RA-21-0307, Subsequent License Renewal Application Response to NRC Requests for Confirmation of Information - Set 1

Subsequent License Renewal Application Response to NRC Requests for Confirmation of Information - Set 1

ML21336A001
Person / Time
Site:	Oconee
Issue date:	12/02/2021
From:	Snider S Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA-21-0307
Download: ML21336A001 (31)

Text

Steven M. Snider Vice President Oconee Nuclear Station Duke Energy ON01VP l 7800 Rochester Hwy Seneca, SC 29672 o: 864.873.3478 f: 864.873.5791 RA-21-0307 Steve.Snider @duke-energy.com December 2, 2021 10 CFR 50.4 10 CFR Part 54 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Duke Energy Carolinas, LLC (Duke Energy)

Oconee Nuclear Station (ONS), Units 1, 2, and 3 Docket Numbers 50-269, 50-270, 50-287 Renewed License Numbers DPR-38, DPR-47, DPR-55 Subsequent License Renewal Application Response to NRC Requests for Confirmation of Information - Set 1

References:

1. Duke Energy Letter (RA-21-0132) dated June 7, 2021, Application for Subsequent Renewed Operating Licenses, (ADAMS Accession Number ML21158A193)

2. NRC Letter dated July 22, 2021, Oconee Nuclear Station, Units 1, 2, and 3 - Determination of Acceptability and Sufficiency for Docketing, Proposed Review Schedule, and Opportunity for a Hearing Regarding Duke Energy Carolinas Application for Subsequent License Renewal (ADAMS Accession Number ML21194A245)

3. NRC E-mail dated November 4, 2021, Oconee SLRA - Requests for Confirmation of Information

- Set 1, ADAMS Accession Number ML21313A232 Ladies and Gentlemen:

By letter dated June 7, 2021 (Reference 1), Duke Energy Carolinas, LLC (Duke Energy) submitted an application for the subsequent license renewal of Renewed Facility Operating License Numbers DPR-38, DPR-47, and DPR-55 for the Oconee Nuclear Station (ONS), Units 1, 2, and 3 to the U.S. Nuclear Regulatory Commission (NRC). On July 22, 2021 (Reference 2), the NRC determined that ONS subsequent license renewal application (SLRA) was acceptable and sufficient for docketing. In an email from Angela X. Wu (NRC) to Steve Snider (Duke Energy) dated November 4, 2021 (Reference 3), the NRC transmitted specific requests for confirmation of information (RCl) to support completion of the Safety Review. contains the responses to the RCIs. Enclosure 2 contains updates to sections of the SLRA affected by the responses to the RCIs. For clarity, deletions are indicated by strikethrough and inserted text by underlined red font in Enclosure 2.

Commitment 34 in Table A6.0-1 has been revised as a result of the response to RCI B2.1.34-A. The updated table is provided in Enclosure 2.

U.S. Nuclear Regulatory Commission December 2, 2021 Page 2 Should you have any questions regarding this submittal, please contact Paul Guill at (704) 382-4753 or by email at paul.guill@duke-energy.com.

I declare under penalty of perjury that the foregoing is true and correct. Executed on December 2, 2021.

Sincerely, Steven M. Snider Site Vice President Oconee Nuclear Station

Enclosures:

Responses to Requests for Confirmation of Information : SLRA Updates

U.S. Nuclear Regulatory Commission December 2, 2021 Page 3 CC: W/O

Enclosures:

Laura A. Dudes Regional Administrator U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, Georgia 30303-1257 Angela X. Wu, Project manager (by electronic mail only)

U.S. Nuclear Regulatory Commission Mail Stop 11 G3 11555 Rockville Pike Rockville, Maryland 20852 Shawn A. Williams, Project Manager (by electronic mail only)

U.S. Nuclear Regulatory Commission Mail Stop 8 B1A 11555 Rockville Pike Rockville, Maryland 20852 Jared Nadel (by electronic mail only)

NRC Senior Resident Inspector Oconee Nuclear Station Anuradha Nair-Gimmi, (by electronic mail only: naira@dhec.sc.gov)

