ML23241A909

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Enclosure 2 - Non-Proprietary - Review of License Renewal Commitment Number 10 Safety Evaluation
ML23241A909
Person / Time
Site: Byron, Braidwood  Constellation icon.png
Issue date: 09/19/2023
From: Lauren Gibson, Tyree C
NRC/NRR/DNRL/NLRP
To: Rhoades D
Constellation Energy Generation
References
EPID L-2022-RNW-0010, EPID L-2022-RNW-0011
Download: ML23241A909 (13)
v • d • e


Text

BRAIDWOOD STATION, UNIT 1 AND BYRON STATION, UNIT 1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION STEAM GENERATOR LICENSE RENEWAL RESPONSE TO COMMITMENT NO. 10 DOCKET NOS. 50-456 AND 50-454

1.0 INTRODUCTION

By letter dated June 7, 2022 (Reference 1), as supplemented by letter dated April 20, 2023 (Reference 2), Constellation Energy Generation, LLC (CEG or the licensee), submitted to the U.S. Nuclear Regulatory Commission (NRC) a response to Commitment No. 10 of the Braidwood Station, Unit 1 and Byron Station, Unit 1 license renewal (LR). In December 2015, the NRC published the final Safety Evaluation Report Related to the License Renewal of Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, as NUREG-2190, Volume 1 (Reference 3) and NUREG-2190, Volume 2 (Reference 4).

Appendix A, Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, License Renewal Commitments, in NUREG-2190, Volume 2 (Reference 4), Commitment No. 10, Parts 1 and 2, each provide for three options to fulfill this commitment. Commitment No. 10 is related to enhancing the Steam Generators (SGs) Aging Management Program (AMP) by validating that primary water stress corrosion cracking (PWSCC) of divider plate welds to the primary head and tubesheet cladding is not occurring for Braidwood Units 1 and 2 and Byron Units 1 and 2, and by validating that PWSCC of the tube-to-tubesheet welds is not occurring for Braidwood Unit 1 and Byron Unit 1. For Braidwood Unit 1 and Byron Unit 1, the licensee selected Option 2: Analysis, and Option 2: Analysis - Susceptibility, to fulfill Commitment No. 10, Parts 1 and 2, respectively. The licensees response does not include Braidwood Unit 2 and Byron Unit 2.

Although the stated purpose of this commitment is validating that PWSCC is not occurring, updated NRC staff guidance instead addresses aging management for PWSCC of divider plate assemblies and tube-to-tubesheet welds if they are occurring. This is discussed in more detail below in Section 2.1.

2.0 REGULATORY EVALUATION

2.1 PWSCC of SG Divider Plate Assemblies and Tube-to-Tubesheet Welds The divider plate within the channel head (i.e., bottom head) of a recirculating SG separates the primary water flow into a cold leg and hot leg so that the incoming primary water is directed into the hot leg portion of the tubes. Divider plates are approximately 1 - 2 inches thick. The divider plate is welded to the channel head cladding and to the tubesheet, but the design details and materials vary among SG models. The tubesheet is a circular plate, typically low-alloy steel and clad with corrosion-resistant material on the primary side, with drilled holes for the SG tubes to pass through. Tubesheets are approximately 21 - 27 inches thick, including the primary-side cladding. During fabrication, the SG tubes are expanded within the tubesheet and then welded to the tubesheet on the primary side. The tube-to-tubesheet welds are part of the reactor coolant pressure boundary (RCPB) unless an alternate repair criteria has been approved by the NRC at a given unit to redefine the RCPB.

Enclosure 2

PWSCC has occurred in SG divider plate assemblies fabricated with Alloy 600 in foreign operating SGs that are similarly designed to the Westinghouse Model 51. This foreign operating experience occurred even with proper primary water chemistry. However, there has been no PWSCC in Alloy 600 divider plate assemblies in U.S. operating SGs. While PWSCC may occur in Alloy 600 SG tube-to-tubesheet welds, there has been none in U.S. operating SGs, and the NRC staff is unaware of any in foreign operating SGs.

For LR, Revision 2 of NUREG-1801, Generic Aging Lessons Learned (GALL) Report (Reference 5), recommends managing PWSCC of nickel alloy SG divider plate assemblies and SG tube-to-tubesheet welds by the Water Chemistry AMP. However, given that the Water Chemistry AMP may not be sufficient to manage PWSCC in divider plate assemblies based on the foreign operating SG experience and that PWSCC may occur in nickel alloy SG tube-to-tubesheet welds, Reference 5 also recommends evaluating the effectiveness of the Water Chemistry AMP to ensure PWSCC is not occurring. Revision 2 of NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants (Reference 6), includes guidance for the evaluation.

