ML22223A145

(Naps), Units 1 and 2 - Subsequent License Renewal Application (SLRA) Second 10 CFR 54.21(b) Annual Amendment

ML22223A145
Person / Time
Site:	North Anna
Issue date:	08/11/2022
From:	James Holloway Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
22-200
Download: ML22223A145 (37)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 August 11, 2022 10 CFR 50 10 CFR 51 10 CFR 54 United States Nuclear Regulatory Commission Serial No.: 22-200 Attention: Document Control Desk NRA/DEA: RO Washington, D.C. 20555-0001 Docket Nos.: 50-338/339 License Nos.: NPF-4/7 VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION (NAPS) UNITS 1 AND 2 SUBSEQUENT LICENSE RENEWAL APPLICATION {SLRA)

SECOND 10 CFR 54.21{b) ANNUAL AMENDMENT By letter dated August 24, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20246G703), Virginia Electric and Power Company (Dominion Energy Virginia or Dominion) submitted an application for the subsequent license renewal of Renewed Facility Operating License Nos. NPF-4 and NPF-7 for North Anna Power Station (NAPS) Units 1 and 2, respectively.

10 CFR 54.21(b) requires Dominion to report changes to the current licensing basis (CLB) that materially affect the contents of the subsequent license renewal application (SLRA),

including the UFSAR supplement. These changes are required to be submitted each year and at least 3 months prior to the scheduled completion of the SLRA review by the NRC.

The first annual review results were provided by letter dated August 5, 2021 (ADAMS Accession No. ML21217A187). The second annual review covers the period from July 1, 2021, through July 1, 2022. Dominion has completed the second annual review and concluded one change to the CLB has been implemented that materially affects the contents of the NAPS SLRA. This letter provides an amendment to various sections of the NAPS SLRA to address this change.

A description of the change and the associated SLRA sections are provided in Enclosure

1. Mark-ups of the affected SLRA pages are provided in Enclosure 2. To aid the staff in assessing changes, Enclosure 2 shows new text as underlined and deleted text as lined through.

Serial No.22-200 Docket Nos. 50-338/339 NAPS SLRA 2nd Annual Amendment Page 2 of 6 If you have any questions or require additional information regarding this submittal, please contact Mr. Paul Aitken at (804) 273-2818.

Sincerely, James E. Holloway Vice President - Nuclear Engineering and Fleet Support

'-r- Kathryn Hill Barret Notary l'ublic Cemmonwealth of Virginia P'teg. No. 7905256 My Commlsalon Expires January 3\ 2!2f COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by James E. Holloway, who is Vice President - Nuclear Engineering and Fleet Support of Virginia Electric and Power Company. He has affirmed before me that he is duly authorized to execute and file the foregoing document on behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this \ \ t\\ day of AuB ys t , 2022.

My Commission Expires: JClJ\l.to.<':f '6 I *, 2021-(

Commitments made in this letter: None

Enclosures:

1. Current Licensing Basis Change that Impacts the SLRA - Second 10 CFR 54.21 (b) Annual Amendment

2. SLRA Mark-ups - Second 10 CFR 54.21(b) Annual Amendment

Serial No.22-200 Docket Nos. 50-338/339 NAPS SLRA 2nd Annual Amendment Page 3 of 6 cc: U.S. Nuclear Regulatory Commission, Region II Marquis One Tower 245 Peachtree Center Avenue, NE Suite 1200 Atlanta, Georgia 30303-1257 NRG Senior Resident Inspector North Anna Power Station Mr. Emmanuel Sayoc NRG Branch Chief U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 0-11 F1 11555 Rockville Pike Rockville, Maryland 20852-2738 Ms. Lauren K. Gibson NRG Branch Chief U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 0-11 F1 11555 Rockville Pike Rockville, Maryland 20852-2738 Mr. Tam Tran NRG Project Manager U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 0-11 F1 11555 Rockville Pike Rockville, Maryland 20852-2738 Mr. G. Edward Miller NRG Senior Project Manager - North Anna U. S. Nuclear Regulatory Commission One White Flint North Mail Stop O-9E3 11555 Rockville Pike Rockville, Maryland 20852-2738

Serial No.22-200 Docket Nos. 50-338/339 NAPS SLRA 2nd Annual Amendment Page 4 of 6 Mr. L. John Klos NRG Project Manager - Surry U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 09 E-3 11555 Rockville Pike Rockville, MD 20852-2738 Mr. Marcus Harris Old Dominion Electric Cooperative Innsbrook Corporate Center, Suite 300 4201 Dominion Boulevard Glen Allen, Virginia 23060 State Health Commissioner Virginia Department of Health James Madison Building - 7th Floor 109 Governor Street Suite 730 Richmond, Virginia 23219 Mr. Mike Rolbard, Director Virginia Department of Environmental Quality P.O. Box 1105 Richmond, VA 23218 Ms. Melanie D. Davenport, Director Water Permitting Division Virginia Department of Environmental Quality P.O. Box 1105 Richmond, VA 23218 Ms. Bettina Rayfield, Manager Office of Environmental Impact Review Virginia Department of Environmental Quality P.O. Box 1105 Richmond, VA 23218 Mr. Michael Dowd, Director Air Division Virginia Department of Environmental Quality P.O. Box 1105 Richmond, VA 23218

Serial No.22-200 Docket Nos. 50-338/339 NAPS SLRA 2nd Annual Amendment Page 5 of 6 Ms. Kathryn Perszyk Land Division Director Virginia Department of Environmental Quality P.O. Box 1105 Richmond, VA 23218 Mr. James Golden, Regional Director Virginia Department of Environmental Quality Piedmont Regional Office 4949-A Cox Road Glen Allen, VA 23060 Mr. Joseph Guthrie, Commissioner Virginia Department of Agriculture & Consumer Services 102 Governor Street Richmond, Virginia 23219 Mr. Jason Bulluck, Director Virginia Department of Conservation & Recreation Virginia Natural Heritage Program 600 East Main Street, 24th Floor Richmond, VA 23219 Mr. Ryan Brown, Executive Director Director's Office Virginia Department of Wildlife Resources P.O. Box 90778 Henrico, VA 23228 Ms. Julie Henderson, Director Virginia Department of Health Office of Environmental Health Services 109 Governor St, 5th Floor Richmond, VA 23129 Ms. Julie Langan, Director Virginia Department of Historic Resources State Historic Preservation Office 2801 Kensington Ave Richmond, VA 23221 Mr. Jamie Green, Commissioner Virginia Marine Resources Commission 380 Fenwick Rd.

Building 96 Fort Monroe, VA 23651

Serial No.22-200 Docket Nos. 50-338/339 NAPS SLRA 2nd Annual Amendment Page 6 of 6 Ms. Angel Deem, Director Virginia Department of Transportation Environmental Division 1401 East Broad St Richmond, VA 23219 Mr. Jason El Koubi, President Virginia Economic Development Partnership 901 East Byrd St Richmond, VA 23219 Mr. William F. Stephens, Director Virginia State Corporation Commission Division of Public Utility Regulation 1300 East Main St, 4th Fl, Tyler Bldg Richmond, VA 23219 Ms. Lauren Opett, Director Virginia Department of Emergency Management 9711 Farrar Ct North Chesterfield, VA 23226 Mr. Mark Stone, Chief Regional Coordinator Virginia Department of Emergency Management 13206 Lovers Lane Culpeper, VA 22701

Serial No.22-200 Page 1 of 3 Enclosure 1 CURRENT LICENSING BASIS CHANGE THAT IMPACTS THE SLRA SECOND 10 CFR 54.21 (bl ANNUAL AMENDMENT Virginia Electric and Power Company (Dominion Energy Virginia or Dominion)

North Anna Power Station Units 1 and 2

Serial No.22-200 Page 2 of 3 CURRENT LICENSING BASIS CHANGE THAT IMPACTS TH E SLRA SECOND 10 CFR 54.21(b) ANNUAL AMENDMENT By letter dated August 24, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20246G703), Virginia Electric and Power Company (Dominion Energy Virginia or Dominion) submitted an application for the subsequent license renewal of Renewed Facility Operating License Nos. NPF-4 and NPF-7 for North Anna Power Station (NAPS) Units 1 and 2, respectively.

10 CFR 54.21 (b) requires Dominion to report changes to the current licensing basis (CLB) that materially affect the contents of the subsequent license renewal application (SLRA),

including the UFSAR supplement. These changes are required to be submitted each year and at least 3 months prior to the scheduled completion of the LRA review by the NRG.

The first annual review results were provided by letter dated August 5, 2021 (ADAMS Accession No. ML21217A187). The second annual review covers the period from July 1, 2021, through July 1, 2022. Dominion has completed the second annual review and concluded the following change to the CLB materially affects the contents of the NAPS SLRA and requires the SLRA to be supplemented:

Current Licensing Basis Change By letter dated September 9, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21252A514), as supplemented by letter dated December 16, 2021 (ADAMS Accession No. ML21350A408), Dominion submitted a request for an amendment to the Technical Specifications (TSs) for NAPS Units 1 and 2.

As requested by Dominion, the Nuclear Regulatory Commission (NRG) approved the proposed changes to the TS for NAPS Units 1 and 2 to adopt Technical Specification Task Force (TSTF) Traveler TSTF-577, Revision 1, "Revised Frequencies for Steam Generator Tube Inspections," on March 22, 2022 (ADAMS Accession No. ML22068A071 ). As part of adopting TSTF-577, the SG tube inspection frequency in NAPS Technical Specification (TS) 5.5.8, "Steam Generator (SG) Program," was revised from at least every 72 effective full power months (EFPM) to at least every 96 EFPM. This allows for the SG primary-side inspections to be performed commensurate with the approved TS inspection frequency of at least 96 EFPM for the SG tubes.

NUREG-2191,Section XI.M19, Steam Generators, Element 3 specifies that SG primary-side inspections will be performed at least every 72 EFPM or every third refueling outage, whichever results in more frequent inspections.

