ML25079A202

Steam Generators License Renewal Responses to Commitment Nos. 34 and 35

ML25079A202
Person / Time
Site:	Callaway
Issue date:	04/01/2025
From:	Mahesh Chawla Plant Licensing Branch IV
To:	Diya F Union Electric Co
Chawla M
References
EPID-L-2024-LRO-0009
Download: ML25079A202 (16)

Text

April 1, 2025 Fadi Diya Senior Vice President and Chief Nuclear Officer Ameren Missouri Callaway Energy Center 8315 County Road 459 Steedman, MO 65077

SUBJECT:

CALLAWAY PLANT, UNIT NO. 1, - STEAM GENERATOR LICENSE RENEWAL RESPONSES TO COMMITMENT NOS. 34 AND 35 (EPID: L-2024-LRO-0009)

Dear Fadi Diya:

By letter dated December 20, 2023 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML23354A244), as supplemented by letters dated December 5, 2024, and December 12, 2024 (Packages ML24340A120 and ML24347A225, respectively), Union Electric Company, doing business as Ameren Missouri (the licensee),

submitted to the U.S. Nuclear Regulatory Commission (NRC) a response to Commitment Nos. 34 and 35 of the Callaway Plant, Unit No.1, license renewal. In March 2015, the NRC published the final Safety Evaluation Report Related to the License Renewal of Callaway Plant, Unit 1, as NUREG-2172 (ML15068A342). In NUREG-2172, appendix A, Callaway Plant Unit 1 License Renewal Commitments, Commitment No. 34 provides for three options to fulfill the commitment, and Commitment No. 35 provides for two options to fulfill the commitment. For both commitments, the licensee selected Option 2: Analysis.

The NRC staff has completed its review of the licensees commitment response. As documented in the enclosed safety evaluation, the staff concludes that the response is acceptable.

If you have any questions, please contact William Orders at 301-415-3329 or by email at William.Orders@nrc.gov.

Sincerely,

/RA William Orders for/

Mahesh Chawla, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-483

Enclosure:

Safety Evaluation cc: Listserv

Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION STEAM GENERATOR LICENSE RENEWAL RESPONSE TO COMMITMENTS 34 AND 35 UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT NO. 1 DOCKET NO. 50-483

1.0 INTRODUCTION

By letter dated December 20, 2023 (Reference 1), as supplemented by letters dated December 5, 2024, and December 12, 2024 (References 2 and 3, respectively), Union Electric Company, doing business as, Ameren Missouri (the licensee), submitted to the U.S. Nuclear Regulatory Commission (NRC) a response to Commitment Nos. 34 and 35 of the Callaway Plant, Unit No.1 (Callaway) license renewal (LR). In March 2015, the NRC published the final Safety Evaluation Report Related to the License Renewal of Callaway Plant, Unit 1, as NUREG-2172 (Reference 4).

Appendix A, Callaway Plant Unit 1 License Renewal Commitments, of NUREG-2172, includes Commitment Nos. 34 and 35, each with three options to fulfill the commitment. Commitment No. 34 is related to potential cracking at the divider plate welds to primary head and tubesheet cladding. To fulfill Commitment No. 34, the licensee selected Option 2: Analysis, to perform an analytical evaluation to establish a technical basis for concluding that the steam generator (SG) reactor coolant system (RCS) pressure boundary is adequately maintained with the presence of SG divider plate cracking. Commitment No. 35 is related to potential failure of the primary-to-secondary pressure boundary due to primary water stress corrosion cracking (PWSCC) of tube-to-tubesheet welds. To fulfill Commitment No. 35, the licensee also selected Option 2:

Analysis, to perform an analytical evaluation of the SG tube-to-tubesheet welds to demonstrate that the welds are not susceptible to PWSCC.