Following development of References 5 and 6, the industry performed analyses to assess the significance of cracks in divider plate assemblies and the potential for the cracks to propagate to RCPB components such as tube-to-tubesheet welds and the channel head. These industry analyses are documented in Electric Power Research Institute (EPRI) Report 3002002850, Steam Generator Management Program: Investigation of Crack Initiation and Propagation in the Steam Generator Channel Head Assembly (Reference 7).

NRC LR Interim Staff Guidance (ISG), LR-ISG-2016-01, Changes to Aging Management Guidance for Various Steam Generator Components (Reference 8), revised the LR guidance for aging management of SG divider plate assemblies and tube-to-tubesheet welds. LR-ISG-2016-01 accepted the analyses in EPRI Report 3002002850 (Reference 7) as a way for LR applicants to determine if aging management activity beyond those under the Water Chemistry and SGs AMPs are warranted for PWSCC of divider plate assemblies and tube-to-tubesheet welds. The aging management guidance is based on the divider plate assembly and tube-to-tubesheet weld materials and whether the analyses in EPRI Report 3002002850 are applicable and bounding for the licensees SGs.

Compared to Revision 2 of NUREG-1801 (Reference 5), a significant change in the NRC guidance for divider plate assemblies and tube-to-tubesheet welds is that there is no longer an expectation that licensees can demonstrate PWSCC is not occurring, or that the materials can be called not susceptible to PWSCC. LR-ISG-2016-01 (Reference 8) considers plant-specific PWSCC susceptibility and whether the visual inspections of SG primary head interior surfaces under the SG program, along with the primary water chemistry program, are adequate for managing cracking of divider plate assemblies and tube-to-tubesheet welds if it is occurring.

The update was based on additional evaluations, tests, and analyses by industry, NRC staff consideration of the industry work, and further NRC staff review of operating experience.

In response to LR-ISG-2016-01 (Reference 8), EPRI provided a checklist to assist licensees with screening their SGs for PWSCC susceptibility of the divider plate assembly and tube-to-tubesheet welds.

2.2 Regulatory Requirements and Guidance Section 54.21(a)(3) of Title 10 of the Code of Federal Regulations (10 CFR) requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. One of the findings that the NRC staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis.

Revision 2 of NUREG-1801 (Reference 5) identifies aging management reviews for systems, structures, and components (SSCs) that may be in scope of LR and identifies AMPs that are acceptable to manage aging effects of the SSCs. Revision 2 of NUREG-1801 recommends managing PWSCC of stainless steel, steel1 (with nickel alloy cladding), and nickel alloy SG divider plates, and nickel alloy SG tube-to-tubesheet welds with the Water Chemistry AMP. In addition, for nickel alloy SG divider plates and SG tube-to-tubesheet welds, it recommends evaluating the need for a plant-specific AMP to verify the effectiveness of the Water Chemistry AMP.

Revision 2 of NUREG-1800 (Reference 6) provides guidance to the NRC staff for the review of LR applications. Section 3.1.2.2.11 provides guidance for evaluating the need for a plant-specific AMP to verify the effectiveness of the Water Chemistry AMP to ensure PWSCC is not occurring in the nickel alloy SG divider plates and SG tube-to-tubesheet welds. The guidance for the nickel alloy SG tube-to-tubesheet welds include acceptance criteria to assist with determining if a plant-specific AMP is needed. The acceptance criteria are based on SG tube material and whether a permanent alternate repair criteria has been approved. This guidance was the basis for using the term not occurring in LR Commitment No. 10.

LR-ISG-2016-01 (Reference 8) describes subsequent changes made to Revision 2 of NUREG-1801 and NUREG-1800 (References 5 and 6) related to PWSCC of SG divider plate assemblies and SG tube-to-tubesheet welds and loss of material due to boric acid corrosion of SG heads and tubesheets. LR-ISG-2016-01 recommends managing PWSCC of steel (with nickel alloy cladding) and nickel alloy SG divider plates and nickel alloy SG tube-to-tubesheet welds with the Water Chemistry and SGs AMPs (it previously specified only the Water Chemistry AMP).