Therefore, as a result of the CLB change described above, this submittal amends the NAPS SLRA Sections below to indicate the Steam Generators program exception to the NUREG-2191,Section XI.M19, inspection frequency requirement of at least every 72 EFPM.

• SRP Table 3.1.1 - Items 3.1.1-025, 3.1.1 -069, 3.1.1-070, 3.1 .1-071, 3.1.1-072, 3.1.1-074, 3.1.1-076, 3.1.1-077, 3.1.1-111, 3.1.1-125, 3.1.1-127 are revised to

Serial No.22-200 Page 3 of 3 indicate exceptions now apply to the NUREG-2191 Aging Management Program (AMP).

• Table 3.1.2-4 is revised to indicate exceptions now apply to the NUREG-2191 AMP.

• Table B2-1 is revised to indicate the Steam Generators program has exceptions to NUREG-2191.

• Section B2.1.10 is revised to update the SG inspection frequency to at least every 96 effective full power months and include the exception to NUREG-2191,Section XI.M19.

• Section A 1.10 is revised to update the SG inspection frequency to at least every 96 effective full power months.

Serial No.22-200 Page 1 of 28 Enclosure 2 SLRA MARK-UPS SECOND 10 CFR 54.21(b) ANNUAL AMENDMENT Virginia Electric and Power Company (Dominion Energy Virginia or Dominion}

North Anna Power Station Units 1 and 2

Serial No.: 22-200 Page 2 of 28 Attachment 2 Table 3.1.1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL-SLR Report Item Aging Aging Ma nagement Further Evaluation Component Discussion Number Effect/Mechanism Program Recommended 3.1.1-019 Stainless steel reactor Cracking due to sec Plant-specific aging Yes (SRP-SLR Consistent with NUREG-2191. Cracking of stainless steel vessel bottom-mounted management program Section 3.1.2.2.6.1) reactor vessel bottom-mounted instrument guide tubes instrument guide tubes (external to reactor vessel) exposed to reactor coolant is (external to reactor vessel) managed by the ASME Code,Section XI lnservice exposed to reactor coolant Inspection, Subsections IWB, IWC, and IWD (82.1.1) program and the Water Chemistry (82.1.2) program.

See further evaluation in Section 3.1.2.2.6 .1.

3.1.1-020 Cast austenitic stainless cracking due to sec AMP XI.M2, Water Chemistry Yes (SRP-SLR Consistent with NUREG-2191. Cracking of cast steel Class 1 piping, piping and plant-specific aging Section 3.1 .2.2.6.2) austenilic stainless steel Class 1 piping, piping components exposed to management program components exposed to reactor coolant is managed by reactor coolant the ASME Code,Section XI lnservice Inspection, Subsections IW8, IWC, and IWD (82.1.1) program and the Water Chemistry (82.1.2) program. See further evaluation in Section 3.1.2.2.6.2.

3.1.1-021 Steel and stainless steel Cracking due to cyclic AMP XI.M1, ASME Code, Yes (SRP-SLR Not applicable - BWR only.

isolation condenser loading Section XI lnservice Section 3.1.2.2.7) components exposed to Inspection, Subsections IWB, reactor coolant IWC, and IWD 3.1.1-022 Steel steam generator Loss of material due Plant-specific aging Yes (SRP-SLR Not applicable. NAPS has no in-scope steel steam feedwater impingement to erosion management program Section 3.1.2.2.B) generator feedwater impingement plate and support plate and support exposed exposed to secondary feedwater in the Reactor Vessel, to secondary feedwater Internals, and Reactor Coolant System. The associated NUREG-2191 aging items are not used.

3.1.1-025 Steel (with nickel alloy Cracking due to AMP XI.M2, Water Yes (SRP-SLR Consistent with NUREG-2191 with exceptions.

cladding) or nickel alloy primary water sec Chemistry, and AMP XI.M19 , Sections 3.1.2.2.11.1 Eis!:eotio s app~ to lbe ~UBEG-2Hl:l re@rnrnemiatio s steam generator primary Steam Generators. In and 3.1.2.2.1 1.2) foe l:llearn Geecatocs (B2 1 1Q) ocogcarn irnoletm: lalio, side components: divider addition, a plant-specific A plant-specific program is not needed for the divider plate and tube-to-tube sheet program is to be evaluated. plate or for the tube-to-tubesheet weld. See further welds exposed to reactor evaluation in Section 3.1.2.2.11.1 and 3.1.2.2.11.2.

coolant North Anna Power Station, Units 1 and 2 Page 3-48 Second Annual Update Application for Subsequent License Renewal

Serial No.: 22-200 Page 3 of 28 Attachment 2 Table 3.1.1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL-SLR Report Item Aging Aging Management Further Evaluation Component Discussion Number Effect/Mechanism Program Recommended 3.1.1-068 Nickel alloy steam Changes in dimension AMP XI.M19, Steam No Not applicable. NAPS has no carbon steel tube support generator tubes exposed to (denting) due to Generators, and AMP XI.M2, plates in the Reactor Vessel, Internals, and Reactor secondary feedwater or corrosion of carbon Water Chemistry Coolant System. The associated NUREG-2191 aging steam steel tube support items are not used.

plate 3.1.1 -069 Nickel alloy steam Cracking due to outer AMP XI.M19, Steam No Consistent with NUREG-2191 with exceptjons generator tubes and diameter SCC, Generators, and AMP XI.M2, Exceptions aPPlll to tbe ~UREG-2191 recommendations sleeves exposed to intergranular attack Water Chemistry for Steam Generatocs (E!2 l l Q) pcogcarn implementatioo, secondary feedwater or steam 3.1.1-070 Nickel alloy steam Cracking due to AMP XI.M19, Steam No Consistent with NUREG-2191 with exceptions.

generator tubes, repair primary water sec Generators, and AMP XI.M2, Exceptions aPPlll to the tl!UBEG-2191 recommendations sleeves, and tube plugs Water Chemistry foe Steam Generators (62,l l Q) 12cogram impleme!a!io ,

exposed to reactor coolant 3.1.1-071 Steel, chrome plated steel, Cracking due to sec AMP XI.M19, Steam No Consistent with NUREG-2191 with exceptions and with stainless steel, nickel alloy or other Generators, and AMP XI.M2, a different program for some component. The ASME steam generator U-bend mechanism(s); loss of Water Chemistry Code,Section XI lnservice Inspection, Subsections IWB, supports including anti- material due general IWC, and IWD (B2.1.1) program is used instead of the vibration bars exposed to (steel only), pitting, Steam Generators (B2. 1.1 0) program to manage secondary feedwater or crevice corrosion cracking and loss of material for the feedwater nozzle steam thermal sleeve. Eisceptions aPPlll to !be ~UBEG-219]

recommendations foe Steam Generators (E!2 1 1Q) peogrnm implemetatio 3.1.1-072 Steel steam generator tube Loss of material due AMP XI.M19, Steam No Consistent with NUREG-2191 with exceptions.

support plate, tube bundle to general, pitting, Generators, and AMP XI.M2, Eisception::. aPPlll to the tl!UBEG-2191 recommendations wrapper, supports and crevice corrosion, Water Chemistry (corrosion foe Steam Genecatoc::. (E!2 1 1Q) pcogcam implementation.

mounting hardware erosion, ligament based aging effects and exposed to secondary cracking due to mechanisms only) feedwater or steam corrosion North Anna Power Station, Units 1 and 2 Page 3-66 Second Annual Update Application for Subsequent License Renewal

Serial No.: 22-200 Page 4 of 28 Attachment 2 Table 3.1.1 Summary of Ag ing Management Prog rams for Reacto r Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL-SLR Report Item Aging Aging Ma nagement Further Evaluation Component Discussion Number Effect/Mechanism Program Recommended 3.1.1-073 Nickel alloy steam Loss of material due AMP XI.M19, Steam No Not applicable. NAPS has no in-scope nickel alloy steam generator tubes and to wastage, pitting Generators, and AMP XI.M2, generator tubes and sleeves exposed to phosphate sleeves exposed to corrosion Water Chemistry chemistry in secondary feedwater or steam in the phosphate chemistry in Reactor Vessel, Internals, and Reactor Coolant System.

secondary feedwater or The associated NUREG-2191 aging items are not used.

steam 3.1.1-074 Steel steam generator Wall thinning due to AMP XI.M19, Steam No Consistent with NUREG-2191 with exceptions upper assembly and flow-accelerated Generators, and AMP XI.M2, Exi;;eQtio s ii!QRlli! to the l'::l!.!BEG-21 !11 cei;;ommem!ii!tions separators including corrosion Water Chemistry for Steam Generi;!tors (B2 1.1 Q) grog ram im12Iementation.

feedwater inlet ring and support exposed to secondary feedwater or steam 3.1.1-075 Steel steam generator tube Wall thinning due to AMP XI.M19, Steam No Not applicable. NAPS has no in-scope steel steam support lattice bars exposed flow-accelerated Generators, and AMP XI.M2, generator tube support lattice bars exposed to secondary to secondary feedwater or corrosion, general Water Chemistry feedwater or steam in the Reactor Vessel, Internals, and steam corrosion Reactor Coolant System. The associated NUREG-2191 aging items are not used.

