ML19164A333

Final Request for Additional Information for the Safety Review of the Surry Power Station, Units 1 and 2 Subsequent License Renewal Application Set 2 June 11, 2019

ML19164A333
Person / Time
Site:	Surry
Issue date:	06/11/2019
From:	Sayoc E NRC/NRR/DMLR/MRPB
To:	Stoddard D Dominion Energy Co, Virginia Electric & Power Co (VEPCO)
Sayoc E, NRR-DMLR 415-4084
References
L-2018-RNW-0023/000951
Download: ML19164A333 (42)

Text

From: Sayoc, Emmanuel To: "Daniel.g.stoddard@dominionenergy.com" Cc: Paul Aitken; "Eric A Blocher"; Wu, Angela; Oesterle, Eric; Tony Banks

Subject:

FINAL REQUESTS FOR ADDITIONAL INFORMATION FOR THE SAFETY REVIEW OF THE SURRY POWER STATION, UNITS 1 AND 2 SUBSEQUENT LICENSE RENEWAL APPLICATION (L-2018-RNW-0023/000951) - SET 2 Date: Tuesday, June 11, 2019 10:15:34 PM Attachments: Attachment 1 - Surry SLRA Final RAI Summary Index - Set 2 v1.pdf Attachment 2 - Surry SLRA Final RAIs Package Set 2.pdf Importance: High Docket No. 50-280 and 50-281

Dear Mr. Stoddard,

By letter dated October 15, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18291A842), as supplemented by letters dated January 29, 2019 (ADAMS Accession No. ML19042A137), and April 2, 2019 (ADAMS Accession No. ML19095A666), Virginia Electric and Power Company (Dominion Energy Virginia or Dominion) submitted to the U.S. Nuclear Regulatory Commission (NRC or staff) an application to renew the Renewed Facility Operating License Nos. DPR-32 and DPR-37 for the Surry Power Station, Unit Nos. 1 and 2. Dominion submitted the application pursuant to Title 10 of the Code of Federal Regulations Part 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants, for subsequent license renewal.

From April 3, 2019 through June 11, 2019, the U.S Nuclear Regulatory Commission (NRC) staff sent Dominion the draft Requests for Additional Information (RAIs) for various technical review packages (TRP). Dominion subsequently informed the NRC staff that clarification calls were needed to discuss the information requested. Between May 29, 2019 through June 11, 2019, clarification calls were completed for all the draft RAIs unless Dominion declined having a call. The specific dates of the draft RAI transmittals and the RAIs clarification calls are summarized in Attachment 1. The final RAIs resulting from these calls are enclosed in Attachment 2.

Paul Aitken of your staff agreed to provide a response to these RAIs within 37 days of the date of this email. The NRC staff will be placing a copy of this email and attachments in the NRCs ADAMS.

Sincerely, Emmanuel Sayoc, Project Manager License Renewal Projects Branch (MRPB)

Division of Materials and License Renewal Office of Nuclear Reactor Regulation Docket No. 50-280 and 50-281 Attachments:

As stated OFFICE PM:MRPB:DMLR BC: MRPB:DMLR PM: MRPB:DMLR

NAME ESayoc EOesterle ESayoc DATE 05/29/2019 06/6/2019 06/11/2019 OFFICIAL RECORD COPY

Surry SLRA RAI Set 2 Index Date - Draft Date -

RAI Sent To Clarification Clarification Call Attendees - Clarification Call Attendees -

Item No RAI Set TRP RAI Number Issue Branch Reviewer Applicant Call Applicant NRC Issue Date Emmanuel Sayoc, Angela Wu, Pat Purtscher, Eric Blocher, Ganesh Cheruvenki, Thermal Aging Charles Tomes, Richard Eagan, Seung Min, Embrittlement of Cast Thomas Zalewski, Anees David Dijamco, Austenitic Stainless Steel Udyawar, Alexandria Carolan John Tsao 1 2 12 B.2.1.6-1 (CASS) MPHB Cheruvenki 5/15/2019 5/30/2019 6/11/2019 Emmanuel Sayoc, Angela Wu, Pat Purtscher, Eric Blocher, Ganesh Cheruvenki, Thermal Aging Charles Tomes, Richard Eagan, Seung Min, Embrittlement of Cast Thomas Zalewski, Anees David Dijamco, Austenitic Stainless Steel Udyawar, Alexandria Carolan John Tsao 2 2 12 B.2.1.6-2 (CASS) MPHB Cheruvenki 5/15/2019 5/30/2019 6/11/2019 Paul Aitken, Eric Blocher, Keith Miller, Ron Burner, William Holston, James Mark Pellegrino, Gavula, Brian Allik, Emmanuel 3 2 15 B.2.1.28-2 Internal Coatings MCCB Gavila, Allik 6/6/2019 6/10/2019 Scott Bray, Sayoc 6/11/2019 Eric Blocher, Tom Snow, Bryan McCarter, James Zaborowski 4 2 17 B.2.1.8-1 Flow-Accelerated Corrosion MCCB Chereskin 5/22/2019 5/30/2019 Angela Wu, Alex Chereskin 6/11/2019 Eric Blocher, Tom Snow, Bryan McCarter, James Zaborowski 5 2 17 B.2.1.8-2 Flow-Accelerated Corrosion MCCB Chereskin 5/22/2019 5/30/2019 Angela Wu, Alex Chereskin 6/11/2019 Eric Blocher, Tom Snow, Bryan McCarter, James Zaborowski 6 2 17 B.2.1.8-3 Flow-Accelerated Corrosion MCCB Chereskin 5/22/2019 5/30/2019 Angela Wu, Alex Chereskin 6/11/2019 Emmanuel Sayoc, Angela Wu, Chereskin, Eric Blocher, John Thomas Alexander Chereskin 7 2 19 B.2.1.10-1 Steam Generator MCCB Huynh 5/15/2019 5/30/2019 6/11/2019 Emmanuel Sayoc, Angela Wu, Chereskin, Eric Blocher, John Thomas Alexander Chereskin 8 2 19 B.2.1.10-2 Steam Generator MCCB Huynh 5/15/2019 5/30/2019 6/11/2019 Emmanuel Sayoc, Jim Bavila, Andrew Jihnson Generic Letter 89-13 Gavula, Paul Aitken, Eric Blocher, Ron Alexander Chereskin 9 2 20 B.2.1.11-1 Commitments MCCB Johnson 6/11/2019 6/11/2019 Berner 6/11/2019 Emmanuel Sayoc, Jim Bavila, Andrew Jihnson AMR Items for Open-Cycle Gavula, Paul Aitken, Eric Blocher, Ron Alexander Chereskin 10 2 20 B.2.1.11-2 Cooling Water System MCCB Johnson 6/11/2019 6/11/2019 Berner 6/11/2019 Eric Blocher, John Thomas, Pratt William Holston, Alan Huynh, Holston, Cherry Angela Wu 11 2 26 B.2.1.15-1 Fire Protection MCCB Huynh 5/13/2019 5/29/2019 6/11/2019 Eric Blocher, John Thomas, Pratt William Holston, Alan Huynh, Holston, Cherry Angela Wu 12 2 26 B.2.1.15-2 Fire Protection MCCB Huynh 5/13/2019 5/29/2019 6/11/2019 Holston, William Holston, Emmanuel 13 2 29 B.2.1.17-1 Atmospheric Metallic Tanks MCCB Huynh 6/4/2019 6/6/2019 Mark Pelegrino, Eric Blocher Sayoc 6/11/2019 Paul Aitken, Eric Blocher, Keith Andrew Johnson, Jim Gavula, Johnson, Miller, Diane Aitken Emmanuel Sayoc 14 2 36 B.2.1.23-1 External Surfaces MCCB Gavula 4/3/2019 6/3/2019 6/11/2019 Paul Aitken, Eric Blocher, John Angela Wu, Juan Lopez, Brian Structures Monitoring Disosway, Jim Johnson Wittick 15 1 46 B.2.1.34-1 Program ESEB Lopez 4/30/2019 5/9/2019 6/11/2019 Eric Blocher, Keith Miller, James Johnson, Charles Tomes, Danila Madden, Mat Yoder, Alex Cherskin, Brian Derreberry, Andrew Prinaris, George Thomas, Jim Hester, Thomas, Mark Yoo, Irradiation Effects on CBS Prinaris, Greg Imrogno (Westinghouse) Emmanuel Sayoc, Scott 16 2 76 3.5.2.2.2.6-1 and RV Steel Support ESEB Buford 6/4/2019 6/6/2019 Krepel 6/11/2019

Eric Blocher, Keith Miller, James Johnson, Charles Tomes, Danila Madden, Mat Yoder, Alex Cherskin, Brian Derreberry, Andrew Prinaris, George Thomas, Jim Hester, Thomas, Mark Yoo, Irradiation Effects on CBS Prinaris, Greg Imrogno (Westinghouse) Emmanuel Sayoc, Scott 17 2 76 3.5.2.2.2.6-2 and RV Steel Support ESEB Buford 6/4/2019 6/6/2019 Krepel 6/11/2019 Eric Blocher, Keith Miller, James Johnson, Charles Tomes, Danila Madden, Mat Yoder, Alex Cherskin, Brian Derreberry, Andrew Prinaris, George Thomas, Jim Hester, Thomas, Mark Yoo, Irradiation Effects on CBS Prinaris, Greg Imrogno (Westinghouse) Emmanuel Sayoc, Scott 18 2 76 3.5.2.2.2.6-3 and RV Steel Support ESEB Buford 6/4/2019 6/6/2019 Krepel 6/11/2019 Eric Blocher, Keith Miller, James Johnson, Charles Tomes, Danila Madden, Mat Yoder, Alex Cherskin, Brian Derreberry, Andrew Prinaris, George Thomas, Jim Hester, Thomas, Mark Yoo, Irradiation Effects on CBS Prinaris, Greg Imrogno (Westinghouse) Emmanuel Sayoc, Scott 19 2 76 3.5.2.2.2.6-4 and RV Steel Support ESEB Buford 6/4/2019 6/6/2019 Krepel 6/11/2019 Eric Blocher, Keith Miller, James Johnson, Charles Tomes, Danila Madden, Mat Yoder, Alex Cherskin, Brian Derreberry, Andrew Prinaris, George Thomas, Jim Hester, Thomas, Mark Yoo, Irradiation Effects on CBS Prinaris, Greg Imrogno (Westinghouse) Emmanuel Sayoc, Scott 20 2 76 3.5.2.2.2.6-5 and RV Steel Support ESEB Buford 6/4/2019 6/6/2019 Krepel 6/11/2019 Eric Blocher, Keith Miller, James Johnson, Charles Tomes, Danila Madden, Mat Yoder, Alex Cherskin, Brian Derreberry, Andrew Prinaris, George Thomas, Jim Hester, Thomas, Mark Yoo, Irradiation Effects on CBS Prinaris, Greg Imrogno (Westinghouse) Emmanuel Sayoc, Scott 21 2 76 3.5.2.2.2.6-7 and RV Steel Support ESEB Buford 6/4/2019 6/6/2019 Krepel 6/11/2019 Eric Blocher, Keith Miller, James Johnson, Charles Tomes, Danila Madden, Mat Yoder, Alex Cherskin, Brian Derreberry, Andrew Prinaris, George Thomas, Jim Hester, Thomas, Mark Yoo, Irradiation Effects on CBS Prinaris, Greg Imrogno (Westinghouse) Emmanuel Sayoc, Scott 22 2 76 3.5.2.2.2.6-8 and RV Steel Support ESEB Buford 6/4/2019 6/6/2019 Krepel 6/11/2019 Eric Blocher, Keith Miller, James Johnson, Charles Tomes, Danila Madden, Mat Yoder, Alex Cherskin, Brian Derreberry, Andrew Prinaris, George Thomas, Jim Hester, Thomas, Mark Yoo, Irradiation Effects on CBS Prinaris, Greg Imrogno (Westinghouse) Emmanuel Sayoc, Scott 23 2 76 3.5.2.2.2.6-9 and RV Steel Support ESEB Buford 6/4/2019 6/6/2019 Krepel 6/11/2019 Eric Blocher, Keith Miller, James Johnson, Charles Tomes, Danila Madden, Mat Yoder, Alex Cherskin, Brian Derreberry, Andrew Prinaris, George Thomas, Jim Hester, Thomas, Mark Yoo, Irradiation Effects on CBS Prinaris, Greg Imrogno (Westinghouse) Emmanuel Sayoc, Scott 24 2 76 3.5.2.2.2.6-10 and RV Steel Support ESEB Buford 6/4/2019 6/6/2019 Krepel 6/11/2019 25 2 147.3 4.7.3-7 Leak-Before-Break MPHB Min, Tsao 6/7/2019 6/10/2019 No Call Per Dominion No Call Per Dominion 6/11/2019 TLAA - , Cracking Associated with Weld Meeting Cancelled by 26 2 147.7 4.7.7-1 Deposited Cladding MVIB Medoff, Yoo 5/22/2019 6/3/2019 Meeting Cancelled by Dominion Dominion 6/11/2019 PWR Vessel Internals, And the Gap Analysis for Reactor Internals Provided in SLRA Appendix C Angela Wu, Jim Medoff 27 2 16A B.2.1.7-2 MVIB Medoff, Yoo 5/22/2019 6/4/2019 Eric Blocher, Tom Snow 6/11/2019 PWR Vessel Internals, And the Gap Analysis for Reactor Internals Provided in SLRA Appendix C Angela Wu, Jim Medoff 28 2 16A B.2.1.7-3 MVIB Medoff, Yoo 5/22/2019 6/4/2019 Eric Blocher, Tom Snow 6/11/2019 Scoping and Screening of 29 2 SS 2.3.1.3 Pressurizer Spray Head SRXB Peng 6/5/2019 6/6/2019 No Call Per Dominion No Call Per Dominion 6/11/2019