For nickel alloy SG divider plates and SG tube-to-tubesheet welds it recommends evaluating the need for a plant-specific AMP to verify the effectiveness of the Water Chemistry and SGs AMPs.

In addition, LR-ISG-2016-01 added acceptance criteria to assist with determining if a plant-specific AMP is needed for nickel alloy SG divider plates based on fabrication material. The acceptance criteria for both the nickel alloy SG divider plates and SG tube-to-tubesheet welds includes that if the industry analyses (Reference 7) are applicable and bounding, then a plant-specific AMP is not required.

1 In NUREG-1801, Revision 2 (Reference 5), Table IX.C, carbon steel, alloy steel, gray cast iron, ductile iron, malleable iron, and high-strength low-alloy steel are generally grouped under the broad term steel.

3.0 TECHNICAL EVALUATION

3.1 Background

In December 2015, the NRC published the final Safety Evaluation Report Related to the License Renewal of Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, as NUREG-2190, Volume 1 (Reference 3) and NUREG-2190, Volume 2 (Reference 4). In NUREG-2190, Volume 2, Appendix A, Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, License Renewal Commitments, Commitment No. 10, Parts 1 and 2, each provide for three options to fulfill this commitment. Commitment No. 10 is related to enhancing the SG Program by validating that PWSCC of divider plate welds to the primary head and tubesheet cladding is not occurring for Braidwood Units 1 and 2 and Byron Units 1 and 2, and by validating that PWSCC of the tube-to-tubesheet welds is not occurring for Braidwood Unit 1 and Byron Unit 1.

Although the stated purpose of Commitment No. 10 is validating that PWSCC is not occurring, updated NRC guidance instead addresses aging management for PWSCC of divider plate assemblies and tube-to-tubesheet welds if they are occurring, as discussed in Section 2.1 of this safety evaluation.

By letter dated June 7, 2022 (Reference 1), CEG submitted to the NRC a response to Commitment No. 10 of the Braidwood Unit 1 and Byron Unit 1 LR. For Braidwood Unit 1 and Byron Unit 1, the licensee selected Option 2: Analysis, and Option 2: Analysis -

Susceptibility, to fulfill Commitment No. 10, Parts 1 and 2, respectively. The licensees response does not include Braidwood Unit 2 and Byron Unit 2.

CEG performed an analytical evaluation specific to the Braidwood Unit 1 and Byron Unit 1 SGs because it could not be determined if the SGs were bounded by the industry analyses, EPRI Report 3002002850 (Reference 7), that the NRC staff accepted for the divider plate and tube-to-tubesheet PWSCC evaluation in LR-ISG-2016-01 (Reference 8). To better understand the technical basis for concluding that the SG RCPB would be maintained in the presence of divider plate cracking, the NRC staff conducted an audit from October 3 to November 10, 2022. During the audit the NRC staff reviewed relevant technical reports and detailed calculations supporting the licensees Braidwood Unit 1 and Byron Unit 1 diver plate analysis. The audit plan (Reference 9) and audit summary (Reference 10) were issued on September 28, 2022, and February 1, 2023, respectively.

Following review of the information submitted by the licensee and information during the audit, the NRC staff requested additional information and confirmation of information related to the upper shelf fracture toughness (KIC) of the SG primary head base material, the source of the material properties used for the flaw tolerance evaluation, and the minimum channel head thickness. The licensee submitted their response on April 20, 2023 (Reference 2).

3.2 Evaluation of Response to License Renewal Commitment No. 10 Commitment No. 10 provided three options for enhancing the SGs AMP to address PWSCC of the divider plate welds and the tube-to-tubesheet welds. Commitment No. 10, Part 1 was to validate that PWSCC of the divider plate welds to the primary head and tubesheet cladding is not occurring for Braidwood Units 1 and 2 and Byron Units 1 and 2 by performing one of the following:

  • Option 1 (Inspection): A one-time inspection under the SGs AMP.
  • Option 2 (Analysis): An analysis that concludes the SG RCPB is maintained in the presence of divider plate weld cracking.
  • Option 3 (Industry/NRC Studies): Revise the commitment if future industry and NRC studies and operating experience determine PWSCC of the divider plate welds is not a credible concern for RCPB failure.