3.1.1-076 Steel, chrome plated steel, Loss of material due AMP XI.M19, Steam No Consistent with NUREG-2191 with exceQUoos and with stainless steel, nickel alloy to wear, fretting Generators a different program for some component. The ASME steam generator U-bend Code,Section XI lnservice Inspection, Subsections IWB, supports including IWC, and IWD (B2.1.1) program is used instead of the anti-vibration bars exposed Steam Generators (B2. 1. 10) program to manage loss of to secondary feedwater or material for the feedwater nozzle thermal sleeve._

steam Exi;;eQtioos apQlli! to the l'::l!.!BEG-2191 cei;;ommendatio s foe Stearn Genecatocs (B2 1 ml iicogcam imiilementatjon 3.1.1 -077 Nickel alloy steam Loss of material due AMP XI.M19, Steam No Consistent with NUREG-2191 with exceplioos and~ with generator tubes and to wear, fretting Generators a TLAA evaluation included. Wear of steam generator sleeves exposed to tubes at tube support plates is a TLAA, evaluated in secondary feedwater or Section 4.7.8, Steam Generator Tube Wear Evaluation._

steam Exi;;eptioos apQlli! to the l'::!UBEG-2191 cecommendatioos foe Steam Genecatocs (B2 11 Q) ii cog cam imiilemetatioo North Anna Power Station, Units 1 and 2 Page 3-67 Second Annual Update Application for Subsequent License Renewal

Serial No.: 22-200 Page 5 of 28 Attachment 2 Table 3.1.1 Summary of Aging Managemen t Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL-SLR Report Item Aging Aging Management Further Evaluation Component Discussion Number Effect/Mechanism Program Recommended 3.1.1 -111 Nickel alloy steam Reduction of heat AMP XI.M2, Water No Consistent with NUREG-2191 with exceptions.

generator tubes exposed to transfer due to fouling Chemistry, and AMP XI.M19, Exceptions ii!PPI~ to the ~!.!REG-2191 recommern;!;;!lions secondary feedwater or Steam Generators foe Steam Geoeratorn (62 :I :I Ql pcog ram im12Iemeotatioo.

steam 3.1.1 -113 Steel reactor vessel Loss of material due AMP XI.M1, ASME Code, No Not applicable - BWR only.

external attachments to general, pitting, Section XI lnservice exposed to indoor, crevice corrosion, Inspection, Subsections IWB, uncontrolled air wear IWC, and IIM) 3.1 .1 -114 Reactor coolant system Cracking due to sec, AMP XI. M1, ASME Code, No Not applicable. Cracking and loss of material of reactor components defined as IGSCC (stainless Section XI lnservice coolant system components defined as ASME Code, ASME Code,Section XI steel, nickel alloy Inspection, Subsections IWB, Section XI Code Class components (ASME Code Class 1 Code Class components components only), IWC, and IWD, and AMP reactor coolant pressure boundary components or core (ASME Code Class 1 cyclic loading; loss of XI.M2, Water Chemistry support structure components. or ASME Code Class 2 or reactor coolant pressure material due to (water chemistry-related or 3 components - including ASME defined appurtenances, boundary components or general corrosion corrosion-related aging effect component supports, and associated pressure boundary core support structure (steel only), pitting mechanisms only) welds, or components subject to plant-specific equivalent components , or ASME corrosion, crevice classifications for these ASME code classes) is Code Class 2 or 3 corrosion, wear addressed by rows 3. 1.1-020, 3. 1.1-033, 3.1.1-035, components - including 3.1 .1-036, 3.1.1-037, 3.1.1-039, 3.1.1 -042, 3.1 .1-045, ASME defined 3. 1.1 -088, and 3. 1.1-116. The associated NUREG-219 1 appurtenances, component aging items are not used.

supports, and associated pressure boundary welds, or components subject to plant-specific equivalent classifications for these ASME code classes) 3.1 .1 -115 Stainless steel piping, piping None None Yes (SRP-SLR Not applicable. NAPS has no in-scope stainless steel components exposed to Section 3.1.2.2.15) piping, piping components exposed to concrete in the concrete Reactor Vessel, Internals. and Reactor Coolant System.

The associated NUREG-2191 aging items are not used.

North Anna Power Station. Units 1 and 2 Page 3-74 Second Annual Update Application for Subsequent License Renewal

Serial No.: 22-200 Page 6 of28 Attachment 2 Table 3.1.1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL-SLR Report Item Aging Aging Management Further Evaluation Component Discussion Number Effect/Mechanism Program Recommended 3.1 .1-120 Stainless steel core plate Loss of preload due to AMP XI.M9, BWR Vessel Yes (SRP-SLR Not applicable - BWR only.

rim holddown bolts exposed thermal or Internals, and TLAA Section 3.1 .2.2.14) to reactor coolant and irradiation-enhanced SRP-SLR 4.7 Other neutron flux stress relaxation Plant-Specific TLAAs [if an analysis is performed as part of the aging management basis and conforms to the definition of a TLAA in 10 CFR 54.3(a))

3.1.1-121 Stainless steel jet pump Loss of preload due to AMP XI.M9, BWR Vessel No Not applicable - BWR only.

assembly holddown beam thermal or Internals bolts exposed to reactor irradiation-enhanced coolant and neutron flux stress relaxation 3.1.1-124 Steel piping, piping Loss of material due AMP XI.M36, External No Consistent with NUREG-2191.

components exposed to to general, pitting, Surfaces Monitoring of air-indoor uncontrolled, crevice corrosion Mechanical Components air-outdoor, condensation 3.1.1 -125 Nickel alloy steam Cracking due to AMP XI.M19, Steam No Consistent with NUREG-2191 with exceptions generator tubes at support flow-induced vibration, Generators Excei;1tioos ;;ippl~ to tbe ~!.!BEG-2191 cecommeod;;itio s plate locations exposed to high-cycle fatigue foe Steam Geoerntors (62 1 10) progrnm implemeot;;ition.

secondary feedwater or steam North Anna Power Station, Units 1 and 2 Page 3-76 Second Annual Update Application for Subsequent License Renewal

Serial No.: 22-200 Page 7 of 28 Attachment 2 Table 3.1.1 Summary of Aging Managemen t Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL-SLR Report Item Aging Aging Management Further Evaluation Component Discussion Number Effect/Mechanism Program Recommended 3.1.1 -127 Steel (with stainless steel or Loss of material due AMP XI. M2, Water No Consistent with NUREG-2191 with exceptions and a nickel alloy cladding) steam to boric acid corrosion Chemistry, and AMP XI.M19, different aging management program is credited for generator heads and Steam Generators some components. The ASME Code,Section XI tubesheets exposed to lnservice Inspecti on, Subsections IWB, IWC, and IWD reactor coolant (B2.1.1) program will manage loss of material of the steel with stainless steel cladding steam generator primary inlet and outlet nozzles, and the stainless steel primary inlet and outlet nozzle safe ends, exposed to reactor coolant, instead of the Water Chemistry (B2. 1.2) and Steam Generators (B2.1 .1 0) programs. Exceptions

m

illl to the t::lUBEG-2llll cecommem:lalions foe Steam Genecatocs (B2 l lQ) pcogrnm implementation 3.1 .1-128 Stainless steel, nickel alloy Cracking due to sec, AMP XI. M7, BWR Stress No Not applicable - BWR only.

nozzles safe ends and IGSCC Corrosion Cracking, and welds: high pressure core AMP XI.M2, Water Chemistry spray; low pressure core spray; recirculating water, low pressure coolant injection or RHR injection mode exposed to reactor coolant 3.1.1 -129 Steel and stainless steel Cracking due to cyclic AMP XI.M1 , ASME Code , No Not applicable - BWR only .

piping, piping components loading Section XI lnservice exposed to reactor coolant: Inspection, Subsections IWB, welded connections IWC, andlWD between the re-routed control rod drive return line and the inlet piping system that delivers return line flow to the reactor pressure vessel exposed to reactor coolant North Anna Power Station, Units 1 and 2 Page 3-77 Second Annual Update Application for Subsequent License Renewal

Serial No.: 22-200 Page 8 of 28 Attachment 2 Table 3.1.2-4 Reactor Vessel, Internals, and Reactor Coolant System - Steam Generator - Aging Management Evaluation Intended Aging Effect Requiring NUREG-2191 Table 1 Subcomponent Material Environment Aging Management Programs Notes Function(s) Management Item Item Anti-vibration bar ss Stainless (E) Treated water Cracking Steam Generators (82. 1.10) IV.D1.RP-384 3.1.1-071 A,Ei steel >60°C (>140°F) Water Chemistry (82. 1.2) IV.D1.RP-384 3. 1.1-071 A Loss of material Steam Generators (82.1. 10) IV.D1.RP-226 3. 1.1-07 1 A.6.

Water Chemistry (82. 1.2) IV.D1.RP-226 3. 1.1-071 A Steam Generators (82.1. 10) IV.D1 .RP-225 3. 1.1-076 A.6.

Channel head PB Steel with (E) Air - indoor Loss of material External Surfaces Monitoring of Mechanical IV.C2.R-431 3. 1.1-124 C stainless uncontrolled Components (82. 1.23) steel (E) Air with borated Loss of material Boric Acid Corrosion (82.1.4) IV.D1.R-17 3. 1.1-049 A cladding water leakage (I) Reactor coolant Cracking Steam Generators (82. 1.10) IV.D1 .RP-232 3. 1.1-033 E, 2 Water Chemistry (82.1.2) IV.0 1.RP-232 3. 1.1-033 C Cumulative fatigue damage TLAA IV.D1 .R-221 3. 1.1-008 A Loss of material Steam Generators (82 .1.10) IV.01 .R-436 3. 1.1-127 A,Ei Water Chemistry (82. 1.2) IV.D1.R-436 3.1.1-127 A Channel head FD Nickel alloy (E) Reactor coolant Cracking Steam Generators (82 .1.10) IV.D1.RP-367 3. 1.1-025 A.6.

divider plate Water Chemistry (82. 1.2) IV.D1.RP-367 3. 1.1-025 A Cumulative fatigue damage TLAA IV.01 .R-221 3. 1.1-008 C Feedwater FD Steel (E) Treated water Loss of material Steam Generators (82. 1.10) IV.0 1.RP-161 3. 1.1-072 GD.

distribution ring Water Chemistry (82. 1.2) IV.D 1.RP-161 3. 1.1-072 C and J-nozzles Steam Generators (82.1. 10) IV.0 1.RP-225 3. 1.1-076 GD.