SURRY POWER STATION, UNITS 1 AND 2 Subsequent License Renewal Application (SLRA)

Request for Additional Information (Set 2)

Regulatory Basis:

10 CFR 54.21(a)(3) requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. One of the findings that the staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis. In order to complete its review and enable making a finding under 10 CFR 54.29(a), the staff requires additional information in regard to the matters described below.

TRP 12: Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS)

RAI- B.2.1.6-1

Background:

Surry SLRA AMP B2.1.6, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS) program consists of the determination of the susceptible piping and piping components in the reactor coolant pressure boundaries with respect to thermal aging embrittlement based on the casting method and chemical composition of the CASS materials. The aging management of the susceptible piping and piping components is accomplished through a component-specific flaw tolerance evaluation in accordance with ASME Code,Section XI. As part of the aging management program, the applicant submitted the following documents addressing the flaw tolerance evaluation in the CASS materials in the reactor coolant piping and piping components at the Surry, Units 1 and 2. The documents submitted in the portal are: (1) WCAP-18258, Flaw Evaluation for Susceptible Reactor Coolant Loop Cast Austenitic Stainless Steel (CASS),

(2) In-house audit response-NRC Audit for SPSS SLR Information for TRP 12 CASS 3 4 19 Tomes.

Issue:

In Item 1 (7) of the report in the portal, In-house audit response-NRC Audit for SPSS SLR Information for TRP 12 CASS, 3 4 19 Tomes, the applicant stated that a postulated fatigue crack is located in the weld region at the ends of the elbow. The staff noted that there are locations within the elbow (such as the intrados, extrados, and cheek locations) that could have higher stresses than the ends of the elbow. The staff also noted that CASS pipes and elbows have a higher delta ferrite content and a lower strength than the weld metal. If the fatigue crack were to occur, it is more likely to occur in the lower strength region near the CASS base metal adjacent to the weld, but not in the weld.

Request:

Based on the issues the staff identified above, stresses could be higher in other locations within the elbow and these locations could have a higher delta ferrite content (and thus subject to a greater degree of thermal embrittlement than the locations the applicant selected for 1

evaluation). The staff requests that the applicant justify the selection of the weld region at the ends of the elbows as the bounding locations for evaluation.

RAI- B.2.1.6-2

Background:

CASS with greater than 20% ferrite is subject to a greater degree of thermal embrittlement and thus lower fracture toughness. The staff noted that the applicant is applying the limit load methodology modified with Z-factors. The staff requests that the applicant take into account the following items (i) and (ii), while addressing the applicability of the limit load methodology for CASS with greater than 20% ferrite:

Issue:

1. On page 70, in Chapter 4 of NUREG/CR-4513, Revision 2, item (c) states that For CASS materials, adequate toughness for the pipe to reach limit load after aging shall be demonstrated. The staff requests that the applicant demonstrate that after aging of CASS with greater than 20% ferrite will have adequate toughness such that limit load methodology is applicable. Confirm that the Z factor used for the limit load analysis will be conservative compared with a full elastic-plastic analysis.

2. The flowchart for evaluating austenitic piping in Figure C-4210-1 of Appendix C of Section XI of the ASME Code indicates that the evaluation criteria for CASS with delta ferrite content greater than 20% is in the course of preparation. Furthermore, the acceptance criteria (Element 6) in XI.M12 of the GALL-SLR Report states that evaluation of CASS piping containing delta ferrite greater than 20% must be approved by the NRC staff on a case-by-case basis. The staff noted that the applicant applied the Z-factor methodology for CASS piping with delta ferrite greater 20% in C-6000 of Appendix C of Section XI of the ASME Code, even though C-6000 can only be applied to wrought austenitic steels and CASS with less than 20% ferrite (per Figure C-4210-1).

Request:

1. Provide justification for the value of the Z factor in the limit load methodology and how that relates to the lower bound fracture toughness value in CASS piping/elbows at Surry Units 1 and 2.

2. The staff will be using the following documents to make a safety determination for the subject AMP. Therefore, the staff requests that the applicant submit these documents officially. The documents are: (1) WCAP-18258, Flaw Evaluation for Susceptible Reactor Coolant Loop Cast Austenitic Stainless Steel (CASS), (2) In-house audit response-NRC Audit for SPSS SLR Information for TRP 12 CASS 3 4 19 Tomes.

2

TRP 15: Internal Coatings / Linings RAI- B.2.1.28-2

Background:

GALL-SLR AMP XI.M42, Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks, provides recommendations, in part, for managing the aging effects of the underlying metallic pressure boundary material due to the loss of coating integrity.

SLRA Section B2.1.11, Open-Cycle Cooling Water System, states that the Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks program in Section B2.1.28 will manage the aging effects of internal surface coatings including those of metallic surfaces coated with Carbon Fiber Reinforced Polymer [CFRP] that is used as a pressure boundary. SLRA Section B2.1.28 states that after enhancements, the program will be consistent with GALL-SLR AMP XI.M42.

Regarding the CFRP lining, relief request (ADAMS Accession No. ML16355A347 (proprietary))

associated with installation of the CFRP repair includes a reference to an American Society of Mechanical Engineers (ASME) Code Case Repair of Class 2 and 3 Piping by Carbon Fiber Reinforced Polymer Composite, and notes that it was in development. The relief request also discusses the project team associated with the CFRP system application and identifies multiple team members who were active members on the ASME Task Group developing the Code Case for Repair of Class 2 and 3 Piping by CFRP Composite. The NRCs associated Safety Evaluation (ADAMS Accession No. ML17303A037 (proprietary)) clarifies that although, at that time, there were no available standards for CFRP repair of pipe, ASME Code Case N-871, Repair of Buried Class 2 and 3 Piping Using Carbon Fiber Reinforced Polymer Composite, was under development. The staff notes that according to the NRCs above cited safety evaluation, the CFRP piping will be inspected over its service life in accordance with station procedures in compliance with Generic Letter (GL) 89-13, Service Water System Problems Affecting Safety-Related Components, to ensure the condition of the piping system is suitable for continued service.

Issue:

SLRA Section B2.1.28, which credits the use of GALL-SLR AMP XI.M42 to manage the effects of aging for CFRP material that functions as the pressure boundary, appears to be beyond the conditions and operating experience of those for which the GALL-SLR AMP XI.M42 was evaluated. The staff notes that, since the submittal of the relief request discussed above, the ASME code committees have approved Code Case N-871. If the requested relief for installation of the CFRP at Surry had occurred today, then the staff would consider the specific inservice inspection (ISI) requirements given in ASME Code Case N-871 as providing adequate actions for managing the effects of aging of CFRP during the subsequent period of extended operation.

However, an alternate industry consensus document, other that ASME Code Case N-871, could be considered if appropriate technical bases are provided.

3

In addition, based on the loads for which the CFRP system was designed, portions of the 30-inch and 36-inch piping encased in concrete appear to be credited for continuing to provide anchorage to portions of the piping routed above ground. Consequently, information regarding the following staff observations is needed for the staff to complete its review:

1. The acceptance criteria specified in the Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks program do not appear to be consistent with the acceptance criteria specified in Code Case N-871 for similar post-installation indications identified in the CFRP lining.

2. The corrective actions specified in the Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks program, including potential alternative actions which allow return-to-service, do not appear to be consistent with the corrective actions specified in Code Case N-871 for similar post-installation defects identified in the CFRP lining.

3. The periodic visual inspections of the CFRP, conducted either through the Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks program or Generic Letter 89-13, do not appear to be consistent with the ISI visual examinations specified in Code Case N-871, regarding the type, extent, and frequency.

4. The training and qualification for individuals involved in coating/lining inspections specified in the Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks program do not appear to be consistent with the corresponding training and qualification requirements given in Code Case N-871, for personnel performing visual examinations and acoustic tap examinations.

5. The optional adhesion testing discussed in the Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks program does not appear to be consistent with the Mandatory Appendix VI Acoustic Tap Examination specified in Code Case N-871, Mandatory Appendix V, Inservice Inspection, for the accessible surfaces of the CFRP at each terminal end.

6. The relief request for the CFRP repairs states that the design objective of the CFRP system is to provide the necessary strength to carry all design loads even if the host steel pipes continue to degrade. However, piping anchor loads from the attached 30-inch and 36-inch piping do not appear to have been included in the CFRP system design. Consequently, some portions of the 30-inch and 36-inch piping encased in concrete appear to be credited for continuing to provide anchorage to portions of the piping routed above ground, during the period of extended operation. Crediting portions of the piping encased in concrete as providing structural support does not appear to be consistent with the design objective of the CFRP system. In addition, existing aging management activities do not address how the continued degradation of the piping encased in concrete, which is being credited as an anchor, will ensure the structural capacity of the host steel piping will be maintained during the subsequent period of extended operation.