Bureau Environmental Health Services Department of Health & Environmental Control 2600 Bull Street Columbia, South Carolina 29201

U.S. Nuclear Regulatory Commission December 2, 2021 Page 4 BCC: W/O

Enclosures:

T.P. Gillespie K. Henderson S.D. Capps T.M. Hamilton P.V. Fisk H.T. Grant D.A. Wilson M.C. Nolan S.M. Snider R.K. Nader G.D. Robison T.M. LeRoy P.F. Guill R.V. Gambrell File: (Corporate)

Electronic Licensing Library (ELL)

ENCLOSURE 1 OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 SUBSEQUENT LICENSE RENEWAL APPLICATION RESPONSES TO REQUESTS FOR CONFIRMATION OF INFORMATION

Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application Responses to Requests for Confirmation of Information Enclosure 1 Subsequent License Renewal Application Responses to Requests for Confirmation of Information Regulatory Basis:

Part 54 of Title 10 of the Code of Federal Regulations (10 CFR), Requirements for Renewal of Operating Licenses for Nuclear Power Plants, is designed to elicit application information that will enable the U.S. Nuclear Regulatory Commission (NRC) staff to perform an adequate safety review and the Commission to make the necessary findings. Reliability of application information is important and advanced by requirements that license applications be submitted in writing under oath or affirmation and that information provided to the NRC by a license renewal applicant or required to be maintained by NRC regulations be complete and accurate in all material respects. Information that must be submitted in writing under oath or affirmation includes the technical information required under 10 CFR 54.21(a) related to assessment of the aging effects on structures, systems, and components subject to an aging management review. Thus, both the general submission requirements for license renewal applications and the specific technical application information requirements require that submission of information material to NRCs safety findings (see 10 CFR 54.29 standards for issuance of a renewed license) be submitted by an applicant as part of the application.

Background:

By letter dated June 7, 2021 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML21158A193), as supplemented by letter dated October 22, 2021 (ADAMS Accession No. ML21295A035), Duke Energy Carolinas, LLC (Duke Energy) submitted to the U.S. Nuclear Regulatory Commission (NRC or staff) an application to renew the Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55 for Oconee Nuclear Station (ONS), Units 1, 2, and 3. Duke Energy submitted the application pursuant to 10 CFR Parts 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants, for subsequent license renewal.

Between July 26 and October 8, 2021, the NRC staff conducted audits of Duke Energys records to confirm information submitted in the ONS subsequent license renewal application.

1

Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application Responses to Requests for Confirmation of Information Request:

During the audit, the staff reviewed several documents that contain information which will likely be used in conclusions documented in the Safety Evaluation Report (SER). To the best of the staff's knowledge, this information is not on the docket. Any information used to reach a conclusion in the SER must be included on the docket by the applicant. We request that you submit confirmation that the information gathered from the documents and listed below is correct or provide the associated corrected information.

2

Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application Responses to Requests for Confirmation of Information Request for Confirmation of Information (RCI) B2.1.3-A:

Based on the review of SLRA Section B2.1.3 and the audit of the applicants breakout question responses posted in the audit portal, the staff needs confirmation of the results of the volumetric examinations in accordance with the ASME Code,Section XI, Table IWB-2500-1 for the 60 reactor closure head studs (Examination Category B-G-1, Item No. B6.20) and the 60 threads-in-flange (Examination Category B-G-1, Item No. B6.40) of each Oconee Nuclear Station (ONS) unit from the last required examinations performed for the unit.

Confirm that there were no relevant indications or issues identified during the last volumetric examinations performed, as required by the ASME Code,Section XI, Table IWB-2500-1, for the 60 reactor closure head studs (Examination Category B-G-1, Item No. B6.20) and the 60 threads-in-flange (Examination Category B-G-1, Item No. B6.40) of each ONS unit.

Response to RCI B2.1.3-A:

This information has been confirmed to be correct as stated.

