ML21281A141

Authorization and Safety Evaluation for Alternative Reactor Pressure Vessel Inservice Inspection Intervals

ML21281A141
Person / Time
Site:	Oconee
Issue date:	11/19/2021
From:	Shawn Williams Plant Licensing Branch II
To:	Snider S Duke Energy Carolinas
Williams S
References
EPID L-2021-LLR-0004
Download: ML21281A141 (16)

Text

November 19, 2021 OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 - AUTHORIZATION AND SAFETY EVALUATION FOR ALTERNATIVE REACTOR PRESSURE VESSEL INSERVICE INSPECTION INTERVALS (EPID L-2021-LLR-0004)

LICENSEE INFORMATION Recipients Name and Address: Steven M. Snider Site Vice President, Oconee Nuclear Station Duke Energy Carolinas, LLC 7800 Rochester Highway Seneca, SC 29672-0752 Licensee: Duke Energy Company (the licensee)

Plant Name and Unit: Oconee Nuclear Station (ONS), Units 1, 2 and 3 Docket Nos.: 50-269, 50-270, and 50-287 APPLICATION INFORMATION Submittal Letter (Relief Request No.) and Date: Serial Letter No. RA-20-0328, dated January 19, 2021 (Reference 1)

Submittal Agencywide Documents Access and Management System (ADAMS) Accession No.: ML21019A276 (Reference 1)

Supplement Date: Serial Letter No. 21-0219, dated August 5, 2021 Supplement ADAMS Accession Nos.: ML21217A293 (Reference 2) Non-Proprietary Version)1, ML21217A294, (Reference 3) Proprietary Version)2 Applicable Inservice Inspection (ISI) Program Interval and Interval Start/End Dates:

The alternative is applicable to the Fifth and Sixth 10-year inservice inspection (ISI) intervals for ONS, Units 1, 2, and 3. The Fifth 10-year ISI interval began on July 15, 2014, and is scheduled to end on July 15, 2024. The Sixth 10-year ISI interval is scheduled to start on July 16, 2024, and end on July 15, 2034.

Alternative Provision: The applicant requested the alternatives for ONS, Units 1, 2, and 3, in accordance with Title 10 of the Code of Federal Regulations (10 CFR), paragraph 50.55a(z)(1),

on the basis that the alternatives provide an acceptable level of quality and safety.

1 A non-proprietary record is available to the public that includes the cover letter (Serial Letter No. RA-21-0219 and non-proprietary responses to requests for additional information (RAIs) Nos. 1 - 3 and the publicly available, redacted response to RAI 4.

2 Proprietary record that includes the proprietary response to RAI No. 4 (Proprietary Enclosure 2 to Serial Letter No.

RA-21-0219).

ISI Requirement and Affected Components: The American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Division 1 (ASME Section XI), Paragraph IWB-2411 (Reference 4), requires the licensee to perform volumetric inspections of essentially 100%

of the reactor pressure vessel (RPV) pressure-retaining welds or nozzle radius sections in the following ASME Section XI, Table IWB-2500-1 Examination Categories and Inspection Items listed below once every 10-year ISI interval:

Unit 1 Affected Components:

• Examination Category B-A, Inspection Item B1.11, Circumferential Shell Welds

• Examination Category B-A, Inspection Item B1.12, Longitudinal Shell Welds

• Examination Category B-A, Inspection Item B1.21, Circumferential Head Welds

• Examination Category B-A, Inspection Item B1.30, Shell-to-Flange Weld

• Examination Category B-D, Inspection Item B3.90, Nozzle-to-Vessel Welds

• Examination Category B-D, Inspection Item B3.100, Nozzle Inside Radius Section Units 2 and 3 Affected Components:

• Examination Category B-A, Inspection Item B1.11, Circumferential Shell Welds

• Examination Category B-A, Inspection Item B1.21, Circumferential Head Welds

• Examination Category B-A, Inspection Item B1.30, Shell-to-Flange Weld

• Examination Category B-D, Inspection Item B3.90, Nozzle-to-Vessel Welds

• Examination Category B-D, Inspection Item B3.100, Nozzle Inside Radius Section

Applicable Code Edition and Addenda

Fifth ISI Interval - ASME Section XI, 2007 Edition through 2008 Addenda (Reference 5)

Sixth ISI Interval - ASME Section XI, Edition and Addenda applicable at the time of examination (Reference 6)

Proposed Alternative and Basis for Use: The licensee requested an alternative to the requirement in ASME Code Section XI, Paragraph IWB-2411, for RPV components or welds in inspection categories and items in IWB-2500 and Table IWB-2500-1. The licensee stated that IWB-2411 requires volumetric examination of essentially 100% of reactor vessel pressure-retaining welds identified in Table IWB-2500-1 once each ISI interval. In order to meet the requirement for the Fifth 10-year ISI interval, the licensee would need to complete the volumetric inspections of the specified RPV weld types and inside radius components by July 15, 2024. The licensees schedule would perform these inspections in Fall 2022 for Unit 1, Fall 2023, for Unit 2, and Spring 2024 for Unit 3, respectively, to meet the Fifth interval. The licensee requests NRC approval to defer performance of the required Fifth 10-year ISI interval volumetric inspections to the Sixth 10-year ISI interval that ends July 16, 2034. If authorized, this alternative would allow the licensee to perform the volumetric inspections in 2032, 2033, and 2034 for Oconee Units 1, 2, and 3, respectively. The licensee stated that extension of the interval between examinations of Category B-A and B-D welds from 10 years to up to 20 years will result in a reduction in radiological exposure and examination costs. The licensees performance of risk-informed RPV fracture analyses that was included in the unit-specific relief requests and the licensees determination that the extension of the ISI interval results in only a negligible change in risk, as demonstrated by satisfying the change in the large early release frequency (LERF) criterion specified in Regulatory Guide (RG) 1.174 Revision 3 (Reference 7).

The NRC staffs safety evaluation dated July 26, 2011(Reference 8), and WCAP-16168-NP-A, Rev. 3 (Reference 9) forms the basis for the licensees unit-specific relief requests and NRC staffs approval. Per WCAP-16168-NP-A, the licensee performed a risk-informed RPV fracture

analyses and demonstrated that the required parameters for each ONS unit are bounded by the results for the Babcock and Wilcox (B&W) pilot plant.