Request:

1. Provide the technical bases for applying the acceptance criteria, regarding the acceptability of blistering, cracking, and flaking, specified in the Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks program that do 4

not appear to be comparable to the acceptance criteria specified in Code Case N-871 for similar degradation (i.e., blistering, cracking and flaking).

2. Provide the technical bases for applying the corrective actions, regarding return to service of coatings with indications of peeling and delamination, specified in the Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks program that do not appear to be comparable to the corrective actions specified in Code Case N-871 for similar degradation.

3. Provide information to show that the periodic visual inspections of the CFRP will be performed to comparable standards as the visual inspections specified in Code Case N-871, Mandatory Appendix V, Inservice Examination for visual inspections.

4. Provide information to show that personnel performing visual inspections or other inspections specified in the Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks program will be qualified equivalent to the provisions of Code Case N-871, Subarticle 5400, Qualification of Examination and QC Inspection Personnel.

5. Provide technical bases to show that the minimum bond length at the terminal ends of the CFRP does not need to be periodically verified to ensure it remains bonded to the steel substrate equivalent to that specified in Code Case N-871, Mandatory Appendix VI, Acoustic Tap Examination.

6. For the portions of 30-inch and 36-inch pipes encased in concrete that are credited to function as anchors for the piping routed above ground, provide information to show that aging effects of the piping will be managed to ensure that the continued degradation of the piping has not caused the structural capacity of the host pipe to be exceeded.

If the anchor points for the 30-inch and 36-inch piping routed above ground do not credit structural integrity of the piping encased in concrete, then provide the anchor loads induced by the piping routed above ground and show that the minimum bond lengths at the terminal ends are adequate to transfer these loads into the CFRP system.

5

TRP 16A: PWR Vessel Internals, and the Gap Analysis for Reactor Internals Provided in SLRA Appendix C RAI B.2.1.7-2 (Clarifications for Programmatic Enhancement No. 7)

Background:

The program in SLRA AMP B.2.1.7, PWR Vessel Internals, includes programmatic Enhancement No. 7. In this enhancement, the applicant states that procedures will be revised to address expansion criteria when degradation occurs for clusters of baffle-former bolts. The enhancement also includes the following additional statement:

MRP 2018-002 identifies expansion criteria as a Needed requirement (per NEI 03-08) to include one-time visual (VT-3) examination of barrel-former bolts if large clusters of baffle-former bolts are found during the initial volumetric (UT) examination. 1 Additional Expansion criteria for performing ultrasonic test (UT) inspections of barrel-former bolts are given in Table 5-3 of EPRI Report No. 3002005349, Revision 1 (MRP-227, Rev. 1).

Issue:

The staff understands that the program currently references two different sources for the acceptance criteria that will be applied to potential contingency inspections of the barrel-former bolts. As a result, the application does not clearly identify whether MRP-2018-002, MRP-227 Revision 1, or some other industry report will be used to establish the acceptance criteria to assess the inspection needs for the barrel-former bolts.

Request:

Clarify whether the acceptance criteria for initiating and performing potential Expansion-based inspections of the barrel-former bolts will be based on: (a) MRP-227, Revision 1, (b) MRP-2018-002, (c) the combination of the two reports (i.e., MRP-227, Revision 1, for UT inspections of the bolts and MRP-2018-002 for initiating VT-3 visual inspections of the bolts), or (d) an alternative report that provides the basis for inspecting the barrel-former bolts. If it is an alternate report, identify the source (report reference) that will be used to provide the acceptance criteria for initiating Expansion-based inspections of the barrel-former bolts, and clarify whether the reports methodology has been endorsed for use by the NRC or provide an appropriate justification for its use.

1 The staff acknowledges that the term clusters in the enhancement is referring to a cluster of degraded bolts, as defined in NSAL 16-1, Revision 1 or in MRP-2017-009.

6

RAI B.2.1.7-3 (Minimum Inspection Coverages for Core Barrel Assembly Expansion-Category Welds Referenced in Enhancement Nos. 9 and 16)

Background:

The program in SLRA AMP B.2.1.7, PWR Vessel Internals, includes programmatic Enhancement Nos. 9 and 16. Collectively, in these enhancements, the applicant states that the minimum EVT-1 visual inspection coverages for the core barrel assembly lower flange welds (LFWs), upper axial welds (UAWs), middle axial welds (MAWs), lower axial welds (LAWs), and upper girth welds (UGWs) is a minimum of 50% of the weld surface.

Issue:

It is not clear to the NRC staff that the proposed minimum inspection coverage of 50% is consistent with MRP 2018-026 which specifies: a minimum coverage of 75% of the weld length on the surface being examined shall be achieved; however, for welds with limited access (Note 4), a minimum examination coverage of 50% of the weld length on the surface being examined shall be achieved.

Requests:

1. For the Surry-specific RVI designs, clarify whether the MAWs and LAWs are restricted by the presence of a thermal shield, thermal panels, or other components located near the welds).

2. Provide the basis for applying a minimum EVT-1 coverage criterion of 50% for potential Expansion-based EVT-1 visual inspections that may be performed on the core barrel assembly UGWs, LFWs, and UAWs. If applicable, identify any components near the UGWs, LFWs, and UAWs that may: (a) restrict access to the components, and (b) limit the ability of Dominion to achieve a minimum 75% coverage criterion for the EVT-1 based contingency inspections of these weld components, as established in MRP 2018-026.

7

TRP 17: Flow-Accelerated Corrosion RAI- B.2.1.8-1

Background:

In SLRA, Section B2.1.8, Flow-Accelerated Corrosion, the applicant claimed consistency with the GALL-SLR Report AMP XI.M17, Flow-Accelerated Corrosion. SLRA Section B2.1.8 states that the erosion activity implements the recommendation of EPRI 3002005530, Recommendations for an Effective Program Against Erosive Attack. The parameters monitored or inspected, detection of aging effects, and monitoring and trending program elements for GALL-SLR Report AMP XI.M17 discuss recommendations to monitor, detect, and trend degradation due to erosion mechanisms (e.g. cavitation, flashing, etc.).

During the In-Office audit, the staff reviewed the program basis document ETE-SLR-2018-1311, Surry Subsequent License Renewal Project - Aging Management Program Evaluation Report

- Flow-Accelerated Corrosion, Revision 1, to evaluate whether the applicant is consistent with the GALL-SLR Report AMP XI.M17 recommendations for the flow-accelerated corrosion (FAC) program. In the document, the applicant stated that the FAC erosion module in CHECWORKS will be used to assist in the development of the inspection plan for the Erosion Control program.

Issue:

The staff has not previously reviewed EPRI 3002005530. Neither the Surry SLRA nor the applicants procedures provide information that describes how the FAC erosion module within the CHECWORKS software will be used to model erosion, and how these results will be used in planning erosion inspections.

Request:

Provide a justification for how the FAC erosion module will meet the recommendations of the GALL-SLR with respect to monitoring effects of wall thinning due to erosive mechanisms (including methods to calculate wear rate), its use in planning inspections for erosive degradation, as well as for monitoring and trending potential degradation due to erosive mechanisms. Additionally, describe how the guidance in EPRI 3002005530 incorporates the use of the FAC erosion module into the Surry erosion program for the program elements described above.

RAI- B.2.1.8-2

Background:

In SLRA, Section B2.1.8, Flow-Accelerated Corrosion [FAC], the applicant claimed consistency with the GALL-SLR Report for the AMP XI.M17, Flow-Accelerated Corrosion. The GALL-SLR Report detection of aging effects program element, states that guidance for 8

inspection scope expansions, when unexpected or inconsistent results are identified in the initial sample scope, is described in the Electric Power Research Institute document NSAC-202L, Revision 4.

Guidance in NSAC-202L, Section 4.4.6 Expanded Sample Inspection states that the reasons for any unexpected or inconsistent inspection results should be investigated by performing an updated FAC predictive analysis, conducting additional inspections, and making material determinations as appropriate. In addition, expanded sample inspections should include any component within two diameters of the affected component and a minimum of the next two most susceptible components from the relative wear ranking in the same train as that containing the piping component displaying significant wear.

During the In-Office audit, the staff reviewed the program basis document ETE-SLR-2018-1311, Surry Subsequent License Renewal Project - Aging Management Program Evaluation Report

- Flow-Accelerated Corrosion, and procedure ER-AA-FAC-102, Flow-Accelerated Corrosion Inspection and Evaluation Activities, to evaluate whether the applicant is consistent with the GALL-SLR Report recommendations for the Flow-Accelerated Corrosion AMP. For the detection of aging effects program element, Section 3.4.2 of the program basis document states that an evaluation is performed to determine the extent of expansion and cites ER-AA-FAC-102, Section 3.9.4. In addition, Section 2.1 of the program basis document states that evaluations documenting various activities including sample expansion are independently reviewed by a qualified FAC engineer. Procedure ER-AA-FAC-102, Section 3.9.4, Inspection Scope Expansion, includes inspection expansion to components upstream and downstream of the degraded component but does not specify any distance. The procedure includes a review of any CHECWORKS model but does not include further discussion regarding the performance of an updated FAC analysis or include, as a minimum, the next two most susceptible components.

To evaluate prior scope expansion documentation, the staff reviewed operating experience associated with the FAC Program outage summary documents ETE-CME-2017-0013, Surry Unit 2, 2017 Refueling Outage, Results of the Flow-Accelerated Corrosion Program, and ETE-CME-2019-0002, Surry Unit 1, 2018, Refueling Outage, Results of the Flow-Accelerated Corrosion Program, which provided examples of where ultrasonic thickness testing has detected unacceptable or inconsistent wall thickness values. The staff also reviewed condition report CR1096902 Significant Wear Observed During FAC Inspection (5-SGS-11-151), to determine the extent of the scope expansion performed by the applicant when unexpected degradation is found as a result of inspections.

Issue:

It is unclear that the requirements of procedure ER-AA-FAC-102, Section 3.9.4 are consistent with the guidance in NSAC-202L, Section 4.4.6, regarding inspection scope expansion. The implementing procedure does not address consideration of performing an updated FAC predictive analysis or making material determinations. In addition, the distance for inspecting upstream and downstream is not discussed and the inclusion of a minimum of the next two most susceptible components from the relative ranking in the same train is not included. In addition, it is not clear that the FAC procedure includes an independent review of sample expansion documentation by a qualified FAC engineer as stated in SLRA Section B2.1.8.

9

The staff notes that its review of operating experience document listed above did not provide information on how far upstream and downstream piping was inspected during a scope expansion, nor did they provide detail on whether the next two most susceptible components in the CHECWORKS model were inspected for potential FAC-related degradation.

Request:

Provide information regarding scope expansion activities to show that either the Surry FAC program implementation includes the guidance in NSAC-202L, Section 4.4.6, or provide bases to show that aging will be effectively managed without being consistent with the guidance in NSAC-202L, Section 4.4.6. Also, provide information regarding the implementation of independent reviews of evaluations documenting sample expansions by qualified FAC engineers, as stated in SLRA Section B2.1.8.

