301
Enclosure:
Safety Evaluation
cc: Listserv SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
REGARDING THE SIXTH 10-YEAR INTERVAL INSERVICE TESTING INTERVAL
10 CFR 50.55a RELIEF REQUEST I6-RR-03
POINT BEACH NUCLEAR POWER PLANT, UNIT 2
NEXTERA ENERGY POINT BEACH. LLC
DOCKET NO. 50-301
1.0 INTRODUCTION
By letter dated June 27, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23178A142), NextEra Energy Point Beach, LLC (NextEra, the licensee) submitted relief requests (RRs) Nos. I6-RR-01, I6-RR-02, and I6-RR-03, to the U.S. Nuclear Regulatory Commission (NRC) for the use of alternatives to the requirements in American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) section XI, Rules for lnservice Inspection of Nuclear Power Plant Components, at Point Beach Nuclear Plant (Point Beach), Units 1 and 2, associated with the sixth 10-year inservice inspection (ISI) interval which began on August 1, 2022, and is scheduled to end on July 31, 2032.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(1),
Acceptable level of quality and safety, the licensee requested to use the proposed alternative in RR I6-RR-03, for a unique weld configuration at Point Beach, Unit 2, on the basis that the proposed alternative will provide an acceptable level of quality and safety. In RR I6-RR-03, the licensee proposed to delay the volumetric ISI of the steam generator (SG) inlet nozzle-to-safe end dissimilar metal (DM) butt welds at the Point Beach, Unit 2. The alternatives contained in I6-RR-01 and I6-RR-02, will not be discussed further in this safety evaluation (SE) and will be handled under separate correspondence.
2.0 REGULATORY EVALUATION
Pursuant to 10 CFR 50.55a(g)(6)(ii)(F), all holders of operating licenses or combined licenses for pressurized-water-reactors (PWRs) as of or after June 3, 2020, shall implement the requirements of ASME Code Case N-770-5, Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N06082 or UNS W86182 Weld Filler Material with or without Application of Listed Mitigation Activities,Section XI, Division 1, instead of ASME Code Case N-770-2, subject to the conditions specified in paragraphs (g)(6)(ii)(F)(2) through (16) of section 50.55a, by no later than 1 year after June 3, 2020.
Enclosure
Regulation 10 CFR 50.55a(z), Alternatives to codes and standards requirements, states, in part, that alternatives to the requirements of 10 CFR 50.55a(g) may be used, when authorized by the NRC, if the licensee demonstrates: (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.
Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request and the NRC to authorize the alternative requested by the licensee.
3.0 TECHNICAL EVALUATION
3.1 Licensees Proposed Alternative
By letter dated December 13, 2019 (ML19339H747), the NRC approved proposed alternative 2-RR-17 (ML19241A492) for Point Beach, Unit 2, authorizing the licensee to perform the volumetric examination of the SG inlet and outlet nozzle-to-safe end DM butt welds in the units fall 2021 refueling outage instead of its spring 2020 refueling outage. It is noted that the licensee completed the baseline volumetric examination of the subject SG nozzle-to-safe end DM butt welds in the October 2021 refueling outage as required by 10 CFR 50.55a(g)(6)(ii)(F).
3.1.1 Affected Component
ASME Code Class 1 SG inlet nozzle-to-safe end DM butt welds are the components for which the alternative is sought. In accordance with table 1 of ASME Code Case N-770-5, the licensee classified the two Point Beach, Unit 2, SG inlet nozzle-to-safe end DM butt welds identified with identification Nos. RC-34-MRCL-AI-05 and RC-34-MRCL-BI-05, as Inspection Item A-2.
The licensee stated that the two SGs of Point Beach, Unit 2, were replaced in the units fall 1996 refueling outage. The materials of construction of the SGs are primarily carbon steel, with the channel head and nozzles clad with austenitic stainless steel, and the safe ends are stainless steel. Alloy 182 buttering and Alloy 82 weld materials joined the SGs nozzles to the safe ends.
During fabrication of the SGs inlet and outlet nozzle-to-safe end DM butt welds at the factory, an Alloy 52 inlay was installed on the inside diameter (ID) surface of the Alloy 82/182 and adjacent base materials as a protective barrier against the primary water stress corrosion cracking (PWSCC). Alloy 52 is known to be less susceptible to the PWSCC than Alloy 82/182.