Commitment No. 10, Part 2 was to validate that PWSCC of the tube-to-tubesheet welds is not occurring for Braidwood Unit 1 and Byron Unit 1 (Commitment No.10, Part 2 does not apply to Braidwood Unit 2 and Byron Unit 2) by performing one of the following:

  • Option 1 (Inspection): A one-time inspection under the SGs program.
  • Option 2 (Analysis - Susceptibility): An analysis that concludes tube-to-tubesheet welds are not susceptible to PWSCC.
  • Option 3 (Analysis - Pressure Boundary): An analysis that determines the tube-to-tubesheet welds are not required to perform an RCPB function.

3.2.1 Commitment No. 10, Part 1 - PWSCC of the Divider Plate Welds to the Primary Head and Tubesheet Cladding To address Part 1 of Commitment No. 10 for Braidwood Unit 1 and Byron Unit 1 CEG selected Option 2 (Analysis). This option requires CEG to do the following for both units:

Perform an analytical evaluation of the steam generator divider plate welds in order to establish a technical basis which concludes that the steam generator reactor coolant pressure boundary is adequately maintained with the presence of steam generator divider plate weld cracking. This analytical evaluation will be submitted to the NRC for review and approval two (2) years prior to entering the associated period of extended operation.

The NRC staff notes that this option does not validate that PWSCC is not occurring, as stated in the commitment. However, it does address the purpose for the commitment of concluding that integrity will be maintained for the part of the RCPB potentially affected by divider plate assembly cracking. Therefore, Option 2 addresses the underlying regulatory requirement in 10 CFR 54.29(a).

The Braidwood Unit 1 and Byron Unit 1 SGs have Alloy 690 divider plates and Alloy 600 type tubesheet cladding material. To meet the guidance in LR-ISG-2016-01 (Reference 8), a plant-specific AMP is necessary unless the applicant confirms the industrys analyses for divider plate cracking are applicable and bounding. The plant-specific AMP is aging management activity beyond control of the reactor water chemistry by the Water Chemistry AMP and visual inspections of the interior surfaces of the channel head performed under the SG AMP.

As part of this evaluation, CEG used a screening checklist (Reference 1, Attachment 3, Appendix A) modeled after the EPRI checklist developed in response to LR-ISG-2016-01 (Reference 8). CEG could not determine if the analyses performed by industry (Reference 7) and accepted in LR-ISG-2016-01 are bounding because the divider plate and channel head designs differ between the Babcock and Wilcox replacement SGs (RSGs) for Braidwood Unit 1 and Byron Unit 1, and the Westinghouse SGs analyzed by industry. In References 1 and 2 CEG described the plant-specific analytical evaluation performed to provide a basis for concluding the SG RCPB is adequately maintained in the presence of divider plate weld cracking. CEG stated that the evaluation approach closely follows that in Reference 7, uses the plant-specific SG geometry and loading conditions, and includes the following key elements:

  • A three-dimensional finite element analysis (FEA) model was applied to a postulated flaw in the divider plate weld buildup. The postulated flaw was considered to extend approximately 40 percent of the SG radius, with one end of the flaw near the primary head inside surface at the triple point (intersection of the divider plate assembly, tubesheet, and primary head). The postulated flaw considered in the Braidwood Unit 1 and Byron Unit 1 analysis is larger than the postulated flaw considered in Reference 7.
  • The FEA model was used for all Service Level A to D transients to calculate the through-wall thickness stress distribution at the triple point due to pressure and thermal loading.
  • The stress distribution was used in fatigue crack growth analyses for both axial and circumferential cracking into the primary head cladding and base material.
  • The crack growth analysis provided the final crack depth and length values at the end of the SG service life. The corresponding maximum stress intensity factor for each transient was compared to the acceptance criteria in the 2017 Edition of American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)

Section XI, IWB-3600 (Reference 11).

  • The primary stress in the primary head for the final crack geometry was compared to the criteria in the 1986 Edition of ASME Code Section III, NB-3000 (Reference 12).

To evaluate CEGs analysis, the NRC staff reviewed CEGs methodology and acceptance criteria, focusing on consistency with the industry analyses in Reference 7 and the basis for differences between the analyses. During the audit, CEG provided the NRC staff with access to detailed information such as the RSG design specification, design transient analysis report, FEA model development, and flaw analysis (Reference 9).

The NRC staff reviewed the materials, identified in Table 1 below, and the material properties used in CEGs analysis.