(I) Treated water Loss of material Steam Generators (82. 1.10) IV.D1 .RP-161 3. 1.1-072 GD.

Water Chemistry (82. 1.2) IV.D 1.RP-161 3. 1.1-072 C Steam Generators (82. 1.10) IV.01 .RP-225 3. 1.1-076 GD.

Wall thinning Steam Generators (82. 1.10) IV.D1.RP-49 3. 1.1-074 A.6.

Water Chemistry (82. 1.2) IV.D1.RP-49 3.1.1-074 A North Anna Power Station, Units 1 and 2 Page 3-109 Second An nual Update Application for Subsequent License Renewal

Serial No.: 22-200 Page 9 of 28 Attachment 2 Table 3.1.2-4 Reactor Vessel, Internals, and Reactor Coolant System - Steam Generator - Aging Management Evaluation Intended Aging Effect Requiring NUREG-2191 Table 1 Subcomponent Material Environment Aging Management Programs Function(s) Management Notes Item Item Feedwater nozzle PB Steel (E) Air- indoor Loss of material External Surfaces Monitoring of Mechanical IV.C2.R-431 3.1.1 -124 C uncontrolled Components (B2 .1.23)

(E) Air with borated Loss of material Boric Acid Corrosion (B2.1.4) IV.0 1.R-17 3.1.1-049 A water leakage (I) Treated water Cumulative fatigue damage TLAA IV.0 1.R-33 3.1.1-005 A Loss of material ASME Section XI lnservice Inspection, IV.01 .RP-368 3.1.1-012 C Subsections IWB, IWC, and IWD (B2.1.1)

Water Chemistry (B2.1.2) IV.0 1.RP-368 3.1.1-012 C Wall thinning Flow-Accelerated Corrosion (B2.1.8) IV .D1.R-37 3.1.1-061 A Feedwaternozzle LTC Stainless (I) Treated water Cracking ASME Section XI lnservice Inspection, IV.01 .RP-384 3.1.1-071 E, 5 thermal sleeve steel >60°c (>140°F) Subsections IWB, IWC, and IWD (B2.1.1)

Water Chemistry (B2.1.2) IV.0 1.RP-384 3.1 .1-071 C Loss of material ASME Section XI lnservice Inspection, IV.01 .RP-226 3.1 .1-071 E, 5 Subsections IWB, IWC, and IWD (B2.1.1)

Water Chemistry (B2.1.2) IV.0 1.RP-226 3.1 .1-071 C ASME Section XI lnservice Inspection , IV.0 1.RP-225 3.1.1-076 E, 5 Subsections IWB, IWC, and IWD (B2.1.1)

Moisture FD Steel (E) Treated water Loss of material Steam Generators (B2 .1 .10) IV.01.RP-1 6 1 3.1.1-072 G.Q separator Water Chemistry (B2 .1.2) IV.01.RP-161 3.1.1-072 C assembly Wall thinning Steam Generators (B2 .1 .10) IV.01.RP-49 3.1.1-074 A.a Water Chemistry (B2.1.2) IV.01.RP-49 3.1 .1-074 A (I) Treated water Loss of material Steam Generators (B2.1.10) IV.01 .RP-161 3.1.1-072 GQ Water Chemistry (B2.1.2) IV.01.RP-161 3.1 .1-072 C Wall thinning Steam Generators (B2.1.10) IV.01.RP-49 3.1 .1-074 Al;i Water Chemistry (B2.1.2) IV.0 1.RP-49 3.1 .1-074 A North Anna Power Station, Units 1 and 2 Page 3-110 Second Annual Update Application fo r Subsequent License Renewal

Serial No.: 22-200 Page 10 of 28 Attachment 2 Table 3.1.2-4 Reactor Vessel, Internals, and Reactor Coolant System - Steam Generator - Aging Management Evaluation Intended Aging Effect Requiring NUREG-2191 Table 1 Subcomponent Material Environment Aging Management Programs Notes Function(s) Management Item Item Primary inlet and PB Steel with (E) Air - indoor Loss of material External Surfaces Monitoring of Mechanical IV.C2.R-431 3.1.1-124 C outlet nozzle stainless uncontrolled Components (B2. 1.23) steel (E) Air with borated Loss of material Boric Acid Corrosion (B2.1.4) IV.D1.R-17 3.1.1-049 A cladding water leakage (I) Reactor coolant Cracking ASME Section XI lnservice Inspection, IV.D1.RP-232 3. 1.1-033 A Subsections IWB, IWC, and IWD (B2. 1.1)

Water Chemistry (B2.1.2) IV.D 1.RP-232 3. 1.1-033 A Cumulative fatigue damage TLAA IV.01 .R-221 3. 1.1-008 A Loss of material ASME Section XI lnservice Inspection, IV.D1 .R-436 3. 1.1 -127 E, 1 Subsections IWB, IWC, and IWD (B2.1. 1)

Water Chemistry (B2 .1.2) IV.01 .R-436 3. 1.1-127 A Primary inlet and PB Stainless (E) Air - indoor Cracking One-Time Inspection (B2. 1.20) V.A.EP-103b 3.2.1-007 A outlet nozzle safe steel uncontrolled Loss of material One-Time Inspection (B2. 1.20) IV.C2.R-452a 3. 1.1-136 C end (I) Reactor coolant Cracking ASME Section XI lnservice Inspection, IV.D1 .RP-232 3. 1.1-033 A Subsections IWB, IWC, and IWD (B2. 1.1)

Water Chemistry (B2 .1.2) IV.01 .RP-232 3. 1.1-033 A Cumulative fatigue damage TLAA IV.01 .R-221 3. 1.1-008 A Loss of material ASME Section XI lnservice Inspection, IV.D1.R-436 3. 1.1-127 E, 1 Subsections IWB, IWC, and IWD (B2.1.1)

Water Chemistry (B2. 1.2) IV.D1.R-436 3. 1.1-127 A Primary inlet and PB Nickel alloy (E) Air - indoor Loss of material One-Time Inspection (B2. 1.20) IV.C2.R-452a 3. 1.1-136 C outlet nozzle uncontrolled weld (I) Reactor coolant Cracking ASME Section XI lnservice Inspection, IV.C2.RP-159 3. 1.1-045 C Subsections IWB, IWC, and IWD (B2.1.1)

Cracking of Nickel-Alloy Components and Loss of IV.C2.RP-159 3. 1.1-045 C Material Due to Boric Acid-Induced Corrosion in Reactor Coolant Pressure Boundary Components (B2. 1.5)

Water Chemistry (B2.1.2) IV.C2.RP-159 3. 1.1-045 C Loss of material Water Chemistry (B2.1.2) IV.C2.RP-23 3. 1.1-088 A North Anna Power Station, Units 1 and 2 Page 3-111 Second Annual Update Application for Subsequent License Renewal

Serial No.: 22-200 Page 11 of 28 Attachment 2 Table 3.1.2-4 Reactor Vessel, Internals, and Reactor Coolant System - Steam Generator - Aging Management Evaluation Intended Aging Effect Requiring NUREG-2191 Table 1 Subcomponent Material Environment Aging Ma nagement Programs Notes Function(s) Management Item Item Primary manway PB Steel with (E) Air - indoor Loss of material External Surfaces Monitoring of Mechanical IV.C2.R-431 3.1.1-124 C stainless uncontrolled Components (B2.1.23) steel (E) Air with borated Loss of material Boric Acid Corrosion (B2.1.4) IV.01.R-17 3.1.1-049 A cladding water leakage (I) Reactor coolant Cracking ASME Section XI lnservice Inspection, IV.01 .RP-232 3.1.1 -033 A Subsections IWB, IWC, and IWD (B2.1.1)

Water Chemistry (B2.1.2) IV.D1 .RP-232 3.1.1-033 A Cumulative fatigue damage TLAA IV.01.R-221 3.1.1 -008 A Loss of material Steam Generators (B2.1.10) IV.D1.R-436 3.1.1-127 A.B.

Water Chemistry (B2.1.2) IV.D1 .R-436 3.1.1-127 A Primary manway PB Steel (E) Air - indoor Loss of material External Surfaces Monitoring of Mechanical IV.C2.R-431 3.1.1-124 C cover uncontrolled Components (B2 .1.23)

(E) Air with borated Loss of material Boric Acid Corrosion (B2.1.4) IV.D1 .R-17 3.1.1 -049 A water leakage Primary manway PB Steel (E) Air - indoor Cumulative fatigue damage TLAA IV.C2.R-18 3.1.1-005 A cover bolting uncontrolled Loss of material Bolting Integrity (B2.1.9) IV.D1 .RP-166 3.1.1-064 A Loss of preload Bolting Integrity (B2.1 .9) IV.D1.RP-46 3.1.1-067 A (E) Air with borated Loss of material Boric Acid Corrosion (B2.1.4) IV.D1.R-17 3.1.1 -049 A water leakage Primary manway PB Nickel alloy (I) Reactor coolant Cracking ASME Section XI lnservice Inspection, IV.C2.RP-159 3.1.1-045 C cover insert Subsections IWB, IWC, and IWD (B2.1.1)

Cracking of Nickel-Alloy Components and Loss of IV.C2.RP-159 3.1.1-045 C Material Due to Boric Acid-Induced Corrosion in Reactor Coolant Pressure Boundary Components (B2.1.5)

Water Chemistry (B2.1.2) IV.C2.RP-159 3.1.1-045 C Loss of material Water Chemistry (B2.1.2) IV.C2.RP-23 3.1.1-088 A North Anna Power Station, Units 1 and 2 Page 3-112 Second Annual Update Application for Subsequent License Renewal