RAI- B.2.1.8-3

Background:

As supplemented by letter dated April 2, 2019, SLRA Table 3.3.2-6 Bearing Cooling, was modified to address the potential for erosion in valve bodies constructed of several different materials. The supplement also states that cavitation in this system could be caused by valve throttling. Additionally, condition report CR1031398, BC Valve - Indication of Cavitation, describes cavitation in a Unit 1 bearing cooling valve and notes that the valve was previously replaced in 2013 due to a pin hole leak in the valve body. This CR also notes that the current non-destructive examination strategy doesnt evaluate the valve body for wall thinning. The staff notes that condition report CR1026621, 2-BC-505 Has a Through-Wall Leak, describes a through-wall leak for the corresponding Unit 2 valve; however, the cause of the leak was not included in the summary documentation.

The applicants erosion susceptibility evaluation (ESE) (ETE-CME-2018-1002, Revision 1, Transmittal of True North Consulting Technical Report BP-2017-0045-TR-01, Erosion Susceptibility Evaluation - Surry, September 2018) designated the bearing cooling system as not being susceptible to cavitation because the cavitation index is greater than 2.5. The ESE states that the bearing cooling system is a closed-loop system which does not have large enough pressure drops for cavitation to occur. The staff notes that comments for other systems in the ESE identify the potential for cavitation and flashing downstream of throttle valves and orifices. The ESE indicates that the criteria for the cavitation index greater than 2.5 is a rule of thumb and cites a reference to a valve manufacturer publication. The associated implementing procedure, ER-AA-FAC-105, Erosion Control Program, Section 3.1.1 states that the ESE is to be periodically updated based on relevant operating experience.

10

Issue:

Although operating experience indicates that valves in the bearing cooling system are susceptible to wall thinning due to cavitation, the ESEs for both units identify the bearing cooling system as not being susceptible to erosive mechanisms, including cavitation. The staff notes that the exclusion criteria for the cavitation index and infrequent operation parameters cited in the ESE are inconsistent with the corresponding criteria provided in the NRC-approved EPRI 112657, Risk Informed Inservice Inspection Evaluation Procedure. Consequently, it is not clear to the staff that there are adequate bases for the exclusion criteria parameters used in the ESE.

Request:

Provide information regarding the bases for the ESE exclusion criteria. Include a discussion about the determination that the bearing cooling system is not susceptible to erosion mechanisms with a specific explanation for why operating experience does not appear to support the ESE determination. Also provide information regarding whether other systems determined to be not susceptible to erosion mechanisms could be similarly affected. Include a discussion regarding how operating experience has been considered by the current ESE.

11

TRP 19: Steam Generators RAI- B.2.1.10-1

Background:

In its SLRA, Section 3.1.2.2.11(1), Cracking due to Primary Water Stress Corrosion Cracking, the applicant stated that the Electric Power Research Insitute (EPRI) Report 3002002850, Steam Generator Management Program: Investigation of Crack Initiation and Propagation in the Steam Generator Channel Head Assembly, dated October 2014, was applicable and bounds the steam generator (SG) divider plates at Surry because Surry has the most limiting SGs analyzed in the Report, namely Alloy 600 Model 51 SGs.

SRP-SLR Section 3.1.2.2.11, Cracking due to Primary Water Stress Corrosion Cracking, recommends actions to manage aging of divider plate assemblies depending on the material of the divider plate assemblies and whether industry analyses (i.e. the EPRI Report) are bounding for the applicants unit(s).

Because the Surry SGs were fabricated with Alloy 600 divider plates, the following recommendations from SRP-SLR are potentially applicable:

1. For units with divider plate assemblies fabricated of Alloy 600 or Alloy 600 type weld materials, if the analyses performed by the industry (EPRI [Electric Power Research Institute] 3002002850) are applicable and bounding for the unit, a plant-specific AMP is not necessary.

2. For units with divider plate assemblies fabricated of Alloy 600 or Alloy 600 type weld materials, if the industry analyses (EPRI 3002002850) are not bounding for the applicant's unit, a plant-specific AMP is necessary or a rationale is necessary for why such a program is not needed. A plant-specific AMP (one beyond the primary water chemistry and the steam generator programs) may include a onetime inspection that is capable of detecting cracking to verify the effectiveness of the water chemistry and steam generator programs and the absence of PWSCC in the divider plate assemblies.

Issue:

SLRA Section 3.1.2.2.11(1) stated that the EPRI analysis is applicable and bounding for the Surry SGs because the divider plates and associated welds are fabricated from Alloy 600 materials, and because Surry has Model 51 SGs which are determined to be the most limiting SG model in the EPRI analysis. The staff recognizes that EPRI Report 3002002850 analyzed the Westinghouse Model 51 SGs as the most limiting SG model; however, due to parameters such as manufacturing tolerances and plant-specific transients/loading, plant-specific parameters may need to be verified in order to demonstrate that EPRI Report 3002002850 is applicable and bounding to the Surry SG divider plates.

Request:

Provide the justification, and supporting evaluation, that demonstrates the Surry SG divider plate assemblies are bounded by industry analyses. Include a discussion of potentially plant-12

specific parameters discussed in EPRI Report 3002002850 (e.g., SG geometry, materials of components, cracking scenarios, plant-specific transient loads and cycles).

RAI- B.2.1.10-2

Background:

SLRA Section B2.1.10 states that the Steam Generators program is consistent with GALL-SLR Report AMP XI.M19, Steam Generators without exceptions and enhancements. As amended by letter dated April 2, 2019, SLRA Table 3.1.2-4 was modified to remove items managing cracking for steel with stainless steel cladding channel heads (and cladding), and loss of material for steel with stainless steel cladding primary inlet nozzle and outlet nozzles (and cladding).

The SRP-SLR Section 3.1, Aging Management of Reactor Vessel, Internals, and Reactor Coolant System, addresses the AMRs associated with certain steam generator components.

This section includes the components discussed above, as well as the recommended AMPs to manage aging effects associated with these components.

Issue:

Table 3.1.2-4, as amended by letter dated April 2, 2019, no longer cites programs to manage cracking for the steel with stainless steel cladding channel head (and cladding), and only cites the Water Chemistry program to manage loss of material for the steel with stainless steel cladding primary inlet nozzle and outlet nozzle (and cladding).

Amended Table 3.1.2-4 no longer includes cracking as an aging effect requiring management for the SG channel head (and cladding). GALL-SLR identifies cracking as an applicable aging effect for steel with stainless steel cladding. For example, GALL-SLR Item RP-232 identifies cracking of steel with stainless steel cladding exposed to reactor coolant as an applicable aging effect to be managed using AMP XI.M1, ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD and AMP XI.M2, Water Chemistry..

Amended Table 3.1.2-4 now references Table 1 Item 3.1.1-088 to manage loss of material for steel with stainless steel cladding primary inlet and outlet nozzles (and cladding) exposed to reactor coolant using the Water Chemistry program. However, GALL-SLR Item R-436 recommends using AMP XI.M2, Water Chemistry and AMP XI.M19, Steam Generators to manage loss of material for steel with stainless steel cladding exposed to reactor coolant.

Request:

1. Explain which program(s) will be used to manage cracking in steel with stainless steel cladding channel heads (and cladding) or state the basis for why a program is not necessary.

2. Are other programs besides the Water Chemistry program used to manage loss of material in steel with stainless steel cladding primary inlet nozzle and outlet nozzles (and cladding)? If not, explain how the Water Chemistry program alone with manage the loss 13

of material without an inspection program (such as the Steam Generator program) to verify effectiveness of the Water Chemistry program.

14

TRP 20 Open-Cycle Cooling Water System RAI- B.2.1.11-1 Generic Letter 89-13 Commitments

Background:

GALL-SLR AMP XI.M20, Open-Cycle Cooling Water System, states that the inspection scope, methods, and frequencies are in accordance with the applicants docketed response to Generic Letter (GL) 89-13, Service Water System Problems Affecting Safety-Related Components.

SLRA Section B2.1.11, Open-Cycle Cooling Water System, states that the program is an existing program that, following enhancement, will be consistent with the GALL-SLR AMP XI.M20. SLRA Section B2.1.11 also states that periodic heat transfer testing, visual inspection, and cleaning of safety-related heat exchangers is performed in accordance with the site commitments to GL 89-13.

ETE-SLR-2018-1314, Aging Management Program Evaluation Report - Open-Cycle Cooling Water System, Revision 2, documents and evaluates the activities in the associated AMP that are credited for managing aging as part of Surrys SLRA. ETE-SLR-2018-1314 discusses a discrepancy between Surrys response to GL 89-13 (letter dated October 2, 1991 (89-572G))

and the maintenance strategy implementation for the charging pump lube oil coolers. The maintenance strategy changed from periodic replacement of charging pump lube oil coolers to performing routine inspection and maintenance. ETE-SLR-2018-1314 states that the discrepancy was evaluated in accordance with the commitment change evaluation process and cites corrective action CA3022000 Submit Commitment Change Paperwork to Update Requirements for Charging Pump LO [Lube Oil] Coolers (March 9, 2017). The staff noted that the change in maintenance strategy affected the scope of Surrys Open-Cycle Cooling Water System program, because components that are periodically replaced are excluded from the scope of an aging management review for license renewal.

In response to staff questions for CA3022000, Surry posted condition report CR1091365, (March 6, 2018) A Commitment Change Evaluation Was Completed and Approved Mistakenly.

The condition report states that the commitment change evaluation was for a change made to a response to the NRC, not a commitment to the NRC. The actions discussed in the condition report included a clarification regarding the difference between a response and a commitment to the NRC.

During its review of ER-SU-5314, Generic Letter 89-13 Program, Revision 2, the staff noted that Attachment 1, Generic Letter 89-13 Components and Commitments, includes a table for each set of components in the program and includes a column labeled Commitment Source.

Every set of components in the list includes Letter to NRC 10/2/91 Serial Number 89-572G as the source of the commitment to perform the specified GL 89-13 activity. However, the tables initial note states that the letter dated April 30, 1991 (Serial 89-572E), summarized the GL 89-13 program and that the response was updated by letter dated October 2, 1991 (Serial 89-572G),

which supersedes Serial Number 89-572E. No new commitments were made.

15

The staff notes that the letter dated April 30, 1991 (Serial 89-572E), states that a detailed revision of [Surrys] initial January 29, 1990, response incorporating the subsequent supplements and the additions integrated into this summary description will be separately forwarded. In addition, the staff notes that none of the GL 89-13 response letters appear to specifically identify the sites activities for periodic heat transfer testing, visual inspection, and cleaning of safety-related heat exchangers as being commitments.

Issue:

Because none of the sites GL 89-13 response letters appear to specifically identify the commitments for periodic heat transfer testing, visual inspection, and cleaning of safety-related heat exchangers, the staff was unable to verify that the Open-Cycle Cooling Water System program would be performed in accordance with the sites commitments to GL 89-13. The program documentation appears to cite the letter dated October 2, 1991 (89-572G), as the source of the sites GL 89-13 commitments. However, the recent condition report (CR1091365) states that because the information in the October 2, 1991, letter was only a response to GL 89-13 and not a commitment, there was no need to perform a commitment change evaluation for changing the approach discussed in the October 2, 1991, letter. Based on the position discussed in CR1091365, the staff is unsure of the sites GL 89-13 commitments.

Request:

Provide additional information to clarify the sites GL 89-13 commitments. Include information about which prior GL 89-13 response letter(s) to the NRC contain(s) the commitments that are discussed in SLRA Section B2.1.11. If the source of Surrys commitments to GL 89-13 are not from the response dated October 2, 1991 (89-572G), also include information regarding the circumstances about why ER-SU-5314, Generic Letter 89-13 Program, Revision 2, cites the letter dated October 2, 1991 (89-572G).