3

Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application Responses to Requests for Confirmation of Information Request for Confirmation of Information (RCI) B2.1.3-B:

Based on the review of SLRA Section B2.1.3 and the audit of the applicants aging management program basis document SLR-ONS-AMPR-XI.M3, Rev. 1, the staff noted that the maximum RC hardness value for the ONS Unit 1 reactor closure head studs is 36 RC, which is less than the value of 41 RC below which stress corrosion cracking does not present a concern as reported in SLRA Section B2.1.3. The staff noted that the document does not have the hardness value for ONS Unit 2. The staff also noted that some of the ultimate strength data (as reported in SLR-ONS-AMPR-XI.M3, Rev. 1) of the material heat applicable to the ONS Unit 1 reactor closure head studs were greater than the 170 ksi recommended in the GALL-SLR report. Similarly, some of the ultimate strength data of the material heat applicable to the ONS Unit 2 reactor closure head studs were greater than 170 ksi. Since ultimate strength correlates to hardness, and both material heats of the ONS Units 1 and 2 reactor closure head studs have ultimate strength data greater than 170 ksi, the maximum RC value of the ONS Unit 2 reactor closure head studs should also be less than 41 RC below which stress corrosion cracking does not present a concern.

Confirm that the maximum RC hardness value for the ONS Unit 2 reactor closure head studs is less than the value of 41 RC below which stress corrosion cracking does not present a concern.

Response to RCI B2.1.3-B:

The hardness data from the certified material test report (CMTR) for the Oconee Unit 2 reactor closure head studs is reported as Brinell hardness rather than Rockwell hardness. Duke Energy confirms with clarification that the maximum reported Brinell hardness value from the CMTR for the Oconee Unit 2 reactor closure head studs is 363 BHN.

4

Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application Responses to Requests for Confirmation of Information RCI B2.1.34-A:

In the SLRA Section B2.1.34, aging management program (AMP) enhancement 14 states that degradation of piles and sheeting are accepted by engineering evaluation or subject to corrective actions, and AMP operating experience 5 states that the sheet piles at the exterior of the condenser circulating water intake structure are within the scope of the inspections performed under this AMP. The staff found that the sheet piles are included as an AMR line item in Table 3.5.2-23, but the staff could not locate AMR line item for sheeting. SLRA Table 3.5.1, item 3.5.1-062 states that ONS has no wooden piles; sheeting is in the scope of subsequent license renewal.

Confirm that there is no in-scope sheeting at Oconee Nuclear Station, Units 1, 2, and 3.

Response to RCI B2.1.34-A:

This information has been confirmed to be correct as stated. The SLRA is being supplemented as a result of this response and the updated sections are provided in Enclosure 2.

5

Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application Responses to Requests for Confirmation of Information RCI 4.6.3-A:

Based on the review of SLRA Section 4.6.3, the staff noted that footnote 3 of SLRA Table 4.6.3-1 states, These governing transients include seismic loads. Based on audit review of Appendix B of SLR-ONS-TLAA-0300, the staff noted that 5 operating basis earthquake (OBE) events of 9 cycles each and 3 OBE events of 9 cycles each were included in the fatigue usage evaluations for the main steam penetration and main feedwater penetration, respectively.

Confirm that the fatigue usage evaluation load combinations included 5 allowable OBE events of 9 cycles each for the main steam penetration evaluation, and 3 allowable OBE events of 9 cycles each for the main feedwater penetration evaluation.

Response to RCI 4.6.3-A:

Duke Energy confirms that the fatigue usage evaluation load combinations described in SLR-ONS-TLAA-0300 actually include 5 operating basis earthquake (OBE) events of 10 cycles each for the main steam penetration evaluation and 3 OBE events of 10 cycles each for the main feedwater penetration evaluation. One cycle per OBE event for each penetration evaluation is conservatively included with the transient combination that results in the highest stress (Load Pair 1). The other 9 OBE cycles per event self-cycle in the hot condition (Load Pair 6 for the main steam penetration evaluation and Load Pair 4 for main feedwater penetration evaluation).