Serial No.: 22-200 Page 12 of 28 Attachment 2 Table 3.1.2-4 Reactor Vessel, Internals, and Reactor Coolant System - Steam Generator - Aging Management Evaluation Intended Aging Effect Requiring NUREG-2191 Table 1 Subcomponent Material Environment Aging Management Programs Notes Function(s) Management Item Item Secondary PB Steel (E) Air - indoor Loss of material External Surfaces Monitoring of Mechanical IV.C2.R-431 3.1.1-124 C closure cover uncontrolled Components (82.1.23)

(E) Air with borated Loss of material Boric Acid Corrosion (82.1.4) IV.D1.R-17 3.1.1-049 A water leakage (I) Treated water Loss of materia l ASME Section XI lnservice Inspection, IV.D 1.RP-368 3.1.1-012 C Subsections IWB, IWC, and IWD (82.1.1)

Water Chemistry (82.1 .2) IV.D1.RP-368 3.1.1-012 C Secondary PB Steel (E) Air - indoor Cumulative fatigue damage TLAA IV.C2.R-18 3.1.1 -005 A closure cover uncontrolled Loss of material Bolting Integrity (82.1.9) IV.D 1.RP-166 3.1.1-064 A bolting Loss of preload Bolting Integrity (82.1.9) IV.D1 .RP-46 3.1.1-067 A (E) Air with borated Loss of material Boric Acid Corrosion (82.1.4) IV.D1.R-17 3.1.1-049 A water leakage Secondary PB Steel (E) Air - indoor Loss of material External Surfaces Monitoring of Mechanical IV.C2.R-431 3.1.1-124 C manway uncontrolled Components (82.1.23)

(includes pad) (E) Air with borated Loss of material Boric Acid Corrosion (82.1.4) IV.D1 .R-17 3.1.1 -049 A water leakage (l)Steam Loss of material ASME Section XI lnservice Inspection, IV.D1.RP-368 3.1 .1-012 A Subsections IWB, IWC, and IWD (82.1.1)

Water Chemistry (82.1.2) IV.D1 .RP-368 3.1.1-012 A ASME Section XI lnservice Inspection, IV.01 .R-3 1 3.1.1-044 A Subsections IWB, IWC, and IWD (82.1.1)

Secondary side PB Steel (E) Air - indoor Loss of material External Surfaces Monitoring of Mechanical IV.C2.R-431 3.1.1-124 C shell (lower shell, uncontrolled Components (82 .1.23) upper shell, (E) Air with borated Loss of material Boric Acid Corrosion (82. 1.4) IV.D1 .R-17 3.1.1-049 A transition cone, water leakage closure weld ,

(I) Treated water Cumulative fatigue damage TLAA IV.D1.R-33 3.1 .1-005 A girth weld)

Loss of material ASME Section XI lnservice Inspection, IV.D1.RP-368 3.1.1-012 A Subsections IWB, IWC, and IWD (82.1.1)

One-Time Inspection (82.1 .20) IV.D1.RP-368 3.1.1-012 E, 4 Water Chemistry (82.1.2) IV.D1.RP-368 3.1.1-012 A North Anna Power Station, Units 1 and 2 Page 3-113 Second Annual Update Application for Subsequent License Renewal

Serial No.: 22-200 Page 13 of 28 Attachment 2 Table 3.1.2-4 Reactor Vessel , Internals, and Reactor Coolant System - Steam Generator - Aging Management Evaluation Intended Aging Effect Requiring NUREG-2191 Table 1 Subcomponent Material Environment Aging Management Programs Notes Function(s) Management Item Item Secondary side PB Steel (E) Air - indoor Loss of material External Surfaces Monitoring of Mechanical IV.C2.R-431 3. 1.1-124 C shell uncontrolled Components (B2. 1.23)

(penetrations) (E) Air with borated Loss of material Boric Acid Corrosion (B2. 1.4) IV.D1.R-17 3. 1.1-049 A water leakage (I) Treated water Cumulative fatigue damage TLAA IV.01.R-33 3. 1.1-005 A Loss of material ASME Section XI lnservice Inspection, IV.01 .RP-368 3. 1.1-012 A Subsections IWB, IWC, and IWD (B2.1.1)

Water Chemistry (B2. 1.2) IV.0 1.RP-368 3. 1.1-012 A Secondary side PB Steel (E) Air - indoor Loss of material External Surfaces Monitoring of Mechanical IV.C2.R-431 3. 1.1-124 C shell (upper uncontrolled Components (B2. 1.23) head) (E) Air with borated Loss of material Boric Acid Corrosion (B2. 1.4) IV.0 1.R-17 3. 1.1-049 A water leakage (l)Steam Cumulative fatigue damage TLAA IV.D1.R-33 3. 1.1-005 A Loss of material ASME Section XI lnservice Inspection, IV.D1.RP-368 3. 1.1-012 A Subsections IWB, IWC, and IWD (B2. 1.1 )

Water Chemistry (B2. 1.2) IV.D1.RP-368 3. 1.1-012 A Stay rod and ss Steel (E) Treated water Cracking Steam Generators (B2.1.10) IV.D1.RP-384 3. 1.1-071 A.6.

spacer Water Chemistry (B2.1.2) IV.01 .RP-384 3. 1.1-071 A Loss of material Steam Generators (B2. 1.10) IV.D1.RP-226 3. 1.1-07 1 AB.

Water Chemistry (B2.1.2) IV.D 1.RP-226 3. 1.1-07 1 A Steam Generators (B2. 1.10) IV.D1.RP-225 3.1.1-076 AB.

Steam flow limiter RF Nickel alloy (E) Steam Cracking Steam Generators (B2. 1.10) IV.0 1.RP-384 3. 1.1-07 1 GQ Water Chemistry (B2. 1.2) IV.D1.RP-384 3.1.1-07 1 C Cumulative fatigue damage TLAA IV.D1.R-46 3. 1.1-002 C Loss of material Steam Generators (B2. 1.10) IV.D 1.RP-226 3. 1.1-071 GQ Water Chemistry (B2.1.2) IV.D 1.RP-226 3. 1.1-071 C North Anna Power Station, Units 1 and 2 Page 3-114 Second Annual Update Application for Subsequent License Renewal

Serial No.: 22-200 Page 14 of 28 Attachment 2 Table 3.1.2-4 Reactor Vessel, Internals, and Reactor Coolant System - Steam Generator - Aging Management Evaluation Intended Aging Effect Requiring NUREG-2191 Table 1 Subcomponent Material Environment Aging Management Programs Notes Function(s) Management Item Item Steam outlet PB Steel (E) Air - indoor Loss of material External Surfaces Monitoring of Mechanical IV.C2.R-431 3.1.1-124 C nozzle uncontrolled Components (B2 .1.23)

(E) Air with borated Loss of material Boric Acid Corrosion (B2.1.4) IV.D1.R-17 3.1.1-049 A water leakage (I) Steam Cumulative fatigue damage TLAA IV.01.R-33 3.1.1-005 A Loss of material ASME Section XI lnservice Inspection, IV.01 .RP-368 3.1.1-012 A Subsections IWB, IWC, and IWD (B2 .1.1)

Water Chemistry (B2.1.2) IV.D1.RP-368 3.1.1-012 A Wall thinning Flow-Accelerated Corrosion (B2.1.8) IV.01.R-37 3.1.1-061 A Support pad ss Steel (E) Air - indoor Loss of material External Surfaces Monitoring of Mechanical IV.C2.R-431 3.1.1 -124 C uncontrolled Components (B2.1.23)

(E) Air with borated Loss of material Boric Acid Corrosion (B2.1.4) IV.01.R-17 3.1.1-049 A water leakage Tube HT;PB Nickel alloy (I) Reactor coolant Cracking Steam Generators (B2 .1 .10) IV.D1.R-44 3.1.1-070 A.6.

Water Chemistry (B2 .1.2) IV.D1.R-44 3.1.1-070 A Cumulative fatigue damage TLAA IV.D1.R-46 3.1 .1-002 A (E) Treated water Cracking Steam Generators (B2.1.10) IV.D1.R-47 3.1.1-069 A.6.

>60°c (>140°F) Water Chemistry (B2.1.2) IV.D1.R-47 3.1.1-069 A Steam Generators (B2.1 .10) IV.01.R-437 3.1.1 -125 A.6.

Loss of material Steam Generators (B2.1.10) IV.01.RP-233 3.1.1 -077 A.6.

TLAA IV.D1.RP-233 3.1.1-077 E, 3 Reduction of heat transfer Steam Generators (B2.1.10) IV.D1.R-407 3.1.1-111 A,6.

Water Chemistry (B2.1.2) IV.D1.R-407 3.1.1-111 A Tube bundle FD;SS Steel (E) Treated water Loss of material Steam Generators (B2.1.10) IV.D1.RP-161 3.1.1-072 A.6.

wrapper Water Chemistry (B2.1.2) IV.D1.RP-161 3.1.1-072 A Tube plug PB Nickel alloy (E) Reactor coolant Cracking Steam Generators (B2.1.10) IV.D1.R-40 3.1.1-070 A.6.

Water Chemistry (B2.1.2) IV.01.R-40 3.1.1-070 A Cumulative fatigue damage TLAA IV.01.R-46 3.1.1 -002 C North Anna Power Station, Units 1 and 2 Page 3-115 Second Annual Update Application for Subsequent License Renewal

Serial No.: 22-200 Page 15 of 28 Attachment 2 Table 3.1.2-4 Reactor Vessel, Internals, and Reactor Coolant System - Steam Generator - Aging Management Evaluation Intended Aging Effect Requiring NUREG-2191 Table 1 Subcomponent Material Environment Aging Management Programs Notes Function(s) Management Item Item Tube support FD;SS Stainless (E) Treated water Cracking Steam Generators (82. 1.10) IV.D1.RP-384 3.1.1-071 AB.

plate steel >60°c (>140°F) Water Chemistry (82. 1.2) IV.D1.RP-384 3. 1.1-071 A Cumulative fatigue damage TLAA IV.C2.R-18 3. 1.1-005 C Loss of material Steam Generators (82. 1.10) IV.D1 .RP-226 3. 1.1-071 AB.