RAI B2.1.11-2 AMR Items for Open-Cycle Cooling Water System

Background:

SLRA Section B2.1.11, Open-Cycle Cooling Water System, states that periodic heat transfer testing, visual inspection, and cleaning of heat exchangers are performed in accordance with the site commitments to GL 89-13 to verify heat transfer capabilities. SLRA Section B2.1.11 also includes an enhancement to the monitoring and trending program element to revise procedures to require trending the inspection results of the emergency service water pump engine heat exchangers.

The staff notes that ER-SU-5314, Generic Letter 89-13 Program, Revision 2, includes the emergency service water pump engine heat exchanger and specifies associated activities for 16

periodic heat transfer testing, as well as inspection and maintenance. In addition, ER-SU-5314 includes the emergency service water pump angle drive and specifies that heat transfer is checked during monthly surveillance testing, and that cooling water flow is verified during inspection and maintenance activities.

Although SLRA Table 3.3.2-4 Service Water - Aging Management Evaluation, includes other emergency service water pump components, it does not appear to include the emergency service water pump engine heat exchanger or the emergency service water pump angle drive.

Issue:

Although SLRA Section B2.1.11 includes an enhancement to trend inspection results associated with emergency service water pump engine heat exchangers, the SLRA does not appear to include a corresponding aging management review item(s). In addition, although Surrys GL 89-13 program appears to specify activities to address heat transfer for the emergency service water pump angle drive, the SLRA does not appear to include a corresponding aging management review item.

Request:

For the emergency service water pump engine heat exchangers and the emergency service water pump angle drives, provide information showing that assessment of the heat transfer capabilities of safety-related heat exchangers (with a heat transfer intended function) will be performed by the SLRA Section B2.1.11, Open-Cycle Cooling Water System program, in accordance with site commitments to GL 89-13. Include information showing either 1) that existing aging management review items with corresponding aging effects are included in the SLRA for these components or 2) that aging management review items are not needed for these components, to demonstrate that the effects of 17

TRP 26: Fire Protection RAI- B.2.1.15-1 BackgroundSLRA Section B2.1.15 states that the Fire Protection program is consistent with GALL-SLR Report AMP XI.M26, Fire Protection, with no exceptions or enhancements. GALL-SLR Report AMP XI.M26, Fire Protection states that the Fire Protection program manages the effects of loss of material and cracking for fire damper assemblies, among other components.

The recommended description in GALL-SLR Report Table XI-01 states that the Fire Protection program requires periodic visual inspection of fire damper assemblies, among other components. GALL-SLR Report Item A-789 (SLRA Table 3.3-1, item 3.3.1-255) identifies the aging effects as [l]oss of material due to general, pitting, crevice corrosion; cracking due to SCC; hardening, loss of strength, shrinkage due to elastomer degradation.

SLRA Section A1.15, Fire Protection and B2.1.15, Fire Protection both use the term fire damper housing. The AMR items in Table 3.3.2-29, Auxiliary Systems - Ventilation - Aging Management Evaluation, that cite Table 3.3.1, Item 3.3.1-255 identify only the housing as a component with aging effects requiring management. In addition, these items cite plant-specific note 3, which states: [t]his row is applicable to fire dampers. Cracking, hardening and loss of strength, and shrinkage are not aging effects requiring management for steel fire dampers exposed to indoor air.

Issue:

The term fire damper assembly includes both the frame and the damper as evidenced by the aging effects requiring management as cited in item A-789. For example, hardening and loss of strength would not be applicable aging effects if the intent of the GALL-SLR Report were to only manage aging effects associated with housings, which are typically constructed of steel materials. Whereas fire damper housing includes just the frame, as evidenced by plant-specific note No. 3. Plant-specific note 3 is not consistent with GALL-SLR Report item A-789. The SLRA lacks a basis for why aging effects will only be managed for the housing versus the damper assembly.

Request:

State the material of construction for the fire damper assemblies other than the housing that perform their intended isolation function in the closed position and the basis for why the aging effects cited in GALL-SLR Report Item A-789 are not applicable to portions of the fire damper assembly other than the housing.

18

RAI- B.2.1.15-2

Background:

SLRA Section B2.1.15 states that the Fire Protection program is consistent with GALL-SLR Report AMP XI.M26, Fire Protection, with no exceptions or enhancements.

The monitoring and trending program element GALL-SLR Report AMP XI.M26 recommends, in part, that results of inspections are trended to provide for timely detection of aging effects and, where identified degradation is projected until the next inspection. In addition, results are evaluated against acceptance criteria to confirm that the timing of subsequent inspections will maintain the components intended functions.

The acceptance criteria program element in GALL-SLR Report AMP XI.M26 recommends specific acceptance criteria for indications of degradation on fire protection components.

Examples include, no visual indications (outside those allowed by approved penetration seal configurations) of cracking, separation of seals from walls and components, separation of layers of material, or ruptures or punctures of seals and no significant indications of cracking and loss of material of fire barrier walls, ceilings.

The corrective actions program element in GALL-SLR Report AMP XI.M26 recommends that, the scope of inspection is expanded to include additional penetration seals in accordance with the plants approved fire protection program should any sign of degradation be detected within the sample of inspected penetration seals. The program element also recommends adjusting inspection frequencies in the event that projected inspection results will not meet acceptance criteria prior to the next scheduled inspection.

Issue:

Based on the staffs review of plant-specific procedures associated with fire protection, the recommendations cited in the three program elements cited above are not included. SLRA Section B2.1.15 does not include enhancements to incorporate these recommendations. The SLRA does not include a basis for why these recommendations have not been addressed.

Request:

Identify the procedures that address the monitoring and trending, acceptance criteria, and corrective actions program elements as described in GALL Report AMP XI.M26 or state the basis as to why the Fire Protection Program is consistent with AMP XI.M26 as-is.

19

TRP 29: Atmospheric Metallic Tanks RAI- B.2.1.17-1

Background:

On April 24, 2019, NRC staff performed a walkdown of the emergency condensate storage tanks (ECSTs). During the walkdown, water was identified around one of the weep drainage holes for the Unit 2 ECST, whereas the remaining weep holes did not have any condensation present. Condition Reports 1121772 and 1121803 state that: (a) a similar condition existed on the Unit 1 ECST; and (b) a sealant will be installed on the missile shield to prevent water intrusion which could cause external corrosion of the tank and potential damage to the external insulation. The condition reports also state that internal inspections of the Unit 1 and Unit 2 ECSTs were completed in 2013 and 2017, respectively, and did not document any concerns regarding the external or internal condition of the tanks.

The detection of aging effects program element in GALL-SLR AMP XI.M29, Outdoor and Large Atmospheric Metallic Storage Tanks states, in part, that [i]f the exterior surface is not coated, visual inspections of the tanks surface are conducted within sufficient proximity to detect loss of material and [i]f the exterior surface of an outdoor tank or indoor tank exposed to condensation is insulated, sufficient insulation is removed to determine the condition of the exterior surface of the tank.

SLRA Section B2.1.17 states an exception to conducting visual and volumetric examinations of the external surfaces of the ECSTs due to the concrete missile shielding and expansion joint filler foam surrounding the tank. The concrete missile shields do not allow visual examinations of the tanks external surfaces as recommended by AMP XI.M29.

Issue:

The duration of the presumably ongoing leakage through the missile shields is unknown. In addition, a review of station drawings indicated that the plug was located above the tank such that any leakage that managed to penetrate the external joint filler foam between the missile barrier and tank could potentially wet the external surface of the tank. Because the tanks are contained within a concrete missile barrier with insulation between, any leakage that penetrates to the surface of the tank could be retained for an extended period, potentially corroding the external surface of the tank.

The summary of the inspections conducted in 2013 and 2017 lacks sufficient detail to justify why external corrosion has not occurred on the tanks as a result of the ongoing leakage. For example, an internal inspection will not detect external corrosion unless a volumetric wall thickness inspection was conducted.

Because of this plant-specific operating experience, SLRA Section B2.1.17 lacks a sufficient basis to justify the exception to AMP XI.M29.

Request:

State the basis for tank integrity will be maintained throughout the SPEO despite the potential for condensation being retained on the surface of the tank and a lack of visual confirmation to prove otherwise.

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TRP 36: External Surfaces Monitoring of Mechanical Components RAI-B.2.1.23-1

Background:

SLRA Section B2.1.23, External Surfaces Monitoring of Mechanical Components, states that after enhancements the existing program will be consistent with GALL-SLR Report XI.M36, External surfaces Monitoring of Mechanical Components. During a review of plant specific operating experience (CR565668 - Pipe Tunnel CC Pipe External Corrosion), the staff noted that loss of material had occurred on the outside surface of the component cooling water (CC) system piping between the pipe and the pipe supports. During clarification discussions, the applicant explained that the general problem was identified as part of the initial license renewal inspections and was addressed through the Infrequently Accessed Area Inspection Activities program. As documented in plant issue S-2002-1794-E1 - Turbine Building to Auxiliary Building Pipe Tunnel, inspection of concrete surfaces at the ends of the turbine/auxiliary-building tunnel revealed ground water in-leakage due to a defect in the tunnel structure. The inspections at that time identified standing water that created an environment conducive to degradation of steel components within the tunnel. Although the external environment in this area would typically be considered as uncontrolled indoor air, Design Change DC-SU-13-00008

- CC Pipe Replacements notes that due to their location near the floor in the turbine/auxiliary-building tunnel, the component cooling water pipes designated as 18-CC-229-121 and 18-CC-235-121 were subject to damp and wet conditions for a number of years, causing corrosion on the outside surfaces of the pipes. DC-SU-13-00008 also notes that the replacement of pipe 18-CC-229-121 was completed in 2015.

The condition report from 2014 (CR565668) notes that wall thickness readings at a pipe support on 18-CC-229-121, which was not accessible until the associated section of piping was removed during scheduled replacement, showed isolated spots below minimum wall thickness.

The condition report states that the overall compensatory measures for the similarly located pipe 18-CC-235-121, which includes yearly wall thickness measurements and quarterly walkdowns of the pipe in the pipe tunnel, should continue until the pipe is restored to maintain piping integrity.

Issue:

As noted in SRP-SLR, Appendix A.1.2.3.10, operating experience for existing programs, including corrective actions that result in program enhancements or additional programs, should be considered. The staff considers the corrective actions to perform the more frequent visual inspections to monitor the environmental conditions in the turbine/auxiliary-building tunnel and the periodic wall thickness measurements of the degraded piping as ongoing condition monitoring activities that manage the effects of aging. Although corrective actions have been initiated to resolve the cause of the degradation, the staff could not determine the overall extent and effectiveness of these actions, based on the documentation provided. In addition, the staff could not determine whether the ongoing aging management activities, which are beyond those specified in the External Surfaces Monitoring of Mechanical Components program, will continue to be performed into the subsequent period of extended operation.

Request:

Provide information discussing the actions taken and their overall effectiveness to address the adverse external environmental conditions in the turbine/auxiliary-building tunnel. Include a discussion whether other activities from the Infrequently Accessed Area Inspection Activities program identified comparable adverse environments that led to significant external corrosion.