The licensees alternative is based on the NRC staff-approved, risk-informed flaw evaluation methods in WCAP-16168-NP-A, which were developed by the Pressurized Water Reactor Owners Group (PWROG) to satisfy the 95th percentile total through-wall cracking frequency (TWCF95-TOTAL) criteria for PWRs in NRC NUREG-1874 (Reference 10) and the LERF criteria specified in RG 1.174, Rev. 3. The pilot plant evaluated in WCAP-16168-NP-A for the fleet of U.S. B&W-designed pressurized water reactors (PWRs) is the ONS Unit 1 reactor vessel. The NRC staff approved methods in WCAP-16168-NP-A conservatively sets the TWCF95-TOTAL less than the corresponding change in CDF and LERF values. The evaluation of CDF, LERF, and RG 1.174 principles are addressed in WCAP 16168-NP-A and are not repeated in the licensees submittal or this safety evaluation.

NRC STAFF EVALUATION Relevant Parameters and Information for These Types of Risk-Informed Alternatives In Section 3.4 and Appendix B of the safety evaluation dated July 26, 2011, the NRC staff established assessment criteria that define the risk-informed parameters, data information, and acceptance criteria for requests to extend RPV ISI intervals from 10 to 20 years.

ISI Implementation Schedules In Section 3.4 and Appendix B of the safety evaluation dated July 26, 2011, the NRC staff stated that the dates for ISI of the RPV welds (and RPV nozzle inside surface location inspections) must be within one refueling cycle of the dates identified for inspection in the implementation plan in PWROG Letter No. OG-10-238 (Reference 11).

In Table 2 of the relief alternative submittal for ONS Units 1, 2, and 3, the licensee identifies that the proposed alternative would defer the resulting ISI inspections of the applicable RPV weld and nozzle components to the year 2032, 2033, and 2034 refueling outage for each Unit respectively. The NRC staff find this acceptable because it is consistent with the ISI schedule for each Unit given in PWROG letter No. OG-10-238.

Based on the above, the NRC finds the licensee has proposed unit-specific ISI implementation schedules for ONS, Units 1, 2, and 3 conform to the with the acceptance criteria in the safety evaluation dated July 26, 2011.

Identification of Limiting Design Basis Transients for Pressurized Thermal Shock (PTS) and Relevant Cladding Information In Appendix B of the safety evaluation dated July 26, 2011, the NRC staff requested confirmation that the dominant transients for PWR pressurized thermal shock (PTS) in NRC PTS Risk Study (i.e., NUREG-1874) are applicable as the dominant design basis transients for the licensees proposed risk-based alternative. The NRC staff also requested information relative to the design of the RPV cladding layer.

Regarding the PTS transients, the licensee identified that the analysis in NRC letter report, Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants, (Reference 12) serves as the plant-specific basis for demonstrating that the limiting design basis transients for the RPV welds are the dominant transients for B&W designed PWRs identified for PTS in NUREG-1874 (and as reflected in WCAP-16168-NP-A). The NRC staff

verified that, for B&W-type PWRs like ONS Units 1, 2 and 3, the PWROGs risk-based crack growth evaluation and TWCF95-TOTAL assessment for the ONS Unit 1 RPV is the bounding evaluation for B&W-designed PWRs, and that the dominant PTS transients are reported in Appendix L of WCAP-16168-NP-A (as based on the ONS Unit 1 pilot plant assessment in the report). Therefore, the NRC staff finds this criterion has been met based on the information in the NRCs PTS Risk Study and the conclusions in the safety evaluation dated July 26, 2011, that establish that the PTS transient characteristics for U.S. PWR-designed light water reactor facilities are generally applicable for plants designed by the same reactor vendor (i.e., B&W for the ONS, Unit 1, 2, and 3).

Regarding the cladding layers, the licensee reported that the cladding for the RPVs at ONS Units 1, 2 and 3 were deposited using a single layer. The NRC staff confirmed that the RPV claddings at ONS Units were deposited using a single layer. The NRC staff also noted that additional assurance against the consequences of postulated, propagating underclad cracks (i.e., as a result of fatigue or transient cycle-induced flaw growth) was analyzed and provided in B&W Non-Proprietary Report No. BAW-2251A, Appendix C (Reference 13) and accepted by the NRC staff in Section 4.2.4.3.4 of NUREG-172 (Reference 14). Based on the above, the NRC staff confirmed that the design of the RPV cladding layers at ONS Units 2 and 3 are consistent with the design and manufacture of the cladding layer in the ONS Unit 1 RPV (as the assessed B&W-design pilot), and that the cladding has been evaluated for consequences of postulated underclad cracks. For the proposed alternative, the NRC staff finds that the licensee did not need to evaluate the impacts that multiple pass layers would have on the design of the RPV cladding at ONS, Units 1, 2, and 3 because the NRC staff confirmed that: (1) the cladding at ONS Units 1, 2, and 3 were all deposited as a single layer austenitic stainless layer, and (2) the design of the cladding layers are consistent and bounded by the NRC staffs assessment of the ONS Unit 1 cladding layer in the safety evaluation dated July 26, 2011.

Based on the above, the NRC staff finds that the licensee has provided sufficient demonstration of the NRC staffs acceptance criteria for identification of limiting PTS transients and RPV cladding design.

Reporting of Past ISI Inspection Results and Flaw Distribution Assessments of Flaw Indications Detected in Past ISI Inspections In its safety evaluation dated July 26, 2011, the NRC staff established that licensees must report the results of the prior ISI of the RPV welds and the proposed schedule for the next 20-year ISI interval.

Further, the NRC staff established that for the flaw populations and flaw sizes within the inner 1 inch or inner 1/10th of the defined RPV design thicknesses, the relief submittal include sufficient demonstration that the flaws detected are bounded appropriately by the pilot plant.

In Table 2 of the individual relief alternatives, the licensee reported the results of the four required RPV weld ISI examinations for the initial licensing of the units. For the ONS Unit 1 assessment, the licensee reported there were a total of 39 RPV subsurface indications that were detected in the RPV beltline and extended beltline region and that the flaws were located in either the intermediate-to-upper shell, upper shell-to-lower shell, lower shell-to-dutchman circumferential welds, the intermediate shell longitudinal welds, or the plate components. For the ONS Unit 2 assessment, the licensee reported there were a total of 24 RPV indications that were detected in the RPV beltline and extended beltline region, with two of the indications being surface-breaking indications (with one on the inside surface of the RPV and one on the outside surface) and the remaining 22 flaw indications being subsurface flaws that are located in either the lower nozzle belt-to-upper shell, upper shell-to-lower shell, and lower shell-to-Dutchman

circumferential welds or the forging components. For the ONS Unit 3 assessment, the licensee reported there were a total of 47 RPV indications that were detected in the RPV beltline and extended beltline region, with the flaw indications being subsurface flaws that are located in either the lower nozzle belt-to-upper shell, upper shell-to-lower shell, lower shell-to-Dutchman circumferential welds or the forging components.