Table 1, Materials for Components Used in CEGs FEA Component Material Tubes SB-163 UNS N06690 Alloy 690 Divider plate (( ))

Divider plate seat bar/weld build-up (( ))

Cladding - tubesheet (( ))

Cladding - primary head (( ))

Primary head (( ))

Tubesheet (( ))

Secondary shell (( ))

CEG clarified in Reference 2 that the yield and tensile strength values were taken from the 1986 Edition of ASME Code Section III, Division 1 - Appendices (Reference 12). The NRC staff finds this acceptable because the materials used for the relevant components are the same as or representative of the materials used in the industry analyses and the 1986 Edition is the Code of Record for the design of the Braidwood Unit 1 and Byron Unit 1 RSGs.

The primary head of a SG is subjected to design-basis internal pressure and thermal transients during the design life. For Braidwood Unit 1 and Byron Unit 1, CEG calculated the stresses from pressure and thermal loading for Level A, B, C, and D transient conditions (A - normal conditions, B - upset and test conditions, C - emergency conditions, and D - faulted conditions) according to the Certified Design Specification for the RSGs and power uprate structural analysis report. The NRC staff finds these transients acceptable for describing the loading conditions because they represent the design basis conditions for the SGs at the currently licensed power level.

CEG used the three-dimensional finite element model with the transient loads to calculate the through-wall stress distribution at the triple point due to pressure and thermal loading. The analysis assumed a flaw in the divider plate weld buildup, starting at the triple point and extending toward the center of the tubesheet to a length of approximately 40 percent of the SG radius. The analysis assumed that, at the triple point, the initial crack in the inside surface of the primary head was equal to the cladding thickness in depth and equal to the width of the divider plate weld buildup in length. Stress distributions through the primary head wall thickness were calculated from pressure and thermal loading due to the Service Level A and B transients combined with residual stress from the primary head seam weld. The stresses were used to calculate the Stress Intensity Factor (KI) used in the fatigue crack growth analysis for eight different postulated crack paths that considered both circumferential and axial elliptical planar flaws independently. ((

)). The crack growth analysis was performed according to the 2017 Edition of ASME Code Section XI (Reference 11), which was the Code edition incorporated by reference in 10 CFR 50.55(a)(1)(ii)(c)(55) at the time of the analysis.

The fatigue crack growth analysis produced final flaw depth and length values. Then, for each transient, including normal, emergency, and faulted conditions (Service Levels A to D),

maximum KI values for these flaws for each path were calculated and compared to the acceptance criteria in the 2017 Edition of ASME Code Section XI, IWB-3600 (Reference 11).

CEG compared the requirements in IWB-3610 in the 2017 Edition with those in the editions and addenda approved for Braidwood Unit 1 (Reference 13) and Byron Unit 1 (Reference 14) and determined that there were no changes to the technical requirements. The NRC staff also compared the editions and addenda and determined there were no changes to the technical requirements in IWB-3610. Finally, the primary stress in the primary head, considering local area reduction from the final crack geometry, was compared to the requirements in the 1986 Edition of ASME Code Section III, NB-3000 (Reference 12).

The NRC staff evaluated the licensees methodology for evaluating PWSCC in the Braidwood Unit 1 and Byron Unit 1 divider plate assemblies. The staff finds the methodology acceptable based on similarity to the industry methodology referenced in LR-ISG-2016-01 (Reference 8).

Specifically, the licensees methodology, (a) uses FEA modeling with loads from plant transients to determine stress distributions and stress intensity factors from a stress corrosion crack that extends from the triple point (tubesheet/divider plate/primary head), (b) uses the calculated stress intensity to propagate the crack by fatigue into the primary head base material according to Appendix A in the 2017 Edition of ASME Code Section XI (Reference 11), (c) postulates multiple crack paths for axial and circumferential flaws, (d) compares the stress intensity values for the final crack geometries to acceptance criteria approved by the staff in 10 CFR 50.55(a)(1)(ii)(c)(55), and (d) evaluates the primary stress in the primary head including the local area reduction from the final crack geometries according to the criteria in the 1986 Edition of ASME Code Section III, NB-3000 (Reference 12).