21

Also provide information regarding the need to continue the ongoing condition monitoring activities for pipe 18-CC-235-121 into the subsequent period of extended operation, such that a specific aging management review item would be needed to capture this activity in an aging management program.

22

TRP 46: Structures Monitoring Program B.2.1.34-1

Background:

Dominion addressed the age-related degradation of loss of material and change in material properties for wooden power poles by including a plant-specific enhancement to the detection of aging effects program element of the Structures Monitoring Program (SLRA Section B2.1.34) to ensure that wooden power poles are inspected on a 10-year frequency. By letter dated April 2, 2019, Dominion stated that this enhancement follows the EPRI 1010654, Evaluation of Wood Pole Condition Assessment Tools, recommendations for inspection cycles as described in the Wood Pole Assessment Practices section.

SRP-SLR Section A.1.2.3.4 recommends that the discussion for the detection of aging effects program element should provide, in part, justification, including codes and standards referenced, to demonstrate that the technique and frequency are adequate to detect the aging effects before a loss of intended function.

Issue:

The staff notes that the referenced EPRI document describes the ten- to fifteen-year inspection cycle as what is typically performed in North America, but it does not provide a technical bases or justification for the use of such reference as a standard. Thus, it is not clear how the vulnerability of poles to decay, based on the wooden pole locations, were considered for the proposed inspection frequency. Additional justification is needed to demonstrate the adequacy of the proposed 10-year inspection frequency for wooden poles to ensure that the aging effects can be detected before a loss of intended function.

Request:

Provide justification that would demonstrate, pursuant to 10 CFR 54.21(a)(3), that the proposed inspection frequency for wooden poles will be adequate to detect the associated aging effects before a loss of intended function.

23

TRP 76: Irradiation Effects on Carbon Based Steel and RV Steel Support RAI 3.5.2.2.2.6-1 (Concrete and NST Fluence/Dose Estimates)

Background:

Dominions Subsequent License Renewal Application (SLRA) Section 3.5.2.2.2.6, as supplemented by Change Notice 1 dated January 29, 2019 (ADAMS Accession No.

ML19042A137), discusses its Further Evaluation of the aging effects of irradiation on the Concrete Biological Shield (CBS) Wall (or Primary Shield Wall) and the Reactor Vessel (RV)

Support Steel Assembly (consisting of the Neutron Shield Tank (NST) and sliding foot assembly. The SLRA concludes that no plant-specific aging management program to manage the effects of irradiation is required. The SLRA, as supplemented, discusses evaluations in support of Dominions estimation of projected fluence and dose to the end of the subsequent period of extended operation (SPEO) at the CBS wall and at the NST, respectively, for comparison against the applicable threshold criteria for concrete in the SRP-SLR Section 3.5.2.2.2.6, and as input to the fracture mechanics evaluation for embrittlement of the RV support steel assembly.

Issue:

The conclusions made in the SLRA with respect to the need for aging management of the concrete CBS wall, NST, and related RV support structures depends, in part, on the projected fluence/dose at the end of the SPEO. The information presented in the SLRA is not sufficient to allow the NRC staff to determine whether reasonable assurance exists that the limiting fluence/gamma dose values have been identified, with sufficient margin and conservatism to accommodate uncertainties due to the relative lack of validation for fluence analysis methodologies directly applicable to the regions of interest. Therefore, with respect to the fluence/dose values presented in the SLRA and the context stated below, the NRC staff needs additional information:

1. During the audit, the NRC staff reviewed information from calculations performed in 2018 (LTR-REA-18-88 referenced in ETE-SLR-2018-1271) to determine the fluence/gamma dose at selected locations at Surry to the end of SPEO. These values provide additional validation of the fluence/dose values cited in the SLRA and SLRA supplement for the CBS wall and NST. However, the SLRA does not provide details of this model and its results.

2. The SLRA provides information for fluence/gamma dose at the vessel side surface of the CBS wall at the limiting location for the RV traditional beltline region. This location includes attenuation of the fluence through the NST. Based on a review of relevant figures and drawings (e.g., 11448/11548-FV-7A, 11448-FM-1G), there are regions of the CBS wall above and below the NST. The fluence incident on these regions do not appear to the staff to be attenuated by the steel or water present in the NST, so even 24

though these regions are further from the traditional RV beltline, they may experience greater fluence than the part of the CBS wall closest to the RV traditional beltline region.

This is especially true for neutron fluence, since a large number of neutrons would not be moderated by the NST water to energies below the lower threshold for inclusion in the fluence estimates.

Request:

1. Provide a brief summary of the origin, details, and validation of the model used in the calculations in LTR-REA-18-88 referenced in ETE-SLR-2018-1271, including the methodology used and relevant model characteristics, to allow the NRC staff to evaluate the adequacy of the model to compute fluence in areas beyond the traditional beltline region of the RV (i.e., the area of applicability envisioned by the NRC approved methodology in the available regulatory guidance in Regulatory Guide 1.190). In addition, provide a summary of the key limiting results for the CBS wall and the NST.

2. Provide an estimate for the maximum neutron fluences (E > 0.1 MeV) and gamma doses associated with the regions on the vessel side surface of the CBS wall above and below the NST, or a justification for why the fluence/dose in these regions is bounded by other available fluence estimates.

3. If the limiting values of fluence/gamma dose in any portion of the CBS exceed the threshold criteria in SRP-SLR, describe how the aging effects of irradiation on concrete will be adequately managed, pursuant to 54.21(a)(3) in those areas; or, provide a summary of a structural evaluation and its results that demonstrate that the CBS wall will remain capable of performing its intended function through the end of the SPEO.

RAI 3.5.2.2.2.6-2 (Operating Experience Bases)

Background:

One criterion, among others, in SRP-SLR Section 3.5.2.2.2.6 for not requiring a plant-specific program for managing aging effects of irradiation is for the applicant to demonstrate that there is no plant-specific operating experience (OE) of irradiation degradation that may impact intended function(s) of applicable materials and components.

SLRA Section 3.5.2.2.2.6, as supplemented by Change Notice 1 dated January 29, 2019 (ADAMS Accession No. ML19042A137), states no plant-specific OE [operating experience] of concrete irradiation degradation has been identified. The SLRA supplement Section 3.5.2.2.2.6, also states that [t]here is no plant-specific or industry OE of reactor vessel support assembly irradiation degradation that would impact a license renewal intended function.

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Issue:

It is not clear what actions may have formed the bases for SPS to make the above plant-specific OE statements related to irradiation degradation of CBS wall and RV steel support assemblies.

Request:

State what actions (e.g., surveillances, inspections, observations, tests), if any, were taken by SPS to provide justification for the plant-specific OE statements made above for irradiation degradation of CBSW and RV steel support assemblies.

RAI 3.5.2.2.2.6-3 (Whether Structural Consequence Analyses Exists in CLB)

Background:

SLRA Section 3.5.2.2.2.6, as supplemented by Change Notice 1 dated January 29, 2019 (ADAMS Accession No. ML19042A137), states:

The PTR fracture mechanics evaluation on the reactor vessel support steel assembly predated resolution of Generic Safety Issue 15 (GSI-15), Radiation Effects on Reactor Pressure Vessel Supports, in 1996, as reported in NUREG-0933 which states in part:

The preliminary conclusion indicated that the potential problem did not pose an immediate threat to public safety. The tentative results indicated that plant safety could be maintained despite reactor vessel support structures (RVSS) radiation damage. In order to encompass the uncertainties in the various analyses and provide an overall conservative assessment, several structural analyses conducted demonstrated the following:

1. Postulating that one of the four RPV supports was broken in a typical PWR, the remaining supports would carry the reactor vessel and the load even under safe-shutdown earthquake (SSE) seismic loads;

2. If all supports were assumed to be totally removed (i.e., broken), the short span of piping between the vessel and the shield wall would support the load of the vessel.

Issue:

It is not clear if supporting plant-specific structural consequence analyses, that postulate failure of one or more RV support assemblies, like those cited above from NUREG-0933, exists in the current licensing basis (CLB) for SPS Units 1 and 2.

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Request:

State if plant-specific structural consequence analyses, postulating failure of one or more RPV support assemblies, exists in the CLB of SPS Units 1 and 2. If they do exist, describe in sufficient technical detail the consequence analyses performed and its results.

RAI 3.5.2.2.2.6-4 (Apparent Discrepancy of Certain Fluence Values cited in SLRA)

Background:

The criteria in SRP-SLR Section 3.5.2.2.2.6 requires a plant-specific program for managing aging effects of irradiation in concrete if the estimated (calculated) fluence levels or irradiation dose received by any portion of the concrete from neutron (fluence cutoff energy E > 0.1 MeV) or gamma radiation exceeds the respective threshold level stated therein during the subsequent period of extended operation, or if there is plant-specific operating experience (OE) of irradiation degradation that may impact intended functions. SLRA Section 3.5.2.2.2.6, as supplemented by Change Notice 1 dated January 29, 2019 (ADAMS Accession No. ML19042A137), states on page 4 of 6 of Enclosure 2: The maximum neutron fluence at the CBS wall surface of 1.18 x 1013 n/cm2 (E > 1.0 MeV) (emphasis added).

Further, the SLRA supplement Section 3.5.2.2.2.6, under sub-title Irradiation of the RV Support Steel Assembly, of Enclosure 2 states that [t]he PTR [Project Topical Report] was conservatively estimated for 100 years of plant operation (76.8 EFPY [effective full power years]) that yields a fast neutron fluence (E > 1 MeV) of 9.5 x 1019 n/cm2 at the inside surface of the RV and a fast neutron fluence (E > 1 MeV) of 5.0 x 1019 n/cm2 at the outside surface of the RV.

Additionally, Enclosure 2 of the SLRA supplement states: The projected EFPY for SPS SLR is 68 EFPY which yields a fast neutron fluence (E > 1.0 MeV) of 3.42 x 1018 n/cm2 at the inside surface of the NST.

Issue:

1. The estimated neutron fluence level on the CBS wall is cited in the SLRA in terms of cutoff energy E > 1.0 MeV, whereas the neutron fluence acceptance threshold in the SRP-SLR Section 3.5.2.2.2.6 is in terms of cutoff energy E > 0.1 MeV; for appropriate comparison, they need to be stated based on the same cutoff energy as the threshold criteria in the SRP-SLR.

2. The staff audited the Project Topical Report (PTR) 2178-1525314-B4 Unit No. 1 Surry Power Station - Life Extension Evaluation of the Reactor Vessel Support, dated October 10, 1986, and noted that the fast neutron fluence (E > 1.0 MeV) at the outside surface of the RV, used for the evaluation for 100 calendar years of operation (76.8 27

EFPY) is 5.0 x 1018 n/cm2. This fluence value is inconsistent with that of 5.0 x 1019 n/cm2 cited in the SLRA.

3. The staff audited ETE-SLR-2018-1271, Assessment of Radiation Effects on Reactor Vessel Supports for SPS Units 1 & 2, Revision 0, and noted that in its Table 3 the reported fast neutron fluence (E > 1.0 MeV) at the inside vessel side surface of the NST is 3.82 x 1018 n/cm2 for Unit 2 at 68 EFPY. This fluence value is inconsistent with that of 3.42 x 1018 n/cm2 cited in the SLRA.

Request:

1. Provide the maximum calculated neutron fluence values for the CBS wall for SPS Units 1 and 2 based on the cutoff energy for concrete damage as defined in SRP-SLR Section 3.5.2.2.2.6 (i.e., E > 0.1 MeV).