6

Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application Responses to Requests for Confirmation of Information RCI 2.1.8-A:

During its audit, the staff reviewed FAC [Flow-Accelerated Corrosion] Breakout Questions and Responses 9.13.21, ONS SLRA FAC Breakout Session Followup, and Follow-up Breakout Questions - TRP 17 - FAC (TerryYoder) with ONS Responses Rev 1. The staff noted that the Software Quality Assurance (SQA) process includes validation and verification for all new software and any changes to existing software (Procedures AD-IT-ALL-0002, Software Quality Assurance (SQA)

Program Administration, and AD-EG-ALL-1110, Design Review Requirements), and will continue to be performed for FAC software through the subsequent period of extended operation as part of the SQA process. The staff noted that software error notification is also addressed in the SQA process (Procedure AD-IT-ALL-0002). The staff noted that the software supplier and industry support organizations provide error notification. In addition, the staff noted that FAC software error notification will continue through the subsequent period of extended operation.

Confirm that validation and verification and error notification for FAC software will continue to be performed during the subsequent period of extended operation as part of the SQA process, including error notification from software supplier and industry support organizations.

Response to RCI 2.1.8-A:

This information has been confirmed to be correct as stated.

7

Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application Responses to Requests for Confirmation of Information RCI 2.1.15-A:

Based on the review of Revision 1 of SLR-ONS-AMPR-XI.M26, Fire Protection AMP [Aging Management Program] Evaluation Report, the staff noted that inspection results not meeting acceptance criteria are referred to the Oconee Corrective Action Program where the results are subject to trending.

During its audit, the staff reviewed Breakout Questions - TRP 26 - Fire Protection (Terry Gavula)_Final responses for ePortal 9-16 and noted that inspection results are provided to the Fire Protection Engineer to review for degradation trends. In addition to trending the inspection results not meeting acceptance criteria that are entered into the Corrective Action Program, confirm that the Oconee Fire Protection program also trends inspection results not entered in the Corrective Action Program to provide for timely detection of aging effects.

Response to RCI 2.1.15-A:

This information has been confirmed to be correct as stated.

8

Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application Responses to Requests for Confirmation of Information RCI 2.1.16-A:

Based on the review of Revision 1 of SLR-ONS-AMPR-XI.M27, Fire Water System AMP Evaluation Report, the staff noted that results not meeting acceptance criteria are addressed in the Oconee Corrective Action Program. In addition, the staff noted that Revision 1 of SLR-ONS-AMPR-XI.M27 states that results of flow testing and ultrasonic testing are provided to engineering for evaluation and trending.

During its audit, the staff reviewed Breakout Questions - TRP 27 - Fire Water System (Terry Gavula)_Final Responses for ePortal_9-16 and Follow-up Breakout Questions - TRP 27 - FWS (Terry)_Final Responses for ePortal_9-28, and noted that results of flow testing and ultrasonic testing are provided to engineering for trending outside of the corrective action program whether or not acceptance criteria are met. In addition, the staff noted that unexpected flushing results are entered into the corrective action program for evaluation and trending.

Confirm the following: (1) the Oconee Fire Water System program trends flow test and ultrasonic test results not entered into the Corrective Action Program in addition to trending flow test and ultrasonic test results not meeting acceptance criteria that are entered into the Corrective Action Program, and (2) all unexpected results of flushes are entered into the Corrective Action Program (all flush result trending is performed under the Corrective Action Program).

Response to RCI 2.1.16-A:

This information has been confirmed to be correct as stated. This information was also clarified in supplement 2 to the SLRA dated November 11, 2021.

9

Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application Responses to Requests for Confirmation of Information RCI 3.5.2.2.2.6-A:

SLRA Section 3.5.2.2.2.6, Reactor Vessel [RV] Support Steel Evaluation, discusses the reactor vessel (RV) structural support assembly grout.

Confirm that the RV structural support assembly grout is in scope of SLR and subject to aging management review (AMR).

Response to RCI 3.5.2.2.2.6-A:

This information has been confirmed to be correct as stated. The SLRA is being supplemented as a result of this response and the updated sections are provided in Enclosure 2.