CEG stated that their evaluation methodology included some key assumptions that were conservative. For example, the length of the stress corrosion crack assumed to exist in the divider plate weld material was approximately 40 percent of the SG radius. The industry analyses assumed a length of about 25 percent from the triple point, while the cracks reported from international operating experience were short, shallow, and did not start at the triple point nor grow to the triple point in over 20 years of operation (Reference 7). No cracking in this location has been observed in U.S. operating SGs, which further indicates the conservatism of this approach. The assumption of a long crack increases conservatism because it increases the calculated stress intensity that drives the fatigue crack growth. Another example is excluding the

(( )) from the fatigue crack growth analysis. Because the initial depth of the fatigue crack was considered as equal to the cladding thickness ((

)). Based on consideration of the licensees SG design and methodology, and the industry methodology (Reference 7), the NRC staff concluded these assumptions are conservative.

The NRC staff asked the licensee to discuss the basis for using an upper-shelf KIC of ((

)), rather than 200 ksi-in as used in the industry analyses (Reference 7), and how it affects the conservatism in the analysis. The licensees response (Reference 2) included a revision of the technical evaluation of divider plate cracking and primary head flaw tolerance originally provided as Attachment 4 of Reference 1. The revised analyses used an upper-shelf KIC value of 200 ksi-in. In the revised analysis the acceptance criteria were met for both axial and circumferential flaws by partially removing one conservatism for (( )) flaws. The conservatism was removed by reducing the ((

)). The licensee stated that this reduced the final stress intensity values, but it retained the conservative effect of the (( )).

The NRC staff also asked the licensee to discuss the effect of using the lower fracture toughness value of 200 ksi-in on the licensees critical crack size adjustment using ((

)). The licensee explained in Reference 2 that the adjustment was used only for Level C and D transients for circumferential flaws, and it provided the revised results showing the final stress intensity values for circumferential flaws for the Level C and D transients met the acceptance criteria that were revised for the lower KIC value.

Axial flaws were unaffected by the change in KIC value.

As noted in the NRC staffs Request for Additional Information C10-1 in Reference 2, the use of 200 ksi-in is not always conservative. However, the staff finds the licensees response acceptable because using the lower fracture toughness of 200 ksi-in is consistent with the industry analyses, the evaluation includes conservative assumptions as previously discussed, and the flaw tolerance acceptance criteria were met for all cases with only one of the conservatisms in the revised analysis being reduced, and only for (( )) flaws.

As discussed in References 7 and 8, if PWSCC was to occur it is not expected to cause a safety issue. The industry analyses (Reference 7) concluded a full-length, full-depth weld crack does not challenge structural integrity. The cracking observed internationally was shallow and did not challenge structural integrity. Based on the time between inspections, it is likely that some of those cracks had been in service for a long time before detection. Reference 8 recommends managing PWSCC of nickel alloy SG divider plates and nickel alloy SG tube-to-tubesheet welds with the Water Chemistry and SGs AMPs. As discussed in Reference 8, the visual inspections performed under the SG AMP provide opportunities to identify cracking if it grows into pressure boundary components through identification of rust stains, gross cracking, or divider plate assembly distortion.

The NRC staff finds the licensees response, as revised by Reference 2, to Commitment No. 10, Part 1 acceptable for Braidwood Unit 1 and Byron Unit 1 based on the following:

  • The licensees methodology for growth and evaluation of flaws in divider plate assemblies closely follows the methodology referenced by the NRC in Reference 8.
  • Using conservative assumptions, the maximum value of stress intensity is less than the allowable value for all analyzed transients for both axial and circumferential flaws using material properties and acceptance criteria consistent with the design basis.
  • The analyzed transients represent design basis conditions at the currently licensed power level.
  • Consistent with Reference 8, aging of the Braidwood Unit 1 and Byron Unit 1 nickel alloy SG divider plates (PWSCC and loss of material) will be managed through control of the reactor water chemistry by the Water Chemistry AMP and visual inspections of the interior surfaces of the channel head performed under the SG AMP as indicated in Table 3.1.2-4 of Reference 15.

In meeting Commitment No. 10, Part 1 to demonstrate that the SG RCPB is adequately maintained with the presence of SG divider plate weld cracking, the licensee is not required to have a plant-specific AMP. Therefore, the NRC staff notes that the licensees SG AMP, as described in Section B.2.1.10 of Reference 15, conforms to the guidance in LR-ISG-2016-01 (Reference 8) as it relates to divider plate assemblies.

3.2.2 Commitment No. 10, Part 2 - PWSCC of the Tube-to-Tubesheet Welds To address Part 2 of Commitment No. 10 for Braidwood Unit 1 and Byron Unit 1, CEG selected Option 2 (Analysis - Susceptibility). This option requires CEG to do the following for both units:

Perform an analytical evaluation of the steam generator tube-to-tubesheet welds to determine that the welds are not susceptible to primary water stress corrosion cracking. The evaluation for determining that the tube-to-tubesheet welds are not susceptible to primary water stress corrosion cracking will be submitted to the NRC for review and approval two (2) years prior to entering the associated period of extended operation.