2. Clarify the inconsistency between the fast neutron fluence (E > 1 MeV) at the outside surface of the RV, cited in the SLRA with that used in the PTR for 100 calendar years of operation (76.8 EFPY), and provide the correct value to the end of the SPEO.

3. Clarify the inconsistency between the fast neutron fluence (E > 1 MeV) at the inside (vessel side) surface of the NST cited in the SLRA and that reported in ETE-SLR-2018-1271 for 68 EFPY. State which reactor Unit experiences the bounding fluence and provide the bounding fluence value.

RAI 3.5.2.2.2.6-5 (Applied Stresses and Fracture Mechanics Evaluation - Methodology and Results)

Background:

SLRA Change Notice 1, dated January 29, 2019 (ADAMS Accession No. ML19042A137),

supplemented SLRA Section 3.5.2.2.2.6 with a new subsection entitled, Irradiation of the Reactor Vessel Support Steel Assembly, to address the aging effect of loss of fracture toughness due to neutron irradiation embrittlement of the reactor vessel (RV) support steel materials in the neutron shield tank (NST). The applicants evaluation is based up on the audited Project Topical Report (PTR): Reactor Vessel Support for Unit No 1 Surry Power Station, Life Extension Evaluation of the Reactor Vessel Support, including Appendix 3, Resistance to Brittle Fracture of the Neutron Shield Tank Materials, October 10, 1986.

This supplemental discussion in the SLRA states that, in this PTR evaluation, the applied stresses for the area of the NST subject to high neutron fluence were developed and compared to the critical (allowable) stresses derived from the fracture toughness evaluation. These evaluations were performed to determine the structural integrity of the Surry Unit 1 NST through the end of projected plant life or to the end of the SPEO. The applied stresses were updated in the audited report ETE-SLR-2018-1270, Review of Loads on Neutron Shield Tank for SPS Units 1 & 2 Reactor Vessel Supports, Revision 0. The assessment of the PTR to support the supplemented SLRA is discussed in audited report ETE-SLR-2018-1271, Assessment of Radiation Effect on Reactor Vessel Supports for SPS Units 1 & 2, Revision 0. The evaluations 28

concluded that: a) the applied stresses calculated from the peak stress values for the associated loads of the NST were demonstrated to be below the critical (allowable) stress for a through wall flaw and a surface flaw, and b) loss of fracture toughness due to irradiation embrittlement is not an aging effect requiring management for the NST. The supplemental discussion further states that the PTR evaluation was updated for SLR in ETE-SLR-2018-1271, which validated that the original PTR evaluation is bounding for: a) the Surry Unit 2 NST, b) the applied stresses for both units through the subsequent license renewal period, and c) the 80-year projected fluence values at the inner surface of the NSTs.

NUREG-1509, Radiation Effects on Reactor Pressure Vessel Supports, provides an engineering approach, including screening criteria and technical evaluation procedures, to reassess the structural integrity of the reactor pressure vessel supports.

The staff noted that the audited Attachment 2 of CM-AA-ETE-101, Technical Report CE-0087, Condition Monitoring of Structures, Revision 7, includes up to 10 percent (minor) loss of material in the design of all SPS steel structures. The staff also noted that this report is being revised to include the NST steel structure.

Issue:

In the supplemented SLRA, the applicant provided the conclusions from the PTR and the updated evaluation that addresses the SLR period. However, the SLRA did not provide sufficient docketed details regarding the methodology used in the updated evaluation of the PTR, including derivation of the critical (allowable) and controlling applied stresses to assess the NST structural integrity during the SPEO. It is also not clear if this evaluation was performed consistent with the NRC staff guidelines in NUREG-1509. Therefore, the NRC staff needs additional information to determine the adequacy of the fracture mechanics and applied stress evaluations (subject to a 10 percent reduction in cross sectional areas as noted in Technical Report CE-0087) of the NSTs and the evaluations remain valid through the end of the SPEO.

Request:

1. Identify and justify the specific loads (e.g., seismic, LOCA, anticipated thrust forces exerted by friction if any), loading conditions/loading combinations used or omitted as not applicable in the above postulated fracture mechanics evaluation(s) of the NST for all calculated applied stresses. State the controlling load combination, the limiting applied stresses and its location for the NSTs.

2. State whether all applied stresses considered for the fracture mechanics calculations of the NST were augmented to include the 10 percent reduction in steel section for loss of material due to corrosion as promulgated in Technical Report CE-0087.

3. In regard to the update to the PTR evaluation in report ETE-SLR-2018-1271 to support subsequent license renewal,

a. Describe in detail the methodology used to perform the fracture mechanics evaluation and to calculate the corresponding critical (allowable) stresses with flaws for the NST. Include in this summary the key assumptions and inputs 29

used, and how the evaluation accounted for the complete neutron fluence spectrum (i.e., slow and fast neutrons), added factors of safety to satisfy margins if any, alloy metals in NST steel, and other additional applicable variables.

b. Provide the calculated critical (allowable) stresses for both a through-wall flaw and surface flaw.

4. Demonstrate that the fracture mechanics evaluation accounts for the effects of irradiation embrittlement of the weld metals used and developed heat affected zones of the parent metal in NST.

DRAI 3.5.2.2.2.6-7 (Impact of NST Leakage)

Background:

Scoping and Screening results of mechanical systems of SLRA describes the cooling functions of the Neutron Shield Tank (NST) system. SLRA Section 2.3.1.3, Reactor Coolant, states that

[t]he reactor coolant system includes a neutron shield tank located inside the primary shield wall around the reactor vessel, and that aging management of the neutron shield tank is addressed in the mechanical section of the application. SLRA Section 2.3.3.9, Neutron Shield Tank Cooling, states that [t]he neutron shield tank cooling system provides cooling for the neutron shield tank fluid which is heated by attenuation of neutron and gamma radiation in the vicinity of the reactor vessel. Heat removal is provided by the component cooling system. The neutron shield tank cooling system also removes heat from the primary shield wall. SLRA Section 3.5.2.2.2.6, Reduction of Strength and Mechanical Properties of Concrete Due to Irradiation, identifies the heated water to be contained within the 1-1/2-inch-thick steel shell walls of the tank.

Title 10 of Code of Federal Regulations (10 CFR) Part 54.4 requires that systems, structures, and components including those that assure the integrity of the reactor coolant pressure boundary remain functional during and following design-basis events and that their intended functions form the basis for including them within the scope of license renewal and subject to aging management review (AMR) such that they continue to fulfill their intended function consistent with 10 CFR 54.21(a).

Issue:

UFSAR Section 11.3, states that [p]rimary shielding is provided to limit radiation emanating from the reactor vessel. It also states that [t]he primary shield consists of a water-filled neutron shield tank [which] designed to prevent overheating and dehydration of the concrete primary shield wall and to prevent activation of the plant components within the reactor containment. In its OE audit, the staff reviewed CA238320 included in CR479576 for SPS Unit 2 and noted that the NST has been experiencing chromated water leakage of up to two and one-half gallons per day since 1989. It is not clear how the NSTs can perform their radiation and thermal shielding functions to protect the reactor primary shield wall effectively when they 30

experience unmitigated leakage. It is also not clear what corrective actions the applicant has taken to remedy leakage such as that experienced in the Unit 2 NST or plans to take for any potential NST leakages during the SPEO. It is further not clear what AMPs and AMRs address management of relevant aging associated with NST leakage.

Request:

1. Discuss proposed plans to maintain structural integrity of the primary shield wall (PSW)

(i.e., reduce/eliminate overheating, dehydration, and radiation induced degradation of the reactor primary shield wall) when NSTs experience fluid leakage of fluid conducive to shielding of PSW.

2. Clarify the AMPs and AMRs that manage the impact of chromated fluid leakage from NST on external surfaces of affected components.

RAI 3.5.2.2.2.6-8 (NST Water Chemistry Sampling for Corrosion)

Background:

Scoping and Screening results of mechanical systems of SLRA Section 2.3.3.9, Neutron Shield Tank [NST] Cooling describes cooling of the NST fluid heated by attenuation of neutron and gamma radiation near the reactor vessel. SLRA Section B2.1.12., Closed Treated Water Systems [CTWS] describes activities including chemistry of the fluid used to prevent loss of material to the NST. The audited SPS procedure CH-93.400 Closed Cooling Water Chemistry Program, further delineates the fluid chemistry of the steel NST and indicates that it is monitored every refueling outage. SPS CH-93.400 procedure also states that the NST mitigating fluid chemistry is examined for its alkalinity and contents of chlorides, chromates, and iron.

The enhancement to the detection of aging effects program element of SLRA Section B2.1.12 AMP states that a new SPEO procedure will be developed to inspect a 20% sample of various populations (each material, water treatment program, and aging effect combination) every 10 years. The enhancement also states that if opportunistic inspections will not fulfill the minimum number of inspections by the end of each 10-year period, the program owner will initiate work orders as necessary to request additional inspections.

Issue:

The SLRA identifies NST to be subject to corrosion mechanism and its mitigating fluid to heat and radiation. Given that the CTWS program is a sampling program, it is not clear from the SLRA how the chemistry of the NST fluid is sampled (i.e., at the NST or at other components 31

having the same material, environment, and aging effect characteristics). It is also not clear how the adverse localized environment of heat and radiation affect the chemistry of the contained fluid and if such chemistry has affected the NST internal (e.g., steel, seals, and other materials) construction.

Request:

1. Discuss how, where (including location if sampled within NST), and at what frequency the NST fluid is sampled. If chemistry data are not directly obtained at the NST but at other sampled components discuss the relevance of such components in providing accurate data that can be used to interpret loss of material at the NST.

2. Discuss how the chromated fluid chemistry controls have trended over the plant life.

Provide several years trending of relevant NST chemistry data to asses for loss of material OE evaluation. If chromate data has changed since the beginning of plant operation explain why and justify how so.

3. Discuss to what extent heat and radiation affects the NST fluid chemistry.

RAI 3.5.2.2.2.6-9 (Potential Degradation of Lubrite Lubricant)

Background:

In its SLRA, Section 3.5.2.1.36, Component Supports, the applicant stated that Lubrite is a material of construction used in structural support subcomponents within the containment. The applicant also stated that aging effects such as loss of mechanical function require aging management for component support subcomponents. The applicant proposed to manage the effects of aging of Lubrite exposed to air used to lubricate the sliding foot assemblies for the reactor vessel (RV) supports with the ASME Section XI, Subsection IWF AMP. However, the applicant did not identify whether the Lubrite at the sliding foot assemblies is susceptible to degradation when exposed to radiation.

During the On-Site audit, the staff reviewed excerpts from EPRI Technical Report 3002013084, Structural Tools - Long Term Operations: Subsequent License Renewal Aging Effects for Structures and Structural Components, (EPRI Report) and the Project Topical Report (PTR) for Unit No. 1 Surry Power Station, Life Extension Evaluation of the Reactor Vessel Support dated October 10, 1986, (Life Extension Report) and had discussions with the applicants staff. As stated in the Life Extension Report, the applicant uses Lubrite Type II lubricant to lubricate the bottom of the sliding block for the reactor vessel supports. The Lubrite is described as a solid lubricant comprised of graphite and an organic binder.

