ML19339H747

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Approval of Relief Request 2-RR-17 Regarding Steam Generator Primary Nozzle Dissimilar Metal Welds Inspection Interval
ML19339H747
Person / Time
Site: Point Beach NextEra Energy icon.png
Issue date: 12/13/2019
From: Nancy Salgado
Plant Licensing Branch III
To: Moul D
Point Beach
Chawla M
References
EPID L-2019-LLR-0084
Download: ML19339H747 (11)
v • d • e


Text

UNITED STATES WASHINGTON, D.C. 20555-0001 December 13, 2019 Mr. Don Moul Vice President, Nuclear Division and Chief Nuclear Officer NextEra Energy Point Beach, LLC Mail Stop: NT3/JW 15430 Endeavor Drive Jupiter, FL 33478

SUBJECT:

POINT BEACH NUCLEAR PLANT, UNIT 2-APPROVAL OF RELIEF REQUEST 2-RR-17 REGARDING STEAM GENERATOR PRIMARY NOZZLE DISSIMILAR METAL WELDS INSPECTION INTERVAL (EPID L-2019-LLR-0084)

Dear Mr. Moul:

By letter dated August 29, 2019 (Agencywide Documents and Access Management System (ADAMS) Accession No. ML19241A492), NextEra Energy Point Beach, LLC, (the licensee) requested relief from certain requirement of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI. In relief request (RR) 2-RR-17, the licensee proposed to delay the volumetric inservice inspection (ISi) of the steam generator (SG) inlet and outlet nozzle-to-safe end dissimilar metal (OM) butt welds at the Point Beach Nuclear Plant (Point Beach), Unit 2.

Specifically, pursuant to Title 1O of the Code of Federal Regulations ( 10 CFR), Section 50.55a(z)(1 ), the licensee proposed an alternative volumetric examination frequency for the subject SG nozzle-to-safe end OM butt welds on the basis that the alternative provides an acceptable level of quality and safety.

The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the subject request and determined that the proposed alternative provides an acceptable level of quality and safety for the SG inlet and outlet nozzle-to-safe end OM butt welds by providing reasonable assurance that the structural integrity of the subject OM butt welds will be maintained. 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 RR 2-RR-17 for up to and including the fall 2021 refueling outage (U2R38), but not to exceed 9 calendar years from the prior examination. The fifth 10-year ISi interval of Point Beach, Unit 2, which commenced on August 1, 2012, and is scheduled to end on July 31, 2022.

D. Moul All other ASME Code,Section XI, requirements for which relief was not specifically requested and authorized herein by the NRC staff remain applicable, including the third-party review by the Authorized Nuclear In-service Inspector.

If you have any questions, please contact Mahesh Chawla of my staff at (301) 415-8371.

Sincerely, Nancy L. Salgado, Chief Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-301

Enclosure:

Safety Evaluation cc: ListServ

UNITED STATES WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST 2-RR-17 REGARDING STEAM GENERATOR PRIMARY NOZZLE DISSIMILAR METAL WELDS INSPECTION INTERVAL NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNIT 2 DOCKET NO. 50-301

1.0 INTRODUCTION

By letter dated August 29, 2019 (Agencywide Documents and Access Management System (ADAMS) Accession ML19241A492}, NextEra Energy Point Beach, LLC, (the licensee) requested relief from certain requirement of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI. In relief request (RR) 2-RR-17, the licensee proposed to delay the volumetric inservice inspection (ISi) of the steam generator (SG) inlet and outlet nozzle-to-safe end dissimilar metal (DM) butt welds at the Point Beach Nuclear Plant (Point Beach), Unit 2.

Specifically, pursuant to Title 10 of the Code of Federal Regulations ( 10 CFR), Section 50.55a(z)(1 ), the licensee proposed an alternative volumetric examination frequency for the subject SG nozzle-to-safe end DM butt welds on the basis that the alternative provides an acceptable level of quality and safety.

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 August 17, 2017, shall implement the requirements of ASME Code Case N-770-2, "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," instead of ASME Code Case N-770-1, subject to the conditions specified in paragraphs (g)(6)(ii)(F)(2) through (13) of Section 50.55a, by the first refueling outage starting after August 17, 2017.