Water Chemistry (82. 1.2) IV.D1.RP-226 3. 1.1-07 1 A Steam Generators (82. 1.10) IV.D1.RP-225 3. 1.1-076 AB.

Tubesheet PB Steel with (I) Reactor coolant Cumulative fatigue damage TLAA IV.D1 .R-221 3. 1.1-008 A nickel alloy Loss of material Steam Generators (82. 1.10) IV.D1 .R-436 3. 1.1- 127 AB.

cladding Water Chemistry (82.1.2) IV.D1.R-436 3. 1.1- 127 A (E) Treated water Loss of material Steam Generators (82. 1.10) IV.D1.RP-161 3. 1.1 -072 GQ Water Chemistry (82. 1.2) IV.D1.RP-161 3. 1.1-072 C Tube-to- ss Nickel alloy (E) Reactor coolant Cracking Steam Generators (82. 1.10) IV.D1.RP-385 3. 1.1-025 AB.

tubesheet weld Water Chemistry (82. 1.2) IV.D1.RP-385 3. 1.1-025 A Cumulative fatigue damage TLAA IV.D1 .R-221 3. 1.1-008 A Table 3.1.2-4 Plant-Specific Notes:

1. The ASME Section XI lnservice Inspection, Subsections IW8, IWC, and IWD (82. 1.1) program is used instead of the Steam Generators (82. 1.10) program to manage loss of material due to boric acid corrosion for the primary inlet and outlet nozzle and safe end.

2. The Steam Generators (82.1.10) program is used instead of the ASME Section XI lnservice Inspection, Subsections IW8, IWC, and IWD (82. 1.1) program to manage cracking due to stress corrosion cracking for the channel head stainless steel cladding.

3. Wear of steam generator tubes at the tube support plates is a plant-specific TLAA, evaluated in Steam Generator Tube Wear Evaluation (4.7.8).

4. The One-Time Inspection (82. 1.20) program, using magnetic particle test intended to cover essentially 100 percent of the total weld length of each weld on each steam generator, will verify the effectiveness of the Water Chemistry (82. 1.2) program to manage loss of material for the upper shell-to-transition cone girth weld and the new transition cone closure weld.

5. The ASME Section XI Inservice Inspection, Subsections IW8, IWC, and IWD (82. 1.1) program is used instead of the Steam Generators (82. 1.10) program to manage cracking and loss of material for the feedwater nozzle thermal sleeve.

North Anna Power Station, Units 1 and 2 Page 3-116 Second Annual Update Application for Subsequent License Renewal

Serial No.: 22-200 Page 16 of 28 Attachment 2 Tables 3.1.2-1 through 3.1.2-4 Industry Standard Notes:

A. Consistent with NUREG-2191 item for component, material, environment, and aging effect. AMP is consistent with NUREG-2191 AMP.

8. Consistent with NUREG-2191 item for component, material, environment, and aging effect. AMP takes some exceptions to NUREG-2191 AMP.

C. Component is different, but consistent with NUREG-2191 item for material, environment, and aging effect. AMP is consistent with NUREG-2191 AMP.

D. Component is different, but consistent with NUREG-2191 item for material, environment, and aging effect. AMP takes some exceptions to the NUREG-2191 AMP.

E. Consistent with NUREG-2191 item for material, environment, and aging effect, but a different AMP is credited or NUREG-2191 identifies a plant-specific AMP.

F. Material not in NUREG-2191 for this component.

G. Environment not in NUREG-2191 for this component and material.

H. Aging effect not in NUREG-2191 for this component, material and environment combination.

I. Aging effect in NUREG-2191 for this component, material and environment combination is not applicable.

J. Neither the component nor the material and environment combination is evaluated in NUREG-2191.

North Anna Power Station, Units 1 and 2 Page 3-117 Second Annual Update Application for Subsequent License Renewal

Serial No.: 22-200 Page 17 of 28 North Anna Power Station, Units 1 and 2 Application for Subsequent License Renewal Second Annual Update Appendix A - UFSAR Supplement A1.10 STEAM GENERATORS The Steam Generators program is an existing condition monitoring program that manages the aging effects of cracking, loss of material (e.g., wall thinning), and reduction of heat transfer for the steam generators. The scope of the program includes primary-side components (e.g., U-tubes

[tubes], plugs, channel head divider plate, channel head, tubesheet, etc), and secondary-side components that are contained within the steam generator. The program uses volumetric inspections for the tubes, and visual inspections for the other primary-side and secondary-side components. The visual inspections of primary-side components listed above are performed in accordance with the Degradation Assessment (DA) that is prepared as each steam generator is scheduled for examination.

Provisions in the Steam Generators program address reporting criteria, inspection scope and frequency, assessments, plugging criteria, and water chemistry monitoring to maintain consistency with established requirements. NEI 97-06, Revision 3, "Steam Generator Program Guidelines," and associated EPRI guidelines, provide a generic industry program to implement Technical Specifications.

As stated in the steam generator DA, tubing and primary side inspeotions are typically performed every thi rd refueling outage for each steam generator, thus satisfying the guidance for v isual inspections to be performed at least e\*ery 72 effeotive fu ll po,,.1er months or every third refueling outage , whi ohever resu lts in more frequent inspections . By letter dated September 9. 2021 (A ge ncywide Doc uments Acc es s and Mana ge ment Syst em (A DA MS} Acc ess i on No .

ML21252A514}. as supplemented by letter dated December 16. 2021 (ADAM S Access ion No.

ML21350A408}. Dominion submitted a request for an amendment to the Technical Specifications

<TSs} for NAPS Units 1 and 2. As req uested by Dominion , th e Nuclea r Regulatory Commission (N RC} approved the proposed ch anges to th e TS for NAPS Units 1 and 2 to adopt Techn ica l Specification Task Force (TSTF} Traveler TSTF-577. Revision 1. "Revised Frequencies for Steam Generator Tube Inspections," on March 22. 2022 (ADAMS Accession No. ML22068A071}. The change to the TS increased the SG tube inspection frequency from at least every 72 effective full power months (EFPM) to at least every 96 EFPM. This allows for the SG primary-side inspections to be performed commensurate with the approved TS inspection frequency of at least every 96 EFPM for the SG tubes. The DA includes a review of applicable industry operating experience (OE) and plant-specific OE which has occurred since the previous DA was performed. The DA review determines the existence of any unaddressed mechanism that could adversely affect steam generator primary-side or secondary-side integrity, as well as the effects of any chemistry excursions or transients that could affect existing degradation mechanisms.

The Steam Generators program includes preventive measures to mitigate aging related to corrosion phenomena, and through foreign material exclusion as a means to inhibit tube PageA-9

Serial No.: 22-200 Page 18 of 28 North Anna Power Station, Units 1 and 2 Application for Subsequent License Renewal Second Annual Update Appendix A - UFSAR Supplement degradation due to wear. Identification of deposits on the secondary side of the steam generator, and the subsequent removal of sludge deposits help avoid tube degradation.

The Technical Specifications require condition monitoring and operational assessments to be performed to ensure tube integrity will be maintained until the next inspection. The operational assessments are performed after steam generator inspections have been completed to verify structural and leakage integrity.

A1.11 OPEN-CYCLE COOLING WATER SYSTEM The Open Cycle Cooling Water System program is an existing preventive, mitigative, condition monitoring, and performance monitoring program that manages cracking, flow blockage, loss of material, and reduction of heat transfer, for the piping, piping components, and heat exchangers identified in the responses to NRG Generic Letter (GL) 89-13, "Service Water System Problems Affecting Safety-Related Equipment." The program is comprised of the aging management aspects of the Virginia Electric and Power Company response to NRC GL 89-13 and includes: (a) surveillance and control to reduce the incidence offlow blockage problems as a result of biofouling, (b) tests to verify heat transfer of safety-related heat exchangers, (c) routine inspection and maintenance so that loss of material, corrosion, erosion, cracking, fouling, and biofouling cannot degrade the performance of systems serviced by the open-cycle cooling water system. This program includes enhancements to the guidance in NRC GL 89-13 that address operating experience (OE) to provide reasonable assurance that aging effects are adequately managed.

System and component testing, visual inspections, nondestructive examination (i.e., ultrasonic testing and eddy current testing), and chemical injection are conducted to ensure that identified aging effects are managed such that system and component intended functions and integrity are maintained. Periodic heat transfer testing, visual inspection and cleaning of safety-related heat exchangers with a heat transfer intended function is performed in accordance with the Virginia Electric and Power Company commitments to GL 89-13 to verify heat transfer capabilities.

The Internal Coatings/Linings For In-Scope Piping, Piping Components, Heat Exchangers, and Tanks program (A 1.28) will manage the aging effects of internal surface coatings.

A1.12 CLOSED TREATED WATER SYSTEMS The Closed Treated Water Systems program is an existing program that manages cracking, loss of material, and reduction of heat transfer for components exposed to a closed treated water environment.

This is a mitigation program that also includes a condition monitoring program to verify the effectiveness of the mitigation activities. The program consists of: (a) water treatment, including the use of corrosion inhibitors, to modify the chemical composition of the water such that the effects of PageA-10

Serial No.: 22-200 Page 19 of 28 North Anna Power Station, Units 1 and 2 Application for Subsequent License Renewal Second Annual Update Appendix B - Aging Management Programs B2 Ag ing Management Programs Table 82-1 lists the aging management programs described in this appendix and identifies the programs consistency with NUREG-2191. As discussed in Section 81 .4 , both plant specific and industry operating experience has been reviewed and considered as it relates to both new and existing aging management programs.

