ML22129A013

NRR E-mail Capture - Draft RAIs for Requests for Alternatives I4R-17, I4R-23, ISI-05-018, I6R-10

ML22129A013
Person / Time
Site:	Calvert Cliffs, Byron, Braidwood, Ginna
Issue date:	05/06/2022
From:	Joel Wiebe NRC/NRR/DORL/LPL3
To:	Tom Loomis Exelon Generation Co
References
L-2021-LLR-091, L-2021-LLR-092, L-2021-LLR-093, L-2021-LLR-094
Download: ML22129A013 (6)

Text

From: Wiebe, Joel Sent: Friday, May 6, 2022 2:21 PM To: Loomis, Thomas R:

Cc: Paige, Jason; Sreenivas, V

Subject:

Draft RAIs for Requests for Alternatives I4R-17, I4R-23, ISI-05-018, I6R-10 (EPID Nos.: L-2021-LLR-091, L-2021-LLR-092, L-2021-LLR-093, L-2021-LLR-094)

Hi Tom, Here are the draft RAIs for your December 14, 2021, relief request. Let me know within one week if you need a clarification call. Your response to the RAIs is requested within 30 days of the date of this email or within 30 days of the clarification call, whichever is later.

Joel REQUEST FOR ADDITIONAL INFORMATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR ALTERNATIVE NOS. I4R-17, I4R-23, ISI-05-018, I6R-10 BRAIDWOOD STATION, UNITS 1 AND 2 BYRON STATION, UNITS 1 AND 2 CALVERT CLIFFS NUCLEAR POWER PLANT, UNITS 1 AND 2 R.E. GINNA NUCLEAR POWER PLANT CONSTELLATION ENERGY GENERATION, LLC (FORMERLY EXELON GENERATION COMPANY, LLC)

DOCKET NOS.: 50-456, 50-457, 50-454, 50-455, 50-317, 50-318, 50-244 EPID NOS.: L-2021-LLR-091, L-2021-LLR-092, L-2021-LLR-093, L-2021-LLR-094

Background

By letter dated December 14, 2021 (Agencywide Document Access and Management System Accession Number ML21348A078), Exelon Generation Company, LLC (the licensee) submitted to the United States Nuclear Regulatory Commission (NRC), a proposed alternative to the inservice inspection (ISI) requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) of the steam generator (SG) welds of Braidwood Station (Braidword), Units 1 and 2, Byron Station (Byron), Units 1 and 2, Calvert Cliffs Nuclear Power Plant (Calvert Cliffs), Units 1 and 2, and R.E. Ginna Nuclear Power Plant (Ginna). On February 1, 2022 (ADAMS Accession No. ML22032A333), Exelon Generation Company, LLC was renamed Constellation Energy Generation, LLC.

Specifically, pursuant to Title 10 of the Code of Federal Regulations, Part 50 (10 CFR Part 50),

Paragraph 50.55a(z)(1), the licensee is proposing to defer the ISI examinations for the SG welds of the subject units from the current ASME Code,Section XI 10-year requirement to the end of the currently approved periods of extended operation for the units. The licensee referred to the results of the probabilistic fracture mechanics (PFM) analyses in the following Electric Power Research Institute (EPRI) non-proprietary report as the primary basis for the deferral of the ISI examinations: 3002015906 Technical Bases for Inspection Requirements for PWR

[Pressurized Water Reactor] Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds, 2019 (hereinafter referred to as EPRI report 15906 (ADAMS Accession No. ML20225A141). The NRC staff needs additional information to complete its review and approval of the licensees submittal.

Regulatory Basis The NRC has established requirements in 10 CFR Part 50 to protect the structural integrity of structures and components in nuclear power plants. Among these requirements are the ISI requirements of Section XI of the ASME Code incorporated by reference in 10 CFR Part 50.55a to ensure that adequate structural integrity of the SG vessel (including welds) is maintained through the service life of the vessel. Therefore, the regulatory basis for the following requests for additional information (RAIs) has to do with demonstrating that the proposed alternative ISI requirements would ensure adequate structural integrity of the SG welds of the subject units for which EPRI report 15906 is referenced, and thereby would provide an acceptable level of quality and safety per 10 CFR 50.55a(z)(1) for these welds.