NRC guidance issued after the Braidwood and Byron renewed licenses reflects an understanding that these components have a very low susceptibility to PWSCC, but it would not be accurate to call them not susceptible under all potential conditions. For example, LR-ISG-2016-01 (Reference 8) states that tube-to-tubesheet welds made between Alloy 690 tubes and Alloy 690 type tubesheet cladding material, which produces the highest chromium range among the materials used, would be highly resistant to PWSCC. The SGs AMP, as revised by LR-ISG-2016-01, states that visual inspections of the tube-to-tubesheet welds are intended to identify signs that cracking or loss of material may be occurring.

The Braidwood Unit 1 and Byron Unit 1 SGs have thermally treated Alloy 690 SG tubes and Alloy 600 type tubesheet cladding material. According to LR-ISG-2016-01 (Reference 8), a plant-specific AMP is necessary unless the applicant confirms that the industrys analyses (Reference 7) for tube-to-tubesheet weld cracking are applicable and bounding, and the applicant will perform general visual inspections of the tubesheet cladding region to look for evidence of cracking on the cladding as part of the SG AMP. The plant-specific AMP is aging management activity beyond control of the reactor water chemistry by the Water Chemistry AMP and visual inspections of the interior surfaces of the channel head, including tubesheet region, performed under the SG AMP.

In Section 7.1.1 of Reference 7, the industry analyses concluded the following about PWSCC susceptibility based on the chromium content of the weld deposit and the stress on the tubesheet:

  • For Alloy 690 tubes and the Alloy 600 type tubesheet cladding material with the highest chromium content (Alloy 82), the resulting weld deposits are expected to contain more than 24 weight percent chromium and therefore be resistant to PWSCC.
  • For Alloy 690 tubes and the lower-chromium Alloy 600 type tubesheet cladding material (Alloy 182), which is used only near the center of the tubesheet in some designs, the resulting weld deposits are expected to contain about 22 percent chromium. PWSCC initiation cannot be ruled out based on the chromium content alone; however, compressive operating stresses near the center of the tubesheet primary side should prevent PWSCC initiation.
  • These expectations are consistent with operating experience.

CEG evaluated the PWSCC susceptibility of the Braidwood Unit 1 and Byron Unit 1 tube-to-tubesheet welds by determining the minimum chromium content of the weld cladding materials, the minimum chromium content of the tubing material, and the tube-to-tubesheet weld design.

As part of this evaluation, CEG used a screening checklist (Reference 1, Attachment 3, Appendix A) modeled after an EPRI checklist developed in response to LR-ISG-2016-01 (Reference 8). Based on a minimum specified chromium content of (( )) weight percent chromium in the SG tubes, a minimum chromium content of (( )) weight percent chromium in the (( )) tubesheet cladding, and (( )) welding with a dilution factor of (( ))

tubes/(( )) cladding to join the tube ends to the tubesheet cladding, CEG concluded the minimum chromium content of the tube-to-tubesheet welds is (( )) weight percent. Based on the minimum chromium content exceeding the value of 24 weight percent in the guidance, CEG concluded that the Braidwood Unit 1 and Byron Unit 1 SGs are not susceptible to PWSCC because the industry analyses are both applicable and bounding.

The NRC staff reviewed the applicants evaluation for Braidwood Unit 1 and Byron Unit 1 and finds that it demonstrates that the tube-to-tubesheet welds are highly resistant to PWSCC and therefore meets the intent of Commitment No. 10, Part 2. As noted above in this section, the current NRC staff guidance is not based on showing the welds are not susceptible to PWSCC.