Table 82-1 NAPS Program Consistency with NUREG-2191 Program Program has Program has Appendix B Existing NU REG-2191 Exceptions to NUREG-2191 Program Reference orNew Enhancements NUREG-2191 ASME Section XI lnservice Inspection, Subsections IWB, 82.1 .1 Existing X IWC, and IWD Water Chemistry 82.1.2 Existing (Primary and Secondary)

Reactor Head Closure Stud Bolting (addressed by ISi 82.1.3 Existing X program)

Boric Acid Corrosion 82.1.4 Existing Cracking of Nickel-Alloy Components and Loss of Material Due to Boric 82.1.5 Existing Acid-induced Corrosion in Reactor Coolant Pressure Boundary Components Thermal Aging Embrittlement of Cast Austenitic Stainless Steel 82.1.6 Existing (CASS)

PWR Vessel Internals 82.1.7 Existing X Flow-Accelerated Corrosion 82 .1.8 Existing Bolting Integrity 82.1.9 Existing X Steam Generators 82.1.10 Existing X Open-Cycle Cooling Water Existing X 82.1.11 System Closed Treated Water Systems 82.1.12 Existing X Page 8-15

Serial No.: 22-200 Page 20 of 28 North Anna Power Station, Units 1 and 2 Application for Subsequent License Renewal Second Annual Update Appendix B - Aging Management Programs 82.1.10 Steam Generators Program Description The Steam Generators program does not include primary-side sleeves since these components are not used in the steam generators. The program uses volumetric inspections for the tubes, and visual inspections for the other primary-side and secondary-side components. The visual inspections of primary-side components are performed in accordance with the Degradation Assessment (DA) that is prepared as each steam generator is scheduled for examination.

The Steam Generators program utilizes industry endorsed guidance regarding tube inspections, evaluation and repair, and leakage monitoring techniques to ensure tube integrity of the steam generators. Aging is managed through assessment of potential degradation mechanisms ,

inspections, tube integrity assessments, plugging and repairs, primary-to-secondary leakage monitoring, maintenance of secondary side component integrity, primary-side and secondary-side water chemistry, and foreign material exclusion. Implementing procedures specify the performance criteria for tube integrity, condition monitoring requirements, inspection scope and frequency, acceptance criteria for the plugging or repair of flawed tubes, acceptable tube repair methods, leakage monitoring requirements, and operational leakage and accident-induced leakage requirements from the Technical Specifications (TS}.

Provisions in the Steam Generators program address reporting criteria, inspection scope and frequency, assessments, plugging criteria, and water chemistry monitoring to maintain consistency with established requirements. Those requirements appear in the following documents:

• Technical Specifications (and Technical Requirements Manual)

• Maintenance Rule (1 0 CFR 50.65)

• EPRI 3002007571, "Steam Generator Integrity Assessment Guidelines"

• EPRI 3002007572, "PWR Steam Generator Examination Guidelines"

• EPRI Technical Report TR1022832, "PWR Primary-to-Secondary Leak Guidelines"

• EPRI 3002007856, "Steam Generator In-Situ Pressure Test Guidelines."

The EPRI guidelines provide a generic industry program to implement the expectations from NEI 97-06, Revision 3, "Steam Generator Program Guidelines."

The original steam generators were replaced for Unit 1 in 1993 and for Unit 2 in 1995. The steam generator replacement projects involved replacing the lower section of each steam generator and refurbishing the upper section. The replacement steam generators incorporated Alloy 690 thermally-treated tubes to improve reliability and minimize aging.

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Serial No.: 22-200 Page 21 of 28 North Anna Power Station, Units 1 and 2 Application for Subsequent License Renewal Second Annual Update Appendix B - Aging Management Programs The Steam Generators program includes plant-specific steam generator DAs that identify existing and potential degradation mechanisms and associated aging effects that could impact the integrity of the steam generators. The DA identifies qualified tube inspection techniques and defines the scope of inspections that are appropriate for the detection and characterization of those aging effects, which consist of cracking, loss of material (e.g., wall thinning), and reduction of heat transfer. As stated in the D.'\, U tube and primary side inspeetions are normally performed every third refueling outage for each steam generator, thus satisfying the guidance for inspeetions to be performed at least &rel)' 72 effeetive fu ll power months or evel)' third refueling outage, '#hiehever resu lts in more frequent inspections.By letter dated September 9. 2021 (Agencywide Documents Access and Management System (ADAM S} Accession No. ML21252A514}. as supplemented by letter dated December 16. 2021 (ADAMS Accession No. ML21350A408}. Dominion submitted a req uest for an amendment to th e Techni ca l Specificati ons (TSs} for NAPS Units 1 a nd 2. As req uested by Dominion . th e Nuclear Reg ulatory Commission (N RC} approved th e pro posed ch anges to th e TS for NAPS Units 1 and 2 to adopt Technical Specification Task Force (TSTF}

Traveler TSTF-577. Revision 1. "Revised Freq uencies for Steam Generator Tube Inspections." on March 22, 2022 (ADAMS Accession No. ML22068A071}. The ch ange to th e TS increased th e steam generator (SG} tube inspection frequency from at least every 72 effective full power months (EFPM} to at least every 96 EFPM. This allows for the SG primary-side inspections to be performed commensurate with the approved TS inspection frequency of at least every 96 EFPM for the SG tub es . The DA includes a review of applicable industry operating experience (OE) and plant-specific OE which has occurred since the previous DA was performed. The DA review determines the existence of any unaddressed mechanism that could adversely affect steam generator primary-side or secondary-side integrity. as well as the effects of any chemistry excursions or transients that could affect existing degradation mechanisms. An excursion of secondary chemistry could lead to fouling of heat transfer surfaces and a reduction of heat transfer thermal performance.

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Serial No.: 22-200 Page 22 of 28 North Anna Power Station, Units 1 and 2 Application for Subsequent License Renewal Second Annual Update Appendix B - Aging Management Programs The DA indicates that primary-side inspections include video/visual examinations, specifically including:

• Tube plugs

• Tube-to-tubesheet welds

• Stub runner and divider plate

• Stub runner to divider plate weld

• Stub runner to tubesheet clad weld

• Divider plate-to-channel head clad weld

• Tubesheet cladding

• Closure ring welds

• Bottom of the bowl cladding The analysis of the steam generator tube-to-tubesheet welds and the channel head design and loading provided by EPRI Technical Report 3002002850, "Steam Generator Management Program: Investigation of Crack Initiation and Propagation in the Steam Generator Channel Head Assembly" is applicable and bounding. A plant specific aging management program is not required for the primary-side channel head. The steam generator tubesheet is clad with Alloy 82, and the Alloy 690 thermally treated tubes are joined to the tubesheet with autogenous welds. General visual inspections of the tubesheet region looking for evidence of cracking (e.g., rust stains on the tubesheet cladding) are performed as part of this program.

The Steam Generators program includes preventive measures to mitigate aging related to corrosion phenomena through foreign material exclusion as a means to inhibit tube degradation due to wear. Identification of deposits on the secondary side of the steam generator, and the subsequent removal of sludge deposits help avoid tube degradation. Sludge mapping occurs when the steam generator is inspected, and inspections for remaining foreign material are performed after sludge lancing is completed. Sludge lancing, steam drum inspections, and feedring inspections typically are performed at least every third refueling outage. As an additional preventive measure, the Water Chemistry program (82.1.2) monitors and controls reactor water chemistry and secondary water chemistry for the steam generators consistent with EPRI 3002000505, "PWR Primary Water Chemistry Guidelines," and EPRI 3002010645, "PWR Secondary Water Chemistry Guidelines".

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Serial No.: 22-200 Page 23 of 28 North Anna Power Station, Units 1 and 2 Application for Subsequent License Renewal Second Annual Update Appendix B - Aging Management Programs I The TS include the following requirements which have been incorporated in the Steam Generators program:

• Conducting condition monitoring assessments for each refueling outage during which steam generator tubes are inspected or plugged.

• Maintaining steam generator tube integrity by meeting performance criteria for tube structural integrity, accident-induced leakage, and operational leakage.

• Installing plugs in tubes found by inservice inspection to contain flaws with a depth equal to, or exceeding, 40% of the nominal tube wall thickness.

• Performing periodic inspections of steam generator tubes. Inspection scope, methods, and interval ensure that tube integrity is maintained until the next planned inspection.

• Monitoring primary-to-secondary leakage.

• Monitoring secondary water chemistry to ensure controls are in place to inhibit steam generator tube degradation.

Non-destructive examination techniques are used to inspect tubing materials in order to identify tubes that may need to be removed from service in accordance with the TS. The Steam Generators program utilizes volumetric examination techniques for the tubes, and visual examinations for other primary-side and secondary-side components. The Steam Generators program defines specific examination techniques, and describes criteria for the qualification of personnel, and for the acquisition and analysis of data. Assessment of tube integrity and plugging criteria of flawed tubes is in accordance with the TS and the Steam Generators program implementing procedures. Tube plugs with indications of aging are evaluated for corrective actions in accordance with the Corrective Action Program and the Steam Generators program.

Condition monitoring assessments are performed to determine whether structural and accident leakage criteria have been satisfied during the previous operating cycle(s). Operational assessments are performed after inspections are completed to verify that structural and leakage integrity will be maintained for the operating interval between inspections, which is selected in accordance with the TS and EPRI Steam Generator Integrity Assessment Guidelines. Comparison of the results of the condition monitoring assessment with the predictions of the previous operational assessment provides feedback for evaluation of the adequacy of the operational assessment and additional insights that can be incorporated into the next operational assessment.

The condition monitoring, and performance monitoring methods, are effective in detecting the applicable aging effects, and the frequency of monitoring is adequate to prevent significant age-related degradation.

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Serial No.: 22-200 Page 24 of 28 North Anna Power Station, Units 1 and 2 Application for Subsequent License Renewal Second Annual Update Appendix B - Aging Management Programs The Steam Generators program is implemented as a Fleet program at Dominion. The Fleet program requirements and Fleet implementation procedures have been previously reviewed and evaluated by the NRC Staff and a determination was made that the effects of aging will be adequately managed so that the intended function(s) will be maintained consistent with the CLB for the subsequent period of extended operation, as required by 10 CFR 54.21(a)(3) (ADAMS Accession No. ML19360A020).