RAI-1

Issue The licensee referenced probabilistic and deterministic analyses in EPRI report 15906 to estimate potential fatigue growth in the subject SG welds. The licensee presented plant-specific information to demonstrate that the referenced EPRI analyses would bound the subject SG welds, including high-level results from previous ISI of the welds. The licensee also provided limited discussion of performance monitoring, primarily focused on justifying application of the EPRI analyses to the proposed ISI interval extension for the subject SG welds (i.e., that leakage would be detected).

Leveraging PFM analyses to define the basis for risk-informing inspection requirements requires knowledge of both the current and future behavior of the material degradation and the associated uncertainties applicable to the subject SG welds. Confidence in the results of these analyses hinges on the assurance that the PFM model adequately represents, and will continue to represent, the degradation behavior in the subject SG welds. The NRC staff has determined that, when considering extended examination intervals, adequate performance monitoring through inspections is needed to ensure that the PFM model continues to predict the material behavior and that emergent degradation is discovered and dispositioned in a timely fashion.

The licensee discusses the system leakage test as providing further assurance for the proposed alternative. However, the NRC staff notes that the visual examinations performed during system leakage tests may not provide sufficient information to ensure that the PFM model continues to predict the material behavior and that emergent degradation is discovered and dispositioned in a timely fashion. Specifically, visual examinations may not directly detect pertinent integrity conditions (e.g., presence or extent of degradation); may not provide direct detection of aging effects prior to potential loss of structure or intended function; and do not provide sufficient validating data necessary to confirm the modeling of degradation behavior in the subject SG welds.

Request

a. Describe the performance monitoring that will be implemented with this proposed alternative to ensure that the PFM model adequately represents, and will continue to represent, the degradation behavior in the subject components commensurate with the duration of the requested alternative (i.e., plant-specific end date).

b. Justify that this performance monitoring will meet this objective and address the concerns discussed above.

c. Explain how this performance monitoring will provide, over the extended examination interval, (1) direct evidence of the presence and extent of degradation, (2) validation and confirmation of the continued adequacy of the PFM model; and (3) timely detection of novel or unexpected degradation.

d. Describe any actions that will be taken if issues are identified through this performance monitoring to ensure that the integrity of the component is adequately maintained.

RAI-2

Issue The PFM analyses in the EPRI report 15906 assume certain examination histories, e.g.,

preservice inspection (PSI) followed by 10-year inspections and assume that 100% coverage was assumed during the PSI examination. The licensee provided actual examination histories for the subject units in Appendix B to the submittal. However, Appendix B does not provide information on PSI history for the units. Also, there seems to be gaps in the SG weld examinations at the subject units.

Request

a. Confirm that PSI examinations were performed on all SG welds at the subject units and that 100% coverage was achieved for the examinations.

b. Discuss whether any of the subject SG welds at the units contain fabrication defects.

c. Discuss any SG welds of the units that required repairs during the fabrication of the SGs prior to commercial operation at the units.

d. Discuss why for some SG welds of the subject units, the examinations of earlier ISI intervals are not shown in Appendix B to the submittal (e.g., the first ISI interval for Braidwood Unit 1).

RAI-3

Issue The following issues have to do with the plant-specific inspection histories included in Appendix B to the submittal.

Section 1 of Attachment 1 to the submittal shows that each SG at Braidwood Unit 1 contains five welds that are required to be examined. Tables B1, B2, B3, and B4 of Appendix B to the submittal show the welds that have been examined. The NRC staff noted that in the Braidwood Unit 1 second and third ISI interval, the licensee examined only four, not five, subject SG welds.

The licensee did not examine the Upper Secondary Shell - Shell Cone weld (component ID numbers 1SG-05-SGC-05, 1SG-06-SGC-05, 1SG-07-SGC-05, and 1SG-08-SGC-05, item number C1.10) during the second and third ISI intervals. The NRC staff recognizes that footnote 4 to the ASME Code,Section XI, Table IWC-2500-1, Examination Category C-A, Item No.

C1.10 permits welds from only one SG to be examined. However, the ASME Code does not state that a weld in the required SG weld sample can be unexamined.

Also, Section 1 of Attachment 1 to the submittal shows the SG welds at the Byron Units 1 and 2.