However, the staff has not previously accepted the full EPRI Report or the Life Extension Report for use in subsequent license renewal and has not determined the applicability of the statements in these documents to potential aging effects of Lubrite for this application.

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Additionally, both documents discuss the potential for organic materials to degrade when exposed to radiation, and the need to consider this as a potential aging effect.

The EPRI Report contained an excerpt that stated humidity, high temperature, and radiation are not significant in the aging of Lubrite. However, the EPRI Report also states that change in materials properties due to radiation is an applicable aging effect if the gamma radiation exceeds a previously defined limit. Additionally, the EPRI Report recommends that [e]ach plant should review specific material types of manufacturers data for detailed information regarding gamma radiation effects.

The PTR Life Extension Report states that if visual inspections under the ASME Code were implemented, they would not provide an indication of an impending lubrication failure. The PTR goes on to state that due to consequences of binding in the sliding foot assemblies and the potential for lubricant degradation, further study or monitoring for binding is recommended. The PTR Life Extension Report also states that at the time the report was written, it was unknown if radiation tests were performed on the lubricant and that [t]he radiation stability of bonded solid lubricants, like Lubrite II, depends on the properties of the binder. Further, the Life Extension Report states that on-line monitoring to detect stick-slip behavior may be implemented, especially if the long-term properties of the lubricant cannot be reliably ascertained.

Additionally, the PTR Life Extension Report states that it is possible to tolerate radiolytic degradation of the binder if it does not produce an adverse effect on the binders cohesion or adhesion properties. However, the Life Extension Report does not discuss the binders cohesion or adhesion properties for the staff to assess whether the specific binder used in the Lubrite at Surry would be able to withstand radiolytic degradation.

Issue:

As noted in both the EPRI and the PTR Life Extension Reports, Lubrite has the potential to degrade when exposed to radiation. Additionally, audited literature from Lubrite Technologies as provided by the applicant states that radiation can degrade lubricants and therefore each lubricant must be designed to meet the specific conditions encountered. Because the applicant has not provided information that demonstrates the lubricant used at Surry was designed to withstand the expected radiation fluence/dose over 80 years, it is not clear to the NRC staff that the Lubrite used in the construction of RV sliding shoe assemblies will continue to perform its intended function throughout the SPEO, and whether its degradation will not impose additional applied stresses on the NSTs and RVs. Potential loss of lubricating ability of the Lubrite may need to be considered in conjunction with the RAIs dealing with applied stresses for the RPV sliding shoe assemblies.

Request:

1. Clarify which Lubrite lubricant is used in the sliding foot assemblies for the RV supports.

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2. Clarify whether the organic binder is designed to sustain degradation and still ensure the lubricant can perform its intended function for the subsequent period of extended operation:

a. If so, provide the technical justification as to why the binder degradation can be tolerated at Surry. The justification should account for aging effects due to radiation and fluence exposure that would be encountered by the lubricant during the SPEO (60 - 80 year span) at Surry. Discuss whether such degradation would impose additional adverse stress effects and the impact the stresses would have on the ability for the supports to perform their intended function.

b. If necessary, considering any answer to request 2)(a) above, provide qualification data, and compare to site-specific conditions, for the specific lubricant used at Surry that demonstrates the lubricant will not experience significant degradation due to environmental factors such as temperature, accumulated gamma radiation dose and flux, and neutron fluence and flux that this material is projected to receive (or be exposed to) through the SPEO. Note that for any qualification data provided it should include aging effects due to both slow and fast neutrons, if applicable.

c. Considering the answers to a. and/or b. above, is the depletion rate of the lubricant sufficiently low to ensure the lubricant can perform its intended function through the end of the SPEO?

3. If the organic binder for the lubricant contains halogens, provide a discussion on how production of acids may impact corrosion of components in contact with the lubricant and justify why it will not contribute significantly to corrosion of these components.

4. State whether the accumulated gamma radiation dose, and neutron fluence the lubricant is projected to receive through the SPEO will degrade the graphite component of the lubricant. Include qualification data, and compare to site-specific conditions, for the lubricant that demonstrates the graphite component of the lubricant will not experience significant degradation that would impact the intended function of the lubricant or provide a justification for not needing to do so.

5. Based on operating experience data, provide confirmation that no degradation of the Lubrite lubricant (i.e. loss of mechanical function) has been observed at Units 1 and 2 of Surry.

DRAI 3.5.2.2.2.6-10 (Stress Corrosion Cracking of RV Support Sliding Foot Components)

Background:

The NRC staff audited CE-1653, Review of Structural Adequacy of the Reactor Vessel Support Sliding Foot Assemblies - Surry Units 1 and 2 dated May 27, 2003. The report states that major components of each RV sliding foot assemblies (i.e., ball, socket plates, sliding block, stationary saddle block, and hold down plates) are fabricated from high strength maraging Vascomax 300 or 350 steels. The report also states that Vascomax 300 or 350 steels are susceptible to stress corrosion cracking (SCC) subject to environmental conditions.

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The GALL-SLR Report Section IX.F, Aging Mechanisms, and its references state that for certain steels (in particular those containing Nickel) SCC is an aging effect that needs to be managed. SPS UFSAR indicates that Vascomax is a maraging iron-based steel alloy that includes a large percentage of nickel as an alloy strengthening agent.

In addition, the SLRA supplement by letter dated January 29, 2019 (ADAMS Accession No.

ML19042A137), references Project Topical Report (PTR) 2178-1525314-B4 Unit No. 1 Surry Power Station - Life Extension Evaluation of the Reactor Vessel Support, dated October 10, 1986, states that the components of the sliding foot assembly were coated with Heresite' VR 514 (a phenolic coating). The NRC staff audited the PTR and noted that it states that the Heresite' coating may not be needed to prevent stress corrosion cracking of the maraging steel components of the sliding foot assembly unless normal operating loads are exacerbated by lubrication failure. The Surry SLRA as revised, does not appear to discuss the Heresite' coating, or whether it has applicable aging effects requiring management.

Issue:

The Life Extension Report discusses the use of Heresite' VR 514 as a preventive coating to manage Vascomax steels susceptibility for SCC. It is not clear how the applicant would manage SCC of Vascomax steels used in the RV shoe assembly components, if the coating cannot provide the required adequate protection for SCC of Vascomax steels subject to environmental conditions, including radiation exposure, during the SPEO. The staff noted that there was no AMP or AMRs that address the susceptibility of Vascomax steels to SCC. It is also unclear whether the Heresite' VR 514 coating is subject to any aging effects requiring management, and if so, whether degradation of the coating is being managed by any AMPs.

Request:

1. Identify what AMP and AMRs will SPS use to manage the effects of aging due to SCC for the Vascomax steels used in the fabrication of the RV shoe assembly components.

2. State whether the Heresite' coating(s) used, is (are) subject to any aging effects requiring management or credited for corrosion control of components that are in-scope for the SLRA.

3. Clarify and justify if no management of aging effects for Vascomax steels and/or of the Heresite' coating(s) used in the RV shoe assembly components is required.

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TRP 147.3: Leak-Before-Break RAI 4.7.3-7 Regulatory Basis In accordance with 10 CFR 54.21(c)(1), the applicant shall provide a list of time-limited aging analyses (TLAAs), as defined in 10 CFR 54.3. The applicant shall demonstrate that: (i) the analyses remain valid for the period of extended operation; (ii) the analyses have been projected to the end of the period of extended operation; or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

Background

SLRA Section 4.7.3 addresses a TLAA on leak-before break (LBB) for the reactor coolant system (RCS) primary loop. Dominion (applicant) indicated that the LBB analysis for 80 years of operation is documented in WCAP-15550, Revision 2. WCAP-15550, Revision 2 identifies three elbow locations (locations 3, 6 and 15) as critical locations in the LBB analysis.

Issue WCAP-15550, Revision 0 (August 2000) is the basis document for the 60-year LBB analysis of the Surry plant, as indicated in Section IV.1.B.vii.2 of the Surry power uprate application dated January 27, 2010. WCAP-15550, Revision 0 indicates that location 4 is one of the critical elbow locations for the 60-year LBB analysis. In contrast, WCAP-15550, Revision 2 indicates that location 3 is one of the critical elbow locations instead of location 4.

Request Provide the basis for the change to the critical elbow location from location 4 (WCAP-15550, Revision 0) to location 3 (WCAP-15550, Revision 2) to confirm that location 3 is the highest stressed elbow location for the hot leg.

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TRP 147.7: Cracking Associated with Weld Deposited Cladding RAI- 4.7.7-1 Regulatory Basis:

For time-limited aging analyses (TLAAs) that are used as the basis for meeting the aging management requirements of 10 CFR 54.21(a)(3), the regulation in 10 CFR 54.21(c)(1) requires the applicant to demonstrate that the TLAAs are acceptable during the period of operation in accordance with the regulatory criteria in §54.21(c)(1)(i), (ii), or (iii). For TLAAs that are identified and dispositioned in accordance with 10 CFR 54.21(c)(1)(ii), the applicants must demonstrate that the analysis has been projected to the end of period of extended operation.

Background:

The applicant provides its time-limited evaluation of underclad cracking in SLRA Section 4.7.7, Cracking Associated with Weld Deposited Cladding (henceforth, referred to as the TLAA on RPV Underclad Cracking). The applicant states that the current flaw evaluation in WCAP-15338-A, which assessed postulated cladding cracks over a 60-Year licensing basis was reassessed in PWR Owners Group Report No. PWROG-17031-NP, Revision 0, to account for potential flaw growth over an 80-year licensing basis. The applicant states that the TLAA is acceptable in accordance with the criterion stated in 10 CFR 54.21(c)(1)(ii) because the analysis has been projected to the end of the subsequent period of extended operation.

Issue:

In order to demonstrate that the cycle-dependent flaw tolerance or crack growth evaluations of PWROG-17031-NP, Revision 1, do not involve a fluence dependency (as defined for the current operating term in accordance with Criterion 3 in 10 CFR 54.3a), the staff will need further demonstration that the use of a fracture toughness value of 200 ksi-in represents a valid, conservative lower-bound fracture toughness input for the values of KIa and KIc cited in the analysis.

Request:

Please justify the use of a fracture toughness of 200 ksi-in as a conservative, lower bound value for the values of KIa and KIc in the analysis.

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Scoping and Screening RAI- 2.3.1.3

Background:

The systems, structures, and components (SSCs) that are in scope and subject to an aging management review (AMR) are those that perform an intended function as described in 10 CFR 54.4.

Issue:

In Section 2.3.1.3, Reactor Coolant, of the Subsequent License Renewal Application, the applicant stated that the pressurizer spray head does not form part of the reactor coolant pressure boundary or provide structural support of reactor coolant pressure boundary components and is therefore excluded from scope. Staff finds that this statement is not sufficient to determine if the pressurizer spray head should be excluded from scope. As noted in Table 2.3-1 of NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants, some plants rely on the pressurizer spray for pressure control to achieve cold shutdown during certain fire events and, in addition, failure of the spray head should be evaluated in terms of any possible damage to surrounding safety grade components, therefore, this component should be evaluated on a plant-specific basis.

Request:

Staff requests that the applicant provide additional information to justify exclusion of the pressurizer spray head from the scope of AMR by specifically addressing the concerns as noted in Table 2.3-1 of NUREG-2192 as well as the specific criteria of 10 CFR 54.4 (a)(1) - (3).

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