NUREG-2191 Consistency The Steam Generators program is an existing program that is consistent... with exception . to NUREG-2191,Section XI.M19, Steam Generators.

Exception Summary NeAeThe following program element(s) are affected:

Parameters Monitored/Inspected (Element 3) and Detection of Aging Effects (Element 4)

1.Section XI. M19 of NUREG-2 191. Steam Generators. specifies that SG divider plate assemblies. tube-to-tubesheet welds. heads (channel or lower/upper heads). and tubesheets be visually inspected at least every 72 EFPM or every third refueling outage. whichever results in more freq uent in spections. Th e Steam Generators pro gram takes exception t o th e NUREG-2 191 inspection frequency requirement of at least 72 EFPM.

Justification for Exception The Nuclear Regulatory Commission {N RC} approved Amendment Nos. 292 and 275 to the TS for NAPS Units 1 and 2 to adopt Technical Specificati on Task Force (TSTF) Traveler TSTF-577.

Revi sion 1. "Revi sed Freq uencies for Steam Generator Tube Inspections." on March 22. 2022 (ADAM S Accession No. ML22068A071). As part of adopting TSTF-577. th e SG tube inspection frequency in NAPS TS 5.5.8. "Steam Generator (SG} Program." was revised from at least every 72 EFPM to at least every 96 EFPM. This a llows for th e SG prim ary-side inspections to be perform ed commensurate wi th th e approved TS inspecti o n freq ue ncy of at least ev e ry 96 EFPM for the SG tubes.

The revi sion to th e SG tube inspection freq uency to at least every 96 EFPM was based on operating history and justifi ed by a unit-specific operational assessment (OA). Like th e SG tubes, the inspection frequency of at least every 96 EFPM for th e SG prim ary-side inspection s is also based on operating history and justified by a unit-specific OA.

During the Spring 2018 refueling outage for Unit 1 (1 RF26) and the Spring 201 9 refu eling outage for Unit 2 (2RF26). visual inspections of th e prim ary-side components were perform ed and the examination results indicated no age-related deg radation of these components.

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Serial No.: 22-200 Page 25 of 28 North Anna Power Station, Units 1 and 2 Application for Subsequent License Renewal Second Annual Update Appendix B - Aging Management Programs Perform ance of the SG tube inspections and the SG primary-side com ponent inspections both require access via the SG primary-side manways. As such, alignment of at least every 96 EFPM for the visual inspections of SG components with the NRG approved inspection frequency of at least every 96 EFPM for the SG tubes reduces both safety and radiological risk.

Based on the above, th e exception to the NUREG-2191 inspection freq uency requirement of at least every 72 EFPM is justified.

Enhancements None Operating Experience Summary The following examples of operating experience provide objective evidence that the Steam Generators program has been, and will be effective in managing the aging effects for SSCs within the scope of the program so that the intended functions will be maintained consistent with the current licensing basis during the subsequent period of extended operation.

1. In March 2009, a foreign object was found in the Unit 1 'A' steam generator during a foreign object search and retrieval examination. The object was retrieved and the tubes in the vicinity of the foreign object were visually examined. No significant tube wall loss was identified and no further evaluation was required.

2. In September 2009, Regulatory Issue Summary 2009-04 (RIS 2009-04), "Steam Generator Tube Inspection Requirements," provided guidance on the implementation of the steam generator inspection requirements. The Steam Generator Inspections AMA (UFSAR Section 18.2.1 8) was reviewed and found to be consistent with the interpretations provided in RIS 2009-04. No further action was needed.

3. In March 2010, foreign material was found in the Unit 2 'C' steam generator during the post sludge-lancing top-of-tubesheet inspection. The material was found to be weld slag, and was retrieved. Inspection of surrounding tubes noted no damage or marks on tubes. No further action was required.

4. In March 2010, a small metallic piece of material was found in the Unit 2 'A' steam generator during the post sludge-lancing top-of-tubesheet inspection. There were no indications of wear on the surrounding tubes. The piece of foreign material was removed. No further action was needed.

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Serial No.: 22-200 Page 26 of 28 North Anna Power Station, Units 1 and 2 Application for Subsequent License Renewal Second Annual Update Appendix B - Aging Management Programs

5. In April 2010, foreign material was noted in Unit 2 'B' steam generator during the post sludge-lancing top-of-tubesheet inspection. Material was noted as wire-like, and it appeared the ends were embedded in a portion of the remaining sludge. The wire was sized less than 0.020 inch in diameter by less than 0.5 inch in length. The tubes adjacent to the wire, and ten surrounding tubes, were +Point examined and no tube degradation was identified. The exposed portion of the wire broke free during a retrieval attempt. An evaluation was provided in the Condition Monitoring and Operational Assessment document. Due to the small diameter, length, and minimal mass, the remaining portion of the wire was determined to not pose a threat to tube integrity.

6. In September 2011, foreign material was noted protruding from the hot leg end of a tube in the Unit 1 'A' steam generator. An additional piece of foreign material was located in the hot leg bowl. An apparent cause evaluation stated that the presence of the foreign material was a breakdown of the Foreign Material Exclusion boundary for the reactor coolant system. The surface of the steam generator tubesheet did not show any evidence of impacts on the cladding or tube ends. No tube wear was observed in the tube containing the foreign material.

Therefore, there was no evidence that the foreign material caused any damage to the steam generator or reactor coolant system. The foreign material was removed.

7. In September 2013, ultrasonic testing (UT) examination results for Unit 1 'A' steam generator indicated wear in the vicinity of a J-tube nozzle. Additional UT examinations and visual inspections for the internal surfaces of the feedrings for all three steam generators resulted in repairs being planned for several J-tube nozzles during the following outage. An Engineering evaluation concluded that the affected J-tube nozzles were fully capable of performing their design basis function, and would maintain structural integrity for an additional 18-month cycle until the repairs could be performed.

8. In September 2013, during post sludge-lancing visual inspection of Unit 1 'A' steam generator, a loose part was identified at the top of the tubesheet. The part appeared to be weld slag, and was removed. Eddy current testing for the two tubes that were in contact with the loose part determined there was no wall loss on either tube contacted by the loose part. Further eddy current examinations for the row of surrounding tubes did not identify any degradation. No further action was required.

9. In March 2016, a foreign object was identified by eddy current testing in a Unit 2 'C' steam generator cold leg tube at the third tube support plate. Foreign object wear with a maximum wall loss of 33%, believed to be from the same foreign material, was identified on an adjacent cold leg tube. The location was inaccessible for foreign object retrieval. The two tubes were plugged at the cold leg and the hot leg, and were stabilized at the cold leg.

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Serial No.: 22-200 Page 27 of 28 North Anna Power Station, Units 1 and 2 Application for Subsequent License Renewal Second Annual Update Appendix B - Aging Management Programs

10. In December 2016, as part of oversight review activities, a review of procedures credited by initial license renewal AMAs was conducted to confirm the following:

• Procedures credited for license renewal were identified

• Procedures were consistent with the licensing basis and bases documents

• Procedures contained a reference to conduct an aging management review prior to revising

• Procedures credited for license renewal were identified by an appropriate program indicator and contained a reference to a license renewal document Procedure changes were completed as necessary to ensure the above items were satisfied.

11. In May 2017, an assessment was performed to determine the progress and substance of license commitment closure and readiness for the IP 71003 NRC Phase II inspection to be conducted for Units 1 and 2 during November and December of 2017. The conclusion was reached that no areas for improvement or enhancements were identified for the Steam Generator Inspections AMA (UFSAR Section 18.2.18).

12. In October 2017, NSAL-12-1, "Steam Generator Channel Head Degradation", was issued to describe degradation of the Steam Generator channel head cladding in a Westinghouse-designed steam generator. Recommended action from NSAL-12-1 was to perform a visual inspection to identify potential breaches in the cladding. No additional action was necessary since steam generator bowl scans of each channel head are performed during primary-side inspections.

13. In March 2018, foreign material was found on the internal surface of the feedring of Unit 1 'B' steam generator. The piece of foreign material was removed, and was determined to be a backing ring used for initial welding and installation of the feedring. No damage was identified within the feed ring. An inspection of the entire feed ring was conducted and other backing rings and other materials were verified intact. Inspections were also performed within the feedrings for the 'A' and 'C' Steam Generators. Backing rings and other components were intact.

14. In April 2019, an effectiveness review was performed on the Steam Generator Inspections AMA (UFSAR Section 18.2. 18). The AMA was evaluated against the performance criteria identified in NEI 14-12, "Aging Management Program Effectiveness." No gaps were identified by the effectiveness review.

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Serial No.: 22-200 Page 28 of 28 North Anna Power Station, Units 1 and 2 Application for Subsequent License Renewal Second Annual Update Appendix B - Aging Management Programs I The above examples of operating experience provide objective evidence that the Steam Generators program includes activities to perform volumetric and visual inspections to identify cracking, loss of material and reduction of heat transfer for primary-side components and secondary side components contained within the steam generator that are within the scope of subsequent license renewal, and to initiate corrective actions. Occurrences identified under the Steam Generators program are evaluated to ensure there is no significant impact to the safe operation of the plant and corrective actions will be taken to prevent recurrence. Guidance or corrective actions for additional inspections, re-evaluation, repairs, or replacements is provided for locations where aging effects are found. The program is informed and enhanced when necessary through the systematic and ongoing review of both plant-specific and industry operating experience. There is reasonable assurance that the continued implementation of the Steam Generators program will effectively manage aging prior to a loss of intended function.

Conclusion The continued implementation of the Steam Generators program provides reasonable assurance that aging effects will be managed such that the components within the scope of this program will continue to perform their intended functions consistent with the current licensing basis during the subsequent period of extended operation.

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