The licensee also stated that the 39 flaw indications in Unit 1, 24 indications in Unit 2, and 47 indications in Unit 3 were found to be acceptable without need for repair per the acceptance criterion in ASME Code,Section XI, Table IWB-3510-1. Of these reported indications, the licensee reported that: (a) 10 of the indications (four in the weld material and six in the plate material) in Unit 1 were located within the inner 1/10th or inner 1 inch of the RPV wall thickness and required additional flaw distribution evaluation in accordance with the alternate PTS rule (i.e., 10 CFR 50.61a); (b) four (4) of the indications in Unit 2 were located within the inner 1/10th or inner 1 inch of the RPV wall thickness and required additional flaw distribution evaluation in accordance with the alternate PTS rule, (c) five (5) of the indications in Unit 3 were located within the inner 1/10th or inner 1 inch of the RPV wall thickness and required additional flaw distribution evaluation in accordance with the alternate PTS rule.

The licensee reported the results of the four required RPV weld ISI examinations for the 1st, 2nd, 3rd, and 4th 10-year ISI intervals for the units. The licensee provided a summary of its past unit-specific ISI results (including identification total number of flaws detected in the RPVs, identification of RPV base metal or weld components containing the flaws, and component-specific subset population of flaws located within the inner 1 inch or inner 1/10th of the defined RPV design thickness) of the unit-specific TWCF95-TOTAL assessments, as shown in Table 2 of this SE. Table 1 of this SE provides a summary of the past ISI inspection results for ONS Units 1, 2, and 3. Table 2 of this SE provides of summary of the flaw distribution results for flaw indications in the ONS Unit 1, 2, and 3 RPVs that were detected within the inner 1 inch or inner 1/10th of the defined RPV design thicknesses.

In RAI 3 (Reference 15), the NRC staff requested the licensee confirm whether the unit-specific ISI inspections of the RPV during the 4th 10-year ISI intervals were the first ISI inspections that detected the flaw indications and whether there is any site-specific flaw growth data for the flaw indications evaluated in Table 2 of unit-specific assessments. If site-specific flaw growth data for the flaw indications is available, the NRC staff asked the licensee to identify the limiting site-specific flaw growth value that was calculated for the flaws evaluated in the unit-specific Table 2 assessments.

The licensees response to RAI No. 3 (Reference 2) is summarized as follows:

For ONS Unit 1. The licensee explained all ten (10) of the evaluated flaw indications were a subset of the 39 flaw indications detected in the RPV. The licensee stated that, of the evaluated indications, four of the flaw indications were originally detected during the ISI inspections performed during the 3rd 10-year and six of the flaw indications were originally detected during the ISI inspections performed during the 4th 10-year ISI interval (year 2012 ISI for Unit 1).

For ONS Unit 2. The licensee explained all four (4) of the evaluated flaw indications were a subset of the 24 flaw indications detected in the RPV. The licensee stated that, of the evaluated indications, one of the flaw indications was originally detected during the ISI inspections performed during the 3rd 10-year ISI interval and that the remaining three flaw indications were originally detected during the ISI inspections performed during the 4th 10-year ISI interval (year 2013 ISI for Unit 2).

For ONS Unit 3. The licensee explained all five (5) of the evaluated flaw indications were a subset of the 47 flaw indications detected in the RPV. The licensee stated that all of the evaluated flaw indications were originally detected during the ISI inspections performed during the 4th 10-year ISI Interval (year 2014 ISI for Unit 3).

Table 1: Summary of Past ISI Results (Flaw Indications)

Subset Population of Flaws Criteria of ASME Total Number of Flaws Component Locations Within Inner 1 Inch or Section XI Table Detected in RPV of Flaws 1/10th of RPV Design IWB-3510-1 Met?

Thickness ONS Unit 1 RPV 39 Specified RPV Beltline Yes - for all flaw 10 Total:

Girth Welds, Axial indications 1 in RPV Axial Weld, (All are subsurface Welds or Plates 3 in RPV Girth welds, indications) 6 in RPV plates ONS Unit 2 RPV 24 Specified RPV Beltline Yes - for all flaw 4 Total:

Girth Welds, or Forging indications 4 in RPV forging materials (Two indications are Components surface breaking indications; 22 indications are subsurface indications)

ONS Unit 3 RPV 47 Specified RPV Beltline Yes - for all flaw 5 Total:

Girth Welds, or Forging indications 5 in RPV forging materials (All are subsurface Components indications)

Table 2: Flaw Distribution Results of Unit Specific Flaws Detected in the Inner 1 Inch or 1/10th of the Defined RPV Wall Thickness Through-wall Extent (TWE) Scaled Maximum Distribution Distribution Distribution TWEMIN (inches) TWEMAX (inches) Number of Results of Results of Results of Allowable Flaws ONS Unit 1 Flaws ONS Unit 2 Flaws ONS Unit 3 Flaws for Specified TWE Evaluated Evaluated Evaluated Flaw Range Flaw Indications Detected in Weld Components Within the Inner 1 Inch or 1/10th of the Specified RPV Design Thickness 0 0.075 No limit on 0 0 0 number of flaws 0.075 0.475 316 4 0 0 0.125 0.475 172 3 0 0 0.175 0.475 43 1 0 0 0.225 0.475 16 0 0 0 0.275 0.475 7 0 0 0 0.325 0.475 5 0 0 0 0.375 0.475 2 0 0 0 0.425 0.475 1 0 0 0 0.475 Infinite 0 0 0 0 Flaw Indications Detected in Plates or Forgings Within the Inner 1 Inch or 1/10th of the Specified RPV Design Thickness 0 0.075 No limit on 1 0 0 number of flaws 0.075 0.375 78 5 4 5 0.125 0.375 30 3 4 4 0.175 0.375 8 1 3 0 0.225 0.375 2 1 2 0 0.275 0.375 1 for ONS U1 1 0 0 Plates 1 /

0 for ONS U2 or U3 forgings 0.325 0.375 0 0 0 0 0.375 Infinite 0 0 0 0 The licensee stated there are no crack growth data for any of the unit-specific flaw indications evaluated in the unit-specific flaw distribution assessments because the flaw indications are all fabrication flaws associated with slag inclusions (i.e., the flaws are not associated with crack-like indications that might require a flaw growth assessment if unit-specific flaw growth data for the flaws was available).