10

Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application Responses to Requests for Confirmation of Information RCI 3.5.2.2.2.6-B:

Based on reviews of SLRA Section 3.5.2.2.1.2, Reduction of Strength and Modulus Due to Elevated Temperature, dated June 7, 2021, UFSAR Section 3.8.1.1, Description of the Containment, dated December 31, 2019, and audit review of OS-160, Specification for Concrete for the Reactor Building, Duke Power Company 1-3, dated April 5, 1973, the ONS concrete is made with crushed marble aggregate obtained from Blacksburg, South Carolina The design strengths are 5000 psi at 28 days Confirm that the reactor vessel support pedestal and primary shield wall (PSW) concretes have the composition and design strength as described in SLRA Section 3.5.2.2.1.2, UFSAR Section 3.8.1.1, and OS-160.

Response to RCI 3.5.2.2.2.6-B:

This information has been confirmed to be correct as stated.

11

Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application Responses to Requests for Confirmation of Information RCI 3.5.2.2.2.6-C:

Based on the audit review of Oconee DWG No. O-69A, Reactor Building Primary & Secondary Shield E[W]alls Plan at El. 777+6 & 802+0 Concrete, Revision 14, and photos provided on the ePortal in response to breakout questions, the staff noted that the PSW liner at the reactor air cavity is made of corrugated steel and anchored to the PSW with 286 3/8X6 Nelson Studs.

Confirm that the current condition of the PSW liner is anchored to the concrete with 286 3/8X6 Nelson Studs and is not degraded.

Response to RCI 3.5.2.2.2.6-C:

Duke Energy confirms with clarification that the corrugated metal attached to the interior of the primary shield wall (PSW) is anchored to the concrete with 366 3/8X6 Nelson Studs for Unit 1 and 248 3/8X6 Nelson Studs for Units 2 and 3 and is not degraded.

12

Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application Responses to Requests for Confirmation of Information RCI 3.5.2.2.2.6-D:

The audit review of ASME Section XI, Subsection IWF InService inspection (ISI) results for the Fourth 10-year Inspection Interval of the RV support skirt and anchorage, indicates that although the noted ISIs addressed the same inspections areas in all three ONS Units in similar environments, the examination results differed. The disparate results were due to changes made in the examination coverage requirements of ONS NDE-91, NDE Procedures Manual, Volume 1, Reporting Coverage During PreService and InService Inspection, Revision 7, dated 2011, during the course of the ISIs, based on an NRC approved relief.

Confirm that Oconees ASME Section XI, Subsection IWF RV support skirt and anchorage disparate ISI results were solely due to changes made in its NDE-91, Revision 7, inspection procedure acceptance criteria consistent with NRC approved relief for extent of inspection coverage.

Response to RCI 3.5.2.2.2.6-D:

This information has been confirmed to be correct as stated.

13

Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application Responses to Requests for Confirmation of Information RCI 3.5.2.2.2.6-E:

SLRA Section 3.5.2.2.2.6, Reactor Vessel Support Skirt and Reactor Cavity Configuration states [t]he reactor cavity serves as a biological shield wall.

Confirm that:

a. The reactor cavity concrete serves as a biological shield wall (aka PSW) and not the cavity itself.

b. There is no other material within the reactor cavity air gap that would be considered as part of the biological shield wall.

Response to RCI 3.5.2.2.2.6-E:

This information has been confirmed to be correct as stated.

14

Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application Responses to Requests for Confirmation of Information RCI 3.5.2.2.2.6-F:

SLRA Section 3.5.2.2.2.6, Reactor Vessel Support Skirt and Reactor Cavity Configuration states, The reactor consists of reinforced concrete surrounding the reactor vessel...

Confirm that there is a typographical error and the referenced statement should read The reactor cavity consists of reinforced concrete surrounding the reactor vessel...

Response to RCI 3.5.2.2.2.6-F:

This information has been confirmed to be correct as stated.