of the licensees submittal dated December 20, 2023, states that the divider plate cracking analyses are applicable to the Framatome replacement SGs (RSGs) at Callaway; Prairie Island Nuclear Generating Plant, Units 1 and 2 (Prairie Island, Units 1 and 2); and Salem Nuclear Generating Station, Unit No. 2 (Salem, Unit 2), because the transients, design cycles, and support loads in the analysis are the bounding values among the four units. The calculated chromium content in the tube-to-tubesheet welds is also provided in enclosure 1 of the December 20, 2023, submittal, for all four units; however, a supplemental report on the tube-to-tubesheet weld chromium content for Callaway is provided in enclosure 2 of the licensees submittal dated December 20, 2023. The NRC staff reviewed the submittal only with respect to whether it satisfies Callaway LR Commitment Nos. 34 and 35; therefore, no findings were made regarding other plants.

2.0 REGULATORY EVALUATION

2.1 Primary Water Stress Corrosion Cracking of Steam Generator Divider Plate Assemblies and Tube-to-Tubesheet Welds The divider plate within the channel head (i.e., bottom head) of a recirculating pressurized water reactor (PWR) SG separates the primary water flow into a hot leg and cold leg so that the heated primary water from the reactor vessel is directed into the hot leg portion of the U-tubes.

The channel head assembly schematic below (Reference 5), shows the divider plate and the location of cracking observed in some international operating SGs. 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. For example, the Callaway SGs do not have a stub runner, which is a plate connecting the divider plate to the tubesheet. 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 criterion has been approved by the NRC at a given unit to redefine the RCPB.

Schematic of Channel Head Assembly (Reference 5)

PWSCC has occurred in SG divider plate assemblies fabricated with Alloy 600 in some international operating SGs similar in design to the Westinghouse Model 51. This international operating experience occurred with proper primary water chemistry. However, there has been no PWSCC in Alloy 600 divider plate assemblies in U.S. PWRs. Nuclear industry evaluations concluded that PWSCC in a small number of foreign divider plates was caused by a shallow layer of cold work in the plate material and did not propagate after the initial detection. While PWSCC may occur in Alloy 600 SG tube-to-tubesheet welds, there has been none reported in U.S. operating SGs and the NRC staff is unaware of any in international operating SGs.

For LR, NUREG-1801, Generic Aging Lessons Learned (GALL) Report, Revision 2 (Reference 6), recommends managing PWSCC of nickel alloy SG divider plate assemblies and SG tube-to-tubesheet welds by the Water Chemistry aging management program (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, NUREG-1801 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 7), includes guidance for the evaluation.

Following development of NUREG-1801 and NUREG-1800 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 8).

NRC LR Interim Staff Guidance (ISG), LR-ISG-2016-01, Changes to Aging Management Guidance for Various Steam Generator Components (Reference 9), revised the LR guidance for aging management of SG divider plate assemblies and tube-to-tubesheet welds. LR ISG, LR-ISG-2016-01 accepted the EPRI Report analyses (Reference 8) 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 EPRI Report analyses are applicable and bounding for the licensees SGs. In response to LR-ISG-2016-01, EPRI provided licensees a checklist to screen their SGs for PWSCC susceptibility of the divider plate assembly and tube-to-tubesheet welds.

2.2 Regulatory Requirements and Guidance Title 10 of the Code of Federal Regulations (10 CFR) 54.21(a)(3) 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. A necessary finding for the NRC staff 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, Contents of applicationtechnical information, 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 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, 1 In NUREG-1801, Revision 2 (Reference 6), 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.

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, NUREG-1801 recommends evaluating the need for a plant-specific AMP to verify the effectiveness of the Water Chemistry AMP. This is the aging management approach approved for Callaway in NUREG-2172.

Revision 2 of NUREG-1800 provides guidance to the NRC staff for the review of LR applications. Section 3.1.2.2.11, Cracking due to Primary Water Stress Corrosion Cracking, 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 includes acceptance criteria to assist with determining if a plant-specific AMP is needed.

The acceptance criteria are based on SG tube material and whether permanent alternate repair criteria have been approved. This guidance was the basis for using the term not susceptible to PWSCC in Commitment No. 35.