Based on the RAI response, the NRC staff confirmed that licensee evaluated the 10 flaws in the ONS Unit 1, four (4) flaws in ONS five (5) flaws in accordance with the flaw evaluation and flaw distribution criteria in WCAP-16168-NP-A and provided the results in the unit-specific flaw distributions tables (i.e., Table 2 in the unit-specific relief requests). The NRC staff compared the flaw distribution assessments in the unit-specific relief requests to the values in WCAP-16168-NP-A and confirmed that: (1) number of flaws in the specified flaw distribution ranges are bounded by the maximum number of flaws allowed for those flaw ranges in WCAP-16168-NP-A for the B&W-design pilot plant (i.e., ONS Unit 1), and (2) no assessed flaw in the weld, plate, or forging component was projected to exceed the maximum allowable flaw size (i.e., 0.475 inch for assessed weld flaws and 0.375 inch for assessed plate or forging flaws) specified and approved in WCAP-16168-NP-A.

Based on the above, the NRC staff finds that the licensee has provided the required past ISI results and criteria for flaws requiring flaw distribution evaluations. The NRC staff finds the licensee has appropriately identified all past flaws that were detected in the Unit 1, 2, and 3 RPVs, and for unit specific-flaws requiring a flaw distribution assessment, that the flaw distributions and flaw sizes are appropriately bounded for the flaw distribution assessments for B&W designed reactors in WCAP-16168-NP-A, including the reports conservative estimates for flaw growth.

Frequency and Severity of Design Basis Transients for Fatigue In Section 3.4 of the safety evaluation dated July 26, 2011, the NRC staff determined that licensees submitting risk-informed ISI extension alternatives for their RPVs must report whether the frequency of the limiting design basis transients during prior plant operation are less than the frequency of the design basis transients identified in PWROG fatigue analysis that are considered to significantly contribute to fatigue crack growth.

In Table 1 of the individual relief alternatives, the licensee stated that the limiting design transients for fatigue flaw growth of detected flaws are the reactor coolant system (RCS) heatup and cooldown transients. The licensee also stated that the number of allowed heatups and cooldowns at ONS Units 1, 2, and 3 is bounded by the 12 cycles/reactor year allowed for plant heatup and cooldown transients of the associated pilot plant assessed in the WCAP-16168-NP-A for B&W designed PWRs (i.e., for ONS Unit 1, as the designated pilot plant). The licensee also identified that design transients 1B (cooldown from 8% to 0% rated power), 2A (power increase from 0% to 15% rated power), and 2B (power reduction from 15% to 0% rated power) are significant contributors for fatigue flaw growth. In the licensees proprietary response to RAI 4, the licensee provided the current number of amassed cycles for the RCS design transients 1B, 2A, and 2B.3 The NRC staff confirmed that, in WCAP-16168-NP-A methodology, the PWROG established that the RCS heatup and cooldown transients are the limiting design transients for RPV fatigue flaw growth in B&W-designed PWR units. The NRC staff also confirmed that the PWROG established 12 cycles/reactor year as the maximum bounding number of heatup and cooldowns that could occur per reactor year for B&W-designed PWRs. The NRC staff notes that, consistent with RPV weld assessment in WCAP-16168-NP-A, the licensees limiting design 3

The specific number of accumulated cycles for design transients 1B, 2A, and 2B in the response to RAI 4 are designated as Framatome Proprietary per 10 CFR 2.390 requirements. However, the NRC staff confirmed the cycle values are significantly less than the design limit allowable for these transients in Table 5-2 of the UFSAR. The 1B cooldown transient, UFSAR Table 5-2 establishes a maximum allowable of 360 cycles for the transient. Similarly, for the power change transients 2A and 2B, UFSAR Table 5-2 sets a maximum allowable of 1440 cycles for the transients.

transients for fatigue flaw growth are the RCS heatup and cooldown transients for ONS Units 1, 2, and 3.

The NRC staff reviewed Table 5-2 of the ONS updated final safety analysis report (UFSAR),

Revision 28 (Reference 16), to verify the licensees fatigue flaw growth basis in the relief request. The NRC staff noted that the UFSAR Table 5-2 establishes a 360-cycle limit for RCS heatups and cooldowns (i.e., Design Transients 1A and 1B in the UFSAR) over the design life of the plant. For a cumulative 60-year licensing term, this correlates to 6 RCS heatup and cooldown cycles per reactor year for ONS Units 1, 2, and 3, and demonstrates licensees yearly frequency value for RCS heatup and cooldown operations is bounded by the 12 cycles/reactor year of RCS heatups and cooldowns allowed for in WCAP-16168-NP-A. The NRC staff also confirmed the number of accumulated cycles for design power changes transients 2A and 2B are all well within the licensees design allowables for these transients in UFSAR Table 5-2.3 Based on the above, the NRC staff finds that the licensees fatigue flaw growth basis is acceptable because: (1) the licensees fatigue flaw growth basis is bounded sufficiently by that analyzed and established for plant heatups and cooldowns (i.e., design transients 1A and 1B in Table 5-2 of the UFSAR for the ONS-specific design basis) of B&W-design PWR units in the WCAP-16168-NP-A report and (2) the licensees fatigue growth assessment accounts appropriately for the fatigue cycle contributions from the plant design power change transients 2A and 2B, which the NRC staff has confirmed are within the scope of the licensees design limits for those transients in Table 5-2 of the UFSAR.

Clad RPV Forgings that are Susceptible to Underclad Cracking - Potential Need for Additional Unit-Specific Evaluations for the Forging Components In Section 3.4 of the safety evaluation dated July 26, 2011, the NRC staff determined that licensees with RPVs containing forgings that are susceptible to underclad cracking and have RTMAX-FO values exceeding 240 °F (i.e., RTMAX-FO > 699.67 ºR) must submit a plant-specific evaluation because the analyses performed in the WCAP-16168-NP-A are not applicable (i.e.,

the scope of analyses in WCAP-16168-NP-A do not cover the evaluation of RPV forging underclad cracks for forgings with high fluence-dependent adjusted reference temperatures

[with high RTMAX-FO values]).