The licensee stated that the SGs primary nozzles are exposed to the normal operating pressure of 2250 pounds per square inch absolute. The normal operating temperature for the inlet nozzles is 611.1 degrees Fahrenheit ( °F) and 543 °F for the outlet nozzles.
3.1.2 ASME Code of Record
The 2017 Edition of the ASME Code, section XI, is the code of record for the sixth 10-year ISI program interval which began on August 1, 2022, and is scheduled to end on July 31, 2032.
3.1.3 ASME Code Requirements
In accordance with table 1 in Code Case N-770-5, the SG inlet nozzle-to-safe end DM butt welds are classified as Inspection Item A-2 and are required to be volumetrically examined every 5 calendar years, not to exceed 9 calendar years.
3.1.4 Reason for Request
The next required examination for the Point Beach, Unit 2, SG inlet nozzle-to-safe end DM welds will be in spring 2026, with the SG outlet nozzle-to-safe end DM welds required in fall 2030. The alternative is being requested to extend the inlet nozzle inspections to the fall 2030 refueling outage. This is to allow for a coordinated examination schedule between the inlet nozzle and outlet nozzle DM welds. This will allow the licensee to only drain the reactor coolant system (RCS) to low levels and open the steam generator manways once instead of twice, thus minimizing the impact to nuclear, radiological, and industrial safety.
3.1.5 Proposed Alternative
The licensee proposed to delay the volumetric examination for the SG inlet nozzle-to-safe end DM butt welds. The proposed alternative is to perform the volumetric examination of the SG inlet nozzle-to-safe end DM butt welds in the fall 2030 refueling outage and not to exceed 9 calendar years from the prior examination which was completed in October 2021.
3.1.6 Basis for Alternative
The licensee stated that the proposed alternative allows for a coordinated schedule for the SG inlet and outlet nozzle-to-safe end DM butt weld examinations and the planned SG tube examinations. By this coordination, draining of the RCS to low levels (i.e., mid-loop) and opening of the SGs manways would occur once instead of twice in the sixth 10-year ISI interval, thus, the impact of these activities to nuclear, radiological, and industrial safety would be minimized.
As discussed below, the licensees basis for the proposed alternative relied on: (1) acceptable results from prior inspections of the subject DM butt welds and (2) a flaw tolerance evaluation for the inlet nozzle DM butt welds to demonstrate reasonable assurance of the integrity of the welds until the next proposed inspection.
3.1.7 Duration of Proposed Alternative
The duration for which the relief is requested is for up to and including the fall 2030 refueling outage but not to exceed 9 calendar years from the prior examination which was completed in October 2021.
3.2 NRC Staff Evaluation
The NRC staff has evaluated I6-RR-03 pursuant to 10 CFR 50.55a(z)(1). The NRC staff focused on whether the alternative (i.e., accepting deferral of the volumetric examination for the SG inlet nozzle-to-safe end DM butt welds from spring 2026 until fall 2030) provides an acceptable level of quality and safety. To reach a conclusion, the NRC staff reviewed the
licensees inspection history and flaw tolerance evaluation, as well as performing independent flaw tolerance analyses for the bounding case of the Point Beach, Unit 2, SG inlet nozzle-to-safe end DM butt welds.
3.2.1 Prior Inspections of the Subject Welds
The licensee stated that the SG inlet nozzle-to-safe end DM butt welds have been in operation for more than 19 effective full power years (EFPY) with no cracking at hot-leg operating temperatures since their replacement in 1997. The history of the licensees inspection activities on the SG inlet and outlet nozzle-to-safe end DM butt welds is summarized below:
ASME Code, Section Ill-required surface examination using the liquid penetrant testing (PT) and volumetric examination using radiography during fabrication of the DM butt welds;
ASME Code, Section XI-required preservice inspection (PSI) by PT and volumetric examination using ultrasonic testing (UT) prior to putting the SG into service;
Baseline examinations as mandated by 10 CFR 50.55a(g)(6)(ii)(F)(3) with conditions, including ID [inside diameter] surface examination using automated eddy current testing (ECT) qualified to ASME Code,Section XI, Appendix IV and a volumetric examination using automated phased array ultrasonic testing (PAUT) qualified to ASME Code,Section XI, Appendix VIII;
Bare-metal visual examination (VE) as mandated by 10 CFR 50.55a(g)(6)(ii)(E) with conditions;
ASME Code,Section XI required visual examination (VT-2) as part of the RCS pressure or system leakage test at end of each refueling outage [RFO].
The licensee also noted that the subject DM butt welds will continue to receive the required VE and the system leakage test accompanied by the VT-2 for the remainder of the fifth 10-year ISI interval.
The licensee stated that there have not been any unacceptable indications (i.e., surface-breaking and/or subsurface flaws) identified in any of the SG inlet nozzle-to-safe end DM butt welds by the examinations performed. Furthermore, the results of ECT and PAUT ensured that there were no flaws within the inner 1/3 of weld wall thickness which could propagate through the Alloy 52 inlays into the Alloy 82/182 welds and cause pressure boundary leakage or failure.
Regarding repair history of the subject welds, the licensee stated that the fabrication of SG inlet and outlet nozzle-to-safe end DM butt welds and the installation of the Alloy 52 inlays were done at the Westinghouse fabrication facility. A review of the manufacturing records including material disposition reports showed no weld repair dispositions were done for the SG inlet and outlet nozzle-to-safe end DM butt welds (as discussed in previous proposed alternative 2-RR-17 dated December 13, 2019). The NRC staff finds the operating experience at Point Beach provides reasonable assurance that the initial flaw size for the flaw tolerance evaluation is bounding.
Further, the NRC staff note that the lack of previous cracking and overall service history of Alloy 52/690 materials provides additional risk insights that the assumed initial flaw size for the flaw tolerance evaluation is bounding.
3.2.2 Flaw Tolerance Analyses
Additional support for the acceptability of extendi ng the examination interval for the SG inlet nozzle-to-safe end DM butt welds is contained in the plant-specific flaw tolerance analyses documented in Attachment 2 of I6-RR-03, Westinghouse LTR-SDA-19-071-NP, Revision 0, Point Beach Unit 2 Steam Generator Safe-End Dissimilar Metal Weld Alloy 52 Inspection Extension. The licensee used industry guidance in Electric Power Research Institute (EPRI)
Materials Reliability Program (MRP) MRP-287, Primary Water Stress Corrosion Cracking (PWSCC) Flaw Evaluation Guidance.
As documented in the submittal letter, the licensee performed plant-specific flaw tolerance analyses to demonstrate that postulated ID axial and circumferential flaws in the DM butt welds would not grow to the ASME Code allowable flaw size between the planned examinations (i.e.,
between the October 2021 RFO and the fall 2030 RFO). Based on Point Beach, Unit 2, operational data and anticipated RFOs scheduled between October 2021 and fall 2030, the licensee projected the plant to operate at full power for 8.6 EFPY between the proposed examinations. The flaw tolerance analyses were performed for the SG inlet nozzle-to-safe end DM weld locations. Due to their operating at a higher temperature, the primary water stress corrosion cracking PWSCC growth rates for the SG inlet nozzle-to-safe end DM weld locations will be higher than for the SG outlet nozzle-to-safe end locations.
The analyses assumed initial cracks in the inlays due to postulated fabrication defects. The initial postulated defects were a 0.059-inch deep axial flaw with an aspect ratio (i.e., flaws length divided by depth) of 2 and a 0.059-inch deep circumferential flaw with an aspect ratio of 10, which represented welding fabrication flaws through the first, inside layer of the two weld pass layer deep Alloy 52 inlays.
The NRC staff reviewed the licensees inputs for their flaw tolerance calculation. Potential PWSCC growth through the Alloy 52 inlay material, and then through the Alloy 82 weld material, was evaluated using the normal operating temperature and pressure at the SG inlet nozzles, the normal operating steady state piping loads, and weld residual stresses (WRSs). The licensee stated that the WRSs in the SG inlet nozzle-to-safe end DM butt welds were computed using a plant-specific finite element analysis (FEA). In calculating WRS distributions, the licensee conservatively assumed 50 percent ID weld repairs. The FEA modeling included a portion of the low alloy steel nozzle, the stainless steel safe end, a portion of the stainless steel piping, the DM weld attaching the nozzle to the safe end along with an inlay on the inside surface, and the stainless steel weld attaching the safe end to the piping. The NRC staff finds these inputs are consistent with MRP-287 and bound the operating conditions and experience., Therefore they are reasonable for use in these analyses.
The NRC staff notes that for the PWSCC growth in Alloy 52, the licensee applied a factor of improvement (FOI) of 18 to the crack growth rate of Alloy 182 in EPRI MRP-115, Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds, to account for the increased resistance of Alloy 52 to PWSCC. The justification for the FOI of 18 for Alloy 52 was provided in EPRI MRP-375, Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles. However, the NRC staff has not generically approved the use of the FOI factors of
MRP-375. Therefore, the NRC staff performed a series of independent flaw tolerance evaluations utilizing NRC staff FOI values for the Alloy 52 diluted material with Alloy 82/182 materials.
Figures 7-1 and 7-2 of Attachment 2 to proposed alternative I6-RR-03 demonstrate that it would take nearly 8.7 EFPY for the initial postulated 0.059-inch deep axial flaw and 9.4 EFPY for the initial postulated 0.059-inch deep circumferentia l flaw to grow to the ASME Code allowable depth limit of 75 percent through-wall thickness for the bounding SG inlet nozzle-to-safe end welds. Therefore, the licensee concluded that the results provided justify the requested change in the time of the next examinations as proposed in I6-RR-03 for the subject SG inlet nozzle-to-safe end DM butt welds.
3.2.3 Staffs Independent Flaw Tolerance Analyses
The NRC staff performed independent flaw tolerance analyses to evaluate whether the projected growth of assumed PWSCC surface-connected flaws in the subject DM butt welds during the proposed period between inspections (i.e., from October 2021 RFO to fall 2030 RFO) would exceed the ASME Code allowable flaw size limit. The NRC staff began by evaluating aspects of the licensees flaw tolerance analyses - specifically the assumed initial defects, the characterization of WRSs, and the methodology for calculating PWSCC growth - for inclusion in the staffs independent analyses. Of note:
For the postulated initial defect size, the licensee used an aspect ratio of 10 for the circumferential flaw and an aspect ratio of 2 for the axial flaw. The NRC staff finds that postulated depth and aspect ratios used are adequate. The NRC staff, therefore, used the initial flaws proposed by the licensee in its independent analyses.
The axial and hoop WRS distributions provided in I6-RR-03 assumed a 50 percent ID weld repair. To develop these WRS distributions, the licensee used FEA which involves modeling as-built geometry of the nozzle-to-safe end DM weld and safe end-to-pipe weld and simulating the steps of the fabrication and welding process. The NRC staff notes that Section 3.6, Attributes of an Acceptable Residual Stress Analysis, of EPRI MRP-287 identifies the expectation that a 50 percent ID weld repair would be used to support analysis for NRC review. Based on these attributes, the NRC staff found the licensees calculated WRS distributions to be acceptable and used the licensees calculated WRSs in its independent analyses.
For Alloy 82/182, the licensee used the 75 th percentile crack growth rate data for Alloy 182 based on EPRI MRP-115. The NRC staff finds that MRP-115 is a generally acceptable source for PWSCC growth laws for Alloy 82/182 weld metals, and thus is adequate for this analysis. The NRC staff, therefore, used the same PWSCC growth rates for Alloy 82/182 in its independent analyses.
For Alloy 52, the licensee used an FOI of 18 based on EPRI MRP-375. Alternatively, the NRC staff relies upon Alloy 690/52/152 crack growth rate data from two NRC contractors: Pacific Northwest National Laboratory (PNNL) and Argonne National Laboratory (ANL). This data, documented in a data summary report (ML14322A587),
generally supports the contention that Alloy 690/52/152 is more crack resistant to crack growth but differs from the MRP-375 data in some respects. The staff characterized an alternative FOI from MRP-375 for diluted Alloy 52 as documented in the proposed rule
published in the Federal Register, 83 FR 56156 dated November 9, 2018. The NRC staff used the data for weld dilution zones from PNNL and ANL as well as the recommendations of Question and Answer No. 29 in NRC Public Meeting Summary, Summary of Public Meeting Between the Nuclear Regulatory Commission Staff and Industry Representatives on Implementation of ASME Code Case N-770-1, dated August 12, 2011 (ML112240818), regarding inlayed weld FOI of 10 for diluted Alloy 52 weld material.
For circumferential flaws, the NRC staffs analyses confirmed the lic ensees conclusion that the structural integrity of the SG inlet nozzle-to-safe end DM butt welds would be maintained through the period of the proposed volumetric ins pection extension. In addition, the NRC staffs analyses for the circumferential flaws show a significant margin exists for time to the ASME Code allowable depth limit of 75 percent.
For axial flaws, the NRC staffs analyses found t he flaws could potentially grow to exceed the ASME Code allowable depth limit of 75 percent and potentially cause leakage within the proposed period between volumetric inspections (i.e., November 2021 to fall 2030). However, an axially-oriented flaw in the DM weld is bounded by low alloy steel or stainless steel on either end. Since a PWSCC type flaw will not propagate significantly into the stainless steel or the low alloy steel adjacent to the DM weld, an axially-oriented flaw cannot grow sufficiently large in length to cause rupture of the weld and adjacent piping system within the duration of the licensees proposed alternative. Thus, the NRC staff applied risk insights to assess safety implications of piping with axial flaws that exceed the allowable as discussed below.
3.2.4 Risk Insights Consideration
The NRC staff considered risk insights to assess the results of its independent confirmatory analysis since the axial flaws do not exhibit adequate margin to the ASME Code allowable depth limit of 75 percent. The NRC staffs risk insights were based on: (1) prior volumetric and surface examinations as well as periodic visual examinations; (2) level of conservative inputs to the analysis to account for uncertainties; (3) leakage or failure of the welds that could lead to a concern for a loss-of-coolant accident (LOCA); (4) existing plant leak detection and monitoring systems; and (5) operating experience.
While the NRC staffs analysis found that an axially-oriented PWSCC type flaw with conservative inputs could cause leakage during the period of the extended inspection interval, any such leakage would be small due to the morphology of PWSCC type flaws and not directly challenge the safety of the plant. Further, the licensee has existing plant procedures such as plant walkdowns and leakage monitoring systems for the RCS which provide added defense-in-depth measures to monitor the leak tightness of the subject DM butt welds.
The NRC staff also recognizes several conserva tive assumptions in the flaw analyses. The primary conservatism is that the analyses assume that PWSCC has already initiated in Alloy 52 inlay and continued growing immediately after the last volumetric inspection. To date, there have not been any occurrences of PWSCC initiations in Alloy 52 weld materials.
Finally, the growth of an axial flaw would be limited in length by the width of the weld. Beyond the weld, the base materials of the pipe and the SG nozzle are not significantly susceptible to the PWSCC degradation mechanism and, therefore, the axial flaw cannot grow sufficiently large
in length to cause rupture of the weld within the duration of the proposed alternative. Thus, the likelihood of a LOCA occurring due to axial PWSCC flaws in the subject DM butt welds is extremely low.
Given the NRC staffs review of the licensees fl aw evaluation of a hypothetical circumferential flaw provides reasonable assurance of structural integrity of the welds during the period of extended volumetric examination frequency, and based on the above application of risk insights for axial flaw growth, the NRC staff finds that there is reasonable assurance that the licensees proposed alternative has a minimal, if any, impact on safety and provides an acceptable level of quality and safety.
4.0 CONCLUSION
As set forth above, the NRC staff finds that the proposed alternative described in I6-RR-03 provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(z)(1). Therefore, the NRC staff authorizes use of the licensees proposed alternative I6-RR-03 for up to and including the fall 2030 RFO, but not to exceed 9 calendar years from the prior volumetric and surface examination.
All other ASME 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.
Principle Contributor: O. Khan, NRR
Date: April 22, 2024
ML24071A091 *via memo NRR-028 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DNRL/NPHP/BC*
NAME SWall SRohrer MMitchell DATE 03/08/2024 03/12/2024 03/07/2024 OFFICE NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM*
NAME JWhited SWall DATE 04/22/2024 04/22/2024