15

Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application Responses to Requests for Confirmation of Information RCI 3.5.2.2.2.6-G:

Based on the review of SLRA Section 3.5.2.2.2.6 regarding fluence and gamma dose and the audit review of the applicants breakout presentation titled FER Item 3.5.2.2.2.6 Meeting - 9/20/2021, 11:00 am posted in the ePortal, the staff needs confirmation regarding the location of the 5.53E-04 dpa exposure level calculated in ANP-3898NP, Revision 0, in the RPV skirt assembly. The breakout presentation stated that the 5.53E-04 dpa exposure level was calculated at approximately 1.75 inches above the top of the transition forging-to-skirt weld. Section 9.4.4 of ANP-3898NP, Revision 0, states:

[The] projected 72 EFPY dpa is obtained by calculating dpa at the bottom of a RG 1.190 compliant RPV DORT model, which is 17.49 inches above the transition forging-to-RPV skirt weldand extrapolating dpa from the bottom of the DORT model to the RPV skirt to transition forging weld.

Confirm that the elevation of 1.75 inches above the top of the transition forging-to-skirt weld and the elevation to which the dpa from the bottom of the discrete ordinate transport (DORT) model was extrapolated are the same locations.

Response to RCI 3.5.2.2.2.6-G:

ANP-3898NP, Revision 0, Section 9.4.4.1, Page 9-20, states the following:

The projected 72 EFPY dpa of the ONS Unit 1, ONS Unit 2, and ONS Unit 3 RPV support skirt weld and embedment anchor bolts is conservatively estimated at 5.53E-04 dpa. This projected 72 EFPY dpa is obtained by calculating dpa at the bottom of a RG 1.190 compliant RPV DORT model (BAW-2241P-A, Revision 2, Reference 9-5), which is approximately 17.49 inches above the transition forging to RPV skirt weld, and then by generating RPV skirt weld flux and dpa rate profiles on the outer RPV surface in the air cavity region and extrapolating dpa from the bottom of the DORT model to the RPV skirt to transition forging weld. The dpa of 5.53E-04 is conservatively assumed to be applicable to all the RPV support assembly and embedment materials (i.e., skirt, flange, anchor bolts, nuts, washers, and shear pins).

16

Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application Responses to Requests for Confirmation of Information The 17.49 inch elevation stated above is the distance from the bottom of the DORT model to the top of the transition forging-to-RPV skirt weld. Duke Energy clarifies that the actual location of the extrapolated exposure at 72 EFPY, reported as 5.53E-04 dpa, is approximately 1.75 inches above the top of the weld that connects the transition forging to the RPV skirt.

17

Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application Responses to Requests for Confirmation of Information RCI 3.5.2.2.2.6-H:

Based on the review of Section 9.4.4.3 of ANP-3898NP, Revision 0, and the audit review of the applicants breakout question responses posted in the ePortal, the staff noted that there were no plant-specific measured values of initial nil-ductility temperature (NDT) for the ONS Units 1, 2, and 3 RPV skirt assembly components or Charpy V-Notch absorbed energy values from which initial NDT can be derived for the components.

Confirm that there were no plant-specific measured values of initial NDT for the ONS Units 1, 2, and 3 RPV skirt assembly components or Charpy V-Notch absorbed energy values from which initial NDT can be derived for the components, and that therefore generic values from NUREG-1509 and BAW-10046A, Revision 2 were used.

Response to RCI 3.5.2.2.2.6-H:

This information has been confirmed to be correct as stated.

18

Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application Responses to Requests for Confirmation of Information RCI 3.5.2.2.2.6-I:

Based on the review of Section 9.4.4.3 of ANP-3898NP, Revision 0, and the audit review of the applicants breakout question responses posted in the ePortal, the staff noted that the sources of the initial NDT values and corresponding margins of the RPV skirt assembly components were NUREG-1509 and topical report BAW-10046A.

Confirm that the initial NDT values and corresponding margins of the RPV skirt assembly components were obtained from NUREG 1509, Radiation Effects on Reactor Pressure Vessel Supports, 1996 and BAW-10046A, Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50, Appendix G, Revision 2.

Response to RCI 3.5.2.2.2.6-I:

This information has been confirmed to be correct as stated.