The NRC staff noted that the licensee did not include any additional plant-specific evaluations for the ONS units beyond those that were already included in the relief request submittal to demonstrate ONS unit-specific conformance with the risk-informed principles and criteria in WCAP-16168-NP-A. The NRC staff noted that the licensee reported calculated maximum (limiting) RTMAX-FO values of 526.37 ºR, 565.59 ºR, and 567.26 ºR for ONS Units 1, 2, and 3 at 54 EFPY using the methods of 10 CFR 50.61 (without inclusion of margin terms), which is bounded by the values provided in Framatome Document No. 86-9315280-000 (Reference 17), of the Duke Energy Letter No. RA-20-0328 submittal. The NRC staff also noted that the licensee appropriately addressed RPV underclad cracking in the license renewal application for the Oconee reactor units, and that the NRC staff accepted the licensees underclad cracking time-limited aging analysis (TLAA) and aging management bases for RPV forging components in Section 4.2.4 of NUREG-1723. Based on the above, the NRC staff concludes that the licensee did not need to include any supplemental plant-specific evaluations for the units in the relief request submittal (i.e., beyond the unit-specific flaw distribution, RTMAX-XX, limiting TWCF95-XX, and TWCF95-TOTAL included for the units in the submittal) because: (1) the licensee has provided sufficient demonstration that RPVs do not include any RPV forgings with RTMAX-FO values in excess of 699.67 ºR at 54 EFPY, and (2) the licensee has appropriately managed underclad cracking in the RPV forgings through implementation of the applicable TLAA that was approved in NUREG-1723.

Potential Need for Submittal of Additional Information Required by Section (e) in 10 CFR 50.61a.

In Section 3.4 of the safety evaluation dated July 26, 2011, the NRC staff stated, Licensees seeking second or additional interval extensions shall provide the information and analyses requested in Section (e) of 10 CFR 50.61a.

The NRC staff verified that the licensee does not need to include the information required by Section (e) of 10 CFR 50.61a in the ONS unit-specific relief request submittals because the licensee is requesting to extend one interval, from 10 to 20 years, therefore the information is not necessary.

Material Property and Neutron Fluence Data Inputs and Bases for Calculating Component-Specific T30 , RTMAX-X and TWCF95-XX Input Parameters, Limiting Material-Specific Through-Wall Cracking Frequencies (TWCF95-XX Values), and Unit-Specific 95th Percentile Through-wall Cracking Frequencies (TWCF95-XX Values).

In Section 3.4 and Appendix B of the safety evaluation dated July 26, 2011, the NRC staff determined that the maximum adjusted reference temperatures and 30 ft-lb shifts in adjusted referenced temperature (RTMAX-X and T30, as defined in 10 CFR 50.61a) values may be calculated using the methods documented in the latest version of Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, (Reference 18) or other NRC-approved methodology. However, the NRC staff stated that, if the request uses the NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS),"

methodology to calculate T30, then the request should include the analysis described in paragraph (6) of subsection (f) to the alternate pressurized thermal shock rule 10 CFR 50.61a.

The NRC staff also stated that the analysis should be done for all of the materials in the beltline area with at least three surveillance data points. In PWROGs response to RAI 3(a) (Appendix N of WCAP-16168-NP-A), the NRC staff stated that the calculation of the T30 and RTMAX-X values should include all material property and fluence information related to these parameters.

The NRC staff noted that the licensee included the material property and neutron fluence data, T30, values, and RTMAX-XX values at 54 EFPY for the RPV components within the scope of Table 3 of the unit-specific relief requests. The NRC staff reviewed the licensees component-specific material property and neutron fluence exposures that were used as inputs for the licensees unit-specific T30, RTMAX-XX, Limiting TWCF95-XX, and TWCF95-TOTAL value calculations.

For unirradiated adjusted reference temperature values (i.e., reported as RT0 value in 10 CFR 50.61a or RTNDT(U) in 10 CFR 50.61 and the licensees relief request submittal) used in the component-specific RTMAX-XX (i.e., XX = AW for axial welds, PL for plate components, CW for circumferential weld components, and FO for forgings) calculations, the NRC staff verified that the values were either: (1) consistent with the RTNDT(U) values reported for the components in the licensees past measurement uncertainty recapture license amendment request (MUR LAR) and supplements, (Reference 19), (Reference 20), (Reference 21), (Reference 22) and approved in the MUR license amendment (Reference 23)4, or (2) as reported as new component-specific RTNDT(U) values for the ONS Unit 1, 2, and 3 RPV Dutchman forgings and Lower Shell-to-Dutchman forging circumferential welds (i.e., as newly evaluated extended beltline components for the Table 3 RTMAX-XX assessments in the unit-specific TWCF95-CW 4 These same RTNDT(U) values were also approved in a 2014 License Amendment approving the updated 54 EFPY Pressure Temperature Limit Curves for ONS Units 1, 2, and 3 (as approved in License Amendment No. 384 for ONS Unit 1, License Amendment No. 386 for ONS Unit 2, and License Amendment No. 385 for ONS Unit 3 (Ref. 18, ADAMS Accession No. ML14041A093).

assessments). Based on the above, the NRC staff finds the reported RTNDT(U) inputs to be acceptable because they were either:

Based on NRC staff-approved values in the current licensing basis (CLB) (1), or (2) for the case of the new RTNDT(U) values used in the calculations of the RTMAX-FO and RTMAX-CW for the referenced Dutchman forgings and forging welds, the RPV Dutchman forgings and RPV Dutchman welds did not serve as the sources of the component-specific RTMAX-FO and RTMAX-CW values used in the limiting TWCF95-FO, and TWCF95-CW calculations (i.e., is there were other RPV forgings or circumferential welds that were confirmed to be the limiting forgings and circumferential weld components for the calculations).

The NRC staff reviewed the neutron fluence value inputs for the unit-specific TWCF95-TOTAL calculations. For the neutron fluence exposure and fluence factor (FF) value inputs used to develop the RPV component-specific T30 and 54 EPFY RTMAX-XX and TWCF95-TOTAL values for Units 1, 2, and 3, the licensee explained in its response to RAI 1 that the component-specific fluence and FF values for 54 EPFY were based on values provided in the licensees previous MUR LAR.

In RAI 2, the NRC staff asked for clarification on whether the RPV inlet nozzles, RPV outlet nozzles, and associated nozzle forging-to-nozzle belt forging welds will be above the threshold component-specific RTMAX-XX assessment of 1x1017 n/cm2 (E > 1.0 MeV).