Pursuant to 10 CFR 50.55a(z), alternatives to the requirements of paragraph (b) through (h) of Section 50.55a, or portions thereof, may be used when authorized by the Director, Office of Enclosure

Nuclear Reactor Regulation. A proposed alternative must be submitted and authorized prior to implementation. The licensee must demonstrate that: (1) the proposed alternative would provide an acceptable level of quality and safety; or (2) compliance with the specified requirements of Section 50.55a 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 U.S. Nuclear Regulatory Commission (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 Background By letter dated March 22, 2016 (ADAMS Accession No. ML16063A058), the NRC approved RR 2-RR-11 (ADAMS Accession No. ML15225A104) 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 spring 2020 refueling outage (U2R37) instead of the spring 2017 refueling outage (U2R35). It is noted that the licensee completed the baseline volumetric examination of the subject SG nozzle-to-safe end DM butt welds in November 2012 refueling outage (U2R32) as required by 10 CFR 50.55a(g)(6)(ii)(F)(3).

3.2 Components for Which Relief is Requested ASME Code Class 1 SG inlet and outlet nozzle-to-safe end DM butt welds are the components for which relief is sought. In accordance with Table 1 of ASME Code Case N-770-2, the licensee classified the SG inlet nozzle-to-safe end DM butt welds identified with identification nos. RC-34-MRCL-Al-05 and RC-34-MRCL-Bl-05 SG as Inspection Item A-2, and the SG outlet nozzle-to-safe end DM butt welds identified with identification Nos. RC-36-MRCL-All-01A and RC-36-MRCL-Bll-01A as Inspection Item B.

The licensee stated that the two SGs of Point Beach, Unit 2, were replaced in the fall 1996 refueling outage (U2R22). 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, Alloy 52 inlay was installed on the inside diameter (ID) surface of 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 degrees Fahrenheit (°F) and the outlet nozzles is 543 °F.

3.3 Applicable Code Edition and Addenda The code of record for the fifth 10-year ISi interval is the 2007 Edition and 2008 Addenda of the ASME Code.

3.4 Duration of Relief Request The duration for which the relief is requested is not to exceed 9 calendar years from the prior examination which was completed in November 2012, refueling outage (U2R32).

3.5 ASME Code Requirement The ASME Code requirements applicable to this request are in Section XI, Table IWB-2500-1.

In accordance with 10 CFR 50.55a(g)(6)(ii)(F), the NRC has mandated an augmented inspection for the DM butt welds which are the subject of this request, and that is to implement the requirements of ASME Code Case N-770-2 with conditions specified in paragraphs (g)(6)(ii)(F)(2) through (13) of Section 50.55a.

In accordance with Table 1 in Code Case N-770-2, the SG inlet nozzle-to-safe end DM butt welds classified as Inspection Item A-2 are required to be volumetrically examined every 5 calendar years, and the SG outlet nozzle-to-safe end DM butt welds classified as Inspection Item Bare required to be volumetrically examined every second inspection period not to exceed 7 calendar years.

3.6 Proposed Alternative The licensee proposed to delay the volumetric examination for the SG inlet and outlet nozzle-to-safe end DM butt welds. The proposed alternative is to perform the volumetric examination of the SG inlet and outlet nozzle-to-safe end DM butt welds in the fall 2021 refueling outage (U2R38) and not to exceed 9 calendar years from the prior examination which was completed in November 2012, refueling outage (U2R32).

3. 7 Basis for Use of 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 reactor coolant system (RCS) to low levels (i.e., mid-loop) and opening of the SGs manways would occur once instead of twice in the fifth 10-year ISi interval, thus, the impact of these activities to nuclear, radiological, and industrial safety, would be minimized.

As discussed below, the licensee's 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 subject DM butt welds to demonstrate reasonable assurance of the integrity of the welds until the next proposed inspection.

3.7.1 Prior Inspections of the Subject Welds The licensee stated that the SG inlet and outlet nozzle-to-safe end DM butt welds have been in operation for more than 19 effective full power years (EFPY) with no PWSCC at hot-leg temperatures since their replacement in 1997. History of the licensee's inspection activities on the SG inlet and outlet nozzle-to-safe end DM butt welds are 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;
  • 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; and
  • The subject OM butt welds will continue to receive the required VE and the system leakage test accompanied by the VT-2 in 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 and outlet nozzle-to-safe end OM 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 could 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 OM butt welds and the installation of Alloy 52 inlays were done at the Westinghouse fabrication facility and 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 OM butt welds (as discussed in previous RR 2-RR-11 dated August 13, 2015).

3. 7.2 Flaw Tolerance Analyses Additional support for the acceptability of extending the examination interval for the SG inlet and outlet nozzle-to-safe end OM butt welds is contained in the plant-specific flaw tolerance analyses documented in Attachment 2 of 2-RR-17, 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)-287, "Primary Water Stress Corrosion Cracking (PWSCC) Flaw Evaluation Guidance," for its flaw tolerance analyses.