19

Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application Responses to Requests for Confirmation of Information RCI 3.5.2.2.2.6-J:

Based on the review of Section 9.4.1 of ANP-3898NP, Revision 0, and the audit review of document OISI-0169.10-0050 BASIS DOC, Revision 0, and the applicants breakout question responses posted in the ePortal, the staff noted that the transition forging-to-skirt welds (WR36) of the ONS Units 1, 2, and 3 RPV skirt assembly fall under the examination requirements of Division 1, Subsection IWF of the ASME Code,Section XI instead of Subsection IWB.

Confirm that the transition forging-to-skirt welds (WR36) of the ONS Units 1, 2, and 3 RPV skirt assembly fall under the examination requirements of Division 1, Subsection IWF of the ASME Code,Section XI instead of Subsection IWB.

Response to RCI 3.5.2.2.2.6-J:

This information has been confirmed to be correct as stated.

20

Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application Responses to Requests for Confirmation of Information RCI 3.5.2.2.2.6-K:

Based on the review of Section 9.4.3 of ANP-3898NP, Revision 0, and the audit review of the applicants breakout question responses posted in the ePortal with excerpts of proprietary Framatome Document No. 32-9311203-000, Oconee Reactor Vessel Skirt Lowest Service Temperature, the staff noted that air circulation (stagnant or otherwise) through the 9.25-inch ventilation holes in the RPV skirt could affect the lowest service temperature value of 139.05°F determined in Section 9.4.3 of ANP-3898NP, especially in the region of the RPV skirt below the 9.25-inch ventilation holes.

Confirm that air circulation through the 9.25-inch ventilation holes in the ONS 1, 2, and 3 RPV skirt was considered in determining the lowest service temperature value of 139.05°F and that there is no impact of this air circulation on the value.

Response to RCI 3.5.2.2.2.6-K:

This information has been confirmed to be correct as stated.

21

ENCLOSURE 2 OCONEE NUCLEAR STATION SUBSEQUENT LICENSE RENEWAL APPLICATION SLRA UPDATES

Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application SLRA Updates SLRA Updates RCI B2.1.34-A Updates:

Section A2.34, Page A-38 is revised as follows:

A2.34 Inspection of Water-Control Structures Associated with Nuclear Power Plants Program Description The Inspection of Water-Control Structures Associated with Nuclear Power Plants AMP is an existing condition monitoring program that consists of inspection and surveillance of the raw water control structures associated with emergency cooling systems, emergency power production or flood control.

The scope of the program also includes structural steel, and structural bolting associated with water control structures, steel or wood piles and sheeting required for the stability of embankments and channel slopes, and miscellaneous steel, such as sluice gates and trash racks.

Section A2.34, Page A-41 is revised as follows:

A2.34 Inspection of Water-Control Structures Associated with Nuclear Power Plants Enhancements

14. Degradation of sheet piles and sheeting are is accepted by engineering evaluation or subject to corrective actions.

Section B2.1.34, Page B-238 is revised as follows:

B2.1.34 INSPECTION OF WATER-CONTROL STRUCTURES ASSOCIATED WITH NUCLEAR POWER PLANTS Enhancements

14. Degradation of sheet piles and sheeting are is accepted by engineering evaluation or subject to corrective actions. (Element 6) 1

Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application SLRA Updates RCI B2.1.34-A Updates:

Table A6.0-1, page A-109 is revised as follows:

Table A6.0-1: Subsequent License Renewal Commitments

1. Program Commitment AMP Implementation 34 Inspection of Water Control 12. Provide evaluation criteria for structural concrete using B2.1.34 Program enhancements for Structures Associated With quantitative second tier criteria of Chapter 5 in ACI SLR will be implemented no Nuclear Power Plants 349.3R. Base the acceptance criteria for concrete later than 6 months prior to the program surfaces on the second-tier evaluation criteria provided SPEO.

in Chapter 5 of ACI 349.3R.