In its response to RAI 2, the licensee stated that the 54 EPFY neutron fluence exposures for some of the unit-specific RPV inlet and outlet nozzle forgings or nozzle forging-to-belt forging welds are projected to be slightly above a threshold of 1x1017 n/cm2 (i.e., ranging from 1.03x1017 n/cm2 to 2.41x1017 n/cm2 [E > 1.0 MeV]) at 54 EFPY. However, the licensee explained that the unit-specific inlet and outlet nozzle forgings or nozzle forging-to-belt forging welds are not the components that yield the limiting RTMAX-XX and TWCF95-XX values for the calculations because those limiting values would be based on the RTMAX-XX and TWCF95-XX values of non-nozzle RPV beltline shell forging, plate, and weld components that are determined to be most limiting in the unit-specific RPV assessments (i.e., due to the fact that they have 54 EPFY projected neutron fluence exposures at least one order of magnitude higher than the specified fluences projected for the inlet/outlet nozzles or nozzle forging-to-nozzle belt forging welds at 54 EPFY).

The NRC staff confirmed that the unit-specific RPV inlet outlet nozzle forgings and nozzle forging-to-belt forging welds do not serve as the limiting RTMAX-XX component inputs for derivation of the unit-specific limiting TWCF95-FO, TWCF95-AW (for Unit 1 only), and TWCF95-CW calculations. Therefore, the NRC staff finds the licensees response provides a sufficient justification for accepting the licensees basis that it does not need to amend the relief request submittal to include RTMAX-XX calculations and TWCF95-XX calculations for the unit-specific RVP inlet and outlet nozzles forgings or RPV inlet/outlet forging-to-RPV belt forging welds.

The NRC staff assessed the validity of the licensees reported 54 EPFY neutron fluence and FF inputs by: (1) performing a spot-check of the 72 EPFY neutron fluence values reported for the components in the licensees MUR LAR for the units, (2) multiplying those values by a linear extrapolation factor of 0.75 (i.e., by a ratio factor of 60-years/80-years in order to derive the NRC staff-calculated 54 EPFY values), and (3) comparing the NRC staffs calculated 54 EPFY fluence values to the corresponding component-specific 54 EPFY fluence values reported in Table 3 of the licensees unit-specific TWCF95-TOTAL assessments. The NRC staff verified that the licensees reported 54 EFPY fluence values were conservative for use in the component-specific RTMAX-XX calculations (i.e., either equal to or slightly higher than the 54 EPFY fluence values derived by the NRC staff using the licensees 72 EPFY MUR LAR values that were accepted in Reference 23). Based on the above, the NRC staff finds the 54 EPFY fluence

values used the component-specific RTMAX-XX assessments and the FF values derived from the neutron fluence values (i.e. using the approved FF methods in 10 CFR 50.61) to be acceptable for implementation.

For reported component-specific copper weight percent (Cu-Wt-%) and nickel weight percent (Ni-Wt-%) values and chemistry factor (CF) values reported in the Unit 1, 2, and 3 specific RTMAX-XX calculations, the NRC staff confirmed that the values were either: (1) consistent with current licensing basis values reported given in Framatome Report No. ANP-3127, Rev. 1 (Reference 24) and approved in the year-2014 pressure-temperature (P-T) limits license amendment for the units (Reference 25), or (2) reported as new Cu-Wt-%, Ni-Wt-% and CF values for the referenced unit-specific RPV Dutchman forgings and forging welds (i.e., as newly evaluated RPV extended beltline components for the unit-specific assessments). Based on the above, the NRC staff finds the reported RTNDT(U) inputs to be acceptable because they were either based on: (1) NRC staff-approved values in the current licensing basis, or (2) for the case of the referenced Dutchman forgings and forging welds, NRC staff-confirmation that the new Cu-Wt-%, Ni-Wt-% and CF values were reported as new extended beltline component values for the referenced Dutchman components, and NRC staff confirmation that the RTMAX-XX values for the Dutchman forgings and referenced Dutchman welds were not used as the RTMAX-XX inputs for limiting TWCF95-FO, TWCF95-AW, or TWCF95-CW values for the units or for the overall 54 EFPY TWCF95-TOTAL values that were calculated for the units.

The NRC staff also confirmed that, with the exception of the weight-% manganese (Mn-Wt-%)

and phosphorous (P-Wt-%) alloying contents of the components, the licensee provided the applicable RPV component-specific material property and neutron fluence inputs that would need to be included in the unit-specific assessments per the property and fluence criteria specified in Appendix N of WCAP-16168-NP-A. However, the NRC staff found the omission of the Mn and P alloying data (i.e., as potential Wt.-% alloying data sources for the chemistry factors used in the RTMAX-XX calculations) to be an acceptable exception, to meeting Criterion 2, based NRC staffs confirmation that: (1) the licensee still relies on the methods of 10 CFR 50.61 for the licensees 60-year PTS assessment in the CLB (i.e. for 55 EFPY), and (2) the methods in 10 CFR 50.61 do not rely in Mn and P alloying percentages as inputs for the chemistry factors and RTPTS values that were required to be calculated in accordance with 10 CFR 50.61. Therefore, the NRC staff finds that the licensee has provided an acceptable basis of omitting P-Wt-% and Mn-Wt-% values, as general basis inputs, for the component-specific RTMAX-XX values used in the licensees unit-specific Limiting TWCF95-XX and TWCF95-TOTAL calculations.

5 Based on the above, the NRC staff finds that the material property inputs and neutron fluence and fluence factor inputs for the unit-specific TWCF95-TOTAL assessments are acceptable for implementation and that the licensee has demonstrated that the TWCF95-TOTAL methodology and data inputs conform with the acceptance criteria in the safety evaluation dated July 26, 2011, with NRC staff-accepted exceptions on: (1) inclusion of 54 EPFY RTMAX-XX calculations for unit-specific RPV inlet/outlet nozzle forgings or inlet/outlet nozzle forging-to-lower nozzle belt forging welds with projected 54 EFPY neutron fluence exposures exceeding a threshold of 1x1017 n/cm2 (E > 1.0 MeV), and (2) use of applicable Mn-Wt-% and P-Wt-% data.

The NRC staff noted that the licensee reported the following limiting TWCF95-XX values (XX =

AW for axial welds, PL for plate materials, CW for circumferential welds, FO for forging materials) in the Table 3 of the individual ONS Unit 1, 2, and 3 probabilistic through-wall cracking assessments to demonstrate that the TWCF95-TOTAL value for the RPV is less than the 5

The only differences being that, if the methods in 10 CFR 50.61 are used and applied as a basis for calculating RTMAX-XX values: (1) the calculations of RTMAX-XX would not need to include a Margin Term assessment in the RT calculations, and (2) in the relief request submittal, the values of RTMAX-XX are reported in ºR instead of ºF.

limiting TWCF95-TOTAL of 4.42x10-7 per reactor year frequency approved in WCAP-16168-NP-A, as summarized in Table 3 below. Based on these unit-specific limiting TWCF95-XX values, the licensee reported 54 EPFY TWCF95-TOTAL values of: (1) 1.40x10-10 per reactor year for the Unit 1 RPV, (2) 2.82x10-11 per reactor year for the Unit 2 RPV, and (3) 2.43x10-12 per reactor year for the Unit 3 RPV.