As documented in the licensee's submittal, the licensee performed plant-specific flaw tolerance analyses to demonstrate that postulated ID axial and circumferential flaws in the OM butt welds would not grow to the ASME Code allowable flaw size between the planned examinations (i.e.,

between the November 2012 refueling outage (U2R32) and the fall 2021 refueling outage (U2R38)). Based on Point Beach, Unit 2, operational data and anticipated refueling outages scheduled between November 2012, and fall 2021, 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 OM weld locations. Due to their operating at a higher temperature, the PWSCC growth rates for the SG inlet nozzle-to-safe end OM weld locations will be higher than for the SG outlet nozzle-to-safe end locations. All other inputs, which include the weld geometries, weld residual stresses (WRSs), and loads, are consistent between the SG inlet and outlet nozzle-to-safe end OM butt welds, and, thus, the results for the SG inlet nozzle-to-safe end OM butt welds bound the SG outlet nozzle-to-safe end OM butt welds.

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., flaw's 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.

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 WRSs. The licensee stated that the WRSs in the SG inlet nozzle-to-safe end OM 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 OM 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.

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."

Figures 7-1 and 7-2 of Attachment 2 to 2-RR-17 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 circumferential 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 2-RR-17 for the subject SG inlet and outlet nozzle-to-safe end OM butt welds.

3.8 NRC Staff Evaluation The NRC staff has evaluated RR 2-RR-17 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 and outlet nozzle-to-safe end OM butt welds from spring 2020 until fall 2021) provides an acceptable level of quality and safety. To reach a conclusion, the staff performed independent flaw tolerance analyses for the bounding case of the Point Beach, Unit 2, SG inlet nozzle-to-safe end OM butt welds.

3.8.1 Staff's 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 November 2012, refueling outage (U2R32) to fall 2021, refueling outage (U2R38)) would exceed the ASME Code allowable flaw size limit. The staff began by evaluating aspects of licensee's flaw tolerance analyses -

specifically the assumed initial defects, the characterization of WRSs, and the methodology for calculating PWSCC growth - for inclusion in the staff's 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, and consistent with the recommendations in EPRI MRP-115. The staff, therefore, used the initial flaws proposed by the licensee in its independent analyses.
  • The axial and hoop WRS distributions provided in 2-RR-17 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 licensee's calculated WRS distributions to be acceptable and used the licensee's calculated WRSs in its independent analyses.
  • For Alloy 82/182, the licensee used the 75th 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 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, that is documented in a data summary report (ADAMS Accession No. ML14322A587), generally supports the contention that the crack growth rate of Alloy 690/52/152 is more crack resistant but differs from the MRP-375 data in some respects. The staff characterized the Alloy 52 crack growth rate based on an acceptable FOi of 10 as documented in the proposed rule published in the Federal Register, 83 FR 56156, dated November 9, 2018. The NRC staff also 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 (ADAMS Accession No. ML112240818), regarding inlayed weld categorization.

For circumferential flaws, the NRC staff's analyses confirmed the licensee's conclusion thatthe structural integrity of the SG inlet nozzle-to-safe end DM butt welds would be maintained

through the period of the proposed volumetric inspection extension. In addition, the NRC staff's 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 staff's analyses found the 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 2012 to fall 2021 ). 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 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. Thus, the NRC staff applied risk insights to assess safety implications of piping with axial flaws that exceed the allowable as discussed below.

3.8.3 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 staff's 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 staff's 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 conservative 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 susceptible to the PWSCC degradation mechanism, and, therefore, the axial flaw cannot grow sufficiently large in length to cause rupture of the weld. Thus, the likelihood of a LOCA occurring due to axial PWSCC flaws in the subject DM butt welds is low.

Based on the above application of risk insights, the NRC staff finds that there is reasonable assurance that the licensee's proposed alternative has a minimal, if any, impact on safety.

4.0 CONCLUSION

As set forth above, the NRC staff determines that the proposed alternative provides an acceptable level of quality and safety for the SG inlet and outlet nozzle-to-safe end DM butt welds by providing reasonable assurance that the structural integrity of the subject DM butt welds will be maintained. 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 RR 2-RR-17 for up to and including the fall 2021 refueling outage (U2R38), but not to exceed 9 calendar years from the prior examination,.

All other ASME Code,Section XI, requirements for which relief was not specifically requested and authorized herein by the staff remain applicable, including the third-party review by the Authorized Nuclear In-service Inspector.

Principal Contributors: A. Rezai, NRR J. Collins, NRR Dateof issuance: December 13, 2019

ML19339H747 *via email OFFICE D0RL/LPL3/PM D0RL/LPL3/LA DNRL/NPHP/BC* D0RL/LPL3/BC NAME MChawla SRohrer MMitchell NSalgado (SWall for)

DATE 12/09/19 12/09/19 11/22/19 12/13/19

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