13. Clarify that loose bolts and nuts are not acceptable unless accepted by engineering evaluation.

14. Degradation of sheet piles and sheeting are is accepted by engineering evaluation or subject to corrective actions.

15. Revise inspection procedures to base inspection acceptance criteria on quantitative requirements derived from industry codes and standards, including but not limited to ACI 349.3R, ACI 318, SEI/ASCE 11, or the relevant AISC specifications and consider industry and plant OE. Use justified quantitative acceptance criteria whenever applicable.

2

Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application SLRA Updates RCI 3.5.2.2.2.6-A Updates:

Table 3.5.1, page 3-1339 is revised as follows:

Table 3.5.1 Summary of Aging Management Programs for Containments, Structures and Component Supports Evaluated in Chapters II and III of the GALL-SLR Report Further Item Aging Effect/ Aging Management Component Evaluation Discussion Number Mechanism Program Recommended 3.5.1-055 Building concrete at locations of Reduction in concrete AMP XI.S6, Structures No Consistent with NUREG- 2191.

expansion and grouted anchors; anchor capacity due to Monitoring ONS also utilizes the ASME grout pads for support base plates local concrete Section XI, Subsection IWF degradation/service- AMP for grout pads for induced cracking or other support base plates on concrete aging Class 1, 2, and 3 component mechanisms supports.

3.5.1-056 Concrete: exterior above- and Loss of material due to AMP XI.S7, Inspection of No Consistent with NUREG- 2191.

below- grade; foundation; interior abrasion; cavitation Water-Control Structures slab Associated with Nuclear Power Plants or the FERC/ US Army Corp of Engineers dam inspections and maintenance programs.

3.5.1-057 Constant and variable load spring Loss of mechanical AMP XI.S3, ASME Section No Consistent with NUREG- 2191.

hangers; guides; stops function due to corrosion, XI, Subsection IWF distortion, dirt or debris accumulation, overload, wear 3.5.1-058 Earthen water-control structures: Loss of material; loss of AMP XI.S7, Inspection of No Consistent with NUREG- 2191.

dams; embankments; reservoirs; form due to erosion, Water-Control Structures channels; canals and ponds settlement, Associated with Nuclear Power sedimentation, frost Plants or the FERC/ US Army action, waves, currents, Corp of Engineers dam surface runoff, seepage inspections and maintenance programs.

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Oconee Nuclear Station, Units 1, 2, and 3 Subsequent License Renewal Application SLRA Updates RCI 3.5.2.2.2.6-A Updates:

Table 3.5.2-23, page 3-1454 is revised as follows:

Table 3.5.2-23 Containments, Structures, and Component Supports - Miscellaneous Structural Commodities - Aging Management Evaluation Component Intended Material Environment Aging Effect Aging Management NUREG-2191 NUREG- Notes Type Function Program Item 2192 Table 1 Fire Barriers - FB Elastomer, Air (External) Hardening, loss of Fire Protection (B2.1.15) VII.G.A-19 3.3.1- 057 A Penetration Seals Rubber and strength, or shrinkage Other Similar Materials Grout Air (External) Cracking, Loss of Fire Protection (B2.1.15) VII.G.A-90 3.3.1- 060 A Material Structures Monitoring VII.G.A-90 3.3.1- 060 A (B2.1.33)

Grout SS Grout Air - Indoor Reduction in Concrete ASME Section XI, III.B1.1.TP-42 3.5.1-055 E Uncontrolled (External) Anchor Capacity Subsection IWF (B2.1.30)

III.B1.2.TP-42 3.5.1-055 E Structures Monitoring III.B1.1.TP-42 3.5.1- 055 A (B2.1.33)

III.B1.2.TP-42 3.5.1- 055 A III.B2.TP-42 3.5.1- 055 A III.B3.TP-42 3.5.1- 055 A III.B4.TP-42 3.5.1- 055 A III.B5.TP-42 3.5.1- 055 A Air - Outdoor (External) Reduction in Concrete Structures Monitoring III.B1.2.TP-42 3.5.1- 055 A Anchor Capacity (B2.1.33)

III.B2.TP-42 3.5.1- 055 A III.B3.TP-42 3.5.1- 055 A III.B4.TP-42 3.5.1- 055 A III.B5.TP-42 3.5.1- 055 A 4