Table 3: Limiting Unit-Specific TWCF95-XX Results TWCF95-XX Value1 At 54 EFPY (in units of ONS Unit 1 ONS Unit 2 ONS Unit 3 failure events per reactor year)

TWCF95-AW 2.80x10-12 NA 2 NA 2 TWCF95-PL 2.35x10-13 NA 2 NA 2 TWCF95-CW 6.67x10-11 1.35x10-11 9.70x10-13 TWCF95-FO 2.38x10-15 1.05x10-13 1.22x10-13

1. XX = AW for axial welds, PL for plate materials, CW for circumferential welds, FO for forging materials.

2. Column entries are listed as not applicable (NA) because the design of the ONS Unit 2 and 3 RPVs do not include any RPV plate or axial weld components.

The NRC staff performed independent calculations of the T30, values, RTMAX-XX and Limiting TWCF95-XX (XX = CW for axial welds, FO for forging materials) values, and TWCF95-TOTAL values for ONS Unit 2 to provide sufficient confirmation that: (1) the licensee had calculated valid T30, values, RTMAX-XX and Limiting TWCF95-XX values, and 54 EPFY TWCF95-TOTAL values for the ONS Unit 1, 2, and 3 RPVs in accordance with the WCAP-16168-NP-A methods, and (2) the licensees 54 EPFY TWCF95-TOTAL values for the units would meet the PWROGs TWCF95-TOTAL acceptance criterion of 4.42x10-7 component failures per reactor year, as established for B&W-designed PWRs. The NRC staff verified that TWCF95-CW and TWCF95-FO values for the limiting RPV circumferential weld and forging components were at least as conservative as those calculated by the NRC staff for the same limiting RPV components. The NRC staff also verified that the licensees limiting TWCF95-TOTAL was for the ONS Unit 2 RPV was at least as conservative as that calculated by the NRC staff for the RPV.

Based on the above, the NRC staff finds that the licensee had performed valid TWCF95-TOTAL values for the ONS Init 1, 2, and 3 RPVs and that the licensee has provided sufficient demonstration that TWCF95-TOTAL values for the RPVs at 54 EPFY are well within the maximum (upper-bound) TWCF95-TOTAL value of 4.42x10-7 component failures per reactor year established for B&W-designed PWRs in WCAP-16168-NP-A.

The NRC staff also noted that the methodology in WCAP-16168-NP-A conservatively sets the TWCF95-TOTAL equal to the LERF value for the RPV, as initiated by a postulated, limiting PTS event at the plant. Based on independent calculation and verifications, the NRC staff verified that the licensee sets the LERF values for ONS Units 1, 2, and 3 (as induced by changing the ISI frequency from a 10-year inspection basis to a proposed 20-year inspection basis) to LERF values of 1.40x10-10, 2.82x10-11, and 2.43x10-12 per reactor year (respectively), and that these values meet the LERF value criterion of 1x10-7 per reactor year specified in RG 1.174, Rev. 3.

Based on the above the NRC staff finds that:

(1) the licensees TWCF95-XX values for the limiting RPV forging and weld components (and additionally for ONS Unit 1, the limiting RPV axial weld and plate components) were calculated in accordance with the approved methodology in WCAP-16168-NP-A and are, therefore, acceptable for implementation,

(2) the licensees TWCF95-TOTAL values for the ONS Unit 1, 2, and 3 RPVs were calculated in accordance with the methodology in WCAP-16168-NP-A and meet the acceptance criterion of 4.42x10-7 set for TWCF95-TOTAL values of B&W-designed PWRs in WCAP-16168-NP-A (3) the licensees LERF values meet and are bounded by the LERF value of 1x10-7 per reactor year specified in RG 1.174, Rev. 3, and supports authorization of the licensees proposed unit-specific relief alternatives using the risk-informed licensing decision bases established in RG 1.174, Rev.3.

Based on the above, the NRC staff finds the licensee has demonstrated that its proposed alternative provides an acceptable level of quality and safety in lieu of complying with the ASME Code,Section XI, requirements for the inspection items inspection categories and items in IWB-2500 and Table IWB-2500-1 and requirements in IWB-2411 for Examination Category for B-A and B-D specified in the licensees relief request.

CONCLUSION The NRC staff has determined that the proposed alternative in the licensees relief request provides an acceptable level of quality and safety. The NRC staff concludes that the licensee has adequately addressed the regulatory requirements set forth in 10 CFR 50.55a(z)(1).

Based on the above, the NRC staff authorizes the use of proposed alternatives for ONS Units 1, 2, and 3 for the remainder of the Fifth 10-year ISI interval and Sixth 10-year ISI interval for Oconee, Units 1, 2, and 3. All other ASME BPV Code,Section XI, requirements for which alternatives were not specifically requested and authorized remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.

REFERENCES 1 Duke Energy letter to U. S. Nuclear Regulatory Commission (NRC), Oconee Nuclear Station, Units 1, 2 and 3, "Request for Alternative for Implementation of Extended Reactor Vessel Inservice Inspection Intervals for Oconee 1, 2 and 3," dated January 19, 2021 (ADAMS Accession No. ML21019A276).

2 Duke Energy letter to U.S. Nuclear Regulatory Commission (NRC), Oconee Nuclear Station, Units 1, 2 and 3, "Response to Request for Additional Information for Request for Alternative for Implementation of Extended Reactor Vessel Inservice Inspection Intervals for Oconee Units 1, 2 and 3," dated August 5, 2021 (ADAMS Accession No. ML21217A293).

3 Duke Energy - Enclosure 2 of letter to U.S. Nuclear Regulatory Commission (NRC),

Response to NRC Request for Additional Information Related to Duke Energys Proposed Alternative Request No. RA-20-0328, for Deferral of RPV ISI Volumetric Examination Until the 6th 10-year ISI Interval for Oconee 1, 2 and 3., - (PWROG-2003-NP Rev. 1), August 2021 (ADAMS Accession No. ML21217A293 - Proprietary).

4 The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC),Section XI, Division 1, "Welding, Brazing, and Fusing Qualifications: Qualification Standard for Welding, Brazing, and Fusing Procedures; Welders; Brazers; and Welding, Brazing, and Fusing Operators," Paragraph IWB-2411, 2021.

5 The American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code,Section IX,"Welding, Brazing, and Fusing Procedures; Welders; Brazers; and Welding; and Welding, Brazing and Fusing Operators," 2007 Edition, dated July 1, 2007 (5th ISI Interval).

6 The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC),Section XI, Division 1, "Welding, Brazing, and Fusing Qualifications: Qualification

Standard for Welding, Brazing, and Fusing Procedures; Welders; Brazers; and Weld," (6th ISI Interval).

7 U.S. Nuclear Regulatory Commission (NRC), Regulatory Guide (RG) 1.174, Revision 3,"An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," January 2018 (ADAMS Accession No. ML17317A256).

8 U.S. Nuclear Regulatory Commission (NRC) letter to PWR [Pressurised Water Reactors]

Owners Group, Revised Final Safety Evaluation by the Office of Nuclear Reactor Regulation Regarding Pressurized Water Reactor Owners Group Topical Report WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval,"

dated July 26, 2011 (ADAMS Accession No. ML111600295 and ML111600303).

9 Westinghouse Electric Corporation (WEC), WCAP-16168-NP-A, Revision 3,"Risk-Inormed Extension of the Reactor Vessel In-Service Inspection Interval," October 2011 (ADAMS Accession No. ML11306A084).

10 U.S. Nuclear Regulatory Commission (NRC), NUREG-1874,"Recommended Screening Limits for Pressurized Thermal Shock (PTS)," March 2010 (ADAMS Accession No, ML15222A848).

11 Pressurized Water Reactors Owner's Group (PWROG) letter to U.S. Nuclear Regulatory Commission (NRC), "Revision to the Revised Plan for Plant Specific Implementation of Extended Inservice Inspection Interval per WCAP-16168-NP, Revision 1," "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," PA-MSC-0120, dated July 12, 2010 (ADAMS Accession No. ML11153A0330).

12 U.S. Nuclear Regulatory Commission (NRC), Sandia National Laboratories, Science Applications International Corporation, ISL, Inc., and Oak Ridge National Laboratory, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," dated December 14, 2004 (ADAMS Accession No ML042880482).

13 Framatome Technologies, Inc. - B&W Owners Group, Generic License Renewal Program -

Duke Power Company, "BAW-2251A - Demonstration of the Management of Aging Effects for the Reactor Vessel," August 1999 (ADAMS Accession No. ML20212G911).

14 U.S. Nuclear Regulatory Commission (NRC), "NUREG-1723, "Safety Evaluation Report Related to the License Renewal of Oconee Nuclear Station, Units 1, 2 and 3," March 2000 (ADAMS Accession No. ML003695154).

15 Williams, Shawn, U.S. Nuclear Regulatory Commission e-mail to Arthur Zaremba , "Oconee Nuclear Station, Units 1, 2 and 3 - Request for Additional Information Re: Alternative for ISI RPV Weld Examination from 10 to 20 years," dated July 8, 2021 (ADAMS Accession No. ML21190A017).

16 U.S. Nuclear Regulatory Commission (NRC), "Oconee 1, 2 and 3, Revision 28 to Updated Final Safety Analysis Report, Chapter 5, Appendix 5A, Tables," dated June 20, 2020 (ADAMS Accession No. ML20189A081).

17 Duke Energy letter to U.S. Nuclear Regulatory Commission (NRC), Request for Alternative for Implementation of Extended Reactor Vessel Inservice Inspection Intervals for Oconee 1, 2 and 3; Attachment 1, "Framatome Document 86-9315280-000 Equivalent Fatigue Crack Growth for Oconee 1, 2 and 3 Beltline Shell Location," dated January 19, 2021 (Attachment 1 to ADAMS Accession No. ML21019A276).

18 U.S. Nuclear Regulatory Commission, "Regulatory Guide (RG) 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials, ," May 1988 (ADAMS Accession No. ML003740284).

19 Duke Energy letters to U.S. Nuclear Regulatory Commission (NRC), "License Amendment Request for Measurement Uncertainty Recapture Power Uprate (MUR)," dated February 19, 2020 (ADAMS Accession No. ML20050D379).

20 Duke Energy letter to Nuclear Regulatory Commission (NRC), "Supplemental Information for Measurement Uncertainty Recapture Power Uprate (MUR) License Amendment Request (LAR)," dated April 6, 2020 (ADAMS Accession No. ML20097E117).

21 Duke Energy letter to U.S. Nuclear Regulatory Commission (NRC), "Supplemental Information for Measurement Uncertainty Recapture Power Uprate (MUR) License Amendment Request (LAR)," dated July 23, 2020 (ADAMS Accession No. ML20205L403).

22 Duke Energy letter to U.S. Nuclear Regulatory Commission (NRC), "Supplemental Information for Measurement Uncertainty Recapture Power Uprate (MUR) License Amendment Request (LAR)," dated August 17, 2020 (ADAMS Accession No. ML20230A127).

23 U.S. Nuclear Regulatory Commission (NRC) letter to Duke Energy Carolinas, LLC, "Oconee Nuclear Station, Units 1, 2 and 3,- Issuance of Amendment Nos. 420, 422 and 421 Re:

Measurement Uncertainty Recapture Power Uprate," January 26, 2021 (ADAMS Accession No. ML20335A001).

24 AREVA , Topical Report ANP-3127, Revision 1,"Oconee Nuclear Station Units 1, 2 and 3, Pressure-Temperature Limits at 54 EFPY," January 2013 (ADAMS Accession No. ML13058A060).

25 U.S. Nuclear Regulatory Commission (NRC), Oconee Nuclear Station, Units 1, 2 and 3,"Issuance of Amendments Regarding Revised Pressure-Temperature Limits," February 27, 2014 (ADAMS Accession No. ML14041A093).

Principal Contributor: J. Medoff, NRR Date: November 19, 2021 Digitally signed by Michael Michael T. T. Markley Date: 2021.11.19 11:20:04 Markley -05'00' Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation cc: Listserv

ML21281A141 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DNLR/NVIB/BC NAME SWilliams Kay Goldstein JTsao for ABuford DATE 11/4/2021 11/2/2021 10/22/2021 OFFICE NRR/DRA/APLA/BC NRR/DORL/LPL2-1/BC NRR/DORL/LPL2-1/PM NAME RPascarelli MMarkley SWilliams DATE 11/17/2021 11/19/2021 11/19/2021