The guidance in LR-ISG-2016-01 describes subsequent changes made to NUREG-1801, Revision 2, and NUREG-1800, Revision 2, 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. The guidance in 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 Steam Generators AMPs (rather than 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 Steam Generators 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. An acceptance criterion for both the nickel alloy SG divider plates and SG tube-to-tubesheet welds is that if the industry analyses (Reference 8) are applicable and bounding, then a plant-specific AMP is not required.

Compared to NUREG-1801, Revision 2 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. The guidance in LR-ISG-2016-01 considers plant-specific PWSCC susceptibility and whether visual inspections of SG primary head interior surfaces performed as part of 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 the evaluations, tests, and analyses by industry, NRC staff consideration of the industry work, and further staff review of operating experience.

3.0 TECHNICAL EVALUATION

3.1 Background

In March 2015, the NRC published NUREG-2172. In NUREG-2172, appendix A, Commitment Nos. 34 and 35 each provide for multiple options to fulfill the commitments. Both commitments have an option for performing a one-time inspection to determine if PWSCC is occurring and an option to perform an analysis. The purpose of the analysis for Commitment No. 34 is to conclude that the RCPB is adequately maintained if PWSCC occurs in the divider plate assembly welds. The purpose of the analysis for Commitment No. 35 is to either determine that the tube-to-tubesheet welds are not susceptible to PWSCC, or to redefine the pressure boundary of the tubes so that the welds are not part of the RCS. Commitment No. 34 has a third option to revise the commitment accordingly if industry and NRC studies determine that potential failure of the RCPB due to PWSCC in the divider plate assembly welds is not a credible concern.

By letter dated December 20, 2023, the licensee submitted to the NRC a response to Commitment Nos. 34 and 35 of the Callaway LR. For both commitments the licensee selected Option 2: Analysis. The licensee performed analytical evaluations specific to the Callaway SGs because it could not be determined if the Callaway SGs were bounded by the industry analyses in EPRI Report 3002002850. The analyses described in enclosure 1 of the licensees submittal dated December 20, 2023, apply to the Framatome-designed RSGs at Callaway; Salem, Unit 2; and Prairie Island, Units 1 and 2. The supplemental information on tube-to-tubesheet weld chromium content in enclosure 2 of the licensees December 20, 2023, submittal, is specific to the Callaway RSGs.

To better understand the technical basis for the tube-to-tubesheet weld chromium content and for concluding that the SG RCPB would be maintained in the presence of divider plate cracking, the NRC staff conducted an audit from May 20 to December 13, 2024. During the audit the staff reviewed relevant technical reports and detailed calculations supporting the licensees Callaway divider plate and tube-to-tubesheet weld analyses. The audit plan (Reference 10), and audit report (Reference 11) were issued on May 7, 2024, and February 10, 2025, respectively.

Following review of the information submitted by the licensee and additional audit information, the NRC staff requested confirmation of information related to the treatment of residual stress in the channel head cladding and base metal at the postulated cracking location. The licensee submitted its response in the supplement dated December 5, 2024 (Reference 2).

3.2 Evaluation of Proposed Response to License Renewal Commitment Nos. 34 and 35.

Commitment No. 34 provided three options to address the potential for PWSCC to occur in the divider plate welds to the primary head and tubesheet cladding:

Option 1 (Inspection): A one-time inspection capable of detecting PWSCC in the divider plate assemblies and associated welds.

Option 2 (Analysis): An analysis that concludes the SG RCPB is adequately 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. 35 provided two options to address the potential for PWSCC to occur in the tube-to-tubesheet welds:

Option 1 (Inspection): A one-time inspection of a representative number of tube-to-tubesheet welds in each SG capable of detecting PWSCC. Repair or engineering evaluation, and periodic monitoring for the remaining life of the SGs, if cracking is identified.

Option 2 (Analysis): An analysis that concludes tube-to-tubesheet welds are not susceptible to PWSCC, or a redefinition of the RCPB so that the tube-to-tubesheet welds are not required to perform an RCPB function.

3.2.1 Commitment No. 34 - PWSCC of the Divider Plate Welds to the Primary Head and Tubesheet Cladding To address Commitment No. 34 for Callaway, the licensee selected Option 2 (Analysis). This option requires the licensee to do the following:

Perform an analytical evaluation of the steam generator divider plate welds in order to establish a technical basis which concludes that the steam generator RCS 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.

The licensee could not determine if the analyses (Reference 8), performed by industry for a limiting Westinghouse-designed SG and accepted in LR-ISG-2016-01, are bounding for the Callaway Framatome-designed SGs because of geometrical and material differences, and potential differences in loading. Section 2.0 of enclosure 1 of the licensees submittal (Reference 1), states that the Framatome engineering analysis for the Callaway SGs is consistent with the industry analysis, and that the objective is to assess crack growth in the tubesheet to divider plate weld to determine the necessary inspections following LR.

Sections 2.0 and 3.0 of enclosure 1 of the licensees submittal (Reference 1), describe the analytical methodology for evaluating crack growth near the triple point region. The description states that the analysis is applicable to the Framatome RSGs at Callaway; Prairie Island, Units 1 and 2, and Salem, Unit 2, because the transients, design cycles, and support loads in the analysis are the bounding values among the four units.

Three-dimensional thermal and structural finite element models (FEMs) were developed to represent the bottom channel head, divider plate, tubesheet, associated welds, and cladding on the tubesheet and channel head.

The FEMs were used to evaluate a postulated flaw at the triple point intersection of the divider plate assembly, tubesheet, and channel head. The postulated flaw was a separation between the divider plate and tubesheet, extending approximately 25 percent of the SG radius from the triple point.

Pressure and thermal loads for normal, upset, emergency, and faulted conditions were used to calculate stress distributions.

Axial stress, hoop stress, and metal temperatures were extracted along six path lines near the triple point and used in fatigue crack growth analyses for axial and circumferential cracking into the primary head base material.

Final stress intensity factor (SIF) values were evaluated against the acceptance criteria in American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)Section XI, IWB-3600.2 2 American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, various editions as identified in text.

To evaluate the licensees analysis, the NRC staff reviewed the methodology and acceptance criteria, focusing on: (1) the consistency with the industry analyses in EPRI Report 3002002850 (Reference 8) and (2) the technical basis for differences between the analyses. During the audit, the licensee provided the staff with access to detailed documentation on topics including the crack growth evaluation tool, stress analyses, flaw tolerance evaluation, and tube-to-tubesheet weld composition (Reference 11).

The NRC staff reviewed the Callaway SG materials, identified in table 1 below, and the material properties used in the analysis. The materials and properties are listed in section 5.2 of enclosure 1 of the licensees December 20, 2023, submittal. The staff noted that in most cases the Callaway analysis was performed using material properties from the ASME Code. Both the Callaway and EPRI industry analyses used equivalent base metal properties from the ASME Code for weld materials. The assumed value of 0.3 for the Poissons ratio is consistent with the values (0.30 - 0.31) listed in ASME Code Section II. The temperature-dependent alloy densities were based on a Framatome materials property program document. The staff did not audit the values but found they are the same as or only slightly different than values the staff calculated using the linear thermal expansion coefficients in Section II, Part D of the ASME Code.3 Table 1, Materials for Components Used in the Licensees Finite Element Analysis (FEA)

Component Material Tubes SB-163 UNS N06690 Alloy 690 Divider plate SB-168 UNS N06690 Alloy 690 Cladding - divider plate to tubesheet FM 152 Cladding - divider plate to channel head FM 152 Cladding - tubesheet flat surface FM 82/182 Cladding - tubesheet cylindrical surface E308L/E309L, EQ308L/EQ309L Cladding - channel head inside surface EQ308L/EQ309L Channel head base metal SA-508 Cl. 3a or SA-508 Gr.3 Cl.2 Tubesheet SA-508 Cl. 3a or SA-508 Gr.3 Cl.2 Lower shell SA-508 Cl. 3a or SA-508 Gr.3 Cl.2 The primary head of an SG is subjected to design-basis internal pressure and thermal transients during the design life. The licensee calculated the stresses from pressure and thermal loading from normal, upset, and test conditions, using bounding transients and cycle numbers based on review of the transient data for the Framatome RSGs analyzed (Callaway, Salem, Unit 2, and Prairie Island, Units 1 and 2). Axial forces from support lugs attached to the lower outside surface of the channel head were included based on support lugs producing the highest axial loads for the Framatome replacement SGs analyzed.

The analysis assumed residual tensile stress of 2 ksi through the thickness of the channel head cladding. The licensee stated in in its supplement dated December 5, 2024 (Reference 2), that this is a conservative value based on the cladding being a small percentage (approximately 5 percent) of the base metal thickness, the post-weld heat treatment, the preservice hydrostatic testing, and inservice temperature changes. The licensee also stated in its December 5, 2024, supplement, that no residual stress was included in the base material because the nearest 3 American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section II, Materials, Part D, Properties, Tables TE-1 and TE-4, 2021 Edition.

circumferential welds are at least 13 inches from any postulated crack in the triple point region, which is beyond the distance at which the expected weld residual stress would attenuate to near zero. In addition, there are no longitudinal seam welds in the tubesheet rings. The NRC staff reviewed the information in the supplement, and found it supported the licensees treatment of residual stress.

The licensee used the three-dimensional FEM with the loads described above to calculate the through-wall stress distribution at the triple point due to pressure and thermal loading. The analysis assumed a separation between the tubesheet cladding and divider plate extending about 25 percent of the divider plate length towards the center of the tubesheet. The analysis assumed that at the triple point the initial postulated axial or circumferential semi-elliptical crack into the channel head base material was equal to the cladding thickness in depth, 0.2 inch minimal or the actual cladding thickness on the flaw path line. The analysis assumed the initial length of the postulated axial cracks was 6 times the depth of the crack, consistent with ASME Code Section XI, table L-3210-1. The analysis assumed the initial length of the postulated circumferential cracks was equivalent to the divider plate thickness, 2.08 inches. This results in a longer initial circumferential flaw than if the initial length was assumed to be 6 times the depth of the crack, as recommended by ASME Code,Section XI. The NRC staff noted that all aspects of this approach are consistent with the industry analyses in EPRI Report 3002002850 (Reference 8).

Axial stresses, hoop stresses, and metal temperatures were extracted to determine the SIFs for axial and circumferential crack growth along six path lines near the triple point. Four path lines were located on each side of the divider plate near the triple point. The other two path lines were located through the center of the divider plate near the triple point. The licensee stated in its supplement dated December 5, 2024, that the selected path lines passed through the locations with the maximum stresses in terms of magnitude and/or in variations. The axial and hoop residual stress in the cladding, transient stresses, sustained stresses (e.g., support lugs),

and crack face pressure stresses were used to calculate fatigue crack growth. In its supplement, the licensee also provided corrected heat transfer coefficient units (British thermal units per second-square inch-degree Fahrenheit (i.e., BTU/s-in2-°F)). The licensee stated that the stress and crack growth calculations were unaffected by the identification of incorrect units in its submittal (Reference 1).

The crack growth analysis was performed according to Article A-4300 of the ASME Code,Section XI, using the Framatome AREVACGC code. The NRC staffs audit included reviewing AREVACGC documentation, which included verification test cases as noted in section 3.2.4, Methodology for Flaw Growth Analysis, of enclosure 1 of the licensees submittal (Reference 1). Because AREVACGC used A-4300 in the 1995 edition of the ASME Code, the licensee assessed differences between the 1995 revision and the 2019 edition, which was the latest edition incorporated by reference in 10 CFR 50.55a(a)(1)(ii)(c)(56) at the time of the analysis. The assessment concluded there was no effect on the results because there was no difference in the applicable parts of A-4300. The staff evaluated the crack growth methodology described in enclosure 1 licensees submittal (Reference 1), and concluded it was consistent with the 2019 Code edition.

The fatigue crack growth analyses produced final flaw length and depth values for the lifetime load cycles applied to the circumferential and axial flaws, as shown in tables 6-4 and 6-5 of enclosure 1 of the licensees submittal. Finally, for each transient, including normal, emergency, and faulted conditions, maximum stress intensity values for these flaws for each path were calculated using the associated loads and compared to the acceptance criteria in ASME Code Section XI, IWB-3600. In tables 7-1 and 7-2 of enclosure 1 of the December 20, 2023 submittal, the licensee compared the final flaw sizes and stress intensity values to the allowable values from IWB-3600 for circumferential and axial flaws. The comparison showed the acceptance criteria were met, with margin, for the postulated axial and circumferential cracks. The licensee also performed the FEA for the cladding on the tubesheet primary surface in the region perforated for tube installation. The analysis found that the principal stresses were compressive at all nodes during steady state operation at 100 percent power.

The NRC staff evaluated the licensees methodology for evaluating PWSCC in the Callaway SG divider plate assemblies. The staff finds the methodology acceptable based on similarity to the industry methodology referenced and accepted by the staff in LR-ISG-2016-01 (Reference 9).

Specifically, the licensees methodology (a) uses FEA modeling with loads from bounding transients and other bounding loads to determine stress distributions and stress intensity factors from a postulated flaw 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 Article A-4300 of ASME Code,Section XI, (c) postulates multiple crack paths for axial and circumferential flaws with the most challenging stress conditions, and (d) compares the stress intensity values for the final crack geometries to acceptance criteria approved by the staff in 10 CFR 50.55a(a)(1)(ii)(c)(56). In addition, the staff found the technical basis for the differences between the Callaway analysis and the prior EPRI industry analysis to be reasonable and acceptable.

With respect to the conservatism of the overall approach, the licensee postulated a flaw in the form of a separation between the divider plate and tubesheet, with a length of about 25 percent of the tubesheet radius starting from the triple point, that is, consistent with the industry analyses. 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 8). No cracking in this location has been observed in the U.S. PWR fleet, 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. Other conservative assumptions identified by the licensee include the assumption of a SG lower assembly support design that produces the highest axial force and bending moment among the applicable Framatome RSG designs, and the assumption of tensile residual stress throughout the thickness of the channel head cladding. Based on consideration of the licensees SG design and methodology, the NRC staff concluded these assumptions are conservative and therefore acceptable.

As discussed in EPRI Report 3002002850 and LR-ISG-2016-01 (References 8 and 9, respectively), if PWSCC were to occur, it is not expected to cause a safety issue. The industry analyses 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. The guidance in LR-ISG-2016-01 recommends managing PWSCC of nickel alloy SG divider plates and nickel alloy SG tube-to-tubesheet welds with the Water Chemistry and Steam Generators AMPs. As discussed in LR-ISG-2016-01, the visual inspections performed under the SG AMP provide opportunities to identify cracking that 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 the supplements dated December 5 and December 12, 2024, satisfies Commitment No. 34 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 EPRI Report 3002002850 (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 NUREG-1800, Revision 2 (Reference 7), aging of the Callaway 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 does not require a plant-specific AMP.

The licensees aging management review in the LR application does not include the Steam Generator program for managing PWSCC of the divider plate assemblies. As discussed above, the SG program was added to the Water Chemistry program for managing PWSCC of these components as part of LR-ISG-2016-01, after Callaways renewed license was issued.

However, the licensee performs visual inspections of the primary-side cladding (Reference 12) capable of identifying cracking that grows into pressure boundary components through identification of rust stains or other abnormal observations such as boric acid deposits. The NRC staff also notes that PWSCC of divider plate assemblies has not been observed in the U.S. operating fleet, and the industry evaluations concluded that PWSCC in a small number of foreign divider plates was caused by a shallow layer of cold work in the plate material and did not propagate after the initial detection.

3.2.2 Commitment No. 35 - PWSCC of the Tube-to-Tubesheet Welds To address Commitment No. 35 for Callaway, the licensee selected Option 2 (Analysis). This option requires the following:

Perform an analytical evaluation of the steam generator tube-to-tubesheet welds either determining that the welds are not susceptible to PWSCC or redefining the reactor coolant pressure boundary of the tubes, where the steam generator tube-to-tubesheet welds are not required to perform a reactor coolant pressure boundary function. The redefinition of the reactor coolant pressure boundary will be submitted as part of a license amendment request requiring approval from the NRC. The evaluation for determination that the welds are not susceptible to PWSCC and do not require inspection will be submitted to the NRC for review.

NRC guidance issued after the Callaway 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 9) states that tube-to-tubesheet welds made between Alloy 690 tubes and Alloy 690 type cladding, which produces the highest chromium range among the materials used, would be highly resistant to PWSCC. The SG program, 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.

In section 7.1.1 of EPRI Report 3002002850 (Reference 8), 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 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 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.

The Callaway SGs have thermally treated Alloy 690 SG tubes and Alloy 600 type tubesheet cladding material. Due to potential differences from the SGs considered in the industry analyses, it could not be determined that the analyses were bounding for Callaway. The licensee evaluated the PWSCC susceptibility of the Callaway SG tube-to-tubesheet welds based on the minimum chromium content of the weld cladding materials, the minimum chromium content of the tubing material, and the tube-to-tubesheet weld design documented in manufacturing completion reports. Section 2.0 of enclosure 2 to the licensees submittal (Reference 1),contains the formula used to estimate the chromium content of the welds, which is the same formula used in the industry analysis based on testing and analysis (Reference 8).

The licensee found the chromium content in the tube-to-tubesheet welds for all locations in the Callaway SGs exceeded 22 percent. Combined with the determination that the tubesheet primary surface is in compression during operation, the licensee concluded a plant-specific AMP is not necessary to meet the NRC aging management guidance in LR-ISG-2016-01. The NRC staff audited the licensees documentation of weld chemistry and confirmed the minimum calculated chromium content.

The NRC staff finds the licensees response, as revised by the supplements dated December 5 and 12, 2024 (References 2 and 3, respectively), satisfies Commitment No. 35 based on the following:

Review of the licensees methodology for determining the minimum chromium content of the welds. The staff concluded the methodology is consistent with that used in the industry analyses (Reference 8).

The Callaway tube-to-tubesheet welds are autogenous welds joining the tubes to the cladding, as in the industry analyses. Therefore, the methodology used in the industry analyses is applicable to Callaway.

The calculated minimum chromium content of the welds is greater than 22 percent and the stresses at the tube-to-tubesheet welds are compressive during operation. As discussed in LR-ISG-2016-01 (Reference 9), the combination of 22 percent chromium and compressive stress make the likelihood of PWSCC very low.

The licensee showed that the combination of chromium content and compressive stress is such that a plant-specific AMP is unnecessary to meet the guidance in LR-ISG-2016-01.

The licensees aging management review in the LR application does not include the SG program for managing PWSCC of the divider plate assemblies. As discussed above, the SG program was added to the Water Chemistry program for managing PWSCC of these components as part of LR-ISG-2016-01, after Callaways renewed license was issued.

However, the licensee performs visual inspections of the primary-side cladding (Reference 12) capable of identifying cracking that grows into pressure boundary components through identification of rust stains or other abnormal observations such as boric acid deposits. The NRC staff also notes that PWSCC of tube-to-tubesheet welds has not been reported in U.S.

operating SGs and the staff is unaware of any in foreign operating SGs.

3.3 Technical Evaluation Conclusion

Based on the review described above, the NRC staff finds that the licensees response in its submittal, as supplemented, satisfies Commitment Nos. 34 and 35 documented in the license renewal Safety Evaluation Report in NUREG-2172 (Reference 4) 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 CLB during the period of extended operation, in accordance with 10 CFR 54.21(a)(3).

4.0 REFERENCES

1.

Witt, T. A., Ameren Missouri, letter to U.S. Nuclear Regulatory Commission, Callaway Plant License Renewal Resolution for Commitments 34 and 35 - Perform Evaluation of Crack Initiation and Propagation in Steam Generator Divider Plate and Tube-to-Tubesheet Welds, dated December 20, 2023 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML23354A244).

2.

Elwood, T. B., Ameren Missouri, letter to U.S. Nuclear Regulatory Commission, Supplementation Information and Response to Request for Confirmation of Information (RCI) to Resolve Callaway Plant License Renewal Commitments 34 and 35 - Perform Evaluation of Crack Initiation and Propagation in Steam Generator Divider Plate and Tube-to-Tubesheet Welds, dated December 5, 2024 (Package ML24340A120).

3.

Elwood, T. B., Ameren Missouri, letter to U.S. Nuclear Regulatory Commission, Updated Framatome Reports to Resolve Callaway Plant License Renewal Commitments 34 and 35 - Perform Evaluation of Crack Initiation and Propagation in Steam Generator Divider Plate and Tube-to-Tubesheet Welds, dated December 12, 2024 (Package ML24347A225).

4.

U.S Nuclear Regulatory Commission, Safety Evaluation Report Related to the License Renewal of Callaway Plant, Unit 1, NUREG-2172, March 2015 (ML15068A342).

5.

Electric Power Research Institute, Industry Presentation Slides, NRC/Industry Meeting Regarding Tube-to-Tubesheet Weld and Divider Plate Cracking Report, dated July 30, 2015 (ML15211A507).

6.

U.S. Nuclear Regulatory Commission, Generic Aging Lessons Learned (GALL) Report, NUREG-1801, Revision 2, December 2010 (ML103490041).

7.

U.S. Nuclear Regulatory Commission, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, NUREG-1800, Revision 2, December 2010 (ML103490036).

8.

Electric Power Research Report, Steam Generator Management Program: Investigation of Crack Initiation and Propagation in the Steam Generator Channel Head Assembly, EPRI Report 3002002850, Palo Alto, CA, October 2014

9.

U.S. Nuclear Regulatory Commission, Changes to Aging Management Guidance for Various Steam Generator Components, License Renewal Interim Staff Guidance LR-ISG-2016-01, November 2016 (ML16237A383).

10.

Chawla, M. L., U.S. Nuclear Regulatory Commission, letter to F. Diya, Ameren Missouri, Callaway Plant, Unit No. 1 - Audit Plan to Support Review of Steam Generator License Renewal Response to Commitment Nos. 34 and 35 (EPID L-2024-LRO-0009), dated May 7, 2024 (ML24122A150).

11.

Chawla, M. L., U.S. Nuclear Regulatory Commission, letter to F. Diya, Ameren Missouri, Callaway Plant, Unit 1 - Audit Report Regarding the Response to Steam Generator License Renewal Commitment Nos. 34 and 35, dated February 10, 2025 (ML25003A165).

12.

Witt, T. A., Ameren Missouri, letter to U.S. Nuclear Regulatory Commission, Callaway Plant, Unit 1 - Results of Steam Generator Tube In-Service Inspection, dated November 1, 2022 (Package ML22305A679).

Principal Contributor: G. Makar, NRR Date: April 1, 2025

ML25079A202

• concurrence by email OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA*

NRR/DNRL/NCSG/BC*

NAME WOrders PBlechman SBloom DATE 3/20/2025 3/25/2025 3/25/2025 OFFICE NRR/DORL/LPL4/BC*

NRR/DORL/LPL4/PM*

NAME TNakanishi MChawla (WOrders for)

DATE 4/1/2025 4/1/2025