For the materials used for the Braidwood Unit 1 and Byron Unit 1 SGs, the guidance considers a plant-specific AMP unnecessary if an applicant confirms that the industry analyses (Reference

7) are applicable and bounding and the applicant will perform general visual inspections for evidence of cracking. The staffs finding is based on the following:
  • Review of the licensees methodology for determining the minimum chromium content of the welds. The staff concluded the methodology closely follows that in Reference 7.
  • The calculated minimum chromium content of the welds exceeds the 22 percent criterion in Reference 8 and the 24 percent criterion in Reference 7.
  • For the tube-to-tubesheet welds, Reference 7 is applicable and bounding for the Braidwood Unit 1 and Byron Unit 1 SGs.
  • The criterion in Reference 8 for compressive stress in the cladding is not applicable because the tubesheet cladding material is resistant to PWSCC.
  • Consistent with Reference 8, aging of the Braidwood Unit 1 and Byron Unit 1 nickel alloy SG tube-to-tubesheet welds (PWSCC and loss of material) will be managed through control of the reactor water chemistry by the Water Chemistry AMP and visual inspections of the interior surfaces of the channel head, including the tubesheet region, performed under the SG AMP as indicated in Table 3.1.2-4 of Reference 15. As discussed in Reference 8, the visual inspections performed under the SG AMP provide opportunity to identify cracking if it grew into pressure boundary components through identification of rust stains or other abnormal observations such as boric acid deposits.

In meeting Commitment No. 10, Part 2 by demonstrating the tube-to-tubesheet welds are highly resistant to PWSCC and that industry analyses (Reference 7) are applicable and bounding, the licensee is not required to have a plant-specific AMP. Therefore, the NRC staff notes that the licensees SGs AMP, as described in Section B.2.1.10 of Reference 15, conforms to the guidance in LR-ISG-2016-01 (Reference 8) as it relates to tube-to-tubesheet welds.

3.3 Technical Evaluation Conclusion

Based on the review described above, the NRC staff finds that the licensees response to Commitment No. 10 for Braidwood Unit 1 and Byron Unit 1 is consistent with the SG AMP in Reference 5, as modified by LR-ISG-2016-01 (Reference 8), for PWSCC in nickel alloy SG divider plate assemblies and tube-to-tubesheet welds. Therefore, the staff concludes the licensee has demonstrated these aging effects will be adequately managed so that the intended functions will be maintained consistent with the current licensing basis during the period of extended operation, as required by 10 CFR 54.21(a)(3).

4.0 REFERENCES

1. Braidwood Station, Unit 1, and Byron Station, Unit 1, Submittal of License Renewal Response to Commitment No. 10 - Evaluation of Possible PWSCC Crack Initiation and Propagation in the Steam Generator Channel Head Assembly and Tube-to- Tubesheet Welds, dated June 7, 2022 (ML22158A294 (Package)).
2. Response to Request for Additional Information to Braidwood Station, Unit 1, and Byron Station, Unit 1, for Steam Generator License Renewal Response to Commitment No. 10, dated April 20, 2023 (ML23110A123 (Package)).
3. NUREG-2190, Volume 1, Safety Evaluation Report Related to the License Renewal of Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, dated December 2015 (ML15350A038).
4. NUREG-2190, Volume 2, Safety Evaluation Report Related to the License Renewal of Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, dated December 2015 (ML15350A041).
5. NUREG-1801, Generic Aging Lessons Learned (GALL) Report, Revision 2, dated December 2010 (ML103490041).
6. NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, Revision 2, dated December 2010 (ML103490036).
7. EPRI Report 3002002850, Steam Generator Management Program: Investigation of Crack Initiation and Propagation in the Steam Generator Channel Head Assembly.
8. Final LR-ISG-2016 Changes to Aging Management Guidance for Various Steam Generator Components, Final, dated November 2016 (ML16237A383).
9. Braidwood Station, Unit 1 and Byron Station, Unit 1 - Steam Generator License Renewal Response to Commitment No. 10 - Audit Plan, dated September 28, 2022 (ML22265A013).
10. Braidwood Station, Unit 1, and Byron Station, Unit 1, Audit Report Regarding the Response to License Renewal Commitment No. 10 (EPIDs: L-2022-RNW-0010 and L-2022-RNW-0011), dated February 1, 2023 (ML23137A068 (Package)).
11. 2017 Edition of American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, IWB-3600.
12. 1986 Edition of American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section III, NB-3000.
13. 2013 Edition American Society of Mechanical Engineers Boiler and Pressure Vessel Code.
14. 2007 Edition American Society of Mechanical Engineers Boiler and Pressure Vessel Code, with 2008 Addenda.
15. Braidwood and Byron, Units 1 and 2 - Application for Renewed Operating Licenses, dated May 29, 2013 (ML131550528 (Package)).

Principal Contributors: Greg Makar, NRR Leslie Terry, NRR David Rudland, NRR

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