However, Tables B5, B6, B7, and B8 of Appendix B to the submittal, which show the SG welds of the Byron units that were examined, do not show all the welds in Section 1 of Attachment 1 to the submittal for the Byron Units 1 and 2.

The note in Table B10 of Appendix B to the submittal for the Calvert Cliffs units refers to Paragraph IWB-2412(b) of Section XI of the ASME Code (2013 Edition for the Calvert Cliffs code of record stated in the submittal), but there is no IWB-2412(b) in of Section XI of the ASME Code. The same note in Table B10 states that the referenced Paragraph of the ASME Code allowed the licensee to forego the inspection of the W7 SG welds of the Calvert Cliffs units. The NRC staff noted that, however, it appears from Table B10 that other welds were not inspected.

For example, welds SG-12-W6, SG-21-W5, and SG-21-W6 were inspected only during the fourth ISI interval, and not during the third ISI interval per Table B10.

Request

a. Discuss why the Upper Secondary Shell - Shell Cone welds (1SG-05-SGC-05, 1SG SGC-05, 1SG-07-SGC-05, and 1SG-08-SGC-05) in the Braidwood Unit 1 SG weld sample were not examined in the second and third ISI interval.

b. Discuss why Tables B5, B6, B7, and B8 of Appendix B to the submittal do not show all the welds in Section 1 of Attachment 1 to the submittal for the Byron Units 1 and 2.

c. Clarify which Paragraph of Section XI of the ASME Code the note in Table B10 of Appendix B to the submittal is supposed to refer to.

d. Discuss why welds SG-12-W6, SG-21-W5, and SG-21-W6 of the Calvert Cliffs units were inspected only during the fourth ISI interval.

RAI-4

Issue Section 1 of Attachment 1 to the submittal states that the affected components are SG pressure retaining welds and full penetration welded nozzles. However, the welded nozzles are not identified in the affected components tables for the Braidwood, Byron, and Ginna units.

Request

a. Discuss whether Braidwood, Byron, and Ginna have any full-penetration welded nozzles attaching to the SG shell.

b. If the answer is affirmative, discuss whether they are classified under Examination Category B-D and Item Number B 3.130 in accordance with the ASME Code,Section XI, Table IWB-2500-1.

c. If the answer is affirmative, confirm that these full-penetration welded nozzles are not required to be examined.

RAI-5

Issue Tables A4 and A8 of Appendix A to the submittal shows transient cycles for SG secondary side vessel welds at the Braidwood and Byron units. Footnotes 4, 5, and 6 state that pressure values for various transients could not be found in the design documentation; therefore, the licensee selected pressure values from other sources.

Request Discuss whether the pressure values selected for the transients in Footnotes 4, 5, and 6 in Tables A4 and A8 are bounding, i.e., if these pressure values were used in the analysis, the results would still be within the probability of failure values calculated in EPRI report 15906.

RAI-6

Issue For the B2.40 and B3.130 welds of the subject units requested in the submittal, the tables in Appendix A to the submittal that show the transient cycles for each of the units also show the lowest temperatures for the Heatup/Cooldown transient for each unit: 120°F for Braidwood and Byron (Tables A3, A4, A7, and A8), 70°F for Calvert Cliffs (Tables A12 and A13), and 100°F for Ginna (Tables A16 and A17). Section 8.2.2.5 of EPRI report 15906 states that the minimum temperature (200°F) during this transient corresponds to Figure 7-22 of EPRI report 15906, and that therefore a fracture toughness (KIC) set at the upper shelf value of the ASME Code KIC curve, 200 ksiin, may be used. The NRC staff noted that Figure 7-22 of EPRI report 15906 is at the end of heatup (as noted in the figure). During the ramp periods at the beginning and end of the Heatup/Cooldown transient, the temperatures at the subject B2.40 and B3.130 welds of the subject units could be lower than 200°F as indicated in the above tables, and thus, KIC could be lower than 200 ksiin. A similar was issue was addressed for Millstone 2 in letter dated March 19, 2021 (ADAMS Accession No. ML21081A136) through plots that compare the KIC history for the Heatup/Cooldown transient, which were determined based on a plant-specific RTNDT value of 0°F for the Millstone 2 SG, with the corresponding applied stress intensity factor history (Figures RAI-4-1 and RAI-4-2 in the March 19, 2020 letter). The comparison showed that the applied stress intensity factor during anytime of the Heatup/Cooldown transient did not exceed the lowest KIC plant-specific value for Millstone 2 and was accepted by the NRC staff in the safety evaluation dated July 16, 2021 (ADAMS Accession No. ML21167A355).

Request Show that the applied stress intensity factor at the B2.40 and B3.130 SG welds of the subject units during anytime of the Heatup/Cooldown transient does not exceed the lowest plant-

specific KIC value for each subject unit that occur at the beginning and end of the Heatup/Cooldown transient.

RAI-7

Issue Tables A1, A2, A5, A6, A9, A10, A11, A14, and A15 in Appendix A to the submittal, the licensee states that the subject units have not resulted in an event (unheated auxiliary feedwater being introduced into a hot SG) that can result in a thermal shock of the SG vessel. The NRC noted that the auxiliary feedwater event referred to in these tables affects the secondary side welds (Item Nos. C1.10, C1.20, and C1.30) requested for the units. The NRC staff noted that a potential thermal shock event from the primary side can affect the primary side SG welds requested for the subject units in the submittal (Item Nos. B2.40 and B3.130).

Request Similar to the thermal shock applicability discussion in Tables A1, A2, A5, A6, A9, A10, A11, A14, and A15 in Appendix A to the submittal, explain whether each of the subject units in the submittal has experienced a thermal shock event in the primary side of the reactor coolant system that can affect the KIC value assumed in the analysis in EPRI report 15906 referenced for the primary side SG welds requested for the subject units in the submittal.

RAI-8

Issue In Section 5.2 of EPRI report 15906, EPRI stated that the analyses in the report did not consider test conditions beyond a system leak test and that no separate test conditions were included in the evaluation. EPRI noted that pressure testing (i.e., hydrostatic tests or system leakage tests) requirements are in Subsubarticle IWA-4540 of ASME Code,Section XI. Subsubarticle IWA-4540 states in part that pressure-retaining boundary shall include a hydrostatic or system leakage test, prior to, or as part of, returning to service. The NRC staff noted that in many PWRs, the system leakage test is integrated to the heatup process after a refueling outage to meet the pressure testing requirements, and that thus, pressure testing need not be analyzed separately from the Heatup/Cooldown transient. However, it is not clear from the submittal whether the practice of integrating the system leakage test to the heatup process is being performed at the subject units.

Request

a. For the subject units in the submittal, clarify whether the system leakage test is integrated to the heatup process.

b. If the system leakage test is not integrated to the heatup process for the subject units, and since pressure testing is not analyzed separately in EPRI report 15906, explain how the transients in EPRI report 15906 selected for analyses bound pressure testing and confirm that the temperature during the test is high enough such that the assumption of an upper shelf KIC value of 200 ksiin is appropriate.

RAI-9

Issue Section 5 of Attachment 1 to the submittal states that for the Calvert Cliffs units, PSI has been followed by the performance of four 10-year interval inspections for welds W3 and W4, but records were only found for the Third and Fourth Interval examinations.

Request Confirm that the coverages achieved for the 1st and 2nd 10-year ISI intervals for the SG welds W3 and W4 of the Calvert Cliffs units listed in Tables B11 and B12 of Appendix B to the submittal were not less than the smallest coverage shown in the tables.

Hearing Identifier: NRR_DRMA Email Number: 1625 Mail Envelope Properties (MN2PR09MB49712BB555C8082CD753FD388BC59)

Subject:

Draft RAIs for Requests for Alternatives I4R-17, I4R-23, ISI-05-018, I6R-10 (EPID Nos.: L-2021-LLR-091, L-2021-LLR-092, L-2021-LLR-093, L-2021-LLR-094)

Sent Date: 5/6/2022 2:20:35 PM Received Date: 5/6/2022 2:20:00 PM From: Wiebe, Joel Created By: Joel.Wiebe@nrc.gov Recipients:

"Paige, Jason" <Jason.Paige@nrc.gov>

Tracking Status: None "Sreenivas, V" <V.Sreenivas@nrc.gov>

Tracking Status: None "Loomis, Thomas R:" <thomas.loomis@constellation.com>

Tracking Status: None Post Office: MN2PR09MB4971.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 17033 5/6/2022 2:20:00 PM Options Priority